ML17308A468

Rept of Changes Made Under Provisions of 10CFR50.59 for Period Ending 881006. W/890406 Ltr

ML17308A468
Person / Time
Site:	Saint Lucie
Issue date:	10/06/1988
From:	Conway W FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-89-133, NUDOCS 8904210079
Download: ML17308A468 (303)

Text

gccpLKRATED D1STRIBU'nON DEMONSTRA 04 WFTEM REGULATORY FORMATION DISTRIBUTION SY M (RIDS)

/filing ACCESSION NBR:8904210079 DOC.DATE: NOTARIZED: NO , DOCKET ¹ FACIL:50-389 St. Lucie Plant, Unit 2, Florida Power a Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION

-CONWAY,W.F. Florida Power a Light Co.

RECIP.NAME .'ECIPIENT AFFILIATION

SUBJECT:

"Rept of Changes Made Under Provisions of 10CFR50.59 for period ending 881006." W/890406 Itr.

DISTRIBUTION CODE: IE47D COPIES RECEIVED'LTR J ENCL L SIZE'OTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA PD2-2 PD NORRIS4J 4 1, 1 0 0

INTERNAL: ACRS AEOD/DSP/TPAB lw l~ AEOD/DOA NRR/DLPQ/HFB 10 1 1 NRR/DOEA/EAB NUDOCS-ABSTRACT ll 1 1

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1 RGN2 FILE 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 NVZE 10 ALL 'KIDS" RECIPZWIS:

PLEASE HELP US 1Q REDUCE HASTE! CDNZACZ 'IHE DOCUMEZI'ONZROL DESK ROCM Pl-37 (EZZ. 20079) TO EZJMZNATB K%K NAME HKH DZBTRZBOZIQN LISTS FOR DOCUMENTS YOU DON'T NEEDf

/'f TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL

P.O. Box14000, Juno Beach, FL 33408 0420 FPt APRIL' 1989 L-89-133 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D." C. 20555 Centleme Re: St. Lucie Unit 2 Docket No. 50-389

~Re ort.'f 10 CPR 50.59 Plant Chan ee Pursuant to 10 CFR 50.-59(b)(2), the enclosed report contains a brief description of plant changes/modifications (PCM) which were made under the provisions of 10 CFR 50.59. Included with the brief description of each PCM is a summary of the safety evaluation.

'.Phis report includes PCMs completed between'ctober 7, 1987 and October 6, 1988 and correlates with the information included in Revision 4 of the Updated Final Safety Analysis Report.

Very truly yours, I

F. Conway Senior Vice President Nuclear WFC/JRH/gp Enclosure cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant

Re: St. Lucie Unit 2 Docket No. 50-389 Re ort of 10 CFR 50.59 Plant Chan ees F

ST. LUCIE PLANT UNIT 2 REPORT OF CHANGES MADE UNDER THE PROVISIONS OF 10 CFR 50.59 FOR THE PERIOD ENDING OCTOBER 6 g 1988 8904210079 88i 006 05000389 PDR ADOCK PNU

PLANT CHAN6E/MOD REVIEMED FOR PSL2 FSAR AMENDMENT 4 NUHBER REVISION T IT LE 895-283 8-2 SHUTDOMN COOLIN6 PURIFICATION SYSTEM 323-283 SECURITY CONSOLES LED 6RAPHIC DISPLAY-UNIT2 POMER SUPP ANN 488-283 U6S LIFT R16 TRIPOD NOD 827-284 HFIV - HAIN FEEDMATFR ISOLATION VALVE PRESS SMITCH REPL CEDHCS CABINET COOLIN6 845-284'54-284 FOXBORO RECORDER HODEL 226S CHAN6F.

129-284 HAIN PUR6E SYSTEM/LLRT TAPS 156-284 8-1 CONTINUOUS HONITORIN6 EQUIPMENT CABLF. NOOIFICATION 163-284 SPENT FUEL .6ATE STORA6F. AREA MODIFICATION 283-284 HYDR06EN DETECTION INSIDF. EXCITER HOUSIN6

-284 488V BUSES CV-2 UNDERVOLTAGE RELAY HODIFICATION

-284 CONTAINMENT ANNULUS AIR SUPPLY 889-285 ICM SYSTE,H ORIFICES 811-285 STEAN TRAP DRAIN PIPIN6 AS-FAIL REPLACEHENT 835-285 FUEL TRANSFER TUBE SHIELDIN6 848-285 6E SAN RELAY PC CARD REPLACEMENT 858-285 NEM FEED TO 488V POMER CENTER 2AS 885-285 ICM PUHP EXPANSION JOINT REPLACEMENT 893-285 8-2 PRESSURIZER HANMAY LIFT IN6 LU6 HODIFICATION 182-285 REHOVAL TEMP S/U STEAN SUPP PIPIN6 186-285 8-1 PSB- 1 UNDERVOLTA6E RELAY CABINET ENHANCEHENT 138-285 28 CHAR61N6 PUHP DISCHAR6E RESTRAINT ADDIT10N 143-285 LINEAR TRIPTEST POTENTIOMETER REPLACEHENT 49-285 CONDENSATE PUHP HINIHUH RECIRCULATION SYSTEM MODIFICATIONS

PLANT CHAN6F/MOD REVIEWED FOR PSL2 FSAR AMENDMFN1 NUMBER REVISION, TITLE 158-285 e-1 CONDENSATE PUMP MINI-RECIRC PIPIH6 163-285 REACTOR HEAD VEN1 LINE RES1RAINT MODIFICATION 196-285 8-1 ANAL06 DISPLAY SYSTEM 6RAPHIC DISPLAY SPARES 283-285 CCM BACKFLUSH STRAINER DRAIN

'88-285 RTD & THERMOkELL REPLM1-RCS 211-. 285 MAKEUP AIR FOR CON1hlNMENT HYDRO PUR6E SYS TEMP VLV MOD 888-286 e ATMOS DUMP VLV MOV ttOD 811-286 8-1 06 60VE.RNOR POltER SUPPLY 814-286 8-1 1AR6ET ROCK VALVES-STEM- ASSEMBLY UP6RADE 824-286 RDF-RTD TEMPERATURE TRANSMITTER REPLACEMENT

-286 CCN PUMP BEARIN6 'MATL CHAN6E 838-286 8-2 PCB TRANSFORMER REPLACEMENT 842-286 QUENCH TANK PN 1SOLATION VALVE REPLACEMFNT 862-286 8" ADD FLAN6E - PENET P-58 864-286 RCP INSULA1ION REPL e69-286 8 TORQUE SEATIN6-ATMOSPHERIC DUMP VAI.VES e75-286 8-1 'HEATER DRAIN PUMP MECHANICAL SEAL DEMINERALIZED MATER SUPPLY.

887-286 MISAPPLICA1ION OF LIMITORQUE OPERATORS 891-.286 e-1 CLOSF. INTERCEPT VALVE'IRCUIT MODIFICATION e92-286 ADDI1IONAL APPENDIX R FIRE SPRINKLER AND FIRE MRAP IN RAB 188-286 HI6H INITIAL RESPONSE EXCITATION SYSTEM 113-286 DIESEL 6ENERATOR CRANK CASE OIL DEFLECTOR PLATE 128-286 18 CFR 58.49 ENVIRONMENTAL QUALIFICATION LIST REVISION

-286 PRESSURIZER MISSILE SHIELD ACCESS LADDFR SAFETY CA6E 4-286 ICM LUBEMATER FLOMRATOR MODIFICATION

PLANT CHAN6E/NOO REVIEMEO FOR PSL2 FSAR AtiENDtIENT 0 NUtIBER REVISION T IT LE 129-286 S/U XjtIR L/0 DISC SMITCHES 131-286 QSPOS SOFTMARE tiODIF ICATION 882-287 IE BULLETIN 85-83 tIOV SMITCH SETTIN6 886-287 NRC 1E BULLETIN 85-83 NOV POSITION INDICATION 887-287 HP TURBINE INNER 6LAND AND BLAND DIAPHRA6H ENHANCEHENTS 819-287 8-2 OIFSEL 6ENERATOR TORSIONAL VlBRA110N ISOLATION 826-287 FIRE PROTECTION STRUCTURAL S1EEL FIRE, PROOFIN6 829-287 TURBINF. 6ENERATOR ADOITONAL OIL SEAL FOR 1 AND 2 BF.ARIN6 833-287 REPLACEHENT OF VALVE V3734 b

848-287 CONDFNSATE RECIRC 10'OND PNE.UHA1IC SQ. R1 EX1RAC10R REPL 287 HFRV POS1110N 1NDICATOR

, 8-287 8-1 REAC10R CAVllY "SEAL RIN6 SEAL 858-287 CONOENSA1E'UtIP EXPANSlON JOINT REPLACEtIE.NT 851-287 8-1 CONDENSER HOTMELL NITR06EN INJECTION CONNECTIONS 852-287 CONDFNSATE POLlSHER TIE-IN 855-287 HSR PART1110N PLATE NU1 REPLACFtIENT 856-287 8-1 ,488V SMITCH6EAR 2A1 AND 281 TRANSFORtIER REPLACEtIENT 858-287 BORIC ACID HAKEUP SYSTEti RELIEF VALVE HODIFICATIONS 859-287 8-1 LOM POMER FEEDMATER CONTROL SYSTN 861-287 8-1 INSTALL VERNIER NERCURY tIANOtiETERS 862-287 ANNUNC1ATOR NUISANCE ALARtIS 863-287 NEM FUEL CRANE INTERLOCK ADDITION 868-287 tiFIV TE,Rti STRIP REPL 9-287 RELOCATION OF 1HF. SBVF HEAVER CONTROL PANELS

-287 REPLACEHENT OF F1SCHE.R AND POR1ER.CONTROLLE.RS

PLANT CHAN6E/HOD REVIEMED FOR PSL2 FSAR AHENDHENT 4 NUHBER REVISION TITLE 877-287 8-1 ERDADS 879-287 EXTRACTION STEAH PIPE AND FITTIN6 HATERIAL UP6RADE 883-287 HISCELLANEOUS ICN SYSTEH HODIFICATIONS 886-287 'CONDENSER OUTLET TUBESHEET AND MATERBOX COATIN6S 889-287 REHOTE REACTOR VESSEL LEVEL INDICATION 891-287 REACTOR HEAD TORUS RIN6 HODIF ICATION 892-287 8-2 REPLACEHENT OF SAFETY RELATED BATTERIES 2A AND 28 896-287 8-1 PRESSURIZER INSTRUHENT NOZZLE REPLACEHENT 182-287 RCA PROTECTIVE CLOTHIN6 BINS 183-287 TSI THRUST BEARIN6 PROBE RELOCATION

- zsr HISCELLANEOUS SNUBBER HODIFICATIONS 114-287 LC XFRHR VLV PACKIN6 HODS 115-287 INSTRUHENT AIR AFTER FILTER ISOLATION VALVES AND BYPASS LINE 117-287 CONDENSER EXPANSION JOINT IHPIN6EHENT PLATE HODIFICATION 118-287 CABLE SUPPORT STRUCTURE CONNECTION HODIFICATION 128-287 6ROUTIN6 OF HASONRY BLOCK WALLS.

121-287 8 STEAH 6ENERATOR TUBE PLU66IN6 CE DESI6N PLU6S 122-287 '-1 2A/B SPARE STH 6ENERATOR INSTR NOZZLES CLOSURE HODIFICATION 126-287 INCORE INSTRUHENTATION I ICI) PLATE HODIFICATION 131-287 AS-BLD CCM SUPPORT 133-287 ltELD - LINE MH-8-88 136-287 CHECK VALVE V27181 REPLACEMENT 137-287 RCS INSTRUHENT NOZZLE INSULATION TEHPORARY HODIFICATIONS

-287 SPLICE BOXES 82124,34,35 139-287 REPLACEHENT OF I-FCV-25-788 ACCUHULATOR CHECK VALVES

PLANT CHAN6E/HOD REVIEMED FOR PSL2 FSAR AHENDHENT 4 NUHBE.R REVISION TITLE 158-287 CEA H6 SETS LOCKOUT RELAY 155-287 EXCORE S/U S CNTL CHAN HODS 827-288 ROSEHOUNT XHTR F19821 828-288 8 REPL PT-22-23 844-288 EQ LIST REV - SPARF. PARTS 848-288 DM6 CLARIFICATION-6EN PROT RLY 8 ICM INSTR 856-288 ICM 8 CM PUHP PACKIN6 REPL 866-288 EQ DOC PAC LIHITORQUE H010R OPERATORS 888-288 NAHCO LIH SM REPL- PCV-8881- 5 896-288 DM6 CLARlFICATION-EN6 SAFE6RDS CAB 288 e-1 PDIS REPLACEHENT

-288 EHDRAC DRAMIN6 2998-738. RK.V 1 les-2ss FUEL POOL PURIFICATION SYS PUHPS HECH SEAL REPLACEHENT 118-288 CONDENSATE RECOVERY SYSTEH PUHPS HECHANICAL SEAL REPLACEHENT 112-288 TURBINE 6LAND SEAL SYSTEHS PUHPS HkCHANICAL SEAL REPLACEHENT 149-288 DOC CORR PS 29-4,1-1i0-2 227-984 TURBINF. 6ANTRY CRANE SEPARAT10N REQUIREHENTS 887-985 HYPOCHLORITE CELL FLUSH SYSTEH 828-986 INTAKE CANAL DRED6IN6 AND SLOPE. RESTORATION 839-986 BLOMDOMN BUILDIN6 RADIATlON HONITORIN6 SYSTEH 111-986 S IHULATOR TRAININ6 FACIL1 TY PIP IN6 TIE- INS 186-988 S6 BLOMDOMN TREATHENT FACILITY SYSTEH PUHPS HECH SEAL RPLCHT

PCM 095-283 SHUTDOWN COOLING PURIFICATION SYSTEM INTRODUCTION The plant requested via DIR M-43 that the Shutdown Cooling Purification System, which is presently arranged using hoses connected to flanged pipe taps in the Safety Injection and Chemical Volume Control Systems, be hard piped. Such a modification would result in a savings in set-up and maintenance time at the start and completion of a refueling outage.

SAFETY ANALYS1S With respect to Title 10 of the Code of Federal Regulations, Part 50.59

'a.proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The SDC Purification System is composed of piping, supports and manual valves and does not consist of any active components. The system is utilized during shutdown conditions only, when RCS temperatures are less than 140 . The piping and supports are designed to ASME III, Class 3 criteria and appropriate portions of the piping are seismically supported to withstand the applicable loadings listed in Chapter 3 of the FSAR and to maintain the seismic separation criteria from safety related equipment and piping. Therefore, there is no increase in the probability of an accident or malfunction previously evaluated in the safety analysis report.

Isolation valves separating the SDC Purification system from the Safety Injection (SI) and CVCS systems are designated as normally closed and locked closed during normal plant operation. As such, these valves are verified'losed by pla'nt administrative procedures prior to reactor start-up. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the safety analysis report is not created.

The SDC purification system performs no safety function and is used only as a clean-up system for the RCS during plant shutdown.

Therefore, the margin of safety as defined is the basis for any technical specificatin is not reduced.

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

The foregoing constitutes, per 10 CFR 50.59 (b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question, and prior Nuclear Regulatory Commission approval for implementation of the PC/M is not required'

PCM 323-283 SECURITY CONSOLES LED GRAPHIC DISPLAY UNIT 2

'NTRODUCTION The NRC has determined that annunciation of the Security System power supplies is required for compliance with 10CFR Part 73 (i.e. requirements for for nuclear power plants). To meet the intent of this requirement, security'ystems status lights shall be installed on the security system alarm consoles to indicate the "at hand" condition of the power input to the security SUPs and therefore, to the entire security system.

SYSTEM DESCRIPTION The status of the security system is continuously monitored by security personnel, who utilize the central Alarm Station (CAS), and the Secondary Alarm Station (SAS). CAS and SAS are independent, manned stations.

This PC/M package provides the design and installation details that are necessary to install indicating lights on the CAS and SAS consoles, which will provide visual indication of the status and lineup of the power supply to the security SUPS and therefore to the security system. 'these lights provide the

following information:

1. "Normal" Indicates that the security SUPS is powered by the normal operating plant equipment lineup.

2. "Diesel" - Indicates that the diesel generator breaker is closed and the diesel generator is supplying power to the plant loads.

3.~ "Bypass" Indicates that the

~

static transfer switch has been placed in the manual bypass position.

Note: The "Bypass" indication will be accompanie'd by a "Normal" or a "Diesel" indication.

4. ",Battery" Indicates that AC power is not available.

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

The security system is a non-safety related plant system. the Central and Secondary Alarm Stations are components of this system. The modifications presented in this PC/M affect both safety and non-safety related plant equipment.

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PCM 323-283 SAFETY ANALYSIS (Continued)

The modifications to the CAS and SAS control panels, i.e., installation and wiring of the annunciator circuitry, and .the inputs to these annunciators are non-safety related. The alarm stations are non-safety, non-seismic structures. The majority of required cable to these areas will be routed in non-safety related cable tray. The balance of cable will be routed through appropriately dedicated raceway.

Diesel generator breaker position is monitored to provide input to the "Normal" and "Diesel" annunciator circuits. This portion of the diesel generator control circuitry is safety related. Therefore, this signal will be isolated from the non-safety security annunciation circuitry by utilizing existing isolation relays. These relays were provided as part of PC/M 015-283 and have already been qualified to the applicable industry standards.

The balance of the control relays that are required in this modification have been purchased and will be installed as non-safety related equipment.

Control power to all relays is from the associated plant power train (safety to isolation relays, non-safety to the non-safety control relays). All cables will 'be routed through the appropriate raceway and the raceway will be seismically supported as required (i.e. inside the RAB).

This modification has no impact on the plant Technical Specifications.

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

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PCM 408-283 UPPER GUIDE STRUCTURE LIFT RIG TRIPOD MODIFICATION ABSTRACT This engineering package provides the details to modify the UGS lift rig tripod to achieve the required clearance to permit the refueling machine to pass over the lift rig when the lift rig is mounted on the UCS in the refueling pool. This modification is necessary in order to minimize time consuming efforts to disassemble and move the tripod out of the path of the refueling machine during each outage.

Although the UGS lift rig does not perform a safety-related function as defined by FSAR Section 3.2.1, its failure during its operation could result in damage to nearby safety-related equipment. Accordingly, quality-related requirements have been applied to this design.

NUREG 0612, "Control of Heavy Loads", and the associated requirements of ANSI.. NI4.6 have been reviewed and any applicable.'requir'ements of. these documents have been incorporated into the design of the new trip'od.

A safety evaluation of this modification has been performed in accordance with 10CFR50.59. This evaluation indicates that implementation of this Engineering Package does not involve an unreviewed safety question. Furthermore, the implementation of this modification does not require a change to the plant Technical Specifications and has no detrimental effect on plant safety and operation. Therefore, prior commission approval for the implementation of this modification is not required.

SAFETY EVALUATION The new tripod assembly installed by this engineering package does not perform or affect any safety-related function and will only be used during refueling operations. Since failure of the tripod while lifting the UCS or CSB could result in a load drop onto the reactor and irradiated fuel assemblies, this component is considered important to safety. For this reason, quality-related requirements have been applied to the design.

The new tripod assembly has been structurally analyzed for dead and seismic loads subject to the requirements, of NUREC 0612, ANSI N10.6, and the applicable ASME and ASTM codes. The results of the analysis demonstrate that the new components are all within allowable stress levels. Additionally, a revie~ was performed to verify the acceptability of storing the original tripod in the refueling pool while attached to the cor'e sup'peart barrel lift rig.

The implementation of this modification has negligible impact on the containment heat sink, hydrogen generating source, and free volume analyses described in FSAR Section "6.2.

.This modification does not change any assumptions made or conclusions drawn in the St. Lucie FSAR, and there is no new failure mode introduced that has'ot been previously evaluated in the FSAR. However, FSAR Figure 9.1-13, Section 9.1.0.2.2.6 and Tables 6.2-7 and 6.2-3 must be updated to reflect the addition of the new tripod assembly.

l

PCM 408-283 The implementation of this modification does not require a change to the Technical Specifications.

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

The probability of occurrence or the consequences of an accident oi malfunction of equipment important to safety previously evaluated is not increased since this modification does not affect the availability, redundancy, capacity, or function of any equipment to mitigate the effects of an accident.

(ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated since the function of the tripod has not been altered.

(iii) The margin of safety as'efined in the bases of any technical specification is not reduced since the modified tripod performs no safety-related function and is not included in the bases of any

<<chn ical speci fication.

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

PCM 027-284 MAIN FEEDWATER ISOLATION VALVE PRESSURE SWITCH REPLACEMENT SYSTEM DESCRIPTION For each main feedwater isolation, valve (HCV-09-1A, 2A, 1B & 2B), there is a pressure switch that senses air reservoir pressure and another switch that senses air supply pressure. The air reservoir pressure switches, PS-09-1A2, PS-09-2A2, PS-09-1B2 & PS-09-2B2, are presently Barksdale series model B2TA12SS from Transamerica Delaval. Their function is to monitor actuator air.

loss past the air check valve by signaling if pressure drops below 70 psig in the air reservoir. The air supply pressure switches, PS-09-1A3, PS-09-2A3, PS&9-1B3 & PS-09-2B3 (same manufacturer and model number), monitor plant air supply loss to the actuator by signaling if pressure drops b'elow 80 psig at the air filter. These switches are very inaccurate since the setpoints are near the low end of their range (50-1200 psig). In addition, the existing switches are not rated to the DC voltage being supplied.

