L-87-295, Forwards Rept of Brief Description of Plant Changes/Mods (Pcms) Made Under Provisions of 10CFR50.59,including Summary of Safety Evaluation for Each.Rept Includes PCMs Completed Between 860123-870122 & Correlates W/Rev 6 of Updated FSAR

Forwards Rept of Brief Description of Plant Changes/Mods (Pcms) Made Under Provisions of 10CFR50.59,including Summary of Safety Evaluation for Each.Rept Includes PCMs Completed Between 860123-870122 & Correlates W/Rev 6 of Updated FSAR

ML20235W844
Person / Time
Site:	Saint Lucie
Issue date:	07/20/1987
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
L-87-295, NUDOCS 8707230742
Download: ML20235W844 (81)

Text

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P. O. BOX 14000, JUNO DEACH, F L 33408 FLORIDA POWER & LIGHT COMPANY _

MLY, Y., O 10 di L-87-295 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: St. Lucie Unit i Docket No. 50-335 Report of 10 CFR 50.59 Plant Changes Pursuant to 10 CFR 50.59(b), the enclosed report contains a brief description of plant changes / modifications (PCM) which were made under the provisions of 10 CFR 50.59, including a summary of the safety evcluation of each. This report includes PCMs completed between January 23, 1986 and January 22,1987 and correlates with the information included in Revision 6 of the Updated Final Safety Analysis Report.

Very truly yours, AZldQf

.o.Wo Group V President Nuclear Energy COW /GRM/gp Enclosure ec: Dr. J. Nelson Grace, Regional Administrator, Region 11, USNRC Senior Resident inspector, USNRC, St. Lucie Plant 0

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gg7230742870720 1 p ADOCK 05000335 PDR ,

PEOPLE. . SERVING PEOPLE GRM5/027/l

I Page 1 of 2 DOCUMENTS REVIEWED FOR ST LUCIE UNIT 1 FUSAR AMENDMENT 6 PC/M SUPP TITLE 414-78 0 REPAIR ICW DRAINS, VENTS AND INST PIPING 544-79 '0,1 105 TON SHANK HOOK FUEL CASK CRANE 580-79 4 ELIM HYDROTEST SUBMITTED IN SUPP 3 078-80 10 ' QUAL DOC AND EVALUATION FOR BAILEY REC AND PWR SUPP 062-81 0 HTR DRAIN PUMP BUSHING MODIFICATION 129-81 0 INSTALLATION OF TEE IN OPC OIL HEADER 133-81 0 HEATER DRAIN PUMP MOTOR REPLACEMENT i 136-81 12 INST OF MI CABLE FOR QSPDS - REVISE CES SHEET 141-81 0 U2 SEC SYS IMUX TIE 025-82 0 TSC RADIO CONSOLE AND CONTROL ROOM TELEPHONES 082-82 0 TURBINE CROSS-UNDER PIPING REPAIR 082-82 1 INCREASE MANWAY COVER BOLT TORQUE 085-82 0 BAM STATION THERMAL RELIEF VALVE INLET PRESS TAPS 045-183 0 RESIN TRANSFER AREA CANOPY 364-183 0 FUEL CASK CRANE BRAKE REPLACEMENT 003-184 0-3 ELECT PENET E-4 CONTAIN SPEC 024-184 0 STM LINE RAD MONITOR WEATHER ENCLOSURE j 037-184 0 CORE SUPPORT BARREL REPAIR 037-184 1 Rx CAVITY REANALYSIS 037-184 2 THERMAL SHIELD REMOVAL ANALYSIS FINAL REPORT 043-184 2 ELECT PENET MODULE - REVISE THE CES 072-184 0 COND PIT FLOOD MOD 072-184 1 SPEC FOR STORM WATER BASIN DREDGING  ;

072-184 2 REVISE EXC LIMITS FOR SOUTH BASIN 110-184 0 CPFD L & N TRANSMITTER REPLACEMENT 135-984 0,1 SEWAGE TREATMENT PLANT, INST OF GATE VALVE 150-984 0 QSPDS INV/PWR LINE COND MOD 158-984 0 DISCHARGE CANAL EROSION 158-984 1 AS-BUILT U2 EROSION REPAIR AND DESIGN FOR U1 REPAIR 158-984 2 REPAIR THE ADDITIONAL EMBANKMENT EROSION 167-184 0 MSIV BYPASS VALVE CABLE REPAIR 172-984 0 WTP ORG SCAVENGER ACID INJECTION SYSTEM 218-184 1 ROSEMOUNT D/P TRANSMITTER EQ DOCUMENTATION 233-184 0 AFW SYSTEM ACTUATION MODIFICATION 253-184 0 Na0CL GAS DISENGAGING TANK BLOWER REMOVAL 012-185 0 OBCW HEAT EXCHANGER CHANNEL GASKET MODIFICATION - U1 021-185 0 DELETE WAVE RUN UP STOP LOGS 037-185 0,1 REPLACE NAMCO LIMIT SWITCHES 043-185 0 TEST FLANGE FOR LOCAL LEAKRATE TESTING 045-185 0 HPSI PUMP OIL DRAIN 045-185 2 DELETE DRAIN VALVE AND ADD NIPPLE AND CAP 089-185 1 ICW ISOL VALVE REPLACEMENT - DOCUMENTATION 103-185 0 REPAIR LEAK IN LINE

Page 2 of 2 DOCUMENTS REVIEWED FOR ST LUCIE UNIT 1 FUSAR AMENDMENT 6 (Con't)

PC/M SUPP TITLE 114-985 0 GOULDS CENT PUMP OIL SEALS 116-985 1 DEMIN WASTE NEVT SYSTEM 117-185 0,1 ICC SYSTEM - LITTON CONN ENHANCEMENTS 125-185 0 STARTUP XFMR DEL SYS ISOL VALVE TAMPER SWITCH 129-185 0 WTP ORGANIC SCAV TOTALIZER 130-185 0 SG LVL XMTR REPLACEMENT 141-185 0 SG N0ZZLE DAM INSERTS 141-185 1,2 N0ZZLE DAM PROC REV 161-185 0 INST AIR / SERV AIR RELOCATE TO WTR TREATMENT PLANT 169-185 0 TURBINE GANTRY CRANE BRAKE MOD 173-185 0 UPPER GUIDE STRUCTURE LIFT RIG MOD 176-185 0 NAMCO LIMIT SWITCH REPLACEMENT 177-185 0 REPLACEMENT OP SOLEN 0ID VALVE 177-185 1 INSTALL SOLENIODS IN MSIV OPENING AND CLOSING CIRCUITS 184-135 0 PSL-1 MISC ICW SYSTEM UPGRADES 193-]85 0 GLAND STEAM SPILLOVER VALVE STATION INSTR TAPS 195-185 0 MSIV AIR ACCUM CHECK VALVE REPLACEMENT 200-185 0 MISC SEC SIDE RELIEF VALVE REPLACEMENT 201-385 0 FEEDWATER LINE/ SUPPORT MOD 207-185 0 TURBINE AUTO-STOP RELIEF VALVE REPLACEMENT 015-186 0 TELEPHONE SYSTEM UPGRADE 055-186 0 CONDENSATE STORAGE TANK SECURITY WALLS 070-186 0 SG FWP LEAKOFF 073-186 0 TURBINE THRUST BEARING - ADD OIL SEALS 077-186 0 10 CFR 50.49 ENV QUAL LIST REVISIONS 079-186 0 STM GEM TUBE PLUGGING The Florida Power and Light to NRC correspondence dated from January 15, 1986 (L-86-013) to January 21, 1987 (L-87-30) has been reviewed for FSAR Amendment j

6. i The following Safety Evaluations were reviewed for FSAR Amendment 6:

I EPO-86-805-E-1 EPO-86-805-E-2 JPE-L-86-83  ;

JPE-M-86-001 JPE-M-86-015 I i

JPE-M-86-019 JPE-M-86-029 JPE-M-86-044 )

JPE-M-86-045 l JPE-M-86-065 JPE-M-86-086 JPE-M-87-001 i 1

PCM 414-78 REPA1R INIAKE COOLING WATER DRAINS, VENTS AND INSTRUMENTATION PIP 1NG RUNS Introduction This procedure covers the installation of weldolets and insulating kits on pipe flanged joints where dissimilar metal pipes are joined together. This procedure also covers the tests required to inrure the adequacy of the insulation during and after the installation of the insulating kits.

Safety Analysis With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possiblity for an accident or malfunction of a different type that 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 which installs weldolets and insulation kits to reduce galvanic action due to dissimilar l metals does not involve an unreviewed safety question because of the following reasons:

(i) The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report will not be increased by this modification because it does not affect the function of the ICW drains, vents or instrument piping runs. The modification is to minimize galvanic action which has caused problems in the system.

(ii) 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 new material used meets the same Piping Class requirements of the existing material with the purpose of increasing system reliability.

(iii) The margin of safety as defined in the bases for any technical specification is not reduced since this modification installs weldolets and insulating ma:erial as replacements for materials which were subject to galvanic actian The ICW system drains, vents and instrument lines were not changed as to their purpose so the bases of any technical specification are not reduced.

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

The foregoing constitutes, per 10CER50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety quesion and prior Commission approval for the implementation of the PCM is not required.

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PCM 544- 79 105 TON SHANK HOOK FOR FUEL' CASK CRANE Introduction The purpose of this PCM is to fabricate a shank hook to replace the existing

. sister book for the Fuel Cask Crane. The present sister hook cannot be rigged to. pick up a spent fuel cask directly. The hook has to'be first attached to the. bale assembly by hanging the bale assembly on one side of the hook. This results in swiveling the hook up at an angle until the central shaft of the

-hook is jammed against the load block.

Safety Analysis-The proposed modification does not involve a change in the Technical Specification incorporated into the license.

The proposed modification does not involve an unreviewed safety question because:

a) The probability of an occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report will not be increased.

b) The possibility for an accident or malfunction of a different type

than any evaluated previously in the Safety Analysis Report will not be created. I c) This hook is to be designed such that its working stresses result in a design factor of 3 for the material's yield strength and 5 for the  !

material's ultimate strength. This complies with the requirements of 10 CFR 71.31 and the original crane design specification.

Revision 47 of the FSAR, dated 7/9/75 included a tabulation of increased (ea fgn factors (12.6 on yield strength and 21 on ultimate strength)  ! resu.' lof using a single element fuel cask instead of the originally planned multi-element fuel cask. Since the increased  ;

design factors were not included as the basis for any technical specification and are not required by any design criteria, the current design requirements shall govern the replacement hook design. Therefore, it is concluded that the margin of safety as defined in the basis for any technical specification is not reduced. .

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PCM 580- 79 ELIMINATION OF HYDROTEST l

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l Introduction The Main Steam Drain Lines' basic function is to drain any condensation accumulated during warm up and normal operation of the main steam piping system.

The steam generated in the Steam Generator enters the steam main and is partially condensed, due to pipe heat losses. These losses are the result of combined ef fects of heat radiation, convection, and conduction through the pipe walls into the ambient atmosphere, and are present with or without insulation in all temperature conditions. It is possible to reach a zero condensate formation situation in superheated steam system where flow is of sufficient rate to prohibit condensation. Condensate is also formed due to the performance of work in the various applications, and in the dead-end section of piping.

Insulation greatly reduces heat loss but does not eliminate it. Fundamentally speaking, a condensate removal device is required to expel from the' working vapor system all condensate, air, oxygen, CO2, and non-condensate gases with a mini-mum loss of. working vapor. To remove condensate from steam systems, a variety of devices have been used, including traps, drilled gate valves, slotted globe valves, etc.

The nain function of the valves on the drain lines are to isolate the main piping system from the rest of the drain system.

The valves on the steam trap bypass lines provide the capability to bypass steam or condensation from the header lines during maintenance to the steam traps or during the priming process to the steam traps.

The vent valves' main function is to vent the piping system during a hydrostatic test or maintenance.

The check valves downstream of the steam traps prevent condensate back flow.

The steam traps on the MSIV's will be isolated during normal operation leaving the free blow valves to be used during startup.

Safety Analysis This change dogs not involve an unreviewed safety question because:

1.'l With respect to the (probability) of occurance of accidents previously evaluated in the FSAR, steam trap and drain lines are not involved in determination of accident probability. Modifications to them would therefore have no impact.

.PCM 580- 79 1.2 With respect to the consequences of accidents previously evaluated in the FSAR, the Saf ety Class 2, Sesimic Category I boundary of the main steam lines has been maintained. This boundary has been defined as upstream of the MSCV's. With the MSCV in its safety or closed position the drain lines are downstream of the check valve seat, thus they may be considered non-nuclear safety related. Therefore, the modification would have no impact on'the consequences of accidents previously evaluated.

1.3 With respect to the probability of malfunction of equipment

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important to safety previously evaluated in the FSAR, the design configuration and parameters have not been changed and the modified portions of the steam trap and drain lines meet the original design criteria. Thus there is no impact on the probability of equipment malf unction.

1.4 With respect to the consequences of malfunction of equipment important to safety previously evaluated in the FSAR, the original plant design criteria has been maintained. The consequences of equipment malfunc-tion would therefore remain unchanged.