The'urpose of this PC/M is to replace the referenced switches with Static-0-Ring models 6N6-BB5-NX-ClA-JJTTX6 (air reservoir) and 6NN-LL5-C1A (air supply). These switches have an ad)ustable range of 20 to 180 psig and will maintain the same setpoints of 70 and 80 psig respectively.

The function of the air reservoir and air supply pressure switches is not nuclear safety related. The switches are used strictly for annunciation of pressure drops below the assigned setpoints. However, because the air reservoir pressure switches are considered pressure boundary safety related, the PC/M is nuclear safety related. In addition, since the pressure switches are electrically connected to nuclear safety related power supplies these switches will be evaluated and demonstrated that their failure does not preclude the actuation of the main feedwater isolation valves (MFIVs).

SAFETY ANALYSIS

,This PC/M is nuclear safety'elated because the air reservoir pressure switches (PS-09-1A2, PS-09-2A2, PS-09-1B2 & PS-09-2B2) on the Main Feedwater Isolation valves are part of a safety related pressure boundary. The air supply pressure switches (PS-09-1A3, PS-09-2A3, PS-09-1B3 & PS-09-2B3) are considered safety related because they are-electrically connected to a nuclear safety related power supply, however, after performing a failure mode and effects evaluation it is shown that any failures to the air supply pressure switches do not propogate and affect the actuation of the MFIVs. Therefore,.

these switches can be considered non-nuclear safety related. Appendix D provides an evaluation of the failure modes for the air reservoir pressure switches and the air supply pressure switches. The function of the air.

reservoir and air supply pressure switches being replaced by this modification is not nuclear safety related since their function is to provide annunciation and lamp indication whenever pressure drops below their assigned setpoints.

However, since the air reservoir pressure switches are pressure boundary safety related, the portion of the PC/M pertaining to these switches is considered nuclear safety related. The air reservoir pressure switches are qualified seismically to prevent the loss of the safety related pressure boundary.

PCM 027-284 The replacement pressure switches assigned to monitor air reservoir pressure (PS&9-1A2, PS9-2A2, PS-09-1B2 & PS-09-2B2) shall be Static-0-Ring model 5N6-BB5&X-ClA-JJTTX6. These switches have been qualified to IEEE-344-1975 standards as per test report no 17344-82N-D prepared by Acton Environmental Testing Corporation (AETC) for Static-O-Ring, Inc. Mounting of these switches, have been evaluated for seismic category I loadingg, This change is not an unreviewed safety question because: the probability of occurrence or the consequence of an accident of malfunction previously evaluated in the FSAR has not been increased. The new pressure switches are of approximately the same weight as the presently installed switches, and will be installed per the same requirements that applied to the existing switches.

The process inputs will remain the same as the existing switches: 1/4 NPT.

This modification will improve annunciation by replacing the existing switches with new switches which are less subject to setpoint drift and have a wider ad5ustable range. In addition, the new switches will satisfy the specifications with respect to voltage and amperage ratings.

For the same reason, no possibility for an accident or malfunction of a different'ype from any evaluated previously in the FSAR has been created by thi's modification. Additionally, the margin of safety, as defined in the bases for the technical specifications, has not been decreased'n conclusion, this modification does not involve and unreviewed safety question,

PCM 045-284 CONTROL ELEMENT DRIVE MECHANISM/CONTROL SYSTEM (CEDMCS) CABINET COOLING Descri tion of Change With a CEDMCS cabinet area ambient temperature or 76 F, cabinet discharge temperatures in excess of 120 F have been measured, accompanied by persistent cooling failure alarms. Such excessive innermabinet temperatures will reduce component lifetimes, resulting in costly premature failures.

This PC/M will add eighteen (18) exhaust fan assemblies (one per bay) to CEDMCS Cabinets C2 through C5 for the purpose of reducing internal cabinet temperatures.

Safet Anal sis With respect. to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (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; of (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 devined in the basis for any technical specification is reduced.

For the following reasons, C-E concludes this change does not involve any unresolved safety questions as defined in items (i), (ii), or (iii) above:

1. CEDMCS is a non-safety grade system;

2. This modification does not affect any of the isolation devices used to interface the CEDMCS Cabinets with Safety Related Equipment/Systems;

3. The system affected by this modification has not been used as a basis for any technical specifications;

4. This modification will reduce the probability of equipment malfunction by reducing the thermal stresses exerted on elect'ronic components.

5. Since the CEDMCS's not seismically qualified, this modification does not require a seismic reanalysis.

I The implementation of the PC/M does not require a change to the plant technical specifications.

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

'C/M is not required.

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PCM 054-284 FOXBORO RECORDER MODEL 226S CHANGE S stem Descri tion The St. Lucie Unit 2 Instrument List calls for Foxboro Recorder Model 226S for the following tag numbers in the Safety Injection System.

Recorder Pens Board Recorder Pen Board JR-001A 1 RTGB-204 PR-3301 1 RTGB-206

'R-.001B 1 RTGB-204 PR"3302 1 RTGB-206 JR-001C 1 RTGB"204 PR-3305 1 RTGB-206 JR-001D 1 RTGB-204 PR-,3306 1 RTGB-206 LR"110X/PR-1108 '

2 RTGB-203 TR-3303W RTGB"206 FR-3301 1 RTGB-206 TR-33032 1 RTGB-206 FR-3306 1 RTGB-206 tr-3351 2 RTGB-206 FR"3313/FR-3323 2 RTGB"206 TR-3352 2 RTGB-206 FR-3317 1 RTGB-206 PR-8013D/PR-8023D 2 RTGB-206 FR-3327 1 RTGB-206 PR-9013D/LR-9023D ' RTGB-. 206 FR-3333/FR-3343 2 RTGB-206 This PC/M allows the use of either the 'Foxboro Model 226S or Model 227S for the instruments identified above. Either model recorder can be 'removed and replaced with the other model- recorder without any wiring changes or input signal modifications. The change is simply the replacement o'f the Model 226S with the Model 227S or vice versa.

The Instrument List will be "as-built" for these recorders to read Foxboro Recorder Model 226S or Model 227S. See Appendix A for, the specific changes.

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

These recorders are safety'elated since they are for the Safety Injection System. There are no unreviewed safety que'sitons, since both recorder models are seismically qualified Category I and. functionally equivalent. The environmental design of the Foxboro recorders satisfies the mild environment of the Control Room and the design specifications for the control boards, that they are bein installed in. The Model 227 recorders shall also be ordered with a certificate of compliance that they are functionally'nterchangeable with Model'26 recorders that have been previously qualified.

PCM 129-284 MAIN PURGE SYSTEM/LLRT TAPS SYSTEM DESCRIPTION The purge system exhausts to the environment via the plant stack. The system has a capacity of 42,000 cfm and is operated during refueling mode only. The system is not required to operate during short term access to the containment.

During normal refueling purge, the containment ai.r is drawn through penetration P-10 which inc'udes butterfly iso ation valves I-FCV-25-4, <<5 and -6 into a filter casing. 'See FSAR Fig 9.4-8).,A valved (I-V-25-207-324P) test tap and plug is provided in the penetration pipe in the containment side of isolation valve FCV-25-5.

The air makeup side of the purge system includes penetration P-ll and

, isolation butterfly valves I-FCV-25"1, -2 and -3 in the direction of flow. k All six isolation valves in penetrations P-10 and P-ll close automatically on Containment Isolation Actuation Signal (CIAS) ~

Similarly to penetration P-10;= a valved (I-V-25-210-324P) test tap and plug is provided in the penetrati.on in the containment side of isolation valve FCV-25-2.

r Technical Specification 4.6..1.7.3 Surveillance Requirements states that "At least once per 6 months on a .STAGGERED TEST BASIS each sealed closed 48 inch containment purge supply and exhaust isolation valve w.'th resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 La when pressurized to Pa.

To perform the above surveillance requirement with the present facilities, the personnel performing the Local Leak Rate Test, has to enter the annulus, with the test equipment and transport the equipment and himself past hot piping areas and use a ladder to reach the present valves and test plugs.

This PCM will modify the means to perform the Local Leak Rate Tests for penetrations P-10 and P-ll from floor elevation 23 feet by providing test stations inside the annulus near the SM access door.

Each new LLRT station consists of an isolation valve and a plugged test 'connection to duplicate the facilities provided by the original design. A bleed valve open to atmosphere has been added to each test station to provide the means for controlled depressurization of the penetration or additional instrument connection.

i PCM 129-284 SAFETY ANAIYSIS 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 consequence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possiblity for an accident or malfunction of different type than any evaluated previously in the safety analysis a

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

The containment purge system is not a safety related system and is not required to operate following a design basis accident. It is required to purge the containment to allow reouired access time for 'the plant personnel during planned shutdown and refueling operations. The system requires approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> to reduce c/mpc to 1.0. A radiation monitor is provided in the plant stack to monitor the radiation level of gases being discharge Isolation valves and containment penetrations are designed to Quality Group B and seismic Category I. The extended installation to the new location for the LLRT test taps in the annulus is designed as seismic Category I.

The containment purge system penetration is not subject to bypass leakage testing since its penetrations are filtered by the shield building ventilation system. This modi.fication will not change that statement. The addition of the test tubing will not adversely affect the limits allowed by the Technical Specifications.-

This Plant Change Modificati.on does not change the philosophy of the main purge system or the Local Leak Rate Test Tap for penetrations P-10 and P-11. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analsyis Report or the Safety Evaluation Report has not been increased nor has a new situation been created. The margin of safety as defined in the basis for the Technical Specifications has not been changed.

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

PCM 156-284 CONTINUOUS MONITORING EQUIPMENT CABLE MODIFICATION S stem Descri tion The Continuous Mongtoring Equipment is a system that has been en-gineered and purchased by Florida Power and Light Co. for the pur-pose of monitoring and recording the electrical generation para-meters of the St Lucie Plant Unit d2 (a similar installation exists for St Lucie Plant Unit 81). Inputs are taken from various locations in the plant system. This"data is collated and recorded at the Continuous Monitoring Equipment cabinet located at Elevation 43.0 in the Reactor Auxiliary Building.

The following CME cable modifications are addressed under this PC/M package:

1. Cables 20916F and 20917F referenced previously on CWDs 2998-B-327 sheets 916, 917, 918 and 919, were intended to provide metering signal from 4160 volt switchgear to CME cabinet via RTGB '201. These cables, however were not installed. Review of the cable routing indicated that due to inaccessibility of conduits in RTGB 201, it was not feasible to utilize the route via RTGB 201. Revised control wiring diagrams employ the al-ternate route which provides cabling from 4160 volt switchgear to the CME cabinet. This change resul'ts into the routing of four cables 20916F, 20917F, 20918H and 20919 H. Pull cards for these cables are included in this PC/M package.

2. Points D15 and D16 on the Data Acquisition Package (DAP) are being used as junction points to tie the field current ammeter (AM-872) to the CME channel 20 (CWD 8871), this is not as shown on CWD 2998-B-327, sheet 872. The internal wiring from T17-61 and 62 (sh.872) to D15 and D16 (sh.871) on the DAP will be removed. New wiring will be provided from T17-61 and 62'to T13- 90, 91 and 92. The CWDs have been corrected, to incorporate this change.

I

3. Underfrequency relays 81F2 and'81F4 on CWD 882 show an internal jumper between actuating and seal-in contacts. This jumper was missing so an external jumper was added. This is shown on revised CWD.

4. Watt-hour meter WHM/881 on CWD 881 had input connections reversed.

Connections to terminals 3-4 should be 4-3, 7-8 should be 8-7.

The CWD has been revised.

5. Startup transformer MOD indicating lights have connections reversed. Cable connections to lights for startup transformer A and B should be swapped. This will be shown on revised CWD 1105. Cables remain in pl'ace. Internal jumpers will be used to prevent repulling.

PCM 156-284 SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; if (i) the prob'ability of occurrence or the consequences of an accident or malfunction .of equipment important to safety- previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunciton of.a different type than any evaluated previous-ly 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 operation of the Continuous Monitoring Equipment enhances the St Lucie Unit 2 generation monitoring system by providing a capability to record parameters such as output voltage, current and power (KW and kVAR ),

generator field current and circuit breaker status.

The modifications entailed in this PC/M are non safety related and are required for the compl'ete.operation of the Continuous Monitoring Equip-ment.

The cable routings .have been designed in accordance with the St Lucie 2 cable ampacity, tray is seismically fill and support criteria. Furthermore, the existing raceway supported.

The foregoing constitutes, per 10 CFR 50.59 {b), the written which provides the basis that this change does not involve safety'valuation an unreviewed safety question, therefore prior Commission approval is not required for implementation of this PC/M.

PCM 163-284 SPENT FUEL GATE STORAGE AREA MODIFICATION INTRODUCTION PCM 163-284, '"=Spent Fuel Gate Storage Area Modification", was initially implemented to modify the spent fuel cask pit bulkhead storage rack, located on the north wall of the spent fuel pool. The modified design prevented the potential interference between a fuel element as it handled and one or the cross members on the bulkhead storage was'eing rack. A'teel plate was added to the front of the rack to present a smooth surface to any fuel element which should contact it.

Supplement 1 to PCM 163-284 was implemented to replace two level switch support brackets and two temperature element support brackets which pro)ected 18" from the spent fuel pool wall. The new design reduced the potential for interference between the brackets and fuel elements during handling by modifying the brackets so that they were no more than 11" from the pool wall.

Supplement 1 to the PCM introduced a new interference between the level switch supports and the refueling machine trolley above the supports.

Consequently, Supplement 2 was implemented to lower the supports to provide additional clearance between the level switches and the trolley mechanism.

Implementation of Supplement 2 to the PCM introduced calibration and setpoint problems for the existing level switches which resulted in the loss of the fuel pool high level annuniciation function. The scope of Supplement 3 entails the restoration of the spent fuel pool high level detection capability. Implementation includes replacement of the two level switches (LS-4420 and LS-4421) with switches which are short enough in height to clear the refueling machine trolley and are capable of being calibrated to the high and low level setpoints. A new flanged spool piece will be fabricated and added to each support bracket to elevate the switches to the correct height for the fuel pool level.

SYSTEM DESCRIPTION" The redundant level switches are safety-related and meet Class 1E and Seismic Category I requirements. They will be calibrated for their annunciation functions at high and low level setpoints of 60.5 and 59.5 feet respectively. The flanged 4" spool pieces, shown on BCS-163-284.3001 R2 in sections C-C and D-D, rest on the lower flange of the support brackets. They are held in place by (8) 5/8 x 3" stainless steel hexagonal head nuts with bolts tack welded in place. The level switches will be located on top of the spool pieces with the instrument displacement devices suspended through the spool pieces into the spent fuel pool water. The level switches will also be held in place by (8) 5/8 x 3" stainless steel nuts and tack welded bolts. Existing wiring will be used to connect the switches to the annunciators.

PCM 163-284 SAFETY ANALYSIS With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of sa'fety as defined in the basis for any technical specification is reduced .

The replacement of two level switches and the addition of spool pieces to the level switch support brackets will restore the spent fuel pool high level alarm function without losing the low level alarm function and without sacrificing clearance between the instruments and the refueling machine trolley. The new level switches are safety-related and meet Seismic Category I and Class 1E requirements as demonstrated by Wyle Labs test report No. 43235-1 Rev A. The low.level alarm function

'onsists of redundant annunciation in the Control Room which completely meets the requirements- of Technical Specification 3/4.9.11. The high level alarm function is provided solely to identify a high level condition to the operators and is not required by NR" regulation or for accident prevention as a result of.FPL' implementation of PCM 052-283 Transfer Canal Bulkhead Modification" . In PCM 052-283, an opening was cut in the .bulkhead door to allow drainage of spent fuel pool water in the unlikely event of an overfill.

Tne implementation of this PCM does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report since the level switches are not required for accident prevention or equipment protection. In addition, the possibility for creating an accident or malfuncion of a different type than any evaluated previously in the safety analysis-report is not created since the function of the level switches has not been. modified and the new switches are seismically qualified and mounted to ensure that the requirements applicable to Seismic Category I are met and are qualified Class lE.

Also, an analysis of the additional weight, due to the new spool pieces required for the level switch installation, shows that the modifications do not exceed the load limits allowed for the brackets.

The imp).ementation of this PCM does not require a change to the plant technical specifications. The foregoing constitutes, per 10CFR50.59(b),

the written safety evaluation which provides the basis that this change does not involve any unreviewed safety question, therefore prior Commission approval is not required for implementation of this PCM.

PCM 203-284 HYDROGEN DETECTION INSIDE EXCITER HOUSING SYSTEM DESCRIPTION The Generator Exciter Housing Hydrogen Monitoring System, as added by this PC/M is a combustible gas detection system which includes:

l. One (I) combustible gas detector mounted on the top of the exciter housing toward the generator end,

2. One (I) combustible gas detector mounted on the top of the generator removable end cover.

3. Two (2) control/indicating modules located in the turbine building on the mezzanine floor level,

4. One (I) terminal box located. inside the generator appearance skirting. A removable access cover to be installed on the appearance skirting in order to provide terminal box access,

5. Provisions for calibrating the sensors, 6, The design incorporates the easy removal of the sensor and sensor assembly from the turbine-generator exciter housing,

7. Electrical enclosure for the control modules, provided with a viewing window.

The Combustible Gas Detection System, is an instrument package specifically designed to continuously monitor for flammable gases and vapors and to activate a warning or alarm when predetermined concentration levels are reached.

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 mal function of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or mal function of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of sa fety 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.

PCM 203-284 The modifications included in this PC/M affect only the turbine-generator'and exciter. The hydrogen detection system provides an additional margin of safety by providing early warning indication of.

combustible gases. In addition, these components are all non-'safety related and non-seismic.

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

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 implementation for this PC/M is not required ~

PCM 217-284 480V BUSSES CV-2 UNDERVOLTAGE RELAY MODIFICATION" Description of Change Dual contacts in the CV-2 relay could not be adjusted so that both operate at the same voltage.

This PC/M revises the wiring to the Westinghouse CV-2 relay for the 480V Busses 2A2/2B2 and 2A5/2B5 Undervoltage Protection System.

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) the'robability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previouslyin the safety analysis report may be created; or (iii) if the margin of safety in the basis for any technical specification is reduced. as'efined This PC/M includes rewir'ing the contacts on the Westinghouse CV-2 undervoltage relay. No mod'ifications to these Class 1E relays are being made. As a result of these modifications no change in the system operation results. The system operates identically to the previous design and only the wiring external to the CV-2 relay is revised.

No modification to the existing undervoltage protection scheme other than the CV-2 wiring are included in this PC/M..

This change alleviates maintenance procedures and ensures proper operation of these undervoltage relays.

This PC/M does not result in a change to the FSAR or Technical Specifications.

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 PC/M.

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PCM 236-284 CONTAINMENT ANNULUS AIR SUPPLY SYSTEM DESCRIPTION

~eration The operation of the Instrument Air (IA) System is affected by this modification since the filter element needs replacement periodically at intervals recommended by this design document. The inlet, outlet and bypass valves added by this modification should be included in section 8.13 of Operating Procedure 2-1010022.

Function The filter added by this modification will function to ensure that particulates do not interfere with the operation of valves inside the annulus which are supplied by the Instrument Air System. Elimination of these particulates wBI increase the reliability of the affected system loads, (i.e., those valves which must be cycled within Technical Specification limitations).

Desi n Descri tion This modification adds one particulate filter and its associated inlet, outlet and bypass valves to 3/4-IA-73 which supplies- the annulus through penetration 62. This filter will be located near the penetration in the RAB.-

SAFETY EVALUATION 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 portions of the IA System affected by this modification are not within the ASME Class III boundary and the components added by this modification do not perform a safety function. Therefore, this modification is classified as non-nuclear safety-r elated, Quality Group D.

1.2 These modifications do'ot-interact with any safety related systems or components.

-1.3 No safety-related equipment or components are compromised by any assumed failure of existing or new equipment or components.

Therefore, failure of the filter wQI not increase the probability of an accident or malfunction of equipment important to safety previously evaluated.

1.4 No parameters relating to Technical Specifications are adversely affected and no Technical Specifications are altered.

2.0 Care has been taken from the design bases to system design phases to recognize and eliminate, mitigate or control all potential features which could be hazardous to the safety of equipment and/or personnel. This review constitutes, per 10 CFR 50.59, the safety evaluation; thereore this modification does not require prior Commission approved for implementation.

PCM 009-285 ICW SYSTEM ORIFICES MODIFICATION DESCRIPTION Cooling Water (ICW)

This PCM provides guidelines and details for replacing Intake Cooling Water System orifices I-SO-21-1A and 1B (downstream of the Component SO-21-2A and 2B (downstream of the Turbine Cooling and Heat Exchangers) and orifice plates are Open Blowdown Cooling Water Heat Exchangers). The existing known to be deteriorated due to stress corrosion and/or flow erosion.

These orifices were installed early in plant life and are designed to stage the valves (TCVs) to pressure drop downstream of the ICW system temperature control reduce flow erosion (due to cavitation) of piping downstream of the TCYs.

'This PCM provides details for new orifice plates, to be constructed of titanium.

SAFETY ANALY SIS Regarding I-SO-21-1A and 1B:

la. With respect to the probability of- occurrence of an accident previously evaluated in the FSAR:

The replacement of these orifices with new orifices, which have identical flow-pressure drop characteristics but are to be constructed of an upgraded material, will have no impact on the probability of accidents previously evaluated in the FSAR since no system design parameters or margins have been changed.

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

The consequences of accidents previously evaluated in the FSAR have not been made more serious since these orifices will produce the same flow-pressure drop characteristics as those they replace, and ICW system heat

'removal capability has not been reduced or altered.

lc. With respect to the probability of malfunction of equipment important to nuclear safety previously evaluated in the FSAR-Same as la Id With respect to the consequences of malfunction of equipment important to nuclear safety previously evaluated in the FSAR:

Same as la 2a. With respect to the possibility of an accident of a different type then previously evaluated in the FSAR:

There is no possibility of an accident of,a.different type than previously evaluated in the FSAR, since, the modification only upgrades the material of the orifices, making them more reliable. Assuming failure (i.e., degradation) of an orifice, potential ICW flows would only increase, therefore, increasing containment heat removal capability.

,PCH 009-285 2b. With respect to the possibility of equipment malfunction of a different type than analyzed in the FSAR:

Other than a straight replacement of the existing orifices with new orifices of an upgraded material, no new equipment is added by this PC/M.

Additionally, no other existing equipment is modified by this package, therefore, there is no possibility of equipment malfun'ction of a different typed than previously evaluated in the FSAR.

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

No ICW system design parameters have been altered since the new orifices have the same flow-pressure drop characteristics as the existing orifices.

Based on the above arguments, it is concluded that no unreviewed safety question exists as defined by 10 CFR 50.59.

Regarding SO-21-2A and 2B:

The upgrade of material for these orifices is considered non-nuclear safety related for the following reasons:

A) These orifices are installed in a non-safety related portion of the ICW System.

B) Postulated failures of these orifices would have no impact on safe shutdown of the plant.

C) The orifices are not required to prevent postulated accidents, mitigate the consequences of such accidents, maintain safe shutdown conditions or adequately store spent fuel.

PCM 011-285 STEAM TRAP DRAIN PIPING AS-FAIL REPLACEMENT Description of Change Existing carbon steel fittings and piping have experienced'everal failures due to corrosion-erosion effects.

This PC/M provides guidelines for replacing fittings and piping in'he steam

'rap-to condenser drain lines with upgraded materials (chrome-molybdenum) on an "as-fail" basis.

Safety Analysis This modification is conside'red nonnuclear safety related for the following reasons:

A. the steam trap drains are non-safety related.

B. No postulated failures of any of the steam trap drains would have on safe shutdown of the plant or safety related systems.

an'mpact C. The steam trap drains are not used to prevent postulated accidents, mitigate the consequences of such accidents, maintain safe shutdown conditions or adequately store spent fuel.

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PCM 035-285 FUEL TRANSFER TUBE SHIELDING ABSTRACT This Engineering Package (EP) details the requirements for the installation of additional concrete and lead shielding in the vicinity of the fuel The additional shielding is required to reduce personnel dose rates in transfer'ube.

the area during fuel transfer operations.

This modification is classified Safety Related because it involves an attachment to the containment vessel, which is Nuclear Safety Class 2.

This modification has been evaluated in accordance with 10CFR 50.59. The safety evaluation has shown that 'the implementation of thi.s Engineering Package does not constitute an unreviewed safety 'question an'd prior Commission approval for its implementation is not required. This modification will have no effect on plant safety or operation.

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

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

This Engineering Package provides for the installation of additional shielding in the vicinity of the fuel transfer tube to reduce personnel dose rates in the area. It does not involve an unreviewed safety question. The following are the bases 'for this conclusion:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since this modification will be performed in accordance with Safety Related requirements, hence the seismic capability of the existing structures in the area is not compromised. Therefore, there can be no impact on any adjacent safety related structures, systems, or equipment.

PCM 035-285 (ii) There is no possibility for an accident or malfunction of a different type than any evaluated previously since the modification will ensure that the additional shielding will have no interaction with safety related equipment and hence will have no effect on plant safety.

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

The implementation of this Engineering Package does not require a change to plant technical specifications.

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

=implementation of this Engineering Package is not required.

PCM 048-285 GE'AM RELAY,PC CARD REPLACEMENT S stem Descri tion Presently, St. Lucie Plant Unit 42 uses the'AM 11B rehys for circuit .

breaker failure backup protection schemes. The following is a list of their application at St. Lucie Plant, per reference 1.B..

St. Lucie Unit.N2 Switchgear/Cubicle 6.9 KV<A1O1 6.9 KV<A1O2 6,9 KV<B1O4 6,9 KVCB1OS 4.16 KVCA2O1 4.16KV<A2O2 4.16 KV<B2O9 4.16 KV<B2-10 4.16 KV< A4-1 4.16 KVCA4%

4.16 KV<B4-1 4.16 KV<B4%

This PC/M will rephce.the existing printed. circuit board for the above relays with a new PCC N0165B1987G10 printed board. This will eliminate any time delay problems due to an unusual initiating contact bounce while maintaining the existing function.

Safet Anal sis 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; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

C The modification being performed under this PC/M will enhance the operation of the G. E. Sam llB relay assuring that if the unlikely event of an initiating contact bounce occurred, the relay will time out appropriately.

The G. E. Sam 11B rehy affected are utilized for circuit breaker failure backup protection schemes and are not in any safety related circuit or performed a safety related function.

PCM 048-285.

Environmental qualification is justified by the fact that these relays and-thus their internal PC cards are located in a mOd environment.

There is no seismic concerns affected by this'modification, the relays have .,

no seismic requirements associated with them.

Therefore, the probability of a previously reviewed accident is 'not increased, the possibility of an accident of a different type has not been created and the margin of safety has not been reduced. The implementation of this PC/M does not require a change to the plant technical specification. The foregoing constitutes, per 10GPR50.59(b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question, therefore, prior Commission approval for implementation of this PC/M is not necessary.

PCM 058-285 NEW FEED TO 480V POWER CENTER 2A5 SYSTEM DESCRIPTION This PC/M provides the design details to add a new safety related "SA" 4.16kV breaker to feed the 480V Power Center, 2A2, including control, indication and annunciation functions previously associated with the bifurcated feeder breaker. The additional 4.16kV breaker is being located in a new cubicle which is being added to .the existing switchgear 2A3.

To allow this modification, the Isolation Panel IP"283, which is presently located at the end of 4.16kV switchgear 2A3, is being relocated to a new location.

The new 4.16kV switchgepr cubicle complete with 1200A, 250MVA short circuit rating, 80kA momentary interrupting current capacity breaker and all appurtenances (relays, bus and miscellaneous devices) switch are being for procured from Westinghouse. Indicating lights and control installation on RTGB 201 are being procured from General Electric.

Environmental and Seismic Qualifications for the above material have been provided by Westinghouse and General Electric, respectively.

Environmental and Seismic Qualification for the 4.16kV -switchgear cubicle addition have been provided by Westinghouse via the following report: "Westinghouse Qualification Report to Florida Power 6 Light Company for DHP Medium Voltage Metal Clad Switchgear Cubicle Addition at St Lucie Plant Unit 2", dated May 1986.

This report has been reviewed, found acceptable and entered into the EYiDRAC system under dra~ing number 2998-18321.

Seismic analysis to determine the .effect of the new loads on the existing seismic qualifications of RTGB 201 is being provided by Acton Environmental Testing Corporation.

Also as part of this package redundant fuses (per Appendix will be added to the control circuits of the new breaker

'R'equirements) in the 4.16 kV Switchgear 2A3 Cubicle 1A.

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

PCM 058-285 The 4.16kV breaker in Switchgear 2A3 feeding power centers 2A2 and 2A5 as well as the associated controls on RTGB 201 are all safety related train ."SA".

The addition of a new safety related "SA" 4.16kV breaker to feed the 480V power center 2A2 including control, indication and annunciation functions on RTGB 201 previously associated with the bifurcated feeder breaker, constitutes an enhancement to safety. By installing the new 4.16kV breaker for feeding power center 2A2, to'tal loss of the safety related 480V and 120V systems in case of a single fire in the "B" area, where the 2A5 power center is located, is being prevented.

The equipment required to implement this package includes a safety related 4.16kV switchgear cubicle, to be installed at the end of existing switchgear 2A3 and control switch/indicating devices to be installed on the RTGB 201. Isolation Panel IP-283 is being relocated to allow that the additional cubicle be attached to the existing 4.16kV Switchgear 2A3.

This implementation will also require the addition of redundant fuses (per Appendix 'R'equirements) to the control circuit of the breaker in the 4.16 kV Switchgear 2A3. This scheme will permit continued control power to the breaker feeding 480V switchgear 2A2 upon isolation from the control room because of control/cable spreading room fire.

~< switchgear cubicle addition,has been environmentally and seismically qualified (Qualification Report No 2998-18321, Revision 1) for its installed location/configuration. The new conduit runs are seismically supported in accordance with the Flectrical Installation Notes and Details.

The addition of the new devices to the previously qualified RTGB and 4.16 kV switchgear 2A3 has been seismically evaluated with no significant impact on the dynamic characteristics of the RTGB and the 4.16 kV switchgear 2A3.

This modification does not require a revision to the Technical Specifications.

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

PCM 085-285 ICW PUMP EXPANSION JOINT REPLACEMENT SYSTEM DESCRIPTION Function The intake cooling water pumps provide cooling water from the intake structure to the CCW heat exchangers and the TCW heat exchangers. Expansion joints between the pump nozzle and the piping system reduce stresses on the nozzle imposed by the movements of the piping system. Although they are not necessary to meet code design stresses, the expansion joints reduce vibration induced stresses and ease reassembly of the system.

Desi Descri tion This PC/M changes the bellows material from Monel (ASME-SB-127) to Inconel 625 (ASMEWB-443). The liner material is changed fron 316 SS to Inconel 625. The attachment of the liner and bellows is changed from a welded design to a Van Stone flange design. These design changes are made to reduce the current rate of corrosion exhibited by the existing expansion joints. Inconel 625 is superior with respect to corrosion fatigue strength, pitting and crevice corrosion resistance when compared with Monel. The new expansion joints should have a significantly longer service life than the existing Mon el expansion join ts.

h The new expansion joints shall be designed and fabricated in accordance with ASME Section III Class 3 requirements, except no N-Stamp is required.

SAFETY ANALYSIS This change does not represent an unreviewed safety question since it does not affect any accident addressed in the FSAR, present any new accident not previously analyzed- in the FSAR, nor does it affect the margin of safety for any technical specification.

The operation of the intake cooling water pumps or the piping system has not been affected by the use of an alternate material as specified in this.

PC/M package, as this alternate material is equal to'or better than the original material in all aspects. Therefore, this material change does not increase the probabilities or consequences of accidents or equipment malfunction important to the safety of the plant previously evaluated in the FSAR.

PCM 093-285 PRESSURIZER MANWAY LIFTING LUG MODIFICATION SYSTEM DESCRIPTION FUNCTION This modification functions to provide for removal of part of the pressurizer manway cover lifting lugs such that accessibility for the:

Kleiber and Schulz stud tensioning ring is provided.

DESCRIPTION This modifica:ion provides for removal of a small portion from the end of each pressurizer manway lifting lug that presently interferes with <he use o'he Kleiber and Schulz stud tensioning ring.

The lug shall be modified by boring a new 5/8" hole located such that sufficient metal will exist on all sides. Also, this change provides for removing sufficient lug material to preclude usage of the existing lift'ng holes.

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

(i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluation previously in the Safety Analysis Report may be created; or (iii) if the m'argin of safety as defined in the basis for any technical specification is reduced.

This modification provides for removing a small portion of the pressurizer manway cover lifting lug. The intent of this modification is to provide for removal of interference for the manway cover stud tensioning ring.

Modification of this lug in.no way affects the integrity or function of the

'anway cover and, therefore, does not increase the probability of occurrence or consequences of an accident or malfunction previously addressed in the Safety Analysis Report. Additionally, the modification does not affect or require a change in the Technical Specifications.

The forgoing constitutes per IOCFR50.59(b), the written safety evaluation which provides the basis that tnis change does not involve an unreviewed safety question. Therefore, prior commission approval is not required for implementation of this PCM.

PCM 102-285 REMOVAL TEMP S/U STEAM SUPPLY PIPING ABSTRACT This PCM package was developed to support the removal of temporary steam supply lines which were installed during the construction of St. Lucie Unit I/2 and are no longer in use. The removal of these lines and the replacement of the existing eroded flange connections at the points where they tie into the steam lines is a non-nuclear safety related modification.

SAFETY ANALYSIS Because the lines to be modified by this PCM (3-MS-35, 2'-MS-06; and 0-MS-62) are components that are not involved in the FSAR analysis of accidents the probability of occurrence of accidents prev iously addressed in the FSAR is unaf fected, the consequences of the accidents addressed in the FSAR are'unchanged, and the possibility of new accidents not considered in the FSAR is not increased. The three lines are not equipment that is impoitant to safety, thus the modification does not affect the probability of malfunction of equipment important to safety previously evaluated in the FSAR, does not change the consequences of malfunction of equipment important to safety previously evaluated in the FSAR, and does not create the possibility of malfunctions of a different type than those analyzed in the FSAR. The three lines are not equipment that is considered in the bases of the Technical Specifications, so no margin of safety defined therein is affected by the modification. Failure modes evaluated as described in the design analysis also demonstrate that there is no potential interaction with safety related equipment or functions.

Based on the above discussion, it can therefore be concluded that the implementation of non-nuclear safety related PCM 102-285 will not create an unreviewed safety question.

PCM 106-285 PSB-1 UNDERVOLTAQE RELAY CABINET ENHANCEMENT ABSTRACT This Engineering Package (EP) modifies circuits and components in the PSB-1 Undervoltage Relay Cabinets to provide improvements to the cabinets as follows:

1) Replace existing potential transformers (PTs) with those with a center tap to evenly 'divide voltage in the event of umbalanced loads on the secondary windings.

2) Install test switches and indicating lights to facilitate periodic relay testing.

3) Modify existing ITE-27N undervoltage relays to correct operating anomali.es. Brown Boveri letter to USNRC dated March 13,.1984 provided a bulletin regarding relay misoperation.

This EP is classified as Nuclear Safety Related since it provides for modification to Nuclear Safety Related Class 1E equipment.

The safety evaluation will be completed upon review and approval of all outstanding qualification documentation, after all HOLD POINTS have been lifted.

Su lement 1 This EP has been revised to lift all HOLD POINTS. National Technical Services Acton Report No 23462-88N (EMDRAC No 2998-18510) has been reviewed and approved for seismic qualification of the potential transformers (PT-6S) as well as seismic qualification of the PSB-1 cabinets as modified by this PCM.

As a result, all HOLD POINTS have been lifted and the safety "evaluation has been updated. The implementation of this PCM does not affect the Plant Technical Specifications and does not constitute an unreviewed safety question. Therefore, Commission approval is not required prior to implementation of this PCM.

This EP has no impact on plant safety or operation.

PCM 106-285 SAFETY EVALUATION With respect to Title 10 of the =Code of Federal Regulations, Part ,

50.59, a proposed change shall be 'deemed to involve an unreviewed safety question: (i) if the probability of occurrence or. the consequences of an 'accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created, or (iii) i'f the margin of safety as defined in the basis for any technical specification is reduced.

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

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

This EP provides for evenly divided voltage in the event of imbalanced load via the new potential transformers with the center tap.. This modification provides more accurate undervoltage sensing and reduces the chance of unanticipated trip; this aspect of the EP does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.,

This EP provides for modifications to the ITE-27N definite time undervoltage relays in order to improve reliability and assure actuation in the event of a degraded grid voltage condition. This serves to decrease the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSARo This EP provides for the installation of test switches and indicating lights in order to isolate the undervoltage relays for testing purposes wihout the possibility of inadvertent propagation of 4160V switchgear trip. This does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

PCM 106-285 (ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated as confirmed as follows:

Replacement of the existing potential transformers with those with the center tap does not introduce the possibility of an accident or malfunction not previously evaluated in the PSAR since the primary and'secondary circuits are in no way changed and this represents design configuration as originally evaluated in the FSAR.

Modification of the ITE-27N undervoltage relays does not introduce the possiblity of an accident or malfunction of a different type than any previously evaluated in the FSAR since the two failure modes of the relays (fail trip and fail no-trip) have been evaluated in the FSAR.

The installation of the test switches and indicating lights does not introduce the possiblity of an accident or malfunction of a different type than any previously evaluated

.in the FSAR since these are non-active, in-line components foi which failure modes resulting in an accident or malfunction are not postulated.

(iii) These modifications"do not change the margin of safety as defined in the basis for any technical specification since they have no negative effect on undervoltage protection and/or plant onsite AC power.

Since this EP affects equipment that is identified as Nuclear Safety Related (the PSB-1 Cabinets provide undervoltage protection for Class lE buses), this package is considered Nuclear Safety Related.

Seismic qualification of the modification of the PSB-1 Cabinets and the potential transformers have been reviewed and approved; the structural integrity of the PSB-1 cabinets will be maintained with the implementation of this PCM and the potential transformers have been qualified for Nuclear, Safety Related'ervice (see NTS-Acton Report No 23462-88N, EMDRAC No 2998-18510 and Attachment 7.8).

This EP involves equipment on the Essential Equipment List, but does not modify their intended operation or function. This package does not affect safe reactor shutdown or alternate shutdown. There are no other changes to equipment which involves 10CFR50 Appendix "R" fire protection (See Attachment 7.1).

Implementation of Nuclear Safety Related PCM 106-285 does not require a change to the Plant Technical Specifications.

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

PCM 138-285 2B CHARGING PUMP DISCHARGE RESTRAINT ADDITION Introduction Pulsation and vibration testing of 2B Charging Pump discharge'line, conducted by FPL, indicated that line I"2"-CH-136 is experiencing excessive vibration when the pump is running. These vibrations have resulted in the need to make frequent repairs to this line.

S stem Descri tion The additional restraint proposed on BCS 138-285.3000, when implemented, will eliminate the undesirable vibrations which has caused the need for frequent repairs to the 2B charging pump discharge line.

Safet Anal sis With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type'han 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 additional restraint proposed by this PC/M, when implemented, will eliminate the undesirable vibrations experienced by the 2B Charging Pump discharge line and will provide additional restraining capabilities during the seismic event. Accordingly, it does not increase the probability of occurrence or the consequences of any previously analysed accident, nor does it create a.new accident, or reduce the margin of safety of any technical specifications.

The implementation of this PC/M does not require a change to plant technical specifications.

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

PCM 143-285 LINEAR TRIPTEST POTENTIOMETER REPLACEMENT This Engineering Package covers modifications in the Reactor Protective System Safety Channels. The PC/M will eliminate the combination pot/switches that are presently installed to perform functional testing of the linear trip function and replace them with high-resolution 10-turn potentiometers and separate toggle switches. The new components will ensure the required sensitivity to calibrate and perform, testing.

The modifications are classified as nuclear safety, related because the components being, replaced are part of the Reactor Protective System..

Safety Analysis With respect to Title 10 of the Code of Federal ~ations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question:

(a) 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.

(b) 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 (c) if the margin of safety as defined in the basis for any technical specification is reduced.

For the follmrirg reasons, C-E concludes this change does not

~

involve any unresolved safety questions as defined in items (a), (b) ar (c) above:

PCM 143-285

%be ch-mpes to the Reactor Protective Systan described in this PC/M i~rove the testing and calibration characteristics of the PM anl decrease the probability of spurious indications arik achmtions. Geese changes do not adversly affect the

/

a - described in the Technical Specif ications. 'Qmxefore, Technical Specification revisions are not necessary.

'Ihe changes to the Reactor, Protective System described in this PC/H do not affect the performance'f its design safety 1.~on, nor are any other plant operations or design W~cteristics adversely affected. Gherefore, the safety aJMysis transients are not affected and the consequ~ of ac'dents previously evaluated in the safety analysis report arp. not increased I Also, since the RPS safety f~Mons are not affected, since plant opemtion and design are not adv~y affected, arxi sir~ accidents previously evaluated in the safety analysis report are not amzneQ,to be initiated by faults within the RPS, the probability of oocurre'nce of accidents previously evaluated is not increased.

Additionally, the changes to the RPS do not adversely affect t.'I perfonm~, testixxy, calibration, or design features of the RPS, and therefore the possibility for a new kuxl of accident is not created.

Ghe foregoing constitutes the written safety evaluation, per 10 CFR 50.59(b), which provides the basis that this change does not involve an unreviewed safety question and prior ccsmr~sion anal for the implementation of this PC/M is not required.

PCM 149-285 CONDENSATE PUMP MINIMUM RECIRCULATION SYSTEM MODIFICATION ABSTRACT This Engineering Package (EP) is for the replacement of the existing 4 inch Condensate Pump Minimum Recirculation Flow Control Valves (FCV 12-3A, 3B and 3C) with 8 inch valves. It also adds an 8 inch manually operated isolation gate valve upstream" of each of the new flow control valves and replaces the single stage restriction orifices in these lines with multistage orifices.