2.1 With respect to the po$sibility of an accident of a different type any analyzed in the FSAR, no new types of equipment or modified piping configurations we included in this design. Thus no new types.of accicents have been created.

2.2 With respect to the possibility of an equipment malfunction of a different type than any analyzed in the FSAR, fer the the reasons stated in 2.1 above, no different type of equipnent malfunction have been created.

3.0 With respect to the margin of safety as defined'in the basis for any Technical Specification steam trap and drain lines are not- specifically addressed in the Tech Spec's. Modifications to these lines would therefore, have no impact on the Tech Spec's.

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PCM 078 QUALIFICATION DOCUMmIATION & EVALUATION FOR BAILEY RIXDRDER & POWER SUPPLY INTRODUCTION A. Existing System-The existing In Containment Radiation Monitoring System for St. Lucie Unit No. 1 is composed of the CIS radiation moni-tors and _ the post-LOCA 12-Jiation monitors. The four (4) l CIS detectors are located in ' the containment at the four quadrants at elevation 90 feet. The two (2) post-LOCA-radiation monitors are located outside the containment at elevations 85 feet and 147 feet.

B. New Requirements The requirements resulting from UUREG-0578 and its various clarifications, resulting in NUREG-0737, are itemized be-low:

1) The In Containment High Range Radiation Monitor shguld 'have a range from personnel safety levels to 10 R/hr.

2) The monitors should be located so as to " view" a 1crge cegr.cnt of the containmae f ree "ohnne.

3) The monitors should be safety related.

C. Deficiencies in Existing System The existing In Containment Radiation Monitoring System at St. Lucie Unit No. 1 does not meet all of the above re-quirements. Those requirements that are not met are itemized below:

1) The In Containment CIS monitors do not have the 'en-vironmental qualification nor the range required of the, post-accident detectors.

2) The externally mounted post-accident monitors are not sufficient because they are not mounted within the containment, cannot detect the low energy gammas'and do not have the range required of the post-accident deteetors.

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l PCM 078- 80 4 i

SAFEIY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a prcposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurence or the con-sequences of an accident or malfunction of equipment important to saf ety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunc-tion 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 above described In Containment High Range Radiation Monitor-ing System will have no impact on safety with respect to 10CFR50-

59. This system is implemented in order to assess plant condi-tions during and following an accident.

The conformance of the In Containment High Range Radiation Moni-toring System with the requirements of NUREG-0578, its clarifi-cations and NUREG-0737 is detailed in Table 1. This table in-cludes the criteria, safety evaluation and remarks, and applic-able guidance f rom NUREG-07 37, Item 11.F.1, Attachment 3.

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i PCM 062- 81 i HEATER DRAIN PUMP MODIFICATIONS Introduction Bingham informs us that the subject bushing material had been upgraded from 11-4 Cr to 440A (stainless). The shaft sleeve hardness is 375 to 475 Brinell while the 440A bushings are 275 to 325 Brinell, thus giving a minimum hardness difference of 50 Brinell. Bingham also recommends that the throttle and neck bushing clearance be increased from the existing 0.008"-0.012" to 0.016"-0.020". Delivery on the bushings is estimated at 4 to 6 weeks after receipt of order.

Bingham approves and recommends the welding and re-machining of the seal housing rabbet fit to closer clearances which facilitates alignment. This style seal housing is used on a variety of pumps so Bingham leaves the clearance sloppy to enable it to fit universally. As to the alignment procedure itself we concur with the normal FPL practice of centering the shaft and aligning with the motor in place, i

Safety Analysis The proposed changes to the Heater Drain Pumps described in PCM 062-181 have been recommended by the pump manufacturer as improvements to increase pump reliability.

The Heater Drain Pumps are non-nuclear safety related and are not discussed in the FSAR in relation to any accident analyses. There are no technical specifications related to the Heater Drain Pumps.

No nuclear accidents related to the Heater Drain Pumps or the proposed modifications to the pumps can be postulated.

For the reasons given above, the modifications proposed by PCM 062-181 do not constitute an unreviewed safety question as defined by 10 CFR 50.59.

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PCM 129- 81 INSTALLATION OF TEE IN OPC OIL HEADER Introduction The purpose of this PCM is to install a 1 inch diameter welded S.S. Tee in the DEH OPC Oil Header with reducers to 1/2 inch diameter and to install two 1/2 inch isolation valves to be tagged closed during plant operation. This Tee is to become an instrument tap for a pressure switch that will activate the Solenoid Valves on the Instrument Air to the Extraction Non-return Valve Actuators. The controls portion is to be covered by a subsequent PCM.

Safety Analysis This change will serve to improve the reliability and performance of the Extraction Non-Return Valve Actuators upon Turbine Trip /0PC Actuation. The function of this change is an addition to tne present controls and in no way increases the probability of consequences of a malfunction of the Extraction Non-return Valves. This change poses no r.dditional threat to safety, and in no way decreases the margin of safety as defined in the basis for any Technical Specification. The Extraction Non-return Valves, themselves, are non-nuclear safety related. The one consideration within the Safety Analysis that involves the Turbine Generator is the operation of external missiles by Turbine Destruction. The Non-return Valves do protect the Turbine from potential damage resulting from overspeed or water induction, however, for the above' mentioned reasons, this change does not involve an unreviewed safety question, and in fact, improves the safety aspect of these controls.

This PCM involves just the installation of the Tee and Isolation Valves, and therefore, does not affect these Controls. The installation of the remaining Control modifications is to be covered under a subsequent PCM.

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PCM 133- 31 HEATER DRAIN PUMP MOTOR REPLACEMENT Introduction Heater drain pumps are provided for pumping the collected drains of the heater drain tanks to the suction of the feedwater pumps.

The presently lleater Drain Pumps lA 6 10 motors are experiencing a higher duty than predicted due to a hydraulic anomoly of the

.hcater drain pump. This PCM uprate the pump's motor horsepower to meet the higher duty. requirement of the driven pumps.

Safety Analysis The changes involved in this PC/M design package are not Nuclear Safety Related items, and are not a part of a Safety Related Design Fea. ore for the following reasons:

(1) The Heater Drain Pumps are not used to prevent postulated accident or to mitigate the consequences of such accidents or to achieve or maintain safe shutdown conditions, or to adequately cool spent stored -

fuel.

(2) The Heater Drain Pumps are not associated with and can-not interfere with Nuclear Safety Related items.

(3) The Heater Drain Pu[ps are not included in the PTP.

l Q-List (CPP-QI 2.3A).

PCM 136- 81 MI CABLE INSTALLATION INTRODUCTION The qualified Safety Parameter Display System (QSPDS) being in-stalled at St Lucie Unit 1 satisfies the NUREG 0737 requirement for a redundant Class IE Inadequate Core Cooling (ICC) instrumentation processing and display system. The QSPDS consists of a Subcooled Margin Monitor System, Heated Thermocouple System and a Core Exit Thermocouple System.

The QSPDS is a micro-processor based signal _ processing system .with an RTGB mounted plasma display unit and associated keyboard for each of.the two channels. Each channel recieves ~ and processes signals and transmits the output to the plasma display unit.

The NRC requirements for ICC detection : necessitate that the instru-mentation systems to be used for that safety functionsurvival.

be subjectedCon-to a qualifications process related to post accident sequently, the CET's for the detection of ICC requires that the CET in-containment cabling be qualified to post accident environmental conditions.

There are four separate PCM's comprising BFI 75-2 (QSPDS). These are: ,

1. PCM 31 QSPDS Inputs.

2. PCM 32 QSPDS Power Supply ,

3. PCM 136 MI Cable

4. PCM 78 Temporary Installation of Class IE Batteries An associated PCM (part of BFI 128-10) is PCM 59-82 which covers the installation of the plasma display unit and associated equipment in the RTGB.

This PCM of (136-81) deals the Heated with the Junction installation of(HJCT)

Thermocouple the in-containment and the Core portion Exit Thermocouple (CET) Mineral Insulated (MI) Cables.

SAFETY ANALYSIS This modification has been reviewed with respect to Title 10 of the Code of Federal Regulations, Part 50.59 which states that a proposed (i) l change shall be deemed to involve an unreviewed safety questio i

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PCM 136- 81 or malfunction of equipment important to safety previously evaluat'ed in the safety analysis report may be increased; or (ii) if a possi-bility 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 Qualified Safety Parameter Display System is a seismic safety Class IE system. All components associated with the system have been tested and qualified to IEEE 323-1974 and IEEE 344-1975 stan-dards. The QSPDS cabinets have been seismically mounted in the Con-trol Room. Complete divisional separation is maintained by dedica-ting separate QSPDS cabinets, displays, containment penetrations and cable raceways for each redundant safety class. Thus, there is total fire, missile and electrical isolation between divisions. Fiber optic cable provided the necessary isolation between the QSPDS and the non-safety Safety Assessment System.

All instrument modifications and installations are according to FSAR Section 7.1 and 7.5 commitments concerning safety class ins trumen-tation.

The installation of the cable trays and cabling has been done in accordance with the criteria set forth in FSAR Chapter 8.

Additionally, the QSPDS is a monitoring and indicating system only and as such does not interact with the automatic initiation of any protection systems.

The MI cable, Litton connectors and Electrical Penetrations are de-signed to meet the criteria and tests imposed by IEEE Standards 323-1974 and 344-1975 The cable support structure and tube track systems and supports are designed as Seismic Category 1.

There fore an unresolved safety question does not exist and prior Commission approval is not required.

l PCM 141- 81 ST. LUCIE UNIT 1/ UNIT 2-SECURITY SYSTEMS

Introduction l

The ' plant. . Security System being installed at Unit #2 is being

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j extended to Unit #1.for eventual replacement of the existing  ;

Unit'#1 security system. j

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This PCM will cover the following aspects: j

1. The installation; of Unit 1 multiplexer '. (total of seven). 'l IMUX- .01 -will be ' installed : in the Security and . Records l buildings IMUX #2,3, and 4 will be installed in the Reactor l

' Auxiliary Building. . IMUX #5 will be installed in the Tur- l bine Generator Building. IMUX #13 will be installed in the i Steam Generator' Blowdown Building, and IMUX #14 will be in- '

stalled.in the Service Building. {

2. - The Loop with Unit 2 multiplexer is done between IMUX #13  !

(Unit 1) and IMUX #8 (Unit 2). The complete loop will allow i testing of Unit 2 software and incorporation of Unit 1 por-  ;

tions of the system with minimum security disruptions at  !

i Unit 1.

3. 120V AC Power Supply to the IMUX #2,3,4, and 5 from the i existing 120VAC Power Panel PP-137 located in the RAB.

4. 120V AC Power Supply to IMUX #1,13, & 14 from the new 120V AC power panel PP-137B to be installed in the existing Data Gathering Panels (DGP) room. The power feed to the new PP-137B will be from the same circuit #2 of PP-137A. The existing PP-137A will eventually be phased out. The Unit 1 l CAS, located in the Security and Records Building will be '

dismantled to allow room for expansion of the entry aisles.

The power supply to the systems presently being supplied i from PP-137A will eventually be transferred to the new PP-137B (It is expected to be 3 circuits only).

l S. Due to the remote locatien of the different IMUX's and i IDP's (Intrusion Detection Panel) and from voltage drop  ;

consideration it is'necessary to install one 10KVA 120/430v i step up transformer in the RAB near to PP-137 and one 10KVA  !

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4 80/120V . step down transformer near PP-137B in the DGP's room in the Security'and Records Building. ,

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6. The cable layout will utilize the existing duct banks and j raceways, the new duct banks issued per PC/M 96-81, and new l raceways / conduits.

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PCM 141- 81 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 unroviewed safety question; (i) if the probability of occurence or the con-sequences of an accident or malfunction of equipment important to safety previously evaluate 3 in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The industrial security system as referenced in FSAR Chapter 13.7 is not a safety related system and the installation of the seven multiplexer and power panel does not constitute an unreviewed safety question.

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1 PCM 025- 82 l

I TECHNICAL SUPPORT CENTER RADIO CONSOLE l

INTRODUCTION This PC/M will cover the installation of a battery backed 120V AC feeder for a new TSC Radio Console which will be located in the Technical Support Center library in the Reactor Auxiliary Building . Also, this PC/M includes the installation of 4 and 3 inch seismically supported conduits between the TSC Radio Con-sole and the communication tray in the Cable Spreading Room and the relocated telep hone equip ment and the Control Room table console.

SAFETY ANALYSIS This modification has been reviewed with respect to Title 10 of the Code of Federal Regulations , Part 50.59 which states that a proposed change shall be deeeed to involve an unreviewed safety question; (i) if the p robability of occurrence or the conse-quences of an accident or malfunction of equiptent i=p ort a nt to safety p reviously evaluated in the safety analysis rep ort may be increased; or (ii) if a possibility for an accident or mal-function of a different t yp e than any evaluated p reviously in the safety analysis rep ort may be created ; or (iii)if the mar-gin of safety as defined in the basis for any technical speci-fication is reduced.

The installation of the seismically supported conduits and the battery backed 120 V AC have been done in'accordance with the criteria set forth in FSAR Chapter 8 and does not imp act on the functionality of any safety system in the plant. Hence, it does not con s titu te an unreviewed safety question nor require prior approval by the tbclear Regulatory Commission.

PCM 082- 82 TURBINE CROSS-UNDER PIPING REPAIR Introduction The purpose of this modification is to upgrade the present condition of PSL-1 turbine cross under (cold reheat) piping. The turbine cross

!- under piping supplies the four (4) Moisture Separator Reheaters (MSRs) with the high pressure turbine exhaust steam.

l l Safety Analysis The turbine cross under pipe modification outlined in this design package is not nuclear safety related, nor is it part of a safety related design j feature because:

a. The cross under pipe is part of a non-seismic, non-safety related system.

b. Failure of the cross under pipe vill have no effect on safe shutdown of the plant.

c. The cross under pipe is not used to prevent postulated accidents mitigate the consequences of such accidents, maintain safe shut-down conditions or adequately store spent fuel.

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k PCM 085- 82 BAM STATION TilERMAL RELIEF VALVE INLET PRESS TAPS INTRODUCTION The purpose of this modification is to. provide pressure taps on the inlets to Boric Acid Makeup (BAM) Station j thermal relief valves V2173 V2426 and V2436. The pressure j taps are to be used for the BAM Station thermal relief valve j surge pressure test and are temporary connections to be removed l after completion of the test.

SAFETY ANALYSIS This modification installs temporary pressure taps on the inlets to BAM Station thermal relief valves V2426, V2436 and V2173.

The pressure caps are temporary and require the removal of the above listed relief valves for installation. Therefore, valves V2154, V2155 and V2172 must be administratively maintained open to ensure thermal overpressure protection is provided during heat tracing heat up for the piping normally protected by the removed thermal relief valves V2426, V2436 and V2173. I This temporary modification does not involve an unreviewed safety question because:

1. The boric acid makeup system is not utilized in the determination of accident probabilities. Therefore, the probability of occurrence of an accident previously evaluated in the F3AR is not increased.

2. Since the thermal relieving capacity is administrative 1y maintained the boric acid makeup function is not. compromised by this modification, thus, no accident previously discussed in the FSAR would be made more serious by the modification.

3. This modification involves temporarily replacing thermal relief valves while still maintaining proper overpressure protection.

Therefore, the probability of the malfunction of equipment important to safety previously evaluated in the FSAR remains unchanged.

4. This modification does not have any effect upon the consequences i of malfunction of equipment important to safety because of the reasons stated in ite: 2 above.

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PCM 085- 82

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5. -This' codification is stithin the ' design criteria of satety class II system and invo'ves equipment which is installed elsewhere in the plant. Therefore the possibility of an accident not considered in the ISAR is not created.

6, For the reasons stated in 2, 3 and 5 above, this modification does not create the possibility of malfunction of equipment not 1 considered in the FSAR.

7. No basis for any Technical.5 specification is affected by this modification. therefore no margin of safety is decreased.

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l PCM 045-183 1(L >

RESIN TRANSFER AREA CANOPY f

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INTRODUCTION At the present time, resin discharge operations are conducted outside the Reactor Auxiliary Building, exposed to the weather. In the event of inclement weather, these operations must be halted so that no water will enter the resin cask or the space between the cask and the resin cark liner.

To allow for continuous resin discharge, a covered, portable straddle crane will be prcvided for use during these operations. This crane will be capable of removing the cask cover from the cask, removing the cask liner for inspection of the cask, replacing the cask liner, allowing resin transfer to the cask and then replacing the cask cover, with all operations being protected from inclement weather. It should be noted that the crane will be supplied with an aluminum roof cover and with hooks used to hang herculite tarpaulins so that operations will be adequately protected from inclement weather.

Along with provision of the crane itself, this PC/M will provide alllighting and auxiliary equipment needed for use with the crane as well as a storage area with fastening devices to be used to secure the crane against movement during a hurricane.

This PC/M does not address permanent electrical power supply for the crane since the plant indicated they would provide temporary power. Such power is to be provided by means of a temporary cable which is to be removed after each use.

SAFETY ANALYSIS The straddle crane provided by this PC/M is to be used only for resin transfer operations. The spent resin tank, piping and associated equipment are in Quality Group D (Not Safety Related - See Table 3.2-1 of the Unit 1 FSAR). Therefore this PC/M is non-nuclear safety related.

Any possible failures of the spent resin system which would increase radiation .

exposures have been addressed in Section 11.3.3.1 of the Unit 1 FSAR. The use i ci the Straddle Crane in no way increases the probabilities of occurrence or consequences of these failures. i The Straddle Crane will be stored and operated outside of the Reactor Auxiliary Building. The effects of a load drop (10,000 lbs from 21 feet) are enveloped by the high trajectory turbine rotor missile analysis. It can be concluded that the ,

underground safety related equipment will not be damaged and the addition of 1 the new crane will not impact the NUREG 0612 analysis for the St. Lucie Plant. '

PCM 045-183 With regard to this Plant Change / Modification; a) The probability of occurrence or the consequence of a design basis accident or malfunction to equipment important to safety previously i

evaluated in the FSAR has not been increased.

b) The possibility for an accident or malfunction of a different type than any evalaated previously in the FSAR has not been created.

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

Based on the above, it can be concluded that this PC/M does not involve an unreviewed safety question.  :

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l PCM 364-183 REPLACEMENT OF SPENT FUEL CASK CRANE BRAKES i

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I, introduction The purpose of this PCM is to replace existing brakes for the auxiliary hoist and bridge with new brakes. The existing brakes have deteriorated such that they are no longer functional.

Safety Analysis This PCM is for replacing the existing auxiliary hoist brakes and bridge brakee on the fuel cask crane. This change is non-safety related since the fuel cask crane is non-safety related per Section 9.1 of the FSAR.

The new brakes are equivalent to the old brakes, as addressed in the attached vendor part number change requests.

With respect to NUREG 0612, the equivalent old brakes were evaluated by the manufacturer (P & H Harnishfeger), for compliance to ANSI B30.2.0 and to CMAA

1. 70. These brakes were found acceptable.

In accordance with QI Section'3.2, no unreviewed safety question has been introduced.

PCM 003-184 ELECTRICAL PENETRATION E-4 N0ZZLE l

l INTRODUCTION l

The function of this modification is to convert spare electrical penetration E-4 nozzle into an outage services penetration. i This modification installs a welding neck flange and a blind flange on the -

spare penetration E-4 nozzle inside containment. The flanges heve a double Chring seal with a test connection between the O-rings for a local leak rate test (LLRT).

During power operation, the blind flange shall remain in-place. During an outage, the blind flange may be removed to route power or hoses into containment. At the end of the outage, the blind flange must be installed and a LLRT must be performed on.the flange to assure containment ~

integrity.

SAFETY ANALYSIS '

This change does not involve an unreviewed safety question because:

1. a) The probability of occurrence of an accident previously evaluated in the FSAR has not been changed since containment penetrations are not utilized in the determination of probabilities of accidents, b) L The consequences of an accident previously evaluated in the FSAR have not been affected since the design of the modified penetration meets St. . Lucie Unit No. 1 containment design criteria for penetrations. ,

c)' The probability of malfunction of equipment important to safety previously evaluated in the FSAR has not been affected since spare penetration E-4 has no effect on the operation or operability of any pisce of equipment important to safety.

d) The consequences of malfunction of equipment important to safety previously evaluated in the FSAR have not changed for the reasons given in c) above.

2. a) The possibility of an accident of a different type than any analyzed in the FSAR has not been created since the design is similar to other penetrations already installed in the containment.

b) 'The possibility of malfunction of a different type than any analyzed in the FSAR has not been created for the reasons given in 2.a).

3. The margin of safety for any Technical Specification has not been reduced since the containment leakage criteria will remain unchanged.

PCM 024-184

. STEAM LINE RAD MONITOR WEATHER ENCLOSURE INTRODUCTION The main steam line radiation monitors (1A & IB), located on the main steam trestle, have been experiencing operational problems due to their expo *ure to rain water. The purpose of this PC/M is to provide an enclosure, around each monitor, to prevent water from falling on the moniters.

SAFETY ANALYSIS The steam line radiation monit rs do net perform a nuclear safety related function tnd are away from any nuclear safety related equipment. Since the monitors are attacr .c to the steam trestle, which is safety related, it has been evaluated and found adequate t0 withstand the additional loads (approximately 12 lbs.) of the wrather enelesure. Theref;re, it can Oe concluded that the addition of the weather enclosure will not adversely affect the structural integrity cf the steam trestle ner w!!! it affect the safe operation of the plant.

Based on the abeve, the probability of occurrence of 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 ace! dent or tralfunction different than those previoulsy evaluated. Also, there E.re no changes to the technical specification cf the plant. Therefere. it can be concluded that the steam I!ne radiation menitor ene:csure d:es not pese an unreviewed safety question pursuant te :0 CFR f D.59.

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PCM 03/-184 CORE SUPPORT BARREL REPAIR INTRODUCTION During the M arch, 1983 refueling outage, difficulties were  !

encountered during core reload when a fuel assembly would not seat i properly on the core cupport plate. Subsequent inspection determined there was debris of unknown origin on the plete. The fue! was unloaded and the core support barrel was removed to investigate the source of the debris. I J

A visual examination of the core support barrel / thermal shield assembly disclosed the thermal shield support system to be severely damaged. A number of thermal shield sppport pins were fractured and/or missing and damage to the core support barrel was visible.

The debris from the deteriorated support system was subsequently removed from the reactor vessel.

An evaluation of the thermal shield support system concluded that (

refurbishment was impractical. Therefore, a decision was made to remove the thermal shield. Analyses performed to evaluate operation of the plant without a thermal shield for its remaining design life indicated that replacement of the thermal shield was not necessary.

The reactor internals interfaces and fuel were examined and did not show any detrimental effect due to thermal shield support system depedation. It was concluded that the damage was confined to the core support barrel and thermal shield. )

Upon removal of the thermal shield from the core support barrel, a n:nfestructive examination of the core support barrel was coni.eted.

Da age of verying degrees was in evidence at eight of the nine thermal shield support lug locations. Four lugs were separated from the core support barrel and throegh wall cracks were confirmed adjacer.: to nme damaged lug areas.

The fenetion of this repair is to stop propagation of the existing cracks in the core support barrel, to maintain bypass leakage at an acceptable level, and to assure core support barrel structural interity remains.

PCM 037-184 SAFETY ANALYSIS Exxon Nuclear Company had submitted the St. Lucie Unit 1 Cycle 6 Safety 1 in Analysis Report and Plant Transient Analysis for St. Lucie Unit January,1983 under References 2.4.3 and 2.4.4 respectively.

Exxon Nuclear Company has reassessed the sefetyar.d conclusions previously concluced ths; te t;.

ferwerded ir tre Cye:e E ReWe Sofety Evslustice re-Ein valid for nor nal operation in light of the proposed plug repeir and cor.se ;uent increases in bypass flow f.;r the St. Lucie 1 core barrel.

The calculated core bypass leakage increases during Cycle 6 as a result of The increase in Gew around the plugs and through cracks in the barrel.

bypass flow has been conservatively estimated to be less than 0.3%, based on an all Exxon Nuclear core, such as is anticipated for Cycle 8. 1 l

Prior to the discovery of cracks in the core support barrel, Enon Nuclear had completed the Reload Safety Analysis for Cycle 6, and Florida Power and Light (FP&L) submitted it to the Nuclear Regulatory Commission.

Exxon Nuclear had already applied appropriate core flow penalties to Exxon Nuclear assemblies to account for cross flows in the fixed core and higher burnable poison rod loadings for Cycle 6.

In light of possible increases in bypass flow, Exxon Nuclear has reassessed their Reload Safety Evaluation and has concluded that:

a. Prior LOC A submittals prepared by Exxon Nuclear alreedy assumed a bypass flow in excess of that anticipated with the repaired core support barrel. Therefore, the LOC A analyses remain valid.

b. The verification of the LCA curves and TM/LP trip setpoints perferr ed by Euon Nuclear woulc ex5 Dit on:y a sligt: red eticr m me cer orstra:ed m argins. Flonca Power anc LigM chose :e Technice; continae using the curves shcwn in the presen:

Specifiestien (based on C-E Cycle 4 and Cycle 5 *stretet" p wer submi:ta's! and the allowable cgeratin; region for cycle 6 is stin w ei within safe limits because the Technical Specifications are more limiting than these recuired by the Exxon Nuclear analysis,

c. A detailed recalculation of the trasiert en51yses performed by Enan Nucleer using the incressed bypass 'lov: values would result in a DMR red eti r ef not n :re thE- 0.01 t re - thc' previour5 calculatec. h itr. the exception of thc se; zed rotor event, all transient events analyzec heve greater then s 0.01 r argin to the

':DNER li rit.

d. Tx :P t seht r;.or evc r l, tv4 cises wc:e and,c z e :.; cr.e tc maximize the pressure peak, and the second to minimize predicted DNBR. Since the seized rotor event is terminated by low flow differential pressure sensors on the loops, the calculation to maximur press.re woald bc uneffected by any ficw red:strixtior. C m ir/ P.i;C

• T.c CL;Llation occurring in thf cerel'pEsi ared.