This EP is classified non-safety related, since the Condensate Pumps Minimum Recirculation lines, where this modification will be implemented, does not perform any safety function.

The safety analysis has correctly concluded that no unreviewed safety concern exist and no changes to the Technical Specifications are required as a result of this modification. Therefore, prior NRC approval for the implementation of this modification is not required.

This EP has no impact on plant safety and operation.

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

This EP is for the replacement of the existing 4 inch Condensate Pump Minimum Recirculation flow control valves with 8 inch flow control valves and for replacing and relocating the existing restriction orifices in the minimum recirculation lines. It also provides the installation of an 8 inch manually operated isolation gate valve upstream of each of the, new flow control valves.

The portion of the Condensate System where this modification will be implemented does not perform any safety function. Accordingly, components in that portion of the Condensate System are classified non-safety class, Quality Group D; therefore this modification is not safety related.

PCM 149-285 Based on the above, this Engineering Package does not constitute an unreviewed safety question and the following are the basis for this

)ustification'.

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The portions of the Condensate Systems where this modification vill be implemented are not used in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.

ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. The components involved in this modification have no safety related function and no changes have been made to

'he operational design of the system.

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

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

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

PCM 150-285 CONDENSATE PUMP MINI-RECIRC PIPING ABSTRACT The 2B condensate pump mini recirculation line was found'etached from the condenser at the nozzle weldment following the plant trip which occured in December 1984.

Examination of the restraint system for the mini recirc lines revealed that the condensate pump mini recirculation lines are supported for dead weight conditions only and that the condenser nozzles are not protected against the dynamic effects of vibration.

This Engineering Design Package provides engineering and design for additional vibration restraints in order to control the vibrations in the condensate pump minimum recirculation lines and to prevent future weld failure at the condenser nozzles.

The condensate system performs no safety related function. Accordingly, the system and its components, including pipe supports/restraints, are classified as non-nuclear safety related, quality group D and non-seismic.

This PCH does'not constitute an unreviewed safety question and enhances th e existing condensate pump system. The addition of vibration restraints to the condensate pump mini recirculation lines provid e s additional protection for condenser nozzle and does not affect any safety related equipment.

The implementation of this PC/M does not require a change to the plant technical specifications ADDENDUM 1 Supplement 1 provides designs for all support/restraint related additions and modifications needed to.address all the changes in piping which includes routing change, removal, addition/replacement of new-valves, etc, being proposed in. PCM 149-285.

This supplement does not affect the safety evaluation that was performed for Rev. 0 of this PC/M and does not require a change to plant technica ical specifications.

This supplement has no affect on plant safety or operation.

PCM 150-285 Safet Evaluation With respect to Title 10 of the Code of Federal Regulations, Part

'50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluted in the safety'nalysis report may be increased; or (ii) if a possibil'ity for an accident or malfunction of a different type than any evaluted 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:

i Section 10.4.7 of the FSAR states that the condensate system is non-safety and non-seismic. The condensate system neither initiates nor mitigates any of the accidents analyzed in the FSAR, therefore

'his PC/M is non-safety related. The additional restraints provided in this modification will control vibration of the line and reduce the dynamic effect of vibration on the condenser nozzle without adversely affecting thermal flexibility. The pipe stresses are within the limits allowed in ANSI B 31.1 "Power Piping". Therefore, the implementation of this PC/M does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.

ii The pipe stresses have remained within the code allowable-attachments to the pipe, which could affect the limits. Integral pressure boundary of the piping, are not used for this modification.

This modification does not create any possibility for an accident or malfunction of a different type than evaluated previously in the Safety Analysis Report (SAR) iii Since the probability of failure induced by vibration is reduced, there is no decrease in the margin of safety at the condenser nozzle as calculated in the original design or as defined in the bases for technical specifications.

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

PCM 163-285 REACTOR HEAD VENT LINE RESTRAINT MODIFICATION INTRODUCTION Restraint RC-98-R2, which is welded to the Reactor Coolant Gas Vent piping, has to be removed with the pipe during each refueling outage.

This restraint also'nterferes with the temporary Reactor Head shielding curtain. Due to the size and weight of the restraint, the removal requires special rigging and handling to prevent bending of the small vent line. In its present configuration the restraint is removable only from the inside of the reactor cooling shroud.

SYSTEM DESCRIPTION The modification proposed on BCS 163-285.3000, when implemented, will facilitate easy removal of restraint RC-98-R2 from outside the Reactor Cooling Shroud and will also reduce the size and weight of the restraint being removed. This is accomplished by reversing the bolts which attach the restraint to the shroud and by providing a flange type connection near the pipe-end of the restraint.

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

, safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in

. the basis for any technical specification is reduced.

This modification does not involve an unreviewed safety question and the following provides the basis'or this conclusion:

This modification will facilitate easy removal of restraint RC-98-R2 from outside the Reactor Cooling Shroud and will reduce the size and weight of the restraint, which currently requires special ragging and handling for removal. This is accomplished by reversing the bolts, which'ttach the restraint to the shroud and by providing a flange type connection near the pipe-e'nd of the restaint.

The addition of a flange type connection to the restraint and reversing the bolts does not alter the original design configuration/furiction nor does it reduce the safety factor calculated during the original design.

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

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

PCM 196-285 ANALOG DISPLAY SYSTEM GRAPHIC DISPLAY SPARES ABSTRACT This engineering design package covers the replacement of the ADS Video display monitor in the RTG Board Section 204 with Ramtek Model GM-721. The existing video display monitor in the RTGB is a Conrac Model 5211 which is no longer manufactured. The Analog Display Syst: em monitors the vertical positions and movements of the 91 Control Element Assemblies (CEA's),

utilizing the signals from reed switch position transmitters. The CRT in the Control Room provides the operator with one of the two continuous video graphic displays for the CEA positions. The CEA position system is Non-Safety Related (see FSAR Section 7.5.1). However, the associated mounting assembly in the RTGB must be seismically qualified, mandating this PCM to be classified as "Quality Related".

This item does not require revision to the plant technical specifications, nor does it meet the criteria for an unreviewed safety question. Therefore, pursuant to 10CFR50.59 this modification can be initiated without prior commission approval.

SUPPLEMENT 1 This EP revision provides for changes to the seismic CRT housing which will be mounted in the Reactor Turbine Generator Board, and which will contain both the ADS CRT monitor and one ERDADS/SAS CRT monitor. The CRT housing changes are necessary to allow the ERDADS/SAS CRT monitor to fit into the housing.

This item does not require revision to the Plant Technical Specifications, nor does it meet the criteria for an unreviewed safety question. Therefore, pursuant to 10CFR50.59 this modification can be initiated without prior commission approval.

PCM 196-285 SAFETY EVALUATION Pith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or if (iii) the margin of safety as defined in the basis for any technical specification is reduced.

The modification described in this PC/M replaces existing CRT monitor associated with the Analog Display System. The vertical position and movement of the 91 Control Element Assemblies (CEA's) are graphically displayed on the CRT. A CEA backup display panel associated with the ADS is also available for operator's use. No modification to the system is initiated by this PC/M since it utilizes a one for one CRT replacement.

. The failure of this component to function would not affect the safe shutdown of the unit since it is not required to shutdown the reactor, cool the core, or cool another safety system in the reactor containment (after an accident); nor is it part of any system that reduces radioactive release during 'an accident. The housing is required to withstand loadings induced by the design basis earthquake. Therefore, this PC/M is classified "Quality Related".

The modifications to the RTGB-204 is anlayzed as to maintain the seismic integrity of the equipment.

\

The implementation of this PC/M does not require a chang'e of the plant specifications.

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

PCM 203-285 CCM BACKFLUSH STRAINER DRAIN Abstract This engineering design package (EDP) modifies the CCW Strainer Backflush Drain piping. ,Existing cast iron and fiberglass drain piping, which is route d to t h e CCM su mp will be replaced with stainless steel piping which ties, into t h e ICW d isc scharge line. '

This vill eliminate the h flooding pro em in thee CCM p'l it area, which is causing corrosion o ear tthee floor.

oor.

structural steel and piping supports mounted on or near Th EDP classified as nuclear safety related since it modifies a safety related system. The safety evaluation has shown own that this-EDP does not constitute an unreviewed safety question.

This EDP has no impact on plant safety and operation.

SAFETY EVALUATION Safety Anal sis With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a propose d cchange an shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or thee con conse q uences of an e y p reviously accident or malfunction of equipment important to sa fet evaluatea in. the safety analysis report may be increased; or (ii) if a possi'b'1't i ity foror aan accident or malfunction of a different type than any evaluated previously in the safety analysis report m y (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The modification included in this engineering design package do not involve an unreviewed safety question because:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the connection of a CCW strainer ba'ckflush drain line to the ICM discharge line will have no effect on the safety performance of the ICW or CCW systems or any of their components.

(ii) There is no possiblity for an accident or malfunction of a different type than any previously evaluted s'ince no changes have been made to the operational design of the CCW strainer backflush system.

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

Implementation of this engineering design package does not require a change to the plant technical specifications.

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

PCM 208-285 RTD AND THERMOWELL REPLACEMENT FOR REACTOR COOLANT SYSTEM ABSTRACT This Engineering Package (EP) provides for the removal and replacement of the existing resistance temperature devices (RTDs) and thermowells with spring loaded tapered RTDs and tapered thermowells. The purpose of this change is to alleviate difficulties experienced in the maintenance of the existing equipment (e.g. prying of RTD from thermowell and thermowell damage sufficient to necessitate its replacement).

This EP is classified as Nuclear Safety Related since it involves the reactor coolant pressure boundary and components which are part of the Reactor Protection System (RPS). This EP also involves components (RTDs) which are identified as post accident monitoring instrumentation (PAMI) and provide control room indication and recording. A review of the changes to be implemented by this PCM was performed against the requirements of 10CFR50.59.

As, indicated in the Safety Evaluation (Section 3.0), this PCM does not involve an-unreviewed safety question, nor does it require a revision to the plant Technical Specifications or the proposed Revised Plant Technical Specifications. This modification will have no effect on plant safety or operation. Prior Commission approval is not required for the implementation of this PCH.

SAFETY EVAGJATION 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) or if the probability malfunction of of equipment occurrence important or the consequences of an to safety previously accident evaluated in the Safety Analysis Report may be increased, or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created, or if (iii) the margin of safety as defined. in the basis for any technical specification is reduced.

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

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the replacement RTD/thermowells meet the requirements of the FSAR. The replacement RTDs provide the same input as the existing, equipme~t and do not alter the function of any of the components, cabinets, or systems that receive RTD input.

PCM 208-285 There is no possibility for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operational design of any control circuits or associated systems.

Installation of the RTD/thermowell assemblies are c'ontrolled by site procedures and the FPL Welding Control Manual. Welding of the thermowells to the sleeve is to be in accordance with FPL General Welding Standard for Nuclear Piping and Piping Components, Rev 1. The RTD/thermowell assemblies have been sub)ected to non&estructive examination (NDE) per the codes and standards listed in Section 2.2.2 of this EP. By adhering to these codes and standards in the implemention of this PCM, there is n'o possibility for an accident or malfunction different than any previously evaluated involving the Primary Coolant Pressure Boundary. The replacement RTD/thermowells have approximately the same weight as those being replaced. Therefore the insignificant change in weight does not have any affect on the pipe stress and/or the support restraints.

{iii) This modification does not change the margin of safety as defined in the basis for any technical specification. This has been determined based on the fact that the replacement items meet the sane Technical Specification limitations as. the existing items and the fact that the'esign limitations of the reactor coolant-pressure boundary, as delineated in FSAR Section 5.1, are maintained with the implementation of this PCM.

Since this EP affects equipment that is identified as Nuclear Safety Related (the RTDs are class 1E; the thermowells are ASME Class 1), this package is considered Nuclear Safety Related.

No hydrostatic pressure test is required after the installation of the RTD/thermowell assemblies per ASME Section XI, Paragraph IWA-4400. An in-service leak test will be performed to ascertain that the imple-mentation of this PCM has met the requirement of no allowable leakage of reactor coolant.

The only effect this EP has on cables essential to safe reactor shutdown and alternate shutdown components is in the disconnection of the

'xisting RTDs and the reconnection of the new RTDs. There are no other changes to equipment which involves 10CFR50 Appendix "R" fire protection (see Attachment 7.1). Thus, the proposed design of this package is in compliance with the applicable codes and FSAR requirements for fire protection equipment.

Implementation of this PCM does not require a change to the Plant Technical Specifications and may'e implemented without prior Commission approval'he foregoing constitutes, per 10CFR 50,59 (b), the written safety evaluation which provides the bases that this change does not involve an unieviewed safety question and prior Commission approval for the implementation of this PCM is not required.

PCM 211-285 MAKEUP AIR FOR CONTAINMENT HYDRO PURGE SYSTEM TEMPORARY VALVE MODIFICAITON INTRODUCTION The continuous containment purge/hydrogen purge system is designed to:

provide a sufficiently low concentration of radionuclides in the containment atmosphere, relieving of containment pressure buildup, the capability of ensuring that the containment source term contribution to the annual average off-site doses are maintained as low as is reasonably achievable and hydrogen removal capability.

This system provides a direct air path between the containment atmosphere and the outside. Leak rate testing is required for penetration. During the leak rate testing, it was found that the isolation valve (FCV-25-36) was leaking. This PCM is for the temporary modification'to remedy these leaks.

SYSTEM DESCRIPTION The system consists of a purge make-up penetration line and exhaust penetration line. These containment penetrations provide a direct air path between the containment atmosphere and the outside. The isolation valves have been provided for both air path and isolation.

During the leak rate test it was determined that valve FCV-25<<36 was leaking. In order to complete the test successfully the valve was modified by bolting a 1" thick plate to one end of the valve. This modification is considered temporary until the valve is replaced or repaired and the system is restored to its normal operating conditions.

The use of a 1" plate is acceptable for a blind flange for FCV 25-36 for the following reasons:

The required thickness for a 1508 flange for this size is 1-1/8". The required pressure for this application is 44 psi versus an allowable pressure of 150 psi for 1-1/8". By engineering judgement the reduction of 1/8" will not affect the ability of this flange to withstand 44 psi. This plate will meet all the other requirements, including documentation, for ASME Section III Class 2 material.

PCM 211-285 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'he 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) a if a possibility for an accident or malfunction of 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.

'Ihe use of a blind flange does n'ot create any new accident since the valve is used as Containment Isolation and its safety function is to isolate the containment. "The blind flange will perform this function passively. The blind flange will be subjected to the same testing requirements as the valves in that system.

This system is used as a non-safety backup to the redundant safety related hydrogen recombiners which maintain the hydrogen concentration below 4Z after any accident. This system is non-safety and does not need to meet single failure criteria since the hydrogen recombiners are the design basis for the plant. This system is not considered in the design basis, therefore the loss of function of this system does not affect the ability of the plant to mitigate an accident.

The only Tech Specs involved are Containment Isolation and Containment Pressure. This modification increases the margin of safety of the.

Containment Isolation Valve Technical Specification since it replaces an active device with a passive device. With regards to the Containment Pressure (normal) Technical Specification, this Technical Sp'ecification is unaffected since the mini purge exhaust line will still function to reduce pressure inside containment.

The other design basis of'the Continuous Containment Hydrogen Purge System will be affected. However, their impact will be in the form of longer duration of plant outages and in no way impair the safe operation of the plant. 'Ihe longer plant outages will be the result of the inability to purge the containment during operation. Thus purging must be accomplished during shutdown.:

Therefore this modification does not constitute any unreviewed safety question.

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

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

PCM 008-286 ATMOSPHERIC DUMP VALVES MOTOR OPERATED VALVE INDICATION ABSTiQCT This Engineering Package covers modification to the control circuitry of motor operated valves (i~GV) i~V-08-18 A/B and bV-08-19 A/B. This involves the utili ation of the motor operator's 412 limit switch to provide a limit switch controlled back-up to the primary method of closure-torque switch control.

These valves function as atmospheric steam dump valves (ASDV) which are used to relieve/control system pressure. Th'e ASDV's are part of the main steam supply system. The FSAR, in section 10.5.2 classifies these valves as safety related. This modification is considered safety related and deemed not to constitute an unreviewed safety question.

SAFETY EVALUATION The function of the atmospheric steam dump valve system is to provide reactor coolant system heat removal capability. The modification is directed only to valve position indication circuitry to provide more reliable position indication while eliminating rotor adjustment difficulty and does not affect valve power circuits.

The proposed design ensures that ASDV's will be closed with the torque switch with a limit back-up closure switch. The change proposed to the ASD's bGV does not introduce any change to the functional configuration of the system required to meet the safety related design function. This design does not alter the original requirements

PCM 008-286 specified in the St. Lucie Unit 2 FSM, Section 10.3, which specified that dump valves are designed to withstand Design Basis Earthquake (DBE) .loads r

simultaneously with the effects of the discharge thrust of steam passing'hrough them, the effect of dead weight, and the effects of internal pressure loads. The FSM required that these valves shall be po>>ered from a (DC) onsite power source. The proposed modification does not change this feature (FSAR, Section 10.3.3).

The use of switch 412 on rotor P3 of the existing KGV will have no adverse affect on nuclear safety since this modification will not adversely affect the limit backup and torque closure limit circuitry. This modification eliminates the adjustment difficulties of the limit back-up and position indication s>>i:tches, thus enhancing the circuit overall operability requirements.

All modifications will be made >>ithin the motor operated valves. The circuit modification is strictly a hardware modification to exchange the rotor contact presently used for spare contacts on a spare rotor, to aid in contact adjustment. The modified circuit will function as the circuit previous to the modification, and in that no external circuit or cable routing will be required, no Appendix R analysis will be affected.

10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical specification is not required. The design change does not alter equipment circuitry used to mitigate accidents. The change

PCM 008-286 allows proper valve torque closure with limit back-up and appropriate indication. Therefore, the probability of occurrence of analyzed accidents remains unaffected. The margin of'safety as defined in the Technical Specification has not been reduced because the atmospheric steam dump valves system operability has not been affected. The capability remains to provide reactor coolant system heat removal and >>ithstand design basis earthquake loads simultaneously with the effects, of the discharge thrust of steam passing through them, the effect of dead weight, and the effects of internal pressure loads.

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

In conclusion, the change proposed in this desi.'n package is acceptable from the standpoint of nuclear safety; does not involve an unreviewed safety question; does not require NRC approval and issuance of Technical Specification changes prior to implementation.

PCM 011-286 DIESEL GENERATOR GOVERNOR POWER SUPPLY INTRODUCTION A Lambda power supply was installed under PCM 389-283 in order to increase reliability of the emergency diesel generators. This power supply has been unable to maintain the necessary regulation required by the governor. At present, this power supply has 'been bypassed.

As presently installed, .the governor power supply is derived'rom plant 125V DC power. This condition is similiar as prior to the implementation of PCM 399-283. It has been demonstrated during testing that any disturbance on the plant 125V DC system will cause erratic governor operation. This erratic operation produces unstable generator electrical output.

The DC governor power supplies installed in St Lucie Unit 1, are Woodward power supplies and have been performing satisfactorily. These power supplies were install'ed under PCM 372-183. Based on this performance, a modification replacing the Unit 2 Lambda power supply with a Woodward power supply will be implemented by PCM 011-286 supplement l.

SYSTEM DESCRIPTION The purpose of the modifications performed by this PCM is to increase the stability and reliability of the Emergency Diesel Generator by

<<stalling Woodward power supplies Part No 9903%34 which is similar to that, used in St Lucie Unit 1 Diesel Generators.

The existing plant 125V DC power will be used to start the diesel and accelerate it to,full rated speed. At this speed (850 RPM), the speed witch will actuate to disconnect the 125V DC plant power and connect

<< new power supply into the governor circuit. Thus, the governor,

<ich is sensitive to power supply disturbances, will be isolated from

<< plant system.

Thee new power supply circuit is also being modified to include contacts erich will maintain the voltage regulator in a de-energized state during

~rmal plant operation.

PCM 011-286 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; (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 maybe increased; or (ii) if a possiblity 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 possibility of occurence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis is not increased by this PCM supplement. Since the diesel generators will be tested for acceptance in accordance with the Plant Technical specifications for periodic testing 4.8.1. 1.2.3, which verify DG performance in a simulated loss of offsite power.

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

The installation of the Woodward power supply will minimize the erratic operation of the DG governor due to any disturbances in the 125 VDC power supply.

The Woodward power supply'has been seismically and environmentally qualified. The environmental qualification evaluation is attached to this PCM supplement.

cd The margin of safety as defined in the basis of the technical specification is not reduced since as previously discussed the operability of the Diesel Generator will be confirmed by the periodic testing discussed above.

The power supply mounting duplicates the mounting used for seismic testing performed by NTS Acton Labs.

The type of power supply, Woodward Model No 9903034 is a vendor received power supply to be used with 2301 series electrical governor system.

There is no change on the existing technical specification due to the implementation of this PCM supplement.

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 supplement.

PCM 014"286 TARGET ROCK VALVES - STEM ASSEMBLY UPGRADE ABSTRACT The Target Rock Valves described herein are model 75C-002, 2" motor operated globe isolation/throttling valves which are installed in the Safety Injection System. Combustion Engineering infobulletin 84-10 provided information regarding the potential of the stem assembly parts galling under long term throttling duty. The Infobulletin discussed possible solutions to the potential problem, which consisted of upgrading the materials of the stem assembly subcomponents.