DNBR previously performed by Exxer. Nuclear shcwed an XDNBR =

1.189 and significantly less than 1% fuel damage. As with the other

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transient calculations mentioned above, a recalculation would show a further decrease in the MDSSR by less the.n 0.01, which would not appreciably increase the level of fuel damage.

All four conclusions stated are based on the Technical Specification minimum loop flow of 370,000 rpm and take credit for neither the expected increase in system fic v. resulti .; fro ~ removel of tre

'tcr c! shielc nor for the actual p;snt flov saire mer.ts, v luer Erc ir excess of the recuired minimum,

e. With respect to the consequences of a new type of accidert other than those previously anelyzed, the worst case single failure, loss of the largest patch has been ane.lyzed in detai; (Section 3.1.5). The annlysis by ESC showed that the most severe transient coupled with the loss of patch I resulted with predicted offsite doses remaining a sr all freetion of 10 CFR 100 limits. Note itE: the probabilit:. of the

ss of a petch is very low besed or. :e c:r.servetive de "r, assumptions and testing for the plug patch assemches.

PCM 043-184 i

ELECTRICAL PENETRATION MODULE INSTALLATION

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INTRODUCTION This PC/M covers the installation of electrical penetration modules in the existing General Atomic electrical penetrations. The installation ,

of these penetration modules is to replace the ones removed under PC/Ms 334-183 and 245-183 and to provide twenty four (24) additional modules in the spare module openings available in existing penetrations.

SAFETY ANALYSIS With respect to Title 10 of the code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibilty 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 cf safety as defined in the basis for any technical specification is reduced.

Electrical penetratio modules are supplied by Conax corporation. t:ho has supplied all of the St Lucie Unit 2 electrical penetration assemblies as well as four electrical penetration assemblies for St Lucie Unit 1. These electrical penetration assemblies have been qualified seismically and environmentally for application in both St Lucie Plants.

The electrical penetration modules that will be supplied by Conax Corporation for use with General Atomic penetrations are of equal design, material, and manufacturing procedures as the ones ,upplied in the Conax electrical penetrations under Ebasco P.O. No. NY-422562. The interface between the Conax penetration modules and the General Atomic's header plate is accomplished by welding the adaptor module assembly to the header plate in the similar way it is done in the General Atomic penetrations.

The temperature rise in the existing penetration due to the replacement and/or addition of new electrical penetration modules has been analyzed by Conax corporation and is based on the design criteria used by General Atomics in their design of the St Lucie 1 electrical penetrations.

The design, fabrication, test, inspection, and qualification of the assemblies will be done in accordance with IEEE Standard 317-76, the Conax Nuclear Quality Assu*ance Manual, and the AS'fE Boiler le Pressure Vessel code,Section III, Division 1, Subsection NE for Class MC Vessels.

Each c?nductor is sealed a: both cnds of its feed through sheath in a series of polysulfone thermoplastic sealants. The volume between the inner-most seals of each sheath is pressure monitored via the existing penetration monitoring system.

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l PCM 043-184 Aft'r the installation of the electrical, penetration modules, PT tests.

a helium leak test (local leak rate test as' required by Appendix J of' 10CTR$0) are to be performed to enstre integrity of the pressure boundary'in the Reactor containment _ Building.. Following these tests, these electrical. penetration modules'are' considered welded spare modules for future use.- They will not be electrically conne'eted at this' time.

The implementation of this PC/M does not_ require a change to the plant technical specifications.

The foregoing constitutes, per 10CTR50.59(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.

l' PCM 072-984' CONDENSER PIT FLOODING i

INTRODUCTION During periods of heavy rain the condenser pits in both units have flooded.

The standing water level in both the east and west pits has been as high as L elevation +13 ft (2). One of the contributing f actors causing the flooding

'is that the overflow provisions from the two existing storm water basins shown on drawing 8770-G-687, Rev 1 vere not constructed as designed. With the ditches and overflow basin properly installed, the water level in the.

Iowever, storm water basins would probably not exceed elevation +7 ft.

without these provisions, the water level can rise to about elevation + 14 ft causing the storm drainage system to back-up into low points in the power

. block area.

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

With the implementation of this PCM the capacity of the storm water drainage system will be substantially increased. This will prevent backflooding in any power block area and consequently eliminate any damaging effect from a j heavy rainstorm. This change will be such that it will not compromise the integrity of the plant security system. The proposed changes do not involve a change to the technical specification.

The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides the basis that this change does not involve any unreviewed safety questions, therefore prior Commission approval is not required for implementation of this PCM, 1

PCM 110-184 CONDENSATE POLISHER L&N TRANSMITTER REPLACEMENT INTRODUCTION The purpose of cnis PCM is to replace the existing Leeds & Nortarup pressure transmitters with a comparable model transmitters by Rosemount. The replacement is necessary to provide a more reliable and less maintanence requiring devices.

Inese instruments are locally mounted on a pipe stand near the sensing element. These new pressure transmitters will be performing the same runction as cne existing ones, thus measuring flow and pressure throughout various portions of the condensate polisher system. In acottton, new five-valve manifolds will be replacing the existing ones for the associated new pressure transmitters. The instrument tag nummers of concern are: FT-19-il, FT-19-12, FT-19-13, FT-19-14, The instruments, FT-19-15 FT-19-01, FT-19-21, PDT-19-03 & PT-19-04 valves and installation bookups will be non safety and non seismic.

This PCM provides the means for ene implementation of the above mentioned transmitters anc instrument manttolds.

SAJETY ANALYSIS Wien respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed enange shall be deemed to lavolve an unreviewed safety quesr. ton; (i) ti che procaoility of occurrence or tne consequences of an accident or malfunction of equipment important to safety previously evaluacec in tne Saiety Analysis Report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluatec previously in tne Safety Analysis Report may be created; or liii) if the margin of safety as defined in :ne basis for any technical spectrication is reduced.

Ine new pressure transmitters will be more reliable and require less maintenance than the existing L&N transmitters. The materials, valves and transmitters will be non sarety, non seismic category, therefore this modification will not tacrease ene probability of the occurrence of any accident, whetner previously evaluated or of a different type than previously evaluated and will not reduce the safety of the plant.

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

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

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PCM 135-984 SEWACE TREATMENT PLANT-INTRODUCTION The sewage treatment plant consists of two separate 10,000' gal / day trains-(A&B). At present the'two trains do not receive an equal amount of sewage flow. This reduces the ef ficiency of the plant system.

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~ 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; (1) 'if the probability of occurrence or the consequences of an accident or malfunction of equipnent important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an -

accident or malfunction of a'different type than any evaluated.previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced. ,

The sewage treatment plant is located approximately 800 feet south-east . l of the RCB, away. f rom all nuclear safet) related structures, systems and components. Failure of this modification would be completely internal and would not affect the safe shutdown of the St Lucie Plant. Therefore, this modification will not increase the probability of occurrence of an

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accident either previously evaluated or created due to the; installation

.of this modification and will not reduce the margin of safety as defined in the basis for the technical specification. ~This change does not in-volve a change to the technical specification.

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

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l QSPDS INVERTER /POWERLINE CONDITIONER MODIFICATION INTRODUCTION ,

The Powerline Conditioners are part of the Inadequate Core Cooling (ICC) system power supply. There are 2 power supply systems lA and IB for Unit 1 and 2A and 2B f or Unit 2.

The power supply systems are composed of a 7.5 KVA inverter and a 25 KVA Powerline Conditioner. The Powerline Conditioner is the By-pass source used for the maintenance of the 7.5 KVA inverters.

Af ter the installation of the ICC power supply systems, the fuses protecting the current transformer in the Powerline Conditioners were blowing out during power transfers (ie. transfer to off-site power or diesel generator).

SATETY ANALYSIS This modification has been reviewed with respect to Title 10 of the Code rf Federal Regulations. Part 50.59, which states that a proposed change s "11 be deemed to invclve an unreviewed saf ety 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 ac-cident or malf unction of a dif ferent 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.

Modification covered in this PC/M requires the replacement of resistors only and does not degrade the system nor affect the internal integrity or quali-fication of the Powerline Conditioner.

The installation of the class 1E. 68 ohm resistor in the Units 1 and 2 Powerline Conditioners enhances the operation of these equipment by increasing the reliability during power transients.

This PC/M neither poses the possibility for a new or dif ferent accident or malfunction as defined in the safety analysis report nor it affects the Plant Technical Specification, as written.

The foregoing constitutes, per 10CTR50.59 (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.

PCM 158-984 CWS - DISCHARGE CANAL EROSION PROTECTION INTRODUCTION The St Lucie Unit 1 and 2 discharge canal erosion protection has been damaged due to undermining since the initial installation. A detailed report on the nature, extent and possible causes of the erosion is presented in the Ebasco ?;eport "CWS Discharge Canal - Erosion and Repair Recommendation" (Ebasco letter P-0-SL-84-2783 dated July 30, 1984). This PCM implements selected repair and improvement recommendations presented in the report.

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; (t) 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 increasea; or (ii) if a possiblity for an accident or malfunction of a different type than any evaluated previously in the safety analysie report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

This PCM package provides repairs and improvements to the plant discharge canal erosion protection. The discharge canal is part of the circulating water system which is not a safety related system. The repairs will restore the discharge canal erosion protection to its original designed integrity. The modifications, such as additional pressure relief holes, extension of the concrete erosion panels and grouting of the vertical joints, are enhancements which will prolong the life of the canal erosion protection. The implementation of this PCM does not in any way adversely afrect the perf ormance of the circulating water system and it is concluded that the probability of an accident is not increased; that the creation of a ditferent type of accident is not created; anc the margin of safety is not reduced. j l

Tne implementation of this PCM does not require a change to the plant l technical specification.

The f oregoing constitutes, per 10CFR50.59(b), the written safety evaluation wnich provides the bases that this change does not involve an unreviewed saf ety question and prior Commission approval f or tne ,

implementation of this PCM is not required. j i

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PCM 167-184 MSIV BYPASS VALVE CABLE REPAIR INTRODUCTION The purpose of this PCM is to repair cables 10314A-SB, 10314C-SB and 10314H which had their outer jackets thermally damaged due to missing insulation on the bypass steam line. These cables provide power and control to the Main Steam Isolation Bypass valve MV-08-1B and the space heater required for same valve. I The Main Steam Isolation Bypass valve MV-08-1B is located in the Steam Trestle and performs a containment isolation function, but is open only during start up.

The damaged section of the cable will be pulled back from MV-08-1B and will be replaced with new cables which will be spliced with the old ones utilizing Class IE Raychem Splice Kits.

SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question:

(i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previouslypossibility evaluated for in the an Safety Analysis Report may be increased; or than (ii) any if a evaluated previously in ae:ident or malfunction of a different type 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 repairs three damaged cables by utilizing Class IE Raychem Splice Kits. Implementation accident or if this PCM does not pose the possibility for a malfunction as defined in the Safety Analysis new or dif f erent Also Report since the new cables are Class IE and will be routed as before.

it 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 basis that this change does not involve an unreviewed safety question or a Technical Specification change and prior Commission approval for the implementation of this PCM is not required.

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I PCM 172-984 1 WATER TREATMENT PLANT ORCANIC SCAVENGER ACID INJECTION SYSTEM INTRODUCTION The Water Treatment Plant (WTP) As designed to process city water in preparation of demineralized wa,ter for the various St Lucie Flant de-mineralized water users.

Each of the WTP trains is provided with an organic scavenger for the removal of organics from the WTP influent water.

The organic scavenger serves as a guard bed to protect the Water Treat-ment Plant ion exchanger resins from organic

• fouling. The organic scavenger uses a weak base anion resin which acts as an adsorbant for organics in the raw water. The resin adsorbs and holds the organics, removing them from the water. This adsorption mechanism allows the resin to be regenerated, stripping-off the organics, and returning the resin to service. Were these organics to contact the ion exchanger resin they would become irreversibly fouled. This would require re-placement of the ion exchanger resins. The ability of the weak base anion exchanger resin to operate properly as an organic scavenger is a function of the pH of the influent water. The optimum pH range for the resin is between 4 and 7; the resin will still perform acceptably up to a pH of 8. Above this pH the ' raw water will act-as a regenerating media and the organic material will appear to pass right through the bed.

In the past few months the pH of the raw water treated by the Water Treatment Plant has been increasing. This appears to be attributable to a change, by the City of Ft Pierce, in biocide used in the water supply. Raw water pH has been in the range of 8.9 to 9.6 as recorded at the plant. The resulting " blinding" of the organic scavenger has caused rebedding the strong base anion exchanger due to organic foul-ing.

This PC/M is to implement an acid wash program to prevent the blinding of the organic scavenger. The wash consists of a 5% H 2504 wash of the organic scavenger bed, after the regeneration of a decineralizer train.

The wash conditions the resin and results in a lowering of the raw water pH in contact with the resin.

SAFETY ANALYSIS This modification has been reviewed with respect to Title 10 of the Code of Federal Regulations, Part 50.59 which states that 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 malfunc-tion of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an acci-dent 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.

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-The Water Treatment Plant is non-safety related and as..such does not.

L perform any safety function vital to plant safety.

The new piping and valves that will be installed per this'PC/M are de-l signed to the same requirements as the existing piping and valves at the WTP.

i-There is no safety related equipment at the iTTP Area. Therefore, fail--

ure of any.of the new piping and valves at the WTP, vill not increase the '

probability of an accident or malfunction of equipment important to safety previouslyLevaluated.-nor is'the. possibility for an accident'of'a dif ferent type than any evaluated previou' sly in the safety analysirs re-port created. 'The proposed change does not itvolve a change to the plant technical specification. The foregoing constitutes,'per 10CTR50.59(b),

the written safety evaluation'which provides the basis that.this does not involve an unreviewed safety question and prior Commission approval for the implementation of this PC/M is not necessary.

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PCM 218-184 i

ROSEMOUNT D/P TRANSMIT 1ER MODEL CHANGE Introduction

1. To satisfy the requirements of the design basis the Rosemount Series D transmitter has been chosen to replace Series A transmitter. The Series A and D are directly interchangeable. No changes to either the wiring or mounting will be required. Both transmitters are qualified by Rosemount and both will function in the applicable environments.

2. The replacement Series D and the existing Series A transmitters are similar in mass and configt. ration. Any increase in loading resulting from the installation of the new transmitters is negligible. Therefore, the existing supports are considered adequate and require no further evaluation.

3. The environmental qualification for the Series D transmitter is provided in Rosemount Report D8300040, Rev. A.

4. The instrument specification sheets for the Series A and Series D transmitters have been reviewed with regards to accuracy and drift. Both instruments are identical. Each has an accuracy of + .25% over the calibrated span and a drift of

.25% over six months.

Safety Analysis This PC/M is considered nuclear safety related because the containment pressure transmitters PT-07-8A and BB which are used to monitor the consequences of an accident, are being replaced. Since, however, the Rosemount Series D transmitter is a one for one replacement meeting the same qualifiestion requirements and no wiring or mounting modification are required this change does not involve an unreviewed safety question.

For the same reasons, the probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR or of a different type of accident than any previously evaluated in the FSAR has not been created. Furthermore the margin of safety as defined in the basis for a Technical Specification has not been reduced.

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PC'M 2'33-184 -

AUXILIARY FEEDWATER ACTIATOPM SYSTEM MODIFICATIONS INTRODUCTION.

This PCM is for!the modification of the control circuit for four (4) auxiliery feedwater discharge valves. The circuit is designed to allow

.the valves.to have throttling capability before and after actuation of AFAS. This modification is required to regulate the steam generator water level.

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 saf ety question ]

(1) if the probability of occurrence or the consequences of an accident I or malfunction of equipment important . to ' safety previously - evaluated in the saf ety 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 createdt or _(iii) if_the margin of safety as defined in the basis for any technical specification is reduced.

This modification is being made to enhance the AFWS. With the implementation' of this PC/M the operator will be able to manually adjust the AFW discharge.

valves, thus limiting the thermal transients experienced by the inj ec tion-of the relatively cold AFW to' the steam generators and' associated piping.

This modification will allow the operctors, in an orderly manner, to regulate AFW flou and maintain proper steen generttor level.

The addition of these relays'will allow the mentioned valves to be operable before and after the actuation of AFAS. These relays will be Class lE and qualified per IEEE-323-1974 and IEEE-344-1975. All cable and conduit will be routed / supported to Seismic Category I. The changes being addressed' in this PC/M will not disturb the normal function of AFAS. The modifica-tion is to allow for the manual control of the MOY's prior and after the initiation of AFAS. i Af ter an AFAS signal is generated, should the operators manually close the '(

AFW discharge valves to an extreme value, thus prohibiting proper flow of )

Aux Feedwater, numerous alarms (i.e steam generator level, temp, RCS temp) l and. indication Aux Feed flow will alert the operator to his error. Should l l

l the operator allow to much flow of water into the steam generator, causing the-level to rise, automatic controls will override and close the AFV dis-charge valves, until additional water is needed, at which time they will open. Therefore, this modification will not increase the probability of the occurrence of any accident, whether previously evaluated or of a differ-ent type than previously evaluated and will not reduce the safety of the plant.

PCM 233-184 l

This ' PC/M does not reduce the margin of safety as defined in-the basis of any technical' specification, nor does it require a revision of a technical specification.  ;

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

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PCM 253-984'

-Na0CL GAS' DISENGAGING TANK BLOWER REMOVAL INTRODUCTION

. Operation of the sodium hypochlorite system will be affected only in that the hydrogen blowers wnl not be installed and therefore operation of their associated - switches will no longer be part of the startup sequence.

Additionally, all automatic interlocks between the blowers and the remainder of the system have been removed. Therefore, the remaining portions of the system wn! operate as normal.

This modification' functions to physically remove both gas disengaging tank blowers and functionally removes all switches, alarm lights and interlocks associated with these blowers.

- This modification provides for the complete physical removal of the gas

' disengaging tank bloweas and provides for installation of. a protective screen on top of the tank cover. Additionally, this modification provides for removing . all interlocks associated with these blowers . and the remainder of .the system.

SAFETY AN ALYSIS 1.0 This modification has been reviewed with respect to 10 CFR 50.59 and has been deemed not to involve any unreviewed safety question because of the following:

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

1.2 This modification involves removal of the gas disengaging tank blowers which are not required when the system is located outdoors. Therefore, the original design intent of the system does not change.

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1.3 These modifications do not interact with any safety related systems or components.

1.4 No safety related systems or components are compromised by any assumed failure of any existing or new equipment or components.

1.5 No parameters relating to Technical Specifications are adversely affected and no Technical Specifications are altered by this modification.

2.0 Tnis constitutes tne safety evaluation according to 10 CFR 50.59.

Thepefore. prior Commission approval is not required prior to l implementation of this modification.

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.PCM 012-185.

OBCW HEAT EDIANGER CHANNEL GASKET MODIFICATION l

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. IyIRODUCTION The Open Blowdown Cooling Water (OBCW) Heat Exchangers were .

originally. supplied with pressed asbestos type gaskets at the channe!-to-channel cover flanges. This type of gasket is normally provided in high temperature - high pressure . applications 'where the fluid medium is essentially 4 1-corrosive to heat exchanger materials. Due to the stiffness of the asbestos gaskets, they were sensitive to any imperfection in the seating surfaces of the OBCW heat exchangers which resulted from the L corrosive se water medium, and leak tight sea!s depended upon the use of fi!!er mater: .is such as Belzona on the seating surfaces, and/or silicone sealants.' Plant personnel eventually began using rubber gaskets on the

'OBCW heat exchar rs, since the more pliable material was'less sensitive L to seating surf ace in perfections. 'Ihe rubber gaskets were also difficult to ust, since they 3ere structurally. weak and tended to sag under their own weight (the existing gasket!. are not bolt centered).

This PC/M provides the details necessary to machine both the channel head flange and channel cover gasket seating surfaces such that larger, bolt centered " Garlock" gaskets may be used. Th!s modification will result in simplified gasket installation, and should also provide a more leak tight joint, since the gasket bearing surface will be somewhat increased through the use of the larger gasket (see PC/M drawing).

5 AFETY ANA!M3 Tr.e tradificatior. to the OBCT heat exchanger channel cover gaskets is consdered non-Nelear saf ety related for the follewing reasons:

A. Tr.e OBC % est exchangers are non-nuclear safety related and are .n a non-nuclear saf ety related portion of the IC% system.

B. Postustec f aaures of the OBCW heat exchangers would . ave ne impset on safe 5hutdewn of the plant or related systems.

C. The CBC% neat exenangers are not used to prevent postu!ated accidents, m.t: gate the conset;uences of such accidents, ma.ntain safe shutdown conditions or adequately store spent fuel.

PCM 021-185 i

DELETE WAVE LUN-UP STOP LOGS INTRODUCTION:

Flood protection requirements for ths St Lucie Unit I nuclear power plant include provision of stop logs to protect all grade level openings from Elevation 19.5 ft to Elevation 22.0 ft during hurricanes. A detailed an-alysis of hurricane surges, erosion, and wave run-up performed for the Units 1 and 2 FSAR's demonstrated that no entrance to either unit was sub--

ject to direct wave run-up. The NRC required stop logs for only two rel-atively exposed entrances to Unit 2 but required all stop logs for Unit 1.

An analysis reviewing the NRC criteria for requiring stop logs at Unit 2 determined that all but one of the stop logs for Unit I can be eliminated by applying the criteria consistently.

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 ques-tion; (1) if the probability of occurrence or the consequences of an acci-dent 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 pre-viously in the Saf ety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The requirement for flood protection of Unit 1 between Elevations 19.5 and 22.0 ft was based upon an early, approximate wave run-up analysis.

This analysis indicated a maximum wave run-up elevation of 22.0 ft (see SER Section 3.4). Subsequently, more detailed analyses (see Unit 1 FSAR Section 2.4.5.7) determined that wave run-up would not exceed elevation 18.0 ft coincident with the maximum peak surge level of 16.22 ft. This is well below the lowest elevation protected by the stop logs of 19.50 ft. Considerable additional flood protection is already afforded by vir-tue of the layout of the roads, buildings and tornado missile protective structures permanently incorporated into the plant design.

For Unit 2, the NRC cetermined that entrances need only be protected from wave splash and spray, and that most entrances were already protected by other structures. Only two entrances to Unit 2 were required to have stop logs for this reason.

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PCM 021-185 There are thirty-six (36) entrances to Unit 1 building for which stop logs are specified. Of these, twenty-five (25) are entrances to non-safety related buildings, therefore protection from splash and spray is not con-sidered necessary. Of the eleven (11) entrances to safety related areas, all but one are protected from splash and spray by other buildingr. By applying the NRC criteria for stop log requirements used at Unit 2, this is the only stop log required at Unit 1. This stop log (No 19) is located in the Fuel Handling Building east wall.

The stop logs are not safety related components and their deletion does not create an unreviewed safety question. Installation of the stop logs is not required by the Technical Specifications, therefore the implementa-tion of this PCM does not require a change to any Technical Specification.

The foregoing constitutes, per 10CFR50.59(b), the written safety evalua-  !

tion which provides the bases that this change does not involve an unre-viewed saf ety question and prior Commission approval for the implementa- I tion of this PCM is not required.

PCM 037-185 REPLACE NAMCO LIMIT SWITCHES ,

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INTRODUCTION grade NAMCO l At present St. Lucie Unit 1 is utilizing sixteen obsolete commen '

These switches are being replaced and upgrade 4 or maintenance limit switches. The following harsh purposes in accordance with 10 CFR 30.49 requirements.

env;ronri,ent locations for the N AMCO switches are listed below.

Reactor Containment Purge Isolation Valves FCV .25-2-L5 FC V-25-3-L5 Reactor Containment Purge Isolation Valves F C V-23-7-L5 Containment Vacuum Relief Valves FC V-2?J-L5 Containment Vacuum Relief Valves HCV-08-i A LSI Main Steam Isolation Valve HCV-08-1 A-L52 Main Steam Isolation Valve HCV-08-1 A-L53 Main Steam Isolation Valve HCV-08-1B-L51 Main Steam Isolation Valve HCV-08-1B-L52 Main Steam Isolation Valve HCV-08-!B-L53 Main Steam Iso!ation Valve Y2303-L5-C Extraction Aux. Steam , Air Evaluation System Extraction Aux. Steam, Air Evacuat;on System V2305.L5-0 B6301-L5-C Waste Management V6 30 !-L5-0 Taste Management V6302-L5-C Taste Management V6302-L5-0 Waste Management In accordance with Paragraph (KXI) of 10 CFR 50.49 this PC/M w:ll a!iew for the replacement of these N AMCO limi'. switches on the at:ve tag numbers with QL-!

model N AMCO limit switches qualified to 10 CFR 50.29. T us PC:M is co .s :ered nuclear safety related, because several of the sotches are on safety related va:ves such as the Main Steam Isolation Valves.

SAFETY ANALYSIS This PC/M is considered nuclear safety related because there are several limit switches located on the Main Steam Isolation Valves which are being replaced.

Since, however, the Namco limit switches are a one for one replacement of obsolete equipment and no wiring or mounting modifications are required, this change does not involve an unreviewed safety question.

For the same reasons, the probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR or of a different type of accident than any previously evaluated in the FSAR has not been created. Furthermore, the margin of safety as defined in the basis for a Technical Specification has not been reduced.

PCM 043-185 TEST FLANGE FOR LOCAL LEAKRATE TESTING INTRODUCTION 1.1 _ Ope ra t io_n, This design package modifies existing non-pressure boundry blind flanges to allow for direct Local Leak Rate Test (LLRT) without i I

removal and subseq;ent re-installation of system flanges. Therefore no ope ating procedures are effected except the LLRT procedare its:lf.