This PC/M provides the information required to upgrade the .eight affected Target-Rock valves on St. Lucie Unit 2 with manufacturer-redesigned stem assemblies constructed of galling resistant materials The modification described herein is classified as Nuclear Safety Related. No unreviewed safety questions exist as defined by 10 CFR 50.59,. therefore commission approval is not required prior to implementation.

Safety Evaluation This modification involves only the upgrade of materials for the stem assemblies of'he High Pressure Safety Injection pump motor operated isolation valves.

10 CFR 50.59 allows changes to a facility described in the FSAR without prior NRC approval if an unreviewed safety question does not exist and if a change to technical specifications is not involved. The following arguments demonstrate that an'nreviewed safety question does not exist:

i) The probability of occurrence of a design basis accident is not increased since this modification does not alter existing Safety Injection System operation, design parameters, and since no new equipment is added.

Additionally, the new stem assemblies are to be designed, fabricated and inspected to the same code criteria as the existing stems.

ii) The consequences of, accidents previously evaluated in the FSAR are not made more serious for the reason provided in Paragraph (1) above.

iii) The possibility of an accident of a different type than any previously postulated in the FSAR is not created for the same reason provided in Paragraph i above.

iv) The margin of safety as defined in the basis for any technical specification is not reduced since Safety Injection System parameters will not be affected by the material change, and sine'e the manufacturers design will meet the original specification requirements.

Since the above arguments demonstrate that an unreviewed safety question does not exist, and since no changes to technical specifications are involved, the modifications to the affected safety injection system isolation valves do not re q uire prior NRC approval.

0168L

PCM 024-286 RDF-RTD TEMPERATURE TRANSMITTER REPLACEMENT ABSTRACT This engineering package covers the replacement of four (4) RdF tempera'ture transmitters. The presently installed transmitters are no longer being manufactured and suitable replacements are being provided for maintenance and replacement capability'his engineexing design package is considered Quality Related since the replacement temperature devices are being seismically mounted on the RTGB. The instrumentation loops associated with the transmitters are not used to mitigate incidents and accidents and ,therefore, as per FUSAR Chapter 15, this PC/M is not considered to be Safety Related.

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

The modification described in this PC/M is associated loops used for reactor xeactivity control, input to with'nstrumentation date processor and Safety Assessment System, control room indication/recording and annunciation. As per FUSAR Section 7.7, this instrumentation and control system is not essential for the safety of the plant.

The new temperature tran: Mtters are being seismically mounted to RTGB-203. These new transmitters have been addressed by Acton Labs as to the seismic impact o. the RTGB. As per Acton Labs letter Att. 7.3 the replacement transmit. ers will have no impact on the equipment seismic qualification ar the dynamic characteristic of the equipment will not be affected.

This modification is "a o;.:e for one replacement of temperature transmitters only and does not alter or change the original transmitter loop arrangement, as such the implementation of this PC/M does not require a change to the plant specifications.

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

PCM 029-286 CCV PUMP BEARING MATERIAL CHANGE ABSTRACT This engineering package covers replacement of the existing east iron journal bearing shells on the component cooling water pumps 2A, 2B R 2C with shells made of carbon steel. The existing cast iron shells are no longer available and the manufacturer's replacement part is the carbon steel shelL As addressed in the Safety Evaluation, this modification is considered nuclear safety related.

Based on the 10 CFR 50.59 review, it has been demonstrated that this change does not involve an unreviewed safety question, and the change will not affect plant safety. Additionally, no change is required to the Technical Specifications. Accordingly, prior NRC approval is not required for implementation of this design.

SAFETY EVALUATION The Unit 1 Component Cooling Water pumps are nuclear safety related and are classified as ASME Section 111, Class 3 Quality Group C components.

They are required to provide a heat sink for safety related components associated with reactor decay heat removal for safe shutdown or IOCA conditions. The journal bearing shell material change affects both journal bearings in the 2A, 2B and 2C pumps.

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

However, failure of a single pump has been previously evaluated and has been accounted for in the Component Cooling Water System design bases as identified in the FSAR. Measures exist to ensure adequate decay heat removal for safe shutdown or LOCA conditions should a single pump faiL Since the new shell parts are internal to the bearing housing, failure of an additional component cooling water pump simultaneous to the first pump failure is not possible based on single failure criteria. In addition, since the new shell material is functionally equal or better than the existing cast iron material, the probability of pump failure remains unchanged.

Based on the above evaluation and information provided in the Design Analysis, it can be demonstrated that an unreviewed safety question as defined by 10 CFR 50.59 is not created. Since no other accident beyond what has been previously addressed in the FSAR has been identified and no other safety related equipment or components are affected as addressed in the failure modes analysis, the probabilty of occurence analyzed accidents has not been increased. The replacement is equal or better to the equipment replaced. No new accidents or malfunctions are introduced as a result of this design change. Additionaoy, the margin of safety as defined in the Technical Specifications has not been reduced and no Technical Specification changes are required. Therefore an unreviewed safety question does not exist.

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

PCM 038-286 PCB TRANSFORMER REPLACEMENT ABSTRACT Due to environmental concerns a'ttendant to polychlorinated biphenyl (PCB) cooling/insulating liquids, "all transformers filled with a PCB liquid are being eliminated from FP&L's system. The neutral grounding transformer for the main turbine generator is filled with 38 gallons of PCB cooling/insulating liquid. The neutral grounding transformer for each emergency diesel generator (EDG 2A and EDG 2B) is filled with 15 gallons of PCB cooling/insulating liquid. This Engineering Package provides for replacement of the three (3) generator neutral grounding transformers with equivalent silicone liquid-filled or dry-type transformers.

The main generator'eutral grounding transformer does not perform any nuclear safety related function, therefore its replacement is classified as nonnuclear safety related.

Due to their association with the safety related emergency diesel generators, the replacement neutral grounding transformers for the emergency diesel generators are classified as nuclear safety related.-

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

SUPPLEMFNT 1 This supplement incorporates vendor and installation drawings, seismic report, associated engineering design calculation certification, design analysis and safety evaluation, design and. safety verification and Total Equipment Data Base (TEDB) sheets.

A11 "On Hold" and "Later" statements affecting the engineering package sections above under Revision 0 are being removed by this supplement.

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

SUPPLEMENT 2 This supplement incorporates Change Request Notice Nos. 038-286.343 and 038"286.386 which modify the emergency diesel generator neutral grounding transformers'erminal numbers and the wiring to the coil of the ground protection relay, respectively. It also addresses wind loading, electrical clearances and Justification for using the 600 volt rated jumper for the emergency diesel generator neutral grounding transformers. In addition, corrections were made to the drawing list to reflect Unit 2 drawing numbers and to Section 1.3.1.2 to reflect revision 1 of Ebasco Specification FLO"E"002..

The safety evaluation has been revised to incorporate the wiring modification. This revision, however, has not altered the previous conclusion which indicates that the modifications presented by this Engineering Package do not constitute an unreviewed safety question, do not require any changes to the plant Technical Specification and do not require prior Commission approval for the implementation of this PC/M.,

PCM 038-286 SAFETY EVALUATION The replacement neutral'rounding .transformers for EDG 2A and 2B are located inside their respective control cabinets. The replacement transformers have been seismically and environmentally qualified. A flexible connection's provided at the Hl terminal lug of the replacement transformers to minimize stress on the lug under seismic conditions. Ground detection relay coil wiring has been revised in

,accordance with General Electric Power systems Management Department instructions GEH-1814B. This does not affect the operability of the relay for diesel generator ground detection since the circuit has not been functionally modified. The existing seismic qualification of the control cabinets has not been affected by the replacement'transformers.

Based on the preceeding, the following conclusions I

can be made:

(i) The probability of occurrence or the consequences of an accident or, malfunction of equipment important to safety previously evaluated in,.the FSAR will not be increased because the existing transformers are being .-replaced on a one-for-one basis by transformers that are equivalent in form, fit and function.

(ii) This modification does not change th'e operation of the Main Generator or the Emergency Diesel Generators, therefore, there is no possibility that an accident or malfunction of a different type than any evaluated in the FSAR may be created.

(iii) The replacement neutral grounding transformers are equivalent in form, fit and function to the existing transformers and perform no safety related functions. Therefore, -this. modification does not reduce the margin of safety as defined in the bases for any technical specification.

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

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

PCM 042-286 QUENCH TANK PMW ISOLATION VALVE REPLACEMENT INTRODUCTION The quench tank is part of the pressurizer pressure control system..

Excessive pressure in the pressurizer is relieved by discharging steam to the quench tank. The steam is condensed in the quench tank by partially filling it with primary makeup water.

Water is supplied to the hose stations inside containment by the primary makeup water system. A 1 inch solenoid valve isolates the PMW to the quench tank from the hose station supply line, The small size of this valve prevents timely makeup of water addition to the tank. This valve is being replaced with a larger valve to 'eliminate this problem.

SYSTEM DESCRIPTION The quench tank and pressurizer relief discharge system are described in Section 5.4.11 of the FSAR. This modification vill replace the 1 inch s"lenoid operated valve (SE-15-2) in line 2-RC-507, with a 2 inch a'r operated ball valve. The new valve and operator are lighter than the original valve, therefore no additional support/restraints will be required. The air for the operator will be supplied by tying into the air supply for valve V3632. A piston type, spring return operator will be used which will close the valve on loss of ins t rumen t a ir .

PCM 042-286 SAFETY ANALYSES 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 anslysis report may be increased; or (ii) if a. possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

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

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evluated is not increased since the quench tank PMW isolation valve is'on-safety related and this modification will have no effect on equipment performing a safety function. This modification will decrease the probability of overpressurizing the quench tank, since the new valve will supply makeup water to the quench tank with a substantially greater flow rate ~

ii There is no possibility for an accident or malfunction of a different type than any previously evaluated since the quench tank PMW isolation valve has no safety function and no changes have been made to the operational design of the system.

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

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

~

technical specifications.

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

PCM 062-286 ADDITION OF FLANGE TO PENETRATION P-50 ABSTRACT This Engineering package provides for the replacement of the welded pipe cap on

-penetration P-50 with a blind. flange. The pipe cap is on the outboard side of the concrete shield building. This modification is nuclear safety related, because it deals with a change to the structural loads of the Containment Shield Building which is a Seismic Category I structure. This modification does not affect the Containment pressure boundary. Since this modification is not to a piping system, a Quality Level designation is not.applicable. The additional loads are small and do not change the seismic classification of the penetration -or Containment. This modification does not constitute an unreviewed safety question as defined by 10 CFR 50.59. As a result of this modification, penetration P-50 can be readily used to support outage related tasks such as eddy.

current testing.

SAFETY EVALUATION The subject modification provides for replacement of the pipe cap on the outboard side of Containment Shield Building penetration P-50 with a weld-neck flange, gasket and blind flange. As defined in Section 3 of the FSAR, the Containment Shield Building is Seismic Category L This modification is considered nuclear safety related because it alters a Seismic Category I structure. Since this modification does not change a piping system, a Quality Level designation is not applicable.

Per the attached Ebasco letter, this modification does not alter the seismic qualification of the Containment Shield Building or penetration P-50. Also, the containment leak rate is not affected because the outboard cap does not form part of the containment pressure boundary. Even so, the seal integrity of the flange replacing the cap is verified by NDE testing and Quality Control verification of flange bolt torquing.

No active components or other safety related systems and/or compoents are impacted by this modification. Accidents considered in Section 6 of the FSAR bound any abnormal condition that could be caused by failuie of the new penetration flange. No Technical Specification is impacted by replacing the outboard shield building penetration cap with a flange.

Based on the. above evaluation, an unreviewed safety question as defined by 10CFR 50.59 does not exist: (i.e., the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not- increased. The possibility of an accident or malfunction of a different type than any evaluated in the FSAR is not changed. The margin of safety as defined in the Technical Specifications is not reduced.) The preceding argument, coupled with the fact that a Technical Specification is not required, leads to the conclusion that prior NRC approval is not required to implement this modification.

PCM 064-286 RCP INSULATION REPLACEMENT ABSTRACT Th1s Engineering Package provides details for replacing the existing blanket type 1nsulation on the Reactor Coolant Pumps with reflective type insulation developed by Diamond Power Specialty Company.

The insulation around the pumps is Quality Related because it must

• remain in place at all times during plant operation and it will be seismically supported. In addition, the metal reflective design has accounted for effects on the containment recirculation system and sump screen blockage. The insulation design has accounted for potential impact on overall containment heat load to insure that containment ambient temperatures will not increase as a result of this modification.

The safety evaluation has shown that'his EP does not constitute any unreviewed safety question, has no adverse effect on plant safety nor does it require a Technical Specification change.

Implementation of this modification is acceptable w1thout prior Commission approval.

SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety apalysis report may be increased; or (ii) if a poss ibility for an accident or malfunction of a different type than any e valuated previously in the safety analysis report may be create (iii) if d'r the margin of safety as defined in the bases for any Technica cal Specification is reduced.

This EP involves the replacement of the blanket type insulation around the reactor coolant pumps with stainless steel reflective type insulation. As discussed in the Design Bases and Design Analysis, this modification is considered Quality related due to t?ie seismic design considerations, the potent1al impact on the containment recirculation s y s tern and sump design and the potential impact on containment ambient temperatures. Based on the failure modes evaluation the insulat io on dd d b this modification will not adversely effect any safety related equipment or components. Based on this and information provided in the Design Analysis; this modification does not involve an unreviewed safety question because:

PCM 064-286 The pr'obability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the stainless steel encapsulated blanket insulation is being replaced by stainless steel reflective insulation. Both. types of, insulation presently exist inside

• containment. Since the stainless steel reflective insulation is similar to other reflective insulation. used, is equal to or better in insulating quality .to that which it replaces and is seismically supported, it will have no effect on equipment or functions important to safety.

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis is not created. The replacement of one type of insulation with another type of insulation, both acceptable for use inside containment, does not change any existing Design Criteria, Operating Procedure or Technical Specification.

(iii) This modification does not affect the basis for any Technical Specification and therefore does not reduce the margin of safety as defined in the basis for any Technical Specification.

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

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

PCM 069-286 TORQUE SEATING ATMOSPHERIC DUMP VALVES This Engineering'Package covers a modification to the Unit 2 Valves (MV-08-18A, MV-08-18B, MV-08-19A and MV-08-1 Atmospheric Dump 9B). 'This modification changes the control circuit for these valves to allow the closing limited by the torque switch instead of by the limit switch as direction to be This PCM is classified as Nuclear Safety Related, and does presently designed.

not constitute an unreviewed safety question.

f v i The atmospheric dump valves (ADV's) provide a means of decay heat removal and cooldown capability when the MSlV's are closed. The ADV's can also be modulated to control primary plant temperature during startup and shutdown.

The valve manufacturer, CCI, was consulted and concurs that these valves may be torque seated. The proposed change is therefore acceptable for all four ADV valves. The proposed design, although different than the original, does not change the operation of the atmospheric dump system as discussed in the PSL Unit 2 FSAR Sections 5A, 6.3, and 10.3. Since the ability of the ADV's to close has not been adversely affected by this change, the probability of occurrence or the c o n se q uences of a design basis accident or malfunction of equipment important

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to safety as discussed in the FSAR Chapter 15 has not been increased. Previous i ly analyzed failure modes for the ADV's remain valid, and thus the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR Chapter 15 is not created. PSL 2 Technical Specificiations Section 3.7.1.7 provides the "Limiting Condition for Operation" for the ADV's. The Technical Specification requirements are not changed by this modification and the margin of safety as defined in the bases for this Technical Specificiation will not be reduced. The safety evaluation thus demonstrates that an unreviewed safety question does not exist.

e chan g e ro'sed prp in this design package is acceptable from the stan d poin t o f nuclear safety; does not invo ve an and does not require NRC approval prior to <mplemen tation.

PCM 075-286 HEATER DRAIN PUMP MECHANICAL SEAL DEMINERALIZED WATER SUPPLY ABSTRACT This design package provides the necessary engineering for adding permanent piping from the demineralized water system to the Unit 2 heater. drain seals. The piping will make available to the seals the necessary back pumps'echanical up flushing water meeting the appropriate chemistry requirements. This backup flushing water is required during initial plant startup whenever the pumps sit idle.

Based on the failure modes analysis and 10 CFR 50.59 review, this modification does not impact any safety related equipment and is not relied upon for any accident prevention or mitigation. Thus it does not constitute an unreviewed safety question and is correctly classified as Non-Nuclear Safety Related.

Implementation of this modification, therefore, does not require prior NRC approval. 'I Su lement l

. This package revision provides valve drawings for valves added by this PC/M and modifies the expiration date to reflect the correct format. The scope of work specified by this Engineering Package has not been affected by this revision.

The safety classification and the safety evaluation as stated is correct and is not impacted.

SAFETY EVALUATION P

The Unit 1 Heater Drain Pumps are located in a Non-Nuclear Safety Related system and as such are not required to function during any existing analyzed accident scenario. Therefore, m'odifications to these pumps affect only Non-Nuclear Safety Related, Quality Group D equipment.

Based on the failure mode analysis, failure of the demineralized water supply

'piphg could result only in failure of the heater drain pumps. Since the piping and components are located remote from any safety related equipment or components, failure of this equipment will not inhibit operation of any safety related equipment or components.

Based on the above evaluation and information supplied in the design analysis it can be demonstrated that an unreviewed safety question as defined in 10CPR50.59 does not exist.

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

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

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

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

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

This change has no affect on any existing Technical Specifications.

PCM 087-286 MISAPPLICATION OF LIMITORQUE OPERATORS ABSTRACT This Engineering Package (EP) is for the replacement of the motors and gear trains on the following valve motor, operators:

Parts to be Valve Ta No Location I-MV-09-l FM Pump 2A Discharge Motor 6 Gear Train I~-09-2 FW Pump 2B Discharge Motor 6 Gear Train MV-09-3 FW Flow Control Station (Train A) Motor MV-09-4 FR Flow Control Station (Train.B) Motor The replacement of the existing motors with motors having lower RPM and increasing the operator gear train ratio in two of these valves is required to reduce the valve stem speed, to be within the limits

recommended by the valve operator manufacturer (Limitorque) for the type of operator (SMB) involved.

This EP is classified non-safety related since the portions of the main feedwater pump discharge piping and flow control stations where the affected valves are installed, does n'ot perform any safety function and they are in the non-safety class portion of the Main Feedwater System.

The safety analysis has correctly concluded that no unreviewed safety concern exist and no changes to the Technical Specifications are required as a result of this modification. Therefore, prior NRC approval for the implementation of this modification is not required.

This EP has no impact on plant safety and/or operation.

PCM 087-286 Safet Evaluation With respect to Title 10 .of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a:different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

This modification does'ot involve an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report (Section 15.1.2.1) is not increased. The portions of the feedwater system where this modification will be implemented are not considered in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.

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

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

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

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

PCM 091-286 CLOSE INTERCEPT VALVE -- CONTROL CIRCUIT MODIFICATION This Engineering Des1gn Package (EDP),provides for the removal of the Close

. Intercept Valve (CIV) anticipatory control circuit from the Westinghouse Digital Electro"hydraulic (DEH) turbine control system.

The original intent of the CIV anticipatory circuit was to provide a temporary closure of the Interceptor Valves in the event of a load mismatch between turbine steam flow and generated electrical output.

This particular circuit does not take into account the dynamic response of the turbine steam cycles, nor does the DEH model P-2000 contain the necessary programming software to perform the required calculations to automatically adjust the turbine governor valves to the new thermodynamic values.

These features, therefore, will, in moat cases, maintain the Interceptor.

Valves closed with a resultant trip of both the turbine and the reactor.

The CIV control circuit is a downstream eztension of the DEH overspeed control channel. System failure would not impact plant safety, since this system is neither required for safe shutdown nor does it perform any safety related functions. However the DEH Control System is required to be operable by the Technical Specifications. .Since th1s modification impacts the subject control circuit, this engineering design package shall be classified as Quality

'elated.

A review of the changes to be implemented by this PC/M was performed against the requirements of 10CFR50.59. As indicated in Section 3.0 of this PC/M, this PC/M does not involve an unreviewed safety question, nor does 1t require a revision to the techn1cal specification; therefore, pr1or Commission approval is not required for implementation of this EDP..

PCM 091-286 SAFZTr EVALUATION Mith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unrevieved safety question; (i) if the probability of occurrence of the of an accident or malfunction of equipment important to consequences safety previously evaluated in the safety analysis report may be increased; or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be creatd; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The probability of occurrence as the consequences of an accident or malfunction of equipment previously evaluated in the Safety Analysis Report is not increased by this PC/M. This modification to the CIV control circuit does not change'r alter the turbine-generator monitoring and control system.

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

The CIV control circuit is an independent function generated by the DEH control system software.

The removal of the CIV anticipatory function does not alter the of the DZH control .system. 'peration

'This modification, vhich vill remove the partial load mismatch circuit, +111 reduce the number of spurious reactor trips vhich vi11 occur should the Interceptor Valves fail to re-open.

The turbine overspeed protection channels to both the Reheater Stop valves and the Intercept valves and the mechanical overspeed protection channels are not altered by implementation of this circuit modification. Therefore, the margin of safety for turbine rupture due to the probability of turbine overspeed is not reduced.