1.2 Function This modification functions to reduce the time required to perfor.m periodic LLRT's and thereby reduce personnel radiation exposure.

l.3 Descrio: ion Blind flanges identical:o presently installed f;anges will be purchased and modified to incluce :est fittings tha; are directly compatible to existing plan: LLRT equipmen:.

This will reduce tne present " hook-Jp" process which is; a) removal of sys:em (mu!:i-oo;;ed) f:ange, b) ins:all and :orque :est fiarge (perform _; RT),

c) re noval of :es: flange, d) installation of new f;ange gas' set, e) re-insta;;atic, anc :orquing of system 1.ange, to esse :ially a s.ng;e s:ep ope a: ion .e w..' n:: equire replaceme-- ga sk e:s. :es fla ges or f.c g e cor remeva! anc torqu.9g.

SAFETY ANALYSIS The replacement of system blind flanges with ident'. cal b!md flanges modified to include a 3./3 inch test fit ing is not nuclear safety related because:

1) The modified flar.ges will not be installed in :nc safety related, pressare boundry portions of the system af fec:ed.

2) Failure of t",e med:fied flanges or tes: fitti^gs w ould nave no

. rpact oc safe shutdown of the plant.

PCM 043-185 j

3) Tne mass difference between the inodif.ed flarges and origina!

system flanges is small enough, as no: :o ef fect pipir.g stress analysis of conne :ed system p. ping.

4) Tne fla. ges mod.fied by this PC/M are no user :c pres ent pestulated accidents, mitigate :ne :orse:st ces of saem '

aCf. der *s, malr.11tn safe sh'JiO O Wri Co's .* ;ns Or *: ~eQUEte.\

5:Dre Spent fue!.

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L PCM 045-185 HPSI PUMP OIL DRAIN li

~ INTRODUCTION _

The St. Lucie Unit i High Pressure Safety injection (HPSI) @mps I A,18, and 1C Maintenance each ' have an ' inboard and outboard oil- lubricated bearing. The procedures require that the bearing oil reservoirs be drained periodically.

present method consists of removing a drain plug under the reservoir and catching the oil in a bucket. As interferences around the drains make it difficult to catch all of the oil, this work task presents a housekeeping problem.

This modification adds a drain valve assembly to each bearing drain. One end of the valve will be threaded into the bearing housing with a pipe nipple while the other end will be plugged. When an oil change is required, the cap will be removed and the hose connector with flexible tubing will be inserted into the end of the valve. The valve will then be opened and the oil will be collected in a buck et. . Once completed, the valve would be closed, the hose connector removed, and the plug replaced.

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;(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 possiollity for an seeident or malfunction of a different type than any evaluated previously' in the safety analysis report  !

may-be created; or (iii) if the margin of safety defined ir. the basis for any technical specification is reducec.

This modification adds a drain valve assembly to the existing HPSI pump bearing drain taps to simplify periodic oil changes. The modification does not in any way impact or jeopardize the intended normal or emergency operation of the HPS!

pumps because the assemoly has oeen designed leak tight end has double isclation capabilities. Pump Coce requirements and seismic Category I design are maintraned.

The probability of an accident or equipment malfunction previously evaluated in l the FSAR does not exis t. Also. 915 m:<ilfication does not increase The ite I probability of occurrence of an accident not evaluated in the FSAR.

probability of a design basts fire is greatly reduced oy eliminating the probability of an oil spill. The margin of safety as defined in plant techniesi speelfiestions is not reduced. Also, a eninge to plant Technical Specifications is not required.

l The. foregoing constitutes, per 10 CFR 50.59(o), the written safety evaluation whlehptovides the bases that this change does not involve an unreviewed safety question an' d prior Commission approval for the implementation of this PC M is not required.

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PCM 089-185 ICW ISOLATION VALVE REMACEMENT Introduction The Intake Cooling Water (ICW) system removes heat from the Component Cooling, Open Blowdown Cooling and Turbine Cooling Water Heat Exchangers. The system consists of three (3) 14,500 gpm pumps and associated piping and valves. Two redundant supply headers are provided, each aligned with one (1) ICW pump. The reserve pump may be manually aligned with either header by actuating the pump discharge header cross-connect valve. The TCW and OBCW portions of the ICW system are non-nuclear safety related. The CCW portion is nuclear safety related.

There are eleven (11) 30 inch (tag # l-SB-21-3) ocd four (4) 36 inch (tag # l-SB-21-7) cast carbon steel bodied, rubber lined butterfly valves in the safety related portion of the ICW system which perform manual isolation functions. During both the 1981 and 1983 Unit #1 refueling outages, ICW piping inspections have revealed estensive damage to the rubber liners of the 30" and 36" Henry Pratt butterfly valves. Various repair techniques were utilized as interim measures in refurbishing these valves.

The purpose of this PC/M is to encble the replacement of any of the subject valves, as required, and the modification and/or addition of pipe supports as dictated by the piping stress analysis. Two (2) 30" and two (2) 36" cast stainless steel bodied butterfly valves have been procured from Henry Pratt.

Safety Analysis The proposed change replaces some of existing cast carbon steel, rubber lined butterfly valve bodies and bonnets with cast stainless steel butterfly valve bodies and bonnets in locations to be determined during the 1985 refueling outage ICW inspection.

The new valves have been selected because of the enhanced capability of stainless steel bodied valves to withstand the corrosive ef fects of seswater with a maximum two (2) ppm of chlorine. Based upon the greater corrosion resistance of the new valves, the modification does not involve an Unreviewed Safety Question because:

i) The ICW system is not considered in determining the probability of design basis accidents (i.e., LOCA, MSLB, LOOP, etc.).

2) The consequence of a malfunction of equipment important to nuclear safety are not made more serious since the flow rate and heat

- .PCM 089-185 removal capability of the ICWl system are.not altered by the valve replacement.

3)- The possibility for an accident or malfunction of a differei.t type that -

any previously; evaluated in the FSAR is not created because the basic design characteristics and function of the ICW valves has not been changed.' . The - material upgrade will result in a reliability enhancement to the ICW system.

4), ' The - margin of ; safety as . defined in the basis of a Technical Specification remains unchanged because the installation of the new, heavier valves has been evaluated with regard to the piping stresses and restraint loadings. Pipe: supports / restraints are to be modified and/or added to maintain the pipe stresses within code allowables.

The PC/M'is classified as a Nuclear Safety Related change because.the portion of the ICW system in which valves would be replaced is required to withstand all design basis events while supplying suf ficient cooling water to .

the'CCT heat exchangers to fulfill emergency requirements l'

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PCM 103-185 REPAIR OF LINE kH-A28 IlmIGICTIOT

- During nonnal operation, line MA28 developed a leak. LineThis MA28 is 'l-line is 1/2". dia. , 304 SS p'ipe which is heat- traced and insulated. -This condition has been identified located inLthe 19.5 level of the RAB.

through NCR 154.

SAFEIY ANALYSIS This Jnedification is considered non-nuclear safety related for the

.following reasons:

1. Line M A28 is non-safety related, Quality Group D. >

2. 'No postulated failures of this line will have an inpact on safe shutdown of the plant or safety related systens.

3. Line MA28 is not used to prevent postulated accidents mitigate the consequences of such accidents, maintain safe shutdown conditions or adequately store spent fuel.

4. NIC Regulatory Guide 1.143 requirements have been factored into this modification, where applicable.

The above evaluations provide the bases, pursuan' to 10 GR 50.59, for the determination that this PCM does not involve an unreviewed safety question.

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PCM 114-985 SMALL CENTRIFUGAL PUMPS OIL SEAL REPLACEMENT ABSTRACT This engineering package covers replacement of lip type oil seals on small centrifugal pumps throughout Units #1 and #2 with labyrinth type seals. The major feature of this package provides a list of acceptable labyrinth seals and a list cf pumps approved for seal replacement. This package is classified as non-nuclear safety related for all pumps listed with the exception of the Boric Acid Makeup (BAM) and Fuel Pool pumps on both Unit #1 and 12 and the Diesel Oil i

Transfer pumps on Unit I which are classifed as nuclear safety related.

Modifications to these pumps, however, do not constitute an unreviewed safety question.

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PCM 116-985, L ,

DEMINERALIZED WASTE NEUTRALIZATION SYSTEM INTRODUCTION Currently the-Water. Treatment Plant (WTP) demineralized regeneration wastes are transferred to the' neutralization basin where they are neutralized and discharged.- The Resource Conservation and Recovery Act (RCRA) requires that wastes with a pH below 2 or above 12.5 must be handled, stored and disposed of as hazardous wastes. Certain steps in the regeneration process create a vastestream outside pH limitations of 2 or 12.5.. Rather than seek exemption from the RCRA permit requirements it-has-been decided.to install a neutralization tank.- Portions of demineralized regeneration vastes which create wastestreams with a pH value outside the limitations will be routed to this tank. All other portions will be routed to the existing neutralization basin as in the current' practice.

To expedite installation, this. package is being issued in mu l t i- su pplemen t s . ' Supplement 0 deals with the tank foundation and butied piping. Supplement I will deal with the piping,-Supports /

Restraints, Electrical, 16C and Civil portions.

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 consequenceslof 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 dif ferent 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, water-treatment system is a common system for both units and is discussed in the PSL-1 FSAR Section 9.2.5 as part of the makeup water

. system.

'This task modifies the demineralized wastes drain lines and also adds s-neutralization tank. The modification is done to comply with the requirements of RCRA.

The water treatment system does not interface with any safety related system, therefore, it is non-safety, non-seismic. ,

.This sub-system does not impact any safety related component.

PCM 116-985 The WTP is aat considered in any accident analysis in the FSAR nor will this modification create an accident or malfunction t.ot previously evaluated in the FSAR. As a non-safety, non-seismic system it will not reduce the margin of safety as defined in the bases for any technical specifications.

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

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

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PCM 117-185 MI CABLE - GRAF01L GASKET RETROFIT FOR LITTON CONNECTORS INTRODUCTION Mineral Insulated (MI) cable assembliec supplied by Combustion Engineering (CE) for both Core Exit Thermocouple (CET) and Heated Junction Thermocouple (HJTC) systems have been installed at St Lucie Unit i under the scope of work covered in PCM 136-81. The documents for the Class lE qualification of these MI cable assemblies developed by Combustion Engineering were transmitted to FPL via supplement 10 and 11 to PCM 136-81.

Litton connectors used for the HJTC MI cable assembly feature an organic silicone gasket. This gasket material, when subjected to chemical spray did not prevent the ingress of moisture during LOCA/MSLB environmental testing by CE. Moisture ingress resulted in signal anomalies for the HJTC application. The Class 1E qualification of Litton connectors for all other phases of the qualification program has been documented in Combustion Engineering reports No. CE NPSD-271-P and CE NPSD-230-P.

In order to resolve the potential moisture ingress in the unmodified Litton connectors used with HJTC MI cable assembly, Combustion Engineering developed a Litton connector seal retrofit kit which replaces the organic silicone seal with an inorganic "Grafoil" graphite seal. By a combination of analysis and test methods, Combustion Engineering has demonstrated that the Litton connector grafoil gasket retrofit is qualified for its intended application in HJTC MI cable assemblies and will eliminate the moisture ingress concerns.

SAFETY ANALYSIS With respect to Title 10 of the Ccde of Federal Regulations, Part 50.59, a proposed change shc11 be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or =alfunction 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.

A. Presently installed Litton connectors on the HJTC MI cable assemblies employ the organic silicone gasket material. This gasket material when subjected to chemical spray, did not prevent the ingress of moisture during LOCA/MSLB environment testing performed by Combustion Engineering.

PCM 117-185 This modification which replaces the organic silicone seal with a "Crafoil" graphite seal on the existing Litton connectors is an ,

enhancement to the present design. As documented in CE's qualification report #CE NPSD-296-P, the Litton connector grafoil j

gasket retrofit will eliminate the potential ingress of moisture E during LOCA/MSLB' design. basis events and will not lead to signal anomalies for the RJTC application.

B. The Litton connector grafoil gasket retrofit is considered qualified for Class 1E application as demonstrated by analysis and test methode provided in CE's qualification report #CE NPSD-296-P. The report concludes that the subject retrofit, when installed on the existing Litton connectors in the prescribed manner, will provide a leak tight seal and hence is qualified for Class 1E applications.

The unmodified Litton connectors as presently installed had previously met all acceptable criteria for all other phases of the qualification program except for the moisture ingress during LOCA/MSLB test conditions. This has been documented in Com'oustion Engineering qualification report #CE NPSD-271-P, CE NFSD-230-P and CE NPS-275-P. This modification eliminates the potential moisture ingress concerns.

C. Four Litton connector gasket retrofit assembly test specimens were subjected to a seismic testing and all four specimens successfully passed the seismic test (1 tem 7.3.5 of CE report #CE NPSD-296-P).

D. The plant technical specifications are not affected by the implementation of this PCM.

Therefore, 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.

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PCM 125-185 i

STARTUP. TRANSFORMER DELUGE SYSTEM ISOLATION VALVE TAMPER ~ SWITCH INTRODUCTION

.. At present St. Lucie Unit I has two startup transformers, each with its owTi deluge isolation block valve. Each of the two valves has a tamper switch connected to .the Transformer Fire Protection System. The purpose of

' these switches is to cause an alarm condition if either valve is closed.