The implementation of this PC/M does not require a change to the St Lucie Unit 2 Technical Specifications.

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

PCM 092-286 ADDITIONAL APPENDIX R FIRE SPRINKLER AND FIRE WRAP IN RAB ABSTRACT An on site inspection by NRC I~E Inspectors identified that in son fire azeas in the RAB, existing ceiling level sprinklers are obstructed by cable trays, HVAC ducts, etc. This condition does not provide adequate fire pro. ection to the conduits located below these obstructions. In order to ensure complianc" with Appendix "R" requirements and to provide adequate fire protection to the affected conduits, modifications to the e-isting sprinklez system are required.

This Engineering Design Package (EDP) provides the engineering and design for the addition of new sprinklers below obstructions, isolation ior two signal transmitters in the Hot Shutdown Control Panel and revision of the Safe Shutdown Analysis to remove protection requiremerts fzom sevezal cables.

This EDP is classified as nuclear safety re'ated since the inputs to the isolated signal transmitters aze accepted from safety class equipment and the devices (isolated signal transmitters) have to be qualified as Class IE per IEEE-Section 323 (1974). Changes to the sprink'er syste" and Safe Shutdown Analysis are considered Quality Related.

The new isolation devices are being installed into circuits that monitor pressurizer pressure and pressuri"er level and supply 'nput to the Safety Assessment System (SAS).

The Safe Shutdown Analysis is being revised to delete from the Analysis cables which have be n removed from the Essential Cable List or to change the Analysis for cables which are isolated by t ansfer switches and therefore no longer require protection. This is being done instead of installing additional sp inklezs in these areas, or upgrading the conduit wrap from one (1) houz to three (3) hour rat'ng.

The changes to be implemented by th's EDP have been reviewed and found to meet the fire pro ection requirements put for th in 10CFR 50, Appendix "R". As indicated in Section 3.0, this ZDP does not involve an unreviewed safety question, nor does it require a revision to the Technical Specification. There ore, prior Commission approval is not requ'red for implementation nf this EDP.

This EDP has no impact on plant safety and operation.

Safet Evaluation With respect to Title 10 of the Code of Federal Regulations Part 50. 59, a proposed change shall be deemed to involve an unreviewed safety ques'zion; (i) if the probability of occurrence o>> the consequence.. of an accident or malfunction of equipmenz impoztar't to s'afety prev'ously evaluated in the safety an-1vsis report m"y be inczea. ed; oz (ii) if a possibility foz an accident or malfunction of a different tyne than any evaluated previously in the safety ana3ysis report may be created; or (iii) if the =argin of safety as defined in the bas's for any Technical Specificatton is reduced'

PCH 092-286 The modification 'included in this Engineering Design Package does not involve an unre:iewed safety quest'cn. T.".c following are the bas " fcr the gustificationi a) Installation of Isolation Transmitter This modification to pressurizer pressure and pressurizer 1 vcl instrumentation does not chan"e or alter the 'pressurizer instrumentation system or the alternate shutdown procedure, The existin~ pressurizer pressure and pressurizer level instrumentation loops are 1"elated from the cable spread room bv way of a 57. ohm resistor providing isolation for cable shorts and discontinuous circuits (open caMe) in the Safety Assessment System (SAS) isolation cabinet. The installation of the s'gnal isol"tors at the 'hot shutdown panel i>ill prevent a=y potential cable to c"b'e failure from propagating to the pressurizer level and pressure signals at the hot shutdown panel, As shown in Attachment 7.4, rhe RIS isolated signa'ransmitte s, installed in the hot Shutdown Control Panel, are located in a mild environ e..t, Therefore, Environmental C<ualification p" .10CFR50.49 is not recuir d. The addition of t'ne transmitters will nor.

adversely affect t?ie seismic qualificat'on 'of the, HSCP. Tne transmitters rhemselves are seismically qualified in accordance with IZiE 344-1975.

The 0-10 KDC 'ut/output as- provided bv the Rochester Isolate" Signal trasnmitter 's compatible with ez'sting loop r quirement limits and constraints and requires no urther modificatio"s as rhere are no other interface points in'olved in this insrrumentation loop..

This modification ensures that conformance to t) e separation requirements of 10CPR50 Appendiz R is met as committed, The probabi'ity of occurrence of an accident or malfunct1on of equipment and systems prev'ously evaluated in rhe PSAR has not been increased by this change. Compensatory measures in the inrerim are identical tc those invoked for other fire prorection modifications',

1 e

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"p a rov ng fire watch (hourly) and operabiliry o the applicable fire derectors are esrablished ~

The possibility or cn "cc1dent or the nalfunction of a different type than described in t?.e sa'ety analysis is reduced by th1s change since isolation superior to the pr vious design is provided by this PEA.

PCM 092-286 The posslbi.lity of an accident other than th t previously evaluat d is not created sine The pressure level and pressurizer pressure onitoring circuins are redundant control room inputs; only on safety-train is modifi d by this PDf, The rep'acemtnt of the e-,isting isolat'on device (5K ohm resistoz) w'th the RIS model SC='302-323-X represents an enhance"ent of isolation and con:zol room independence and affects ro other syst ms, The margin of safety as defined 'n the ba"es for any Technical Specificat'on is not reduced because isolation for Altezna e Shutdown In-tru en ation i's now provided by a Class IP. component (RIS isolate" signal tra-..smarter) which when implemented improves the s fe ty v erg in.

The RIS "odel SC-'02-323-X iso'ated signal transmitter 'is qualified C'ass IZ pez I'=HE Section 323-197', T.'is ha been identified as an essentia'quipment in the St Lucie Plant Unit 2 Safe Shutdown Analy s s. i The i"plementation of this mcdification doe" not require a change o the plant Technical Specifications.

b) Additional Sprinkler System The proba'bility of occurrence or the consequences of an accident or malfunction of equip ent important to safety pre'viouslv evaluated in the sa ety analysis report is not increased because the preaction sprinkler system is non sa ety related ard does not perform any safety related unction nor does it have a direct connection with any safety related system or wquipmont.

A possibility for an accident nr malfunction of a -different type than any previous'y evaluated 'n the safety analysis report is not created because'there aze no new connections made to any safety related system or equipment. In areas where failire of the piping and/oz supports may cause damage to safety related system or equipment, the piping is seismically analyzed and supports are seismically designed.

The margins'f safetv as defined in the basis for any Technical Specification is not "educed because based. on a hydraulic check spzink'or addit'ons it is determined that design adequacy has o'he been maintained for the propez operation of the fire suppression system.

The imple. entation of this modification does not require a chan, to the pl nt Technical Speci ication ~

PCM 092-286 (c) Revision to the cables in the Safe Shutdown Analys's Tne -following is a list of those cables, which a"e being re;ised 'n the Safe Shutdc a Analysis, ard th'e associated PC.".s by which they were rodified to be, rec;oved fron th 'ss.ntial C"ble List or isolate" by transfer ",~itches.

Cable No Cable Function Hodificaticn ?Pi H21631A I-SE-09-2, AF4 P2A Pwr Removed by PCH 120-285 H20370F PT-1108, Pressurizer (See '.emote 1)

Pressure RTGB H20649A 120V AC, PP201, HSCP (See tiote 2)

H21738N PY-1108-4, Pressur'zer (See tio e 3)

Pressure HSCP H 217 38.'i LY-1105-1, Pressurizer (See Note 3)

Level PSCP C20250B V-,3481 SDC 2A Isolatior. (See emote 4)

Control H21629E V-1474 PORV Control Isolated by TS PCA 130-284 H20253B V-3651 Control (See Note 4)

H21630E V-1475 PORV Control Isol-ted by S P .'i 130-284 hOT""S:

1 ~ Isolated by iso'ation device during front f' of St Lucie Unit 2

2. Cable has been removed frot: PP201 ard reconnected to PP201A by PQ". 121-285

3. Isolated by isolat'n device e'ded via th's PCN.

4, Breaker rac!ced out by operating procedures valve can be operated manually by Hand 'kneel.

The Safety Analysis in t'ne above PC;".s provided the bases for the justification that the i-pie-entatios 0" ti.ese PC.is did not involve

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any unrevieved s"fety questinr. ~ The effects of th '= ale"..en ation of these odifications on the Plant Technical Specification vere also addressed 'n these PC.'!s.

The foregoing const'utes, per 10CFR 50.59 (b), the written safety evaluation which provides the bases that this ch nge does not involve an unreviewed safety question and prior com~isslcn approval for the i=ple~entation of this PCH 's not required.

PCM 108-286 HIGH INITIAL RESPONSE EXCITATION SYSTEM hBSTRA CZ This Engineering Package covers modifications to the Turbine&enerator brushless excitation system. The brushless excitation system will be upgraded to a High Initial Response (HIR) Brushless Excitation System which will allow the generator to respond quickly to changes in system voltage.

A larger permanent magnet generator, a new stator coil in the brushless-exciter, a new voltage regulator and a new voltage regulator enclosure vill be required to modify this system.

The Turbine-Generator does not perform a safety related function. The modifications to the Turbine Generator are classified as non safety related. However, since there will be modifications to the RTG Boards, this package is classified as Quality Related.

This 'EP does nest constitute -an unreviewed safety question and the modifications described were 'reviewed in 'accordance with 10CFR50.59 and were determined to have no adverse impact on plant operations or safety related equipment.

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

This change does riot involve an unrevie'wed .safety question and prior Commission approval for the implementation of this PC/M is not required.

PCM 108-286 With respect to Title 10 of the Code of Pederal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed

'afety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

The Turbine, Generator High Initial Response (HIR) brushless excitation system is not a safety related system, h larger permanent magnet generator (PMG), stator coil in brushless exciter and voltage=

regulator will replace the existing equipment and have ao impact on any plant system or operation. The HIR excitation system allows the generator to respond quickly to changes in system voltage.

Subsection 3.5.3.2 of the FSAR addresses External Missiles with subpart (b) addressing Turbine Missiles, specifically, missiles generated by the high pressure turbine rotor and the low pressure turbine discs. There are no changes to the high pressure turbine rotor nor the low pressure turbine discs. The modifications required to upgrade the system include a new PNG rotor, PMG stator and exciter stator which are located at the exciter end. The consequences of turbine failure and the potential for damage to critical plant structures, systems, and components from the resulting missiles has not been increased by this modifiction.

The modifications to the Turbine Generator, the voltage regulator, the voltage regulator enclosure and the HVAC system in the Turbine Building are not safety related and do not affect any plant systems.

The cables for the lighting, receptacles and power feeds in the voltage reguator enclosure are routed in cable tray and conduit in the Turbine Building. They do not require seismic support and do not affect safety related equipments The modifications to the RTG Boards will involve the replacement of selector switches with an updated version that are the same model sire and have the same characteristics as the existing switches.

Additional modifications involve the relabeling of annunciator windows and the actuation of an existing spare relay. These modifications do ,not effect the safety related functions of the.

affected RTG Boards.

PCM 108-286 Based on the preceeding, the following'onclusions can be made.

The probability of occurrence or the consequence of an accident or .malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased, since the modifications to the Turbine Generator High Initial Response (HIR) System enhances the operability of the equipment. The introduction of the HIR exciter and voltage regulator in the St Lucie Turbine Generator System will have no effect on the turbine generator control system or the steam supply system (See Attachment 7.5). The addition of a larger PMG, a new stator coil in the brushless exciter, and a new voltage regulator will allow the generator to respond quickly to changes in system voltage.

(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. This modification does not affect any safety related equipment. There are no additional missiles generated by the addition of equipment to the Turbine Generator. There is no introduction of any new failure mode for the equipment.

(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification. The safety function that is controlled by the various applicable Technical Specifications, is maintained by this change. The proposed design ensures that the new HIR system will allow the generator to respond quickly to changes in system voltage.

Since the Turbine Generator is a non-safety related piece of equipment the margin of safety provided by the Technical Specification is preserved.

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

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

PCM 113-286 DIESEL GENERATOR CRANK CASE OIL DEFLECTOR PLATE ABSTRACT The. St. 'Lucie Unit 2, 2A, l2 cylinder diesel engine had a problem with false crankcase over-pressurization,. alarms. Analysis indicated this problem was attributed to oil splashing against the diaphragm of the pressure sensor. To resolve this-condition, an oil deflector plate, supplied by EMD the engine vendor, installed, in front of the crankcase pressure detector sensor..This

'as modification has corrected the problem and does not adversely affect the engine operability. NCR 2-028 (Reference 6.4) and its associated Safety Evaluation accommodated the temporary use of the design change. The temporary modification will be made permanent by this Engineering Package.

Based on the FSAR, the diesel generator is safety related but the oil pressure detector performs no safety related function. Since the oil pressure detector is attached to a safety related structure (the diesel engine), the oil pressure detector must be considered nuclear safety related.

Based on a failure mode evaluation and a 10CFR50.59 review, this modification does not involve an unreviewed safety question nor require a change to the technical specifications. Therefore, prior iXRC approval is not required for implementation of the modification. This modification has'no effect on plant safety.

SAFETY EVALUATION This Engineering Package will make the temporary modification, provided in the disposition to NCR-2-208 (Reference 6.4), to the 2A 12 cylinder, diesel engine permanent. The engine vendor, Eh!D, supplied the oil deflector plate, which was installed in front of the crankcase pressure detector.

Although the diesel generator is nuclear safety related, the crankcase pressure detector performs a non-safety related function, which is to trip the engine due to high crankcase pressure during testing situations.

This trip function is overridden when the engine is auto started due to SIAS, CIAS, CSAS, or loss of offsite power.

The oil pressure deflector plate will not in anyway affect the safety related components of the diesel engine. A stress analysis (Ref. 6.3) of the deflector plate demonstrated that failure of the plate is not possible when installed properly.

PCM 113-286 Based on the above and information supplied in the design analysis it can be demonstrated that an unreviewed safety question as defined by LOCFF50.59 does not exist.

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

A failure analysis (Ref. 6.3) was performed for the deflector plate. Based on the results, it was concluded that a failure of the plate was not a credible event. Therefore, the probability of occurrence of accidents previously addressed in the FSAR has not been increased.

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

The failure analysis (Ref. 6.3) has shown that this modification will not result in a credible failure of the diesel generator. Therefore the possibility of an accident of a different type has not been created.

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

Since the intended function of the diesel generator is not affected by this modification, the margin of safety as defined in the basis for any technical specification has not been reduced.

LOCFR50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the technical specification is not required. As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by LOCFR50.59 that pertains to an unreviewed safety question can be positively answered. Also, no change to the technical specifications is required based on the above evaluation.

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

PCM 120-286 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION LIST REVISION ABSTRACT This Engineering Package provides the vehicle for updating several areas of equipment qualification. This package includes corrections to the 10CFR50.49 list, changes in maintenance requirements, and various documentation p'ackage corrections.

This Engineering 'Package (EP) is considered Nuclear Safety Related because it affects equipment falling under the scope of 10CFR50.49. This package does not represent an unreviewed safety question since it deals strictly with enhancing the present documentation used to qualify equipment at St Lucie Unit No 2 and no physical plant modifications are required by the Engineering Package. The safety evaluation of this package indicates that a change to the Plant Technical Specifications is not required. The equipment removed from the 10CFR50.49 list are listed in Section 2.1.2 of this PCM. Removal of equipment from the 10CFR50.49 list does not affect plant safety and operation.

Su lement 1 This supplement adds additional splicing materials to the 10CFR50.49 list and updates EQ Documentation Package 2998-A-451-16.1 "Raychem Corporation Splices". This supplement also revises maintenance note 24 of the 10CFR50.49 list and updates EQ Documentation Package 2998-A-451-35.6 "Target Rock Solenoid Valves". The original safety evaluation is not affected by this supplement.

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

This Engineering Package provides for several changes to the present St Lucie Unit No. 2's 10CFR50.49 list. This documentation will affect the future procurement of various safety related components and assist in validating the components'bility to function before, during and after a design basis event. Therefore, this EP is considered Nuclear Safety Related.

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

PCM 120-286 Based on the above, the modifications included in this Engineering Package do not involve an unreviewed safety question because of the following reasons:

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

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created by this modification. Function, mounting pnd

'the ability to withstand harsh environmental conditions have not been altered and this modification does not affect any other safety related equipment.'iii)

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

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

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

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

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

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

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~q"ipment being removed has been shown to be installed in a mild environment or not required to mitigate and monitor the consequences of an accident. The probability of malfunction of equipment is unc~ ged The probability of malfunction of equipment important to previously evaluated in the FSAR remains unchanged. The consequences of malfunction of equipment important to safety previously valuated in the FSAR are unchanged. The possibility of malfunctions a different type than those analyzed in the FSAR is not created.

PCM 123-286 PRESSURIZER MISSILE SHIELD ACCESS LADDER SAFETY CAGE ABSTRACT Tnis design package consists of the fabrication and installation of a personnel safety cage for the pressurizer missile shield access ladder 'and modification of the ladder, The safety cage will be attached to the ladder.

The modification of the ladder is required to provide safe access to the top of the pressurizer wall as well as to the missile shield.

The personnel safety cage does not perform or affect a safety"related function. However, this Engineering Package is classified Quality Related since there is a potential that, during a seismic event, the personnel safety cage could 'damage safety-related items that are in the vicinity'. Quality Related requirements are applied to this design.

This modification has been evaluated in accordance with 10CFR50.59. This safety evaluation indicates that implementation of this EP does not involve an unreviewed safety question, and prior Commission approval for its implementation is not required.

The implementation of this modification does not re'quire a change to the plant Technical Specifications and has no effect on plant safety and operation.

PCM 123-286 SAFETY EVALUATION Safety Analysis With respect to title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The pressurizer missile shield access ladder and safety cage do not perform or affect any safety-related system or. function. However, this Engineering Package is classified as Quality Related since failure of the access ladder or safety cage during a design basis event (e.g., earthquake) could potentially affect a safety-related system or equipment, since the ladder and cage are located in the containment building which contains safety-related systems.

Consequently, the ladder and safety cage have been designed for the design basis conditions specified in the FSAR and Quality Related design requirements have been implemented, thus assuring the integrity of the installation during any design basis event.

The modifications included in this Engineering Package do not involve any unreviewed safety questions because:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since this modification will have no effect on equipment required to shut down the plant and.monitor the plant in a safe shutdown condition.

(ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated since tie ladder . and cage perform no safety function and no changes have been made to any operational design. Failure of the ladder and cage could not occur since the modification has been designed for the design basis conditions.

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

The implementation of this Engineering Package does not require a change to plant technical specifications.

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

PCM 124-286 ICW LUBEWATER FLOWRATOR MODIFICATION ABSTRACT The, Intake Cooling Water (ICW)'umps Lubewater Flowrators are Brooks, armored magnetically actus'ted, rotameter type indicating switches.

This engineering package covers the rebuilding of (6) six rotameters by replacing damaged internals. These changes will not modify the present

, configuration of the lubewater installation.

The function of each rotameter is to:

l. Assist the operators to ad)ust the lubewater flow rate supplied to each individual pump bearing cooling water flow.

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2. Provide a low flow alarm in the'control room.

The ICW pump lube water flowrators are considered to be an extension of the ICW system which is nuclear safety related ~

The ICW pumps lubewater flow indicating switches indicates flow and actuates a low flow alarm, therefore their failure will not have any effect on the plant operation or safety.

A review of the changes to be implemented .by this PCM was performed against the requirements of 10"FR50.59. As indicated in Section 3,0 of this PCM, this PCM does not involve an unreviewed safety question, nor does it require a revision to the technical specification; therefore, prior Commission approval is not required for implementation of this PCM.

PCM 124-286 SAFETY EVAIlJATION 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 corisequences 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 a'ccident or malfunction of a different type than any evaluated previously in the safety analysis report may be created, or (iii) if,the margin of safety as defined in the bases for any Technical Specification is reduced.

The proposed modification affects ICW pump bear'ing Lube Mater flow instrumentation The eristing flow indicating switches will be removed, rebuilt by the vendor and reinstalled in 'the system.'o configuration changes will be made, however, the range of the instrument will be.

increased to protect 'the float from damage when the flow is higher than normal.

This modification does not involve an unreviewed safety question

'because'.

The probability of occurence or the consequences of an acc'ident or malfunction of equipment important to safety previously evaluated in, the safety analysis report is, not increased. The ICW Lube Mater Flow instrumentation is not used to determine the probability of any accid'ent and as stated in Section 2, failure of this instrumentation cannot block lube water flow and therefore has no consequence for any equipment malfunction.

ii The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not crea'ted. The system function and operation remain unchanged and no new failure modes are introduced. These instruments do not provide a control function, therefore cannot cause the failure of equipment important to safety.

iii The margin of safety as defined in the bases for any technical specification is not reduced. These instruments are not used in the bases of any technical specificatons.

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

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

.PCM 129-286 S/U .XFMR LOCKOUT DISCONNECT SWITCHES ABSTRACT This Engineering Package (EP) provides for the installation of disconnect switches in the plant startup transformers lockout 'relay circuits. The purpose of this change is to facilitate lockout relay maintenance testing while eliminating the possibility of inadvertent plant trip by propagation of a lockout relay trip during lockout relay maintenance test.