At present the switches are connected incorrectly. They are normally closed switches (switch opens when ' valve is closed) which are wired in parallel to' one normally closed alarm input' of the Transformer Fire Protection System. The result is that an alarm condition occu s only when both valves are clcsed, not when a single valve is closed.

The proposed change is to connect the two switches in series. This will cause an alarm condition to occur whenever either valve is closed.

SAFF.TY ANALYSIS This PC/M is considered not nuclear-safety related QA/QC required for the following reasons:

1) All fire protection systems are considered Q /QC required through FPL commitment to NRC.

2) The modifications contained herein affect only the Transformer Fire Protection System. The function of this system is limited to the protection of the Main Transformer, Auxiliary Transformer, and Startup transformer. None of these items are considered Nuclear Safety Related.

3) No new components or wiring are installed and none are removed.

4) The changes involved upgrade the function of the Transformer Fire Protection System to meet original design objectives.

For the same reasons, the probability of occurrence or the consequence of-

. a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR or of a different type of accident than any previously evaluated in the FS AR has not been created. Furthermore, the margin of_ safety as defined in the basis for a Technical Specification has not been reduced.

In conclusion, this modification does not involve ar. unreviewed safety question.

This ' PC/M does not alter any fire z or.es and does not introduce cornbustibles. Consequently, no Fire Pr:tectior. Prog am criteria arc affected.

'PCM 129-185-I WTP ORGANIC SCAVENGER TOTALIZER j

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INTRODUCTION-At present the. Water Treatment Plant at St. Lucie' Unit 1 & 2 contains y several unused alarms.~ 1hese alarms are being disabled because they' are .1 unused and cause distractions. The alarms to be disabled are listed below.

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~Ref. Alarm O Scavenger Cation Flow Totalized - Train A

-2) Scavenger Cation Flow Totalized - Train B

3) Strong Base Anion Flow Totalized - Train A 4). Strong Base Anion Flow Totalized - Train B

5) Main Drain Strainer D/P High j

6) Effluent Strainer D/P High ]

The alarms referenced 1, 2, 3 and 4 will be disabled by disconnecting the i contact wiring at the corresponding watermeter. The alarms referenced 5 and 6 will be disabled ' by disconnecting the contact wiring at the ,

annunciator. ,

SAFETY ANALYSIS This modification"is not nuclear safety related becauw the Water ,

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' Treatment Plant is not nuclear safety related. All modifications performed willbe in the Water Treatment Plant.

The function of the Water Treatment Plant is not required for a safe shutdown of the reactor, is not required to mitigate an accident, and is not necessary to perform a containment isolation. The system does not have any control functions, and does not monitor any safety related parameters.

The system. is in an area that will not impact safety related equipment.

Therefore, the ' probability of occurrence or the consequences of an accident or malfunction previously evaluated in the FSAR has-not been increased.

For the same reasons, no possibility for an accident or malfunction of a different type from any evaluated previously in the FSAR has been created by this modification. Additionally, the margin of safety, as defined in the In

- basis for the technical specifications, has not been decreased.

conclusion, this modification does not involve an unreviewed safety question.

PCM 130-185 1

1

' STEAM GENERATOR TRANSMITTER REPLACEMENT d

I INTRODUCTION j

This PC/M provides_for the replacement of the existing Rosemount transmitters shown on Table No.1, Appendix A with Rosemount model 1154 transmitters. l The existing Rosemount transmitters model li33DD-4 presently installed per

- PCM 1-82 for the Steam Generator level measurement do not support their use beyond 150" WC range. The steam generator water level range is 0-131.6" WC for cold conditions. However, in order to compensate the density difference between hot and cold water, the upper range of the transmitter is subjected to 201.6" WC.

The use of Rosemount 1154DP-5 has been considered because the new 1154 model provides less uncertainties under accident conditions than model 1153.

The new model 1154 DP-3 has a range of 0-750" WC which covers the required range under any operating conditions. The new transmitter will be located in the same location and using the existing mounting bracket. Both transmitters ll53DD and Il54DP have the same dimmensions and weight.

SAFETY ANALYSIS This PC/M is considered nuclear safety related because the transmitter being replaced monitors Steam Generator level which is an important parameter for the safe operation of the Unit. This change does not involve an unreviewed safety question because:

A) The Model 1154 transmitter is a one for one replacement with the model 1153 transmitter. The only difference being that the range of the replacement instrument is increased. No wiring or mounting changes will be required.

B) The specification of the replacement transmitters have been reviewed and found to meet or exceed the specifications of the presently installed transmitters.

C) The consequences of an accident previousiv evaluated in the FSAR have not been changed since this modification does not affect the availability, redundance, capacity, or function of any equipment required to mitigate

.ne ef fects of an accident.

D) The consequences of the malfunction of eculoment important to safety previously evaluated in the FSAR have not been affected. Function, mounting, and the ability to withstand harsh environmental condition have not been altered. Nor does this modification affect any other safety related equipment..v..w. c.. -

E) 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 or operating characteristics of the S/G Level Transmitters.

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. l PCM 141-185  !

t STEAM GENERATOR N0ZZLE DAM INSERTS INIRODUCTIC24 .!

During a typical refueling outage, there are rumeruus planned irspecticn and maintenance activities in the primary head of the steam generator. Eddy current testing of steam w =.etor tubes during routine nuclear plant outages requires entry of

' # M and/or peh into the steam generator primary i In order to make this entzy, the water level in the' t

j head (s) .

l primary. system'is dig to a level belcw the bottom edge of the steam generator nozzles. S e manway cover (s) can then be removed to allow entry. 'Ihe reactor water level must be ,

maintadnad at this artificially.. low level as long as work is being p d ormed in the steam generator head or if the manway cover is rencved. mis situation prevents simultaneous refueling cpentions and steam generator inspection of the primary head.

C-E has developed adams" which can be rapidly installed in the steam generator rozzles to provide an effective isolation of the steam generator nozzles frun the primary system. mis .

allcus reactor watar level conditions to be maintained without concern for the work being performed in the steam generator.

Such an att w .t can result in a significant critical path saving, since both refueling and maintenance can take place simultaneously. l mis PC/M involves the installation of inserts into ea::h rozzle of the steam generator primary head. Se inserts oculd be used in the future to accept support pins for a C-E style nozzle dam, mis will prepare the steam generators for possible future use of s'aam generator nozzle dams.

mis PC/M also provides instructions for installation of the nozzle dams and cperation of the tczzle dam pressurization system.

PCM~141-185 b '

SAFLTY ANALYSIS L

' With rosy.ct to Title 10 of the code of Federal Regulations, Part 50.59, a pwiii.sd change shall be riaamai to irwolve an unreviewed safety questicrl:

(a) if the pr+=h414ty of occurrence or the consequences of an acx:ident or malfunction of aq'4==t important to safety previously _ evaluated in the safety analysis report may be increased, or

.(b)...if.a possibility.for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be creatad, or (c) if the margin of safety as defined in the basis for any technical specification is rectaced.

C-E concludes that the installation of nczzle dam inserts and the use of nozzle dans during nc=al refuelirg operation does not itwolve any unresolved safety gaestions as defined in iteos (a),

(b), or (c) above. -

1. Nozzle Dam InsePJ

. Die nozzle dam inserts do not perform a safety related function.

D)eir installation does however reqaire a verification to a safety related w -it. After welding, the inserts will have very little protrusicri into the reactor coolant flow path. 2)eir effect will be similar to existing clad overlaps. 'Ihe flow and other performance characteristics will not be affected. As described below, the ability 'of the steam generator to meet its plant life transient Based on these items, operatire recyiremen*a it is not affected.

there is' ro impact on plant technical specifications.

21e structural adegaacy of the steam generators is rot affected by the drilled'and welded inserts. Die design adequacy was investigated to the req.ti n t;s of the ASME Boiler and Pressure Vessel Code,Section III,1971 edition, and addenda th%. Sim m r 1972. 211s Structural Analysis is included as AttachneM. (2) .

. ahe W84* "

PCM 141-185 he load carrying capcbility of the rcxiified nozzles and the reactor coolant system stiffness during a design basis accident is not affected since the inserts are installed in a reinforced section of the nozzles. S e r.in M n prope.t es would be at the nozzle to pipe veld which is zu::cte frun the rcxiified area.

Although none is expected, any darage that is inadvertently caused to the steam generator pressure bcr.rdary dring irsert installation will be evaluated and repaind in accordance with ASME Code section XI reqairche.t s.

2. Nozzle Drs A stractural analysis w?.s pe:Tormed an the dam geacet:y to calculate the stresses due to the expec".ed loadings. Since the stnLN vas sy=aetrical, a finite ele aant resh rede,1 of one fourth of the asse:61ed dam was utilized. Several isostress plots were also printed to depict the general stress patterns for various strur'aral sections of interest, such as the d e , side wall, and the cross rib sections. Bis analysis determined that the dam stnLe is conservatively designed for the intended application.

An extensive test s vg u was performed to further verify the structural adequacy of the dam ard to d& Lobate the sealing capabilities of the seal design. Se test setup was a cladded nozzle that was used to test the first prtfatypical dam.Be dam was hydrostatically tested at 80 feet of head which is well above expected operating heads of 30 to 35 ft. Se cacplex dam stnLe was instrumented with strain gages located at the critical areas determined fren the stnLtal analysis. Se results fra:n this gaalification test showed that the dam stru.h. could withstaM the design caniitions with carfortable rargins of safety. Be seal was shcun to perform without leakage qwzating in the gecraetry which replicated the nozzle bors. Each nozzle dam to be irstalled has been hydrostatically tested to 60 ft. of water head.

Based upon the above analysis and testing, and upon retandarcy of the design, failure of the nozzle da~, during rc=al refueling operatic-s is not a credible event.

PCM 161-185 Ib L t

INSTRUMENT AIR AND SERVICE AIR LINES RELOCATION AT THE WTP INTRODUCTION.

Instrument' and Service Air supply lines to the Water Treatment Plant (V!P) are presently routed underground. Due to excessive corrosion the existing supply lines will. be capped and new lines will be routed from the air piping in the vicinity of the carbon filters, over the trestle and to the WTP air headers.

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 questions; (i) if the probability of occurrence or the consequences of i 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 saf ety as defined in the basis for any technical specification is reduced.

The instrument / service air lines relocation does not increase the probability of occurrence of an accident previcusly evaluated in the Ftnal

.Saf ety Analysis Report since the Water Treatment Plant (V!P) is a  ;

Y non-safety.related item.

Previously. evaluated malfunctions of equipment impc rt a r.t to safety are not increased since the WTF and the instrument / service air system is y

non-safety related, and is not located near any safety related equtpment The possibility for an accident or malfunction of a dffferent type than any previously evaluated is not created since the VTP is non-safety related. In addition, this modification does not affect any function er failure mode of the VTP.

The instrument / service air lines relocation does not reduce the margin l of safety as defined in the basis for any Technical Specification.

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

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

1 an unreviewed safety question and prior Commission approval for the implementation of this PCM is not necessary.

PCM 169-185 TURBINE G ANTRY CRANE BRAKE SYSTEM ABSTRACT REA SLN-85-72 requested engineering to be provided to upgrade the Unit I turbine gantry crane brake system to meet the operating capabilities of the existing Unit 2 turbine gantry crane brake system. Based upon the design and hardware provided by the crane vendor (Indusco), a pneumatic hydraulic system functionally equivalent to that utilized on the Unit 2 turbine gantry crane was implemented. To support this modification, a 10 CFR 50.59 review was completed and the respective safety analysis which is now part of this document was transmitted by Ref. 3. This design package functions to endorse the brake modification implemented by the vendor. The modification is considered Non-Nuclear Safety Related and does not create an unreviewed safety question.

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PCM 173-185 4

UPPER CUIDE STRUCTURE LIFT RIG MODIFICATIONS a.

Introducticri Durity removal of the Upper Guide Structure (UGS) in preparation'for refueling, one (1) of the three (3) bolts that attach the UGS to the UGS Lift Rig pulled out due to

=it. 'Ihis PC/M package d~'monts insufficient thread ar-+

the modifications to the IMS Lift Rig that are :w*=ca7 to allow the Lift Rig to be used for a one-time reinstallation of the UGS 'into the reactor following refueling. '1hese modifications are described in Section III, System Description.

Safet/ Aralysis With .W to ~'itle 10 of the Code of Federal Faqulatiers, Part 50.59, a prqr **4 change shall be deemad to involve an unreviewed safer / question:

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a. if the probability of occurrence er the ccrsequences of an i accident er ralfurction of equipne.v i=pertant to safec/

previcusly evaluated in the safety aralysis report ray be inc:: eased, er

b. if a possibility for an accident er malfunction of a different tj)x than any evaluated previci: sly in tre safet/

a alysis report may be created, er

c. if the margin of safety as defired in the basis for any ,

tatnical specification is r=*n M .

24 modification to the Upper Guide St::*rttre Lift Rig described in this PC/M package dces not involve any unreviewed safety question as defined in ite::s (a), (b), er (c) abcVe fer the follcwing reasons:

1. Utis modification does not increase the probability of occuicrence er the ecnsequences of an accident or malfunction of equipment i=portant to safety previcusly evaluated in the safety aralysis report. @A LUS Lift Rig is not safety related, and das not been aralyzed in the FSAR.

PCM 173-185

2. @.e possibility for an accident of malfunct3.cn of a different type than any previcusly evaluated in the safety analysis report has not been created. As stated, the UGs Lift hy is not safety related. Be design of tPA kWary Lifting Bolt Asse=blies Pas been aralyzed to i d kata the admiacy of this design, 'and to ersure l l

that the modifications will perform as required.

3. @A margin of safety, as defined in tPA basis for any technical specifications is not re ca9 The IMS Lift Rig is not used as the basis for any Technical Specificati<:n.

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i PCM 176-185 REPLACEMENT OF LIMIT SWITCHES ON FCV-23-3 & FCV-23-5 INTRODUCTION ~

It has been determined that valves FCV-23-3 and FCV-23-5 do not have

- qualified limit ~ switches for use outside containment in ~ the mechanical type penetration area. These valves presently use NAMCO EA-08022100 limit, switches which. are not qualified to withstand the parameters at NAMCO has a the penetration room for 40 year operation plus 1 year DBE.

one-for-one replacement for these limit switches,.the'NAMCO type EA180, they are designed to withstand a short-term exposure of 1750CThis and are PC/M qualified to IEEE standards 344-1975,. 323-1974 and 382-1972.

lirait switches with the EA180 are provides the method to replace the currentenvironmental qualification requirements limit. svtiches so that the ,

met. ,

SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed' change shall be dee=ed 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 reort may.be inc reased; or (ii) if a -

possibility for an accident or malfunction of a~different type than any evaluated previously . in the safety analysis report may be created; or (iii) if the margin of safety as defined-in the basis for any technical specification is reduced.

The change out of the limit switches perf ormed by this PC/M installs NAMCO EA-180 limit switches which are qualified to IEEE 323-1974, 382-1972 and 344-1975. These switches provide positive indication 'of valve position in the control room. The modifications included in this PC/M do not affect technical specifications. .

of l The four (4)' limit switches being replaced are considered to b6 part the Blowdown Co ntainment Isolation System. The replacement of these limit switches will not compromise the capability of these systems in any manner. The probability of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not-increased.

10CFR50.59(b), the written safety The foregoing constitutes, per i evaluation which provides the basis that this change does not involve any unreviewed safety question, therefore prior Commission approval is not required for implementation of this PC/M. 1 1

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• PCM 177-135 7

REPLACEMENT OF VALVE SOLEN 0 IDS INTRODUCTION .

The purpose of this PC/M is ' to replace sixteen (16) existing ASCO and g

two (2) AVC0 valve solenoid, no longer manuf actured, with " Qualified" ASCO solenoid NP-8316 Series on various valves. The replacement schenoids are . environmentally qualified IEEE-323-1974, ' 1EEE-344-197.5. and IE2E-382-i

. 1972; SA' FETY 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 (1) if the probability of occurrence or the consequences of an' accident

.or malfunction of equipment important to safety previously evaluated inLthe 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 b6 sis for an technical specifica-tions is reduced.

.This modification does.not involve an unreviewed safety question and the following.provides the basis for this conclusion.

The replacement of ASCO valve solenoids have been qualified to IEEE-323-1974, IEEE-344-1975 and IEEE-382-1972 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 Basis Accidents (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 ASCO valve solenoids are a one-for-one replacement of the existing ASCO solenoids. For the AVC0 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 replacement of these solenoids enhance the existing valves by installing

-environmentally qualified solenoids. .

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 evalua-tion which provides the basis that this change does not involve an unre-viewed safety question, therefore prior Commission approval is not requir-ed for implementation of this PC/M.

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PCM 184-185 ltl 1

1 MISCELLANEOUS ICW SYSTEM MODIFICATIONS 1 ABSTRACT This Engineering Package- documents minor. modifications' made to the Intake  ?

Cooling Water (ICW) system resulting from disassembly, inspection, repair and

- reassembly during the 1985 fall refueling outage. The modifications were initiated - via the Field Change Notice / Request - form and dispositioned by engineering. The modifications are classified as nuclear safety related and do  !

not constitute an unreviewed safety question.

- NOTE: THIS PCM IS FOR DOCUMENTATION PURPOSES ONLY.

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PCM 193-185- I

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Gland Steam Spl11over Vd ve Station Instrument Taps ABSTRACT This . design package covers the modification that adds a 1/2" instrument tap .

upstream and downstream of the gland steam spillover valve station. These instrumen'. taps consist of associated piping, root valves and pipe caps.. These ,

i taps ~ will enable . the plant to add . local instrumentation that will moni tor the -

l performance of the gland steam spillover system. This modification is classified as. non-nuclear safety related and does not involve an unreviewed safety l question.

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PCM 195-185 l l

MSIV AIR ACCUMULATOR 1A1 CHECK VALVE REPLACEMENT INTRODUCTION Upon inspection, it was discovered that check valve I-V-18-099 had damage to the valve body and internals and would require replacemec.t.

This is one of two check valves to MSIV Air Accumulator lAl on line 1-2-IA-100, which prevent loss of air from the accumulator in the event of instrument air loss.

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

The replacement of the MSIV Air Accumulator check valve does not increase the probability of occurrence nor does it increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report, since an existing da.maged check valve is being replaced with a valve of greater dependability.

The possibility for an accident or malfunction of a different type than any previously evaluated is not created since there has been no change to the system design, the seismic analysis has been reviewed and the margin of safety as defined in the basis for any technical specification has not been reduced by the subject modification.

The implementation of this PCM 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 ut reviewed safety question and prior Commission approval for the implementation of this PCM is not required.

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PCM 200-185 M[SCELLANEOUS SECONDARY SIDE RELIEF VALVE REPLACEMFJl' Abstract This Engineering Package covers the replacement of relief valves on the Open Blowdown ~ Heat Exchanger and the Steam Jet Air Ejector After Condenser which have beer found damaged._ .The replacement valves have been manufactured in accordance with ASME Section Vill and have been determined to be equal to or better than the existing valves. This modification is classified as non-nuclear safety related and does not involve an unreviewed safety question.

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PCM 201-185-FEEDWATER'LINE SUPPORT MODIFICATION INTRODUCTION Support BFH-76A has been found to have damaged rod and channels supporting this rod. This damage was done.during the current re fueling outage while lifting the' end bell of Reactor Coolant Pump 181 motor.

SAFEIY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety. question; (i) if the probability of occurrence or the con-sequences of an accident or malfunction o f equipment important to -

. safety previously evaluated in the Safety Analys si Report may be

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increased; or (ii) if a possibility for an accident or malfunction of a dif ferent 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 basis for this conclusion.

The modifications provided in this PCM repairs the damaged support assembly, retaining its ' original design function, without reducing its previously evaluated margin of sa fety.

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 evaulation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the ,

implementation o f this PCM is not required.

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i PCM 207-185 TURBINE AUTO-STOP RELIEF VALVE REPLACEMENT ABSTRACT This engineering package cove.s the replacement of the Turbine ~ Auto-Stop Relief Valves from the existing Lunkenhiemer and Teledyne Relief Valves to Fisher Type 98 Valves. The new Fisher valves will alleviate concerns over the inability of the existing relief valves to hold their setpoints. The valve replacement is classified as non-nuclear safety related.

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,,c w 9 PCM 015-936 l

TELEPHONE SYSTEM UPGRADE ABSTRACT 1 l

Tnis Engineering Package covers the modifications and details required to support the installation.(by AT&T) of a new AT&T System 85 PBX Telephone System. The central equipment for System 85 will be located' ,

J in'the Telecommunication Equipment Rooms in tne Unit 1 Service Building and Unit'2 D-13 Building.

Tne modifications and details consist of enlargement of the

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telecommunication rooms to accommodate the new equipment; installation of redundant air conditioner units for each room to satisfy equipment environmental requirements; power supplies with emergency back-up; q raceway between the two telecommunication rooms to install the AT&T '

supplied fiberoptics cable; raceway between Unit 1 PAX Mainframe and the new PBX System 85 and raceway between the D-13 Building, G-3 l Building and Start-up Trailers to accommodate the AT&T supplied multipair telephone cables.

Based on the importance of the telephone system as one of the plant communication means, this Engineering Package has been classified Quality Related to enhance the system design and installation confidence.

Tne new " System 85" will replace the existing Dimension 600 Electronic Stored Program PBX located in the Unit 1 Service Building Telecommunications Room and the Private Automatic Telephone Exchange (PAX) located in Unit 1 Reactor Auxiliary Building (Elev 43'-0). Tnio repl a cement will require modification of Section 9.5.2 " Communication Systems" of tne Unit I and Unit 2 FUSAR, Figures 9.5-1 and 9.5-4, Table 9.5-6 of the Unit 2 FUSAR and Section 3.8 of the Unit I and Unit 2 Nuclear Fire Protection Program and Section 7.1 of the Security Plan.

To energize the System 85 telephone equipment and air conditioners located in the. Unit 1 Service Building upon loss of norma'l off-site power will require manual switching at Power Panel PP-135 located in the Security and Records Building. Resetting may also be required upon returning of normal of f-site power. Tne System 85 telephone equipmen; modules and air conditioners located in tne Unit 2 D-13 Building will be automatically supplied by the Non-Claes IE diesel supplying tne D-13 Building upon loss-of normal power.

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PCM 055-186 CONDENSATE STORAGE TANK SECURITY ~JALLS INTRUMTION When the , covers over the condensate storage tank pipe trenches are removed, access to the condensate storage tank is possible throegt openings in the shield wall.

1 Because the condensate storage tank provides the feedwater for th-auxiliary feedwater system, this area is considered a protectec or

" vital" area under the St Lucie security plan.

SAFETV ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part

'50.59,"a', proposed change shall be deemed to involve an unreviewec safety questibn; (i) if the p robt.bilit y of occurrence or the consequences of an accident or malfunction of equipment important to safet.y 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 my be created; or (iii) if the margin of safety as defined in the basio.for any technical specification is reduced.

1 This PCM provides the details f or additional concrete walls to plug the existing - openings in the condensate storage tank shield wall in the l pipe trenches. This modification does not affect the st ruc tur a l capability of the existing shield wall.

Pipe sleeves have been provided in the new walls for the existing safety. related pipes that penetrate the walls. Accordingly, the walls have been seismically designed so as to not adversely affect this piping; and therefore. ' the walls are classified as Quality Related. i This proposed change does not involve a change to the Technical Specification.

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

PCM 070-186 SOFW PUMP LEAKOFF TANK TO CONDENSER MODIFICATION ABSTRACT Tnis Plant Change / Modification will raise the Steam Generator Feecwater Pump leakoff tank discharge line inside the condenser, in order to reduce its oxygen contribution to condensate.

This PCM is classified as non-safety related since the SGFW Pump leakoff to condenser system performs no safety related function and does not affect any safety related equipment.

This PCM does not. constitute an unreviewed safety question and bas no effect on plant safety.

This PCM has'no impact on plant operation.

Also, tne leakoff tank discharge control valve will be repla:ed witr.

a maculating ball valve to recuce pressure drop in the line, in order to raise the discharge line as high as possibit inside tne condenser.

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PCM 073-186 l

TURBINE THRUST BEARING - ADD OIL SEALS ABSTRACT Due to excessive oil leakage on the LP turbine at St. Lucie Unit 1, an investigation was performed by the site Westinghouse representative. Based upon his review, Westinghouse developed and fabricated a supplemental oil seal to be used in conjunction with the primary oil seal (i.e., outer seaD. This new seal was then installed between the thrust bearing and existing outer . seal'en the LP turbine rotor. To support this modification, a 10CFR50.59 review was completed and the respective safety analysis which is now part of this package was prepared to verify that a safety concern was not created. This design package functions to endorse the use of the additional seal and provide for adequate ? scument upda ting. This modi.'ication is considered Non-Nuclear Safety Related and does not create an unreviewed safety question.

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PCM 077-186 l

ENVIIiONMENTAL QUALIFICATION UPDATE ABSTRACT This engineering design package provides the vehicle for updating several areas of equipment qualification. This package includes corrections to the 10 CF R50.49 list, changes in maintenance requirements, and various documentation corrections.

This design package is considered nuclear safety related because it affects equipment falling under the scope of 10CFR50.40. 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.

PCM 079-186

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STEAM GENERATOR TUBE PLUGGING - CE DESIGN PLUGS ABSTRACT This PCM documents Engineering review and concurrence for the use of Corebustion Engineering expanded type tube plugs in the St. Lucie Unit I steam l generators. This PCM also provides the information necessary to as-build affected documents.

The modification described herein is classified as nuclear safety related. No unreviewed safety questions exist as defined by 10CFR 50.59; therefore, commission approval is not required prior to implementation.

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_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ , _ __ g,m.. _,,_ , ,,,,.m,, %_,  % ",og h% ,