This EP is classified as Quality Related since lockout circuit actuation will trip the startup transformer and would result in plant operation under Technical Specification conditions. Subsequent loss of offsite power to the station buses could affect plant trip, starting and loading emergency diesel generators. A review of the changes to be implemented by- this PCM was performed in accordance with the requirements of 10CFR50.59. As indicated in the Safety Evaluation (Section 3.0), this PCM does not involve an unreviewed safety question, nor does it require a revisi'on to the plant Technical Specifications. This modification will have no effect on plant safety or operation. Prior Commission approval is not required for the implementation of this PCM.

PCM 129-286 SAFETY EVALUATION Pith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be,'deemed, to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if the possibility for an accident or malfunction of a different, type'han 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'educed.

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

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated, in .FSAR Section 8.2.1. 5, is not increased since the startup transformers and their lockout trip circuits are not Nuclear Safety Related equipment. Failure of the test switches will not affect the availability of the Emergency Diesel Generators .in the event of loss of offsite power (LOOP).

ii) There is no.possibility for an accident or malfunction of a different type, than any previously evaluated since the startup transformers are used for plant startup and shutdown; in the event of test switch failure, the emergency diesel generator start and loading will occur as previously evaluated in FSAR Section 8.2.

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

This EP affects equipment that is Non-Nuclear Safety Related.

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

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

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

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

PCM 134-286 QSPDS SOFTWARE MODIFICATION ABSTRACT This Engineering Package covers the modifications to the previously certified software of the gualified Safety Parameter Display System (gSPDS); The modifications consist of additions to assist the plant operator in accident monitoring. There is no major gSPDS hardware modifications as a result of this PC/N. However, the exchange of identical Erasable Programable Read Only Memory (EPROM) chips were required as a result of software modifications.

This Engineering Package is safety r'elated because it involves modifications to a nuclear safety related system gSPDS. The gSPDS is a .safety grade class, 1E processing and display system used, for post-accident monitoring. The

.hardware and software changes of this PC/H were evaluated against 10CFR 50.59.

The results of the evaluation indicate that there is no unreviewed safety ques.i on.

The effect of the mooifications on Technical Specifications was evaluated.

Since the mcdifications improve the system by, for example, enhancing the readability of the display, it is concluded that there is no technical specification changes required.

The effect of the modifications on plant safety and operation was evaluated.

There is no effect on plant safety and there is no effect on normal plant operations other than the operation of the gSPDS itself. The changes of the gSPDS operations is included in the revised version of gSPDS User's Guide.

SAFETY EVALUATION This engineering package is safety related because it involv'es a modification to a safety grade system. Me have evaluated the effects of this PC/H with respect to regulation 10CFR 50.59, and concluded that it:

a) Does not increase the probability of occur rence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

PCM 134-286 There are no major hardware changes due to this PC/M, since the

- -; .;.--.-exchanged hardware (EPROM's) are identical to the original.

"-.The software changes consist of the addition of one display page which is consistent with .the requirements of format, content and visibility of the original design. Therefore, the. e is no increase in the probability of occurrence or conseouence of an accident, or malfunc.ion of equipment because of this modification .o the gSPDS.

b) Ooes not create a possibility for an accident or mal function of a different type than, any evaluated previously in the safety analysis report.

The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR has not been created for the same reasons stated in item (a).

'i c) Ooes not reduce the margin of safety as defined in the basis for any technical specification.-

The margin of safety is not decreased by this PC/M. Instead, the safety margin is considered to be increased due to the increased visibility of the safety parameters to the operator as a result of this PC/M.

The requirements established in the Technical Specification for

+he gSPOS are unaffected by this PC/M. The 'changes of this PCM did not affect design, nor previous function, ft merely improved Human Factors Engineering considerations.

In conclusion, this proposed change does n'ot involve an unreviewed safety question or a Technical Specification change; therefore, prior NRC approval is not required to implement'his modification.

PCM 002-287 IE BULLETIN 85-03 MOV SWITCH SETTING ABSTRACT NRC IE Bulletin 85-03 requires that operating nuclear plants 'develop and implement a program to ensure that switch settings on selected, safety-related motor-operated valves (MOV's) are correctly selected, set and maintained to accommodate the maximum differential pressures expected on these valves during all postulated events within the design basis. Item a) of the bulletin requires that the design basis for those MOV's located in AFW and HPSI systems be reviewed to determine the maximum differential pressure expected during both opening and closing strokes for all postulated events. This effort was performed for St. Lucie Units 1 and 2 by Combustion Engineering as part of the CE Owner's Group (CEOG) Tasks 528 and 531. The results of the Item a) were subsequently transmitted to the NRC via FPL letter L-86-204, dated %lay 15, 1986.

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

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

Because all of the MOV's associated with Bulletin 85-03 are safety-related, this engineering package has been classified-as nuclear safety-related. A review of the switch setting changes to be implemented by this PC/M was performed against the requirements of 10CFR 50.59, and it was concluded that these modifications do not constitute an unreviewed safety question and do not require a change to the plant Technical Specifications.

PCM 002-287 SAFETY EVALUATION

',Uith respect to Title 10 of the Code of Federal Regulations, Part 50.59, the modification described in this engineering package does not constitute an unreviewed safety question because:

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. This engineering package only provides the necessary design information required to set NOV switch settings utilizing MOVATS signature analysis techniques. The recommended switch settings are considered enhancements to the existing settings to further ensure valve operability. Also, FSAR design bases were reviewed to determine the maximum loading conditions on each MOV to ensure the switch settings were properly selected. Furthermore, Item c) of BuQetin 85-03 requires that each 'QOV be stroke tested under maximum differential pressure conditions to ensure valve operability.

ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. No hardware modifications are performed as part of this PC/~>1. The proposed NOV switch settings alter accident mitigating equipment to further enhance operability. However, malfunctions of

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these MOV's do not in themselves initiate an accident. Therefore, no new accidents have been created.

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

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

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

PCM 006-287 NRC IE BULLETIN 85-03 MOV POSITION INDICATION ABSTRACT I

This Engineering Package covers modifications to the safety related. motor operated valves (MOV's) in the Auxiliary Feedwater (AFW) and the High Pressure Safety Infection (HPSI) systems. I, This Engineering Package will provide the engineering and design details required to implement the close to open torque bypass switch and closed position indication wiring modifications for the motor operated valves.

The MOV's in the AFW and HPSI systems are required for plant safe shutdown and classified as Class 1E, are seismically qualified and perform a safety related function. Therefore, this PC/M is considered Nuclear Safety Related.

This EP does not constitute an unreviewed safety question since the modifications described above were reviewed in accordance with 10CFR50.59 and will not have an adverse impact on plant operations or safety related equipment.

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

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

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

This Engineering Package provides the engineering and design details required to install additional rotors and/or internal wiring changes to MOV's in the AFW and HPSI systems. PC/M 002-287 increases the closed to open torque bypass switch settings which impact the closed position indicating light. Increasing the number of rotors from two to four will allow the limit switch for the closed position indicating light to be located on a rotor other than that used for the'orque. bypass switch. Motor-operated valves that have four rotors will only require internal wiring changes. The addition of the new rotors does not affect the existing equipment qualifications.

The implementation of this Engineering Package increases the availability of the MOV's during safe shutdown conditions and improves the MOV position indication provided to the control room operators.

PCM 006-287 NRC Regulatory Guide 1.106, Rev 1'iscusses and provides guidance directed at ensuring the thermal overload device will not needlessly prevent the motor from performing its safety function. To ensure that the safety related motor operated valves will perform their function, the thermal overload protection devices are continuously bypassed and

'temporarily placed in force only when the valve motors are undergoing periodic or maintenance testing. The thermal overload heater devices have been sized using the methodology provided by the "Power Distribution and Motor Data" Sheets.

'he MOV' that are being modified perform safety related functions within the AFW and HPSI systems and are designed for operation under conditions that could be imposed by a Design Basis Accident (DBA).

This EP has been classified as Nuclear Safety Related..

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

.(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased, since the modifications to the MOV's enhances the operability of the equipment. The addition of rotors and/or internal wiring changes to the valves will prevent the possibility of inaccurate remote closed position indication resulting from the increased bypass limit switch settings.

(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. This modification alters accident mitigating equipment to enhance their operation. There was no introduction of any new failure mode for the equipment.

(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification.

The safety function that is controlled by the various applicable Technical Specifications is maintained by this change. The proposed design ensures that the MOV' will function as assumed during an accident. Thus the margin of safety provided by the Technical Specifications is preserved.

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

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

PCM 007-287 HP TURBINE INNER GLAND AND GLAND DIAPHRAGM ENHANCEMENTS ABSTRACT This Engineering Package (EP)'ocuments the original equipment manufacturer's (OEM's) material and design changes to the HP turbine inner gland casings, and gland diaphragms..The changes and bases are as follows:

o HP Turbine Inner Glands Casings The material for the HP inner glands has been changed from a carbon steel to a 12%

Cr stainless steel to minimize erosion potential. Geometrically the design remains the same.

o HP Turbine Gland Diaphragms For these components the material has been changed from a carbon steel to a 12% Cr stainless steel to minimize erosion potential. In addition, the design has been simplified to a single wall vessel versus the previously employed double wall.

This modification has been classified as Non-Nuclear Safety Related because the inner gland casings, and the gland diaphragms are sub-assemblies of'the turbine generator's high pressure turbine. The turbine generator is not required for operation during OBE or SSE and also is classified as non-seismic per FSAR sect. 10.2.1.

Based on a failure mode evaluation and a 10 CFR 50.59 review, these enhancements do not involve an unreviewed safety question nor require a change to the Technical Specifications. Therefore, prior NRC approval is not required for implementation of the modified components. These modifications have no effect on plant safety.

PCM 007-287 SAFETY EVALUATION The components being enhanced by this EP are a part of the turbine generator assembly, s pecifically the high pressure turbine stationary casing. The components directly interface with the turbine gland steam system. The turbine generator assembly and the gland steam system perform no "safety related function, and are non seismic (FSAR sections 10.2.1, 10.4.3.)

A failure mode evaluation has demonstrated that there is no postulated failure of the components being enhanced that would result in, the 'generation of missles from the H.P. turbine. FSAR Section 10.2.3d, supports this analysis by stating that fragments generated by any postulated failure of the HP turbine rotor would be contained by the HP turbine blade rings and casings.

Title 10 of the Code of Federal Regulations Part 50.59 allows changes without prior Commission approval provided the proposed changes does not involve an unreviewed safety question or require changes to the technical specifications.

These proposed component enhancements do not involve an unreviewed safety question because:

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

As stated, these component enhancements affect only non-nuclear safety related equipment, and have no affects on the potential probability of turbine missles be generated.

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

The component enhancements are basically a like kind exchange of existing components and therefore an accident or malfunction of a different type than any evaluated previously in the safety report is not created.

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

The component enhancements affect no technical specification nor are changes to the Technical Specifications required.

In conclusion, the component nhancements performed under this EP are acceptable from the standpoint of nuclear safety since they do not involve an unreviewed safety question as defined by 10CFR50.59 and'do not require changes to the Technical Specifications. Implementation of this modification does not require prior NRC approval.

PCM 019-287 DIESEL GENERATOR TORSIONAL VIBRATION.ISOLATION

~BST$ ~

This engineering package covers modifications to the Diesel.

Generator (D/G) Fan Drive System which will isolate the D/G Cooling Fans and the Fan Drive System from forced torsional vibrations emitted from the diesel engines. The major feature of this package is the installation of a torsionally flexible coupling at the flange between the Power Takeoff (PTO) shaft and the fan drive shaft for each of the 12 and 16 cylinder engines in the 2A and 2B D/G sets. The change proposed by this enginee'ring package is classified as Nuclear Safety Related, is acceptable from the standpoint of nuclear safety,, does not involve an unreviewed safety question and does not require a change to the Technical Specifications.

Su lement 1 This supplement is issued because Morrison-Knudsen, Power System Division (PSD) was contracted to procure and dedicate the flexible couplings to be used in the subject modification. PSD requested the use of Lord Corporation Part No. LCD-0300-20R-C in the 16 cylinder fan drive shafts in lieu of Part No. LCD-200-20R-C as delineated in Supplement 0. Part No. LCD-0300-20R-C has a higher torque rating than the LCD-0200-20R-C which provides a higher margin of safety in the design.

This supplement is issued to document the engineering acceptance of the diesel generator configuration tested per Supplement 1.

This Supplement does not affect, amend, or change the original safety evaluation or Technical Specifications.

Thi's change modifies the Diesel Generators by installing a flexible coupling in the 12 and 16 cylinder diesel engine fan drive shafts. The fan drive shafts are part of the Diesel Generator Cooling Water System which provides sufficient capacity to cool the diesel generator set it serves under postulated loading and ambient conditions. Since the diesel generators are required to provide standby emergency power to Safety Related equipment in the event the preferred power supply is not available, any modification to the D/G's is classified as Safety Related. As demonstrated in the design analysis, this modification has been designed in accordance with the safety and regulatory requirements applicable to the components which comprise the D/G Fan Drive System. Zn addition, the failure modes analysis (paragraph 2.2.1) confirms that this modification will not prevent the diesel generators from performing their design function of providing emergency power;

PCM 019-287 Based on the above and information supplied in the-

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design analysis -it can be demonstrated that an unreviewed safety question as defined by 10CFR50.59 does not exist.

The probability of occurrence or the consequences of an ,accident or malfunction of equipment important to ,safety. evaluated in the safety-analysis report (reference 6.6) has not been increased because the modification has been designed in accordance with the applicable design and safety requirements applicable for the D/G Fan Drive System. Therefore, the probability of a diesel generator failure has not been increased and the consequences. of a diesel generator failure remains the. same.

The. possibility of an accident or malfunction of a different type than my evaluated previously in the safety analysis r(port (reference 6.6) has not been crea ed since this modification does not alter the operatior cl character:sties of the diesel generator: e" s, apart from reducing vibration levels in "he fan drive system. The failure of one die,-.el generator set an'd 'the startup of the redun"ant set is assumed for all accident analyses. This assumption remains unchanged. Finally, t~';s modification affects no other system. Therefo"e, no new accident or malfunction is created.

The margin of safety as defined in the basis for any Technical Specification has not been reduced because the redundancy of the diesel generators required by the Technical Specifications is maintained.

10CFR50.59 allows changes to a facility not described in the FSAR if an unreviewed safety question or a change in the Technical Specifications is not required. As shown in the preceding sections, the proposed change does not.involve an unreviewed safety question because each conce'rn as posed by 10CFR50.59 that pertains to unreviewed safety questions can be positively answered and a change to a Technical Specification is not required.

Zn conclusion, the change proposed by this engineering package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question and does not require a change to the Technical Specifications.

PCM 026-287 FIRE PROTECTION STRUCTURAL STEEL FIRE PROOFING assn@ cz

, In order to enhance compliance with 10 CFR Part 50 Appendix R requirements, this Engineering Package (EP) provides the following:

a) Engineering and design for the addition of new sprinkler heads outside the Aerated Haste Storage Tank (EST) Room to provide conduit fire protection.

b) Engineering and design for fire wrapping conduit support steel in areas where adequate protection is currently not provided.

c) Replacement of nine (9) existing conduit supports which are attached to cable tray supports with four (4) new supports and fire wrapping of these new supports.

This Ep is designated as Nuclear Safety Related because it modifies Safety Related conduit supports by either fire wrapping the existing conduit supports steel or removing the existing conduit supports and adding new Safety Related conduit supports and fire wrapping them.

Changes to the existing sprinkler system are considered Quality Related.

The changes to be implemented by this EP have been reviewed and found to meet the fire protection requirements put forth in 10 CFR Part 50, Appendix "R". As indicated in Section 3.0, this EP neither involves an unreviewed safety question, nor does it require a revision to the Technical Specification. Therefore, prior Commission approval is not required for implementation of this EP, This EP has no impact on plant saXery and operation.

PCM 026-287 Safet Evalu'ation With respect to Title 10 of the Code of Pederal Regulations Part 50.59'J a proposed change shall be deemed to involve an unrevieved safety question; (I) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

This EP provides the required modifications to expand the existing preaction fire sprinkler, to provide adequate protection to exposed support steel of Safety Related conduit supports and additional nev conduit supports in the RAB. This EP is designated Safety Related.

The modification included in thii Engineering Design Package does not involve an unrevieved safety question. The folloving are the bases for the justification.

a) hddition of Sprinkler Heads The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased because the preaction sprinkler system is non safety related and does.not perform any safety related function nor does it have a direct connection vith any safety related system or equipment.

h possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report is not created because there are no net connections made to any safety related system or equipment. In areas vhere failure of the piping and/or supports may cause damage to safety related system or equipment, the piping is seismically analyzed and supports are seismically designed.

The margin of safety as defined in the basis for any Technical Specification is not reduced because based on a review of the hydraulic calculation for the existing sprinkler system these sprinkler head additions do not affect the design adequacy for the proper operation of the fire suppression, system.

PCM 026-287 b) Re lacement'of existin conduit su o'rts attached: to CTRs with new, conduit su orts and wra in of new existin conduit 'su ort steel

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This modification does not increase the probability 'of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report because, it does not change or alter the intended function or design requirements of any safety related conduit originally installed.

This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously because there are no new connections made to any safety related system or equipment by this modification. The existing conduit supports being removed from the CTRs have b'een replaced with new conduit supports.

These new conduit supports have been seismically designed. The construction note (9.6) requires that the new conduit supports shall be installed prior to removing the existing conduit supports from CTRs;

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therefore the structural integrity of the conduits affected by this modification are not compromised. The existing con'duit supports being wrapped have been evaluated for additional loads of fire wrap material and determined to be adequate for these loads. The seismic block wall to which a new brace from a conduit support is added has been evaluated for the additional load and the structural integrity of the masonry block wall is not compromised.

The margin of safety as defined Xn the basis for any Technical Specification is not affected by this modification because the components involved in this modification are not included in the bases of any Technical Specification.

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

The foregoing constitutes, per 10 CPR Part 50.59 (b), the written safety eraluation which provides the bases that this change does not involve an

~reviewed safety question and prior commission approval for the implementation of this EP is not required.

PCH 029-287 TURBINE GENERATOR ADDITIONAL OIL SEAL FOR 1 AND 2 BEARING ABSTRACT This engineering package covers the addition of one supplemental labyrinth to the //l and 82 bearings oil seals. Oil leakage from these seals could lead to fires due to t e proximity of the seals to hot surfaces. To preclude potential leakage, Westinghouse (th tu b ne enerator vendor) has designed and fabricated the supplemental seal which functions as an integral part of the existing seals. Use of the supplemental seal increases sealing capabilities thereby reducing the liklihood of oil leakage. This modification is classified as Non-Nuclear Safety Related'uality Croup D, but the design has been classified as Quality Related due to explicit Quality Control requirements pertaining to the installation effort.

Based on the 10 CFR 50.59 review and the failure modes evaluation,=it has been demonstrated that this change does not involve an unreviewed safety question.

Add't nally no change is required to the Technical Specifications. This modification does not adversely affect plant safety or operability. Prior NRC approval v iis not required for implementation of this design.

PCM 029-287 SAFETY EVALUATION The Unit 2 turbine generator is located in a non-nuclear safety related system and as such is not required to function for accident mitigation. These modifications affect only non-nuclear safety related Quality Group D equipment.

Based on the failure mode evaluation, failure of the 'components added by this modification. will not inhibit the operation of any existing safety related equipment or components. This evaluation is based on the assumption the new seal is 'installed according to design. Adequate Quality Control inspections have been specified to verify proper installation and therefore operation.

Accordingly, this design is classified as Quality Related.

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

This proposed change does not involve an unreviewed safety question because:

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated. in the safety analysis report harlot been increased.

As stated, the modification affects only non-nuclear safety related Quality Group D equipment. In addition, the failure modes analysis demonstrates that no safety related equipment is affected by this mod ification.

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

The failure n:odes analysis has shown that the possibility of an accident or malfunction of a different type than any previously evaluated in the safety analysis is not created.

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

This design modification affects no technical specification nor are changes to the Technical Specifications required.

Based 'n this information, an unreviewed safety question as defined by 10CFR50.59 is not created. Since no accident previously identified in the safety analysis report has 1'een affected, no new accidents or malfunctions are created and no changes to the Technical Specifications are required. An unreviewed safety question does not exit. Prior NRC approva'l is not required for implementation o'his modification.

PCM 033-287 REPLACEMENT OF VALVE V3734 ABSTRACT This Engineering Package (EP) provides for replacement of St. Lucie Unit 2 Safety Injection Tank 2A2 solenoid vent valve V3734. The existing valve manufactured by Garrett Pneumatic Systems has failed and a direct replacement is not available. The Garrett valve will be replaced with a Target Rock Model 80B-001 valve.

This modification is classified nuclear safety related since the Safety Injection Tank Solenoid Vent Valves according to FSAR Section 6.3.2.2.1 are nuclear .safety related. This EP does not have any adverse impact on plant safety and operation. Based on a failure mode analysis and 10 CFR 50.59 review, the change proposed by this EP is acceptable from the standpoint of nuclear safety, it does not involve an unreviewed safety question, and does not require any change to the Technical Specifications. Therefore, prior NRC approval is not required for implementation of the modification.

SAFETY EVALUATION This EP provides for replacement of St. Lucie Unit 2 SIT 2A2 solenoid vent valve V3734-. The existing valve manufactured by Garrett Pneumatic Systems has failed and a direct replacement is not available. The Garrett valve will be replaced with a Target Rock Model 80B-001 valve.

This modification is classified nuclear safety related since the SIT solenoid vent valves according to FSAR Section 6. 3. 2. 2. 1 are nuclear safety related. This Ep does not have any adverse impact on plant safety and operation. The .new SIT solenoid vent valve has been designed to Safety Class 2, Quality Group B, Seismic Category I, and'lass 1E requirements. All safety and regulatory requirements specified in FSAR Section 6.3 have been met.

The function of the SIT vent valves is as follows:

During plant cooldown, the SIT solenoid vent valves may be used.

When the Reactor Coolant System pressure is 650 psia the SITs are depressurized to 235 psia. The SITs can be depressurized by either opening the SIT vent valves or by draining the SIT to not less than 48 percent full.

PCM 033-287 Based on the above and the information supplied in the design

analysis, question it can be demonstrated that an unreviewed as defined by 10 CFR 50.59 does not exist.

safety 0 The probability of occurrence or the consequences of an -

accident or malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased.

The replacement SIT solenoid vent valve meets all safety and regulatory requirements specified n the FSAR. The operating characteristics of the new valve are, shown to be acceptable by Section 2.0. The replacement and original valve are functionally equal. Based on this, the probability of occurrence or the consequences of all analyzed accidents remain unchanged.

0" The possibility of an accident or malfunction of a d'ifferent type than any evaluated previously in the Safety Analysis report has not been created.

The proposed design change alters accident mitigation equipment, Safety In)ection system. There are no accidents that are initiated by malfunctions associated with this system. Therefore, no new accidents have been created.

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

Technical Specification 4.5.2.a requires once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that valve V3734 be verified in a locked close position.

This modification does not affect this Tech. Spec. Thus, the margin of safety provided by valve V3734 and controlled by the Technical Specifications are preserved.

10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specifications are not required. As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by 10CFR50;59 that pertains to an unreviewed. sa ety question can be positively answered. Also, no change to the Technical Specifications is required based on the above evaluation.

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

NRC approval is not required for implementation of the modifications.

PCM 040-287 CONDENSATE RECIRCULATION TO CONDENSER SQ RT EXTRACTOR REPLACEMENT'BSTRACT

.This Engineering Package covers the replacement of one (I) square root extractor.

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

A review of the changes to be implemented by this PC/M was performed against the requirements of 10. CFR 50.59. As indicated in Section 3.0 of this PC/M, this PC/M does not involve an unreviewed safety question, nor does it require a revision to the technical specification, therefore prior commission approval is not required for the implementation of this PC/M.

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

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

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

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

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

The foregoing constitutes, per 10CFR 50.59, 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 042-287 MFRV POSITION INDICATORS REHOVAL ABSTRACT This engineering package covers the removal of two.Main Feedwater Regulating Valve position indicators ZI-9011 and 9021 from,RTG unfastened Board 202 along with associated wiring, cable, and conduit. h steel plate will be to the control board to cover the exposed area.

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

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

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

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

These changes were reviewed against the requirements of 10CFR50.59. hs verified in the Safety Evaluation, this change

- neither requires a Technical Specification revision nor is it an unreviewed safety question. Therefore, prior NRC approval is not required.

hFETY EVhLUhTION This EP is clas'sified as Quality Related because the components being removed, while performing a Non-Nuclear Safety Related function, are installed in'he RTG Board where the potential through exists for impacting Safety RelatedBoard,equipment of modification'f the wiring in the RTG cover:plates the removal that could equipment and the installation of potentially have an effect on'he seismic integtity of the RTG Board.

This design proposes to remove the Main Feedwater Regulating Valve (MFRV) position indicators currently installed in the RTG Board 202.

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

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

PCM 042-287 The change is not an unreviewed safety question because the of an accident or probability of occurrence or the consequences evaluated malfunction important to safety previously in the FSAR has not been increased.

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

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

The restoration of the RTG Board through appropriate. cover plates to replace the removed indicators has been evaluated within this package. This evaluation concluded both that the shismic integrity of the RTG Board will be event which couldno adversely retained and that missiles could be generated during a seismic impact Safety Related equipment.

Based on the above and the information supplied in the design analysis.

question it can as defined be demonstrated that an unreviewed safety by 10 CFR 50.59 does not exist.

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

These indicators are not considered by th'e FSAR in determining %he probab'ility of accidents, possible types of accidents or in the evaluation of consequences of accidents. Also they could provide misleading information which their removal would prevent.

Therefore it can be concluded that the probability of occurr'ence or consequences of accidents previously addressed in the FSAR remains unchanged.

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

As stated above, these indicators are unreliable and could provide 'isleading valve position indication.

Since. these indicators are located in the Control Room, misleading information from them could lead to an accident or malfunction of a different type than any previously evaluated in the FSAR. By removing them, this possibility is eliminated since this chance of

.error is no longer present.

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

These indicators are not required by any technical specification nor are they included in the basis of any technical specification. Therefore, their removal will not reduce the margin of safety as defined in the loses for any technical specification.

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

PCM 048-287 REACTOR CAVITY SEAL RING MODIFICATION ABSTRACT This Engineering Design Package covers modifications to the Reactor Cavity Seal Ring. The pneumatic seals have been modified by the vendor. The male studs used to attach the seals to the seal plate have been changed to female threaded fittings. Also, the seal air lines have been changed from neoprene hose to stainless steel braid hose. These modifications are necessary to Improve the reliability and operability of the seal and the air lines.

Based on. the FSAR, the cavity seal ring is non-nuclear safety related. However, to ensure the Reactor Cavity Seal Ring will perform i'ts intended function, quality requirements are assigned. Therefore, this modification is classified as Quality Related.

Based on a failure mode evaluation and a 10 CFR 50.59 review, these modifications do not involve an unreviewed safety question nor require changes to the technical specifications. Therefore, prior NRC approval is not required for implementation of the modifications. These modifications have no effect on plant safety.

Su lement I Supplement I incorporates a minor drawing revision. Tt e changes made by this supplement are non-technical and administrative in nature. The drawing was revised to change the drawing number and revision. Design Integration has been reviewed and it has been determined that there are no adverse consequences as a result of revising the drawing. There are no other changes to this package. The safety evaluation remains valid since there are no unreviewed safety questions and no changes to the technical specifications are required. This change has no effect on plant safety.

PCM 048-287 SAFETY EVALUATION.',

This Engineering Package covers modifications to the Reactor Cavity Seal Ring. The pneumatic'seals have been modified by the vendor, the studs used to attach the seals to the seal plate have been changed to female threaded fittings and the seal air lines have been changed from neoprene to stainless steel braid.

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

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

Since the reactor cavity seal ring is not considered by the FSAR in determining the probability of accidents, possible types of accidents, or in the evaluation of consequences of accidents, it can be concluded that the probability of occurrence of accidents previously addressed in the FSAR remains unchanged.

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

Since the sealing portion of the cavity seal ring has not been changed, the possibility of an accident of a different type has not been created.

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

Again, since the sealing portion of the cavity seal ring has not changed, the margin of safety as defined in the basis for any technical specification has not been reduced. r 10 CFR 50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the technical specification is not required. As shown in the preceding sections, the change proposed by this design package does not involve an unreviewed safety question because each concern posed by 10 CFR 50.59 that pertains to an unreviewed safety question can be positively answered. Also, no change to the technical specifications is required based on the above e va lua t i on.

ln conclusion, the change proposed in this design package is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question, and does not require changes to the technical specifications. Therefore, prior NRC approval is not required for implementation of the modification.

PCM 050-287 CONDENSATE PUHP EXPANSION JOINT REPLACEMENT ABSTRACT This Engineering Package covers the change out of the St. Lucie Unit 2 Condensate pump expansion joints and the modification to the adjacent pipe supports. The existing expansion joints are made of an elastomeric material which has deteriorated due to aging. The replacement expansion joints are made of stainless steei and will correct the problems associated with the aging deterioration.

These modifications are classified as non-nuclear safety related, according to the FSAR. Based on a failure mode evaluation and a 10CFR50.59 review, these modifications do not involve an unreviewed safety question nor a change to the technical specifications. Therefore, prior NRC approval is not required for implementation of these modifications. These modifications have no effect on plant safety.

PCM 050-287 SAFETY EVALUATION This Engineering Package covers the modifications to the condensate pump expansion joints. The elastomeric expansion joints will be replaced with stainless steel expansion joints.

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

o The probability, of occurrence or the consequences of an accident or

~

malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased.

Due to the location of these expansion joints, their failure would not cause interaction with any safety related equipment. Also since the condensate system is not considered by the FSAR, Section 10.0.7, in determining the probability of accidents, possible types of accidents, or in the evaluation of consequences of accidents, it can be concluded that the probability of occurrence of accidents previously addressed in the FSAR remains unchanged.

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

II The components involved in this modification do not perform any safety related function. The operational design of the condensate system has not been affected by the material change of the expansion joints. Also, due to their location, the failure of these expansion joints would not cause interaction with any safety related equipment. Therefore, the possibility of an accident of a different type has not been created.

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

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

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

PCM 051-287 CONDENSER HOTWELL NITROGEN INJECTION CONNECTIONS ABSTRACT This Engineering Package covers modifications to the St. Lucie Unit 2 condensers to allow the installation-of taps for the purpose of injecting nitrogen into the condenser hotwells. Testing (Ref 6.0) has shown that injecting 1 cfm of nitrogen into a condenser shell reduces condensate dissolved oxygen concentration by approximately 2 ppb. The flow of non-condensibles in the air removal section of the tube bundle becomes inhibited when there is low air in-leakage into the condensers. Oxygen is entrained as the condensate drips through the air pockets which form as a result of the stagnant conditions.

Injecting an'inert gas such as nitrogen enables the air removal section of the condenser to establish the flow required to remove non-condensibles without introducing additional oxygen into the system.

This modification is classified as non-nuclear safety related. Based on a failure mode evaluation and a 10 CFR 50.59 review, this modification does not involve an unreviewed safety question nor require changes to the technical specifications. Therefore, prior NRC approval is not required for implementation of this modification. This modification has no adverse affect on plant safety or operability.

Su lement 1 Supplement 1 adds four.(0) weld numbers to drawing number 3PE-051-287-005.

The changes made by this supplement are non-technical and administrative in nature. The drawing was revised to include the weld numbers and revision change. Design Integration has been reviewed and it has been determined that there are no adverse consequences as a result of revising the drawing. There are no other changes to this package. The safety evaluation remains valid since there are no unreviewed safety questions and no changes to the technical specifications required. This change has no effect on plant safety.

PCM 051<<287

'AFETY EVALUATION .

This Engineering Package'overs the modifications necessary to install condenser taps for the purpose of. injecting nitrogen into the condenser, hotwells. The condensers are classified as non-nuclear safety related, quality group D. A complete failure of these connections could result only in a loss of condenser vacuum and subsequently a turbine trip. However, no safety related equipment, components or safety related functions are affected.

Based on the above and information supplied in the design analysis it can be demonstrated that an unrevlewed safety question as defined by I 0 CFR50.59 does not exist.

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

Due to the location of the piping, valves and control devices associated with this modification, no interaction with safety related equipment will occur in an event of failure. Also, the condenser is not considered by the FSAR in determining the probability of accidents, possible types of accidents, or in the evaluation of consequences of accidents. It can be concluded that the probability of occurrence of accidents previously addressed in the FSAR remains unchanged.

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

In the event of failure, the equipment added by this Engineering Package will not interact with any safety related equipment due to the location of the modifications. Also, the installation of the condenser taps does not change the intended function of the condensers. Therefore, the possibility of an accident of a different type has not been created.

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

Again, since the intended function of the condenser is not affected by this modification, the margin of safety as defined in the basis for any technical specification has not been reduced.

10CFR50.59 allows changes to a facility as described in the FSAR, if an unreviewed safety question does not exist and if a change to the technical specification is not required. As shown in the preceding sections, the change p'roposed by this design package does not involve an unreviewed.

safety question because each concern posed by 10CFR 50.59 that pertains to an unreviewed safety question can be positively answered. Also, no change to the technical specifications is required based on the above evaluation.

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

PCM 052-287 CONDENSER POLISHER TIE-IN ABSTRACT This Engineering Package (EP) is for the installation of the 24 inch tie-in piping and valves required for the future connection of the Condensate Polisher System (CPS) to the Unit 2 Condensate System. It also includes the installation of the by-pass flow control valve required for operating the CPS using Unit 2 condensate and the installation of a connection to the Unit 2 condensate storage tank for providing the capability of using Unit' condensate for backwashing the condensate polishers.

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

The safety evaluation has determined that this EP does not constitute unreviewed safety question and implementati'on of the EP does not

'n require a change to the Plant Technical Specification. Therefore, prior NRC notification for implementing this EP is not required.

This EP has no impact on plant safety and operation.

PCM 052-287 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed change shall be deemed to involve an unreviewed'safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

This Engineering Package (EP) is for the installation of the 24 inch piping and valves required for the future connection of the Condensate Polisher System (CPS) to the Unit 2 Condensate System. It also includes the installation of the by~ass flow control valve required for operating the CPS using Unit 2 condensate and the installation of a connection to the Unit 2 condensate tank for providing the capability of using Unit 2 condensate for backwashing the condensate polishers.

The portions of the Condensate System, Condensate Storage Tank piping and the CPS that this modification will be implementing does not perform any safety function or interact with safety related equipment, therefore this package is classified as non-nuclear safety related.

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

(i) The probability of occurrence or the .consequences of an accident or malfunction of equipment- important to safety previously evaluated in the safety analysis report is not increased. The Condensate System and the CPS are not used in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.

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

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

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 provided the bases that this change does not involve an unreviewed safety question and prior Commisision approval for the implementation of the PCM is not required.

PCM 055-287 MSR PARTITION PLATE NUT REPLACEMENT ABSTRACT This PC/M provides for the replacement of the moisture separator reheater tube bundle hemi-head partition plate nuts with .new Westinghouse nuts made of a different material. The existing nuts were found to be susceptible to stress corrosion cracking and failures have been experienced at various Westinghouse Plants, including Turkey Point. Failure of these nuts can result in degraded MSR performance.

A review of the changes to be implemented by this PC/M was performed against the requirements of 10CFR50.59. As a result, this modification is classified as non-safety related, does not constitute an unreviewed safety question, will not affect plant safety, and does not require a change to the plant Technical Specif ication.

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

This modification does not involve an unreviewed safety question because:

i) The probability of occurance or the consequences of an accident or malfunction of equipment important to safety prevoiusly evaluated in the safety analysis report is not increased. The MSR's are not used in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.

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

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

iii) The margin of safety as defined in the bases for any Technical Specification is not affected by this PCM since the component involved in this modification is not included in the bases of any Technical Specif ication.

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

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

PCM 056-287 480V SWITCHGEAR 2Al and 2Bl TRANSFORMER REPLACEMENT ABSTRACT r

Due to environmental concerns attendant to polychlorinated biphenyl (PCB) cooling/insulating liq'uids', all transformers filled wth PCB are being eliminated from FP&L's system. The station service transformers for 480 volt switchgear 2A1 and 2B1 are filled with PCB cooling/insulating oil. Each transformer contains 254 gallons of PCB liquid. This Engineering Package provides for the replacement of the existing PCB filled station service transformers with equivalent transformers filled with an environmentally acceptable silicone cooling/insulating liquid.

Station service transformers 2A1 and 2B1 do not perform any nuclear safety related functions, however, because of their importance to normal balance of plant operations the replacement transformers are classified as Quality Related in this Engineering Package.

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

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

Su lement 1 Supplement 1 incorporates vendor drawings and associated engineering design calculation certification sheet for information only. -

The original safety evaluation is not affected by this supplement.

PCM 056-287 SAFETY EVAWATION 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'r the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be "created; or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

This Engineering Package addresses the replacement of station service transformers 2Al and 2Bl which are both non-safety related. The FSAR refers to "two non-safety" related transformers (2Al and 2B1) in Subsection 8.3.1.1.1.c, on the bottom of page 8.3%. On FSAR Figure 8.3-2a the 2A1 and 2Bl station service transformers are identified as non-Class 1E, i.e. non-safety related.

Station service transformers 2A1 and 2Bl do not perform any nuclear safety related functions, however, because of their importance to normal balance of plant operations the replacement transformers are classified as Quality Related in this Engineering Package.

The 2A1 and 2B1 station 'service transformers are located on the ground elevation of the turbine building. The transformers will be replaced on a one-for-one basis by transformers essentially identical except for the silicone cooling/insulating liquid.

Based on the preceeding, the foU.owing conclusions can be made:

{i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased because the existing transformers are classified as non-safety related in the FSAR and they are being replaced on a one-for-one basis by transformers that aie identical in form, fit and function.

(ii) This modification does not change the. operation of the non-safety related 480 volt auxiliary power distribution system. Therefore, there is no possibility that an accident or malfunction of a different type than any evaluated in the FSAR may be created.

(iii) The replacement station service transformers are identical in fozm, fit and no safety function to the related functions.

existing transformers .and perform Therefore, this modification does not reduce the margin of safety as defined in the bases for any technical specification.

The implementation of this PC/M does,not require a change to the pl'ant Technical Specifications.

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

PCM 058-287 BORIC ACID'AKEUP SYSTEM RELIEF VALVE MODIFICATION This Design Package covers the installation of lap joint flang'es on the inlet side of twelve (12) 1/2" x 1" relief valves on:he Boric Acid Makeup System; V-2123, V-2160, -V-2171, V-2630, V-2631, V-2632, V-2634, V-2636, V-2637, V"2639, V-2641, 6 V-2648. This will aflow post-maintenance reassembly of the relief valves without regard to inlet flange bolt hole alignment. In addition, the relief valve manual lift levers will be removed and their activating shafts seal welded to eliminate leakage. The relief valves involved provide thermal relief protection for ASME Section ill Class II piping, which makes. this modification safety related. Based on a failure mode analysis and 10CFR 5659 review, the changes proposed by this Engineering Package are acceptable from the standpoint of Nuclear Safety.

This modification does not involve an unreviewed safety question and a Tech Spec change is not required, therefore, prior HRC approval is not required for implementation of this modification. The function of the .-elief valves is not altered by this modification.

Safety Evaluation This modification consists of the replacement of the existing socket weld flange with Iap joint flanges, the removal of the relief valve manual lift lever, and,the seal welding of the lift lever activating shaft. This modification does not affect the d<sign function of the relief valves, and does not introduce any new active components to the system. In fact, the ren.oval of the manual lift lever eliminates one potential failure mode of the relief valves; that of the relief valve inadvertently being manually lifted. Since the system and components modified by this Engineering Package are ASME section III Class II, This package is classified as Nuclear Safety Related.

The fo!lo ing constitutes an evaluation to determine if the implementation of this Engineerii = Package will result in an unreviewed safety question as defined by 10CF RS0.59:

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

'nd consequences are not affected by this modification.

The probability of an accident or malfunction of a different type than previously evaluated in the FSAR has not been created. Since the system Design Bases as described in FSAR sections 9.3 4.1 and 9.3.4.3.2 (h) are not affected by this modification, no new accidents are made possible.

The margin of safety as defined in the basis for any Technical Specification has not been reduced since no system design parameters are being altered.

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

PCM 059-287 LOW POWER FEEDWATER CONTROL SYSTEM ABSTRACT In order to reduce the frequency of reactor trips encountered during start-up with manual control of the Feedwater by~ass regulating valves, a new Low Power Feedwater Control System (LPFCS) will be added to the existing Feedwater Control System. The LPFCS is designed to provide stable and automatic control of the by-pass feedwater regulators, which offers an additional advantage over the present. manual operation at low power loads in the load range of 2-15K.

The inherent design of this equipment is to provide for a smooth and steady output for automatic control of the by-pass regulators and to significantly reduce the frequency of reactor trips during unit start-up. This equipment does not perform any Safety Related functions and is not required for safe shutdown or alternate shutdown functions. However, this equipment will be installed in the Control Room and will be seismically evaluated. Therefore, this package shall be considered Quality Related.

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

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

PCM 059-287 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not 1ncreased since this new Low Power Feedwater Control System (LPFCS) is an extension of the Feedwater Regulating System and as described in FSAR Subsection 7.7.1, this system function is not essential for the safety of the plant. The installation of the LPFCS will provide control improvements to maintain steam generator water level at set point value during unit start-up with significant reduction in the number of reactor trips due to steam generator level excursions.

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

a) The new equipment mountings and added components have been seismically analyzed for additional loading and it has been concluded that these additions will not alter the original stress conditions or the fundamental frequency of the RTG Boards. Consequently, the seismic qualification of the RTG Boards will not be adversely affected.

b) Also, the LPFCS, which is an extension of the Feedwater Regulation System, is neither required for safe shutdown nor for mitigating the consequences of an accident.

iii) The margin of safety as defined in the bases for any Technical Specifications is not affected by this EP since the components involved in this modification are not 1ncluded in the bases of any Technical Specification.

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

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

PCM 061-287 INSTALLATION OF VERNIER MERCURY MANOMETERS ABSTRACT Unit 2 main condenser pressure is measured by two full range, electronic, absolute pressure transmitters connected to Condenser 2A, and'one narrow range, absolute pressure transmitter connected to Condenser 2B. Corresponding electronic receivers located in the. Control Room panels complete the existing monitoring instrumentation.

The main condensers are classified as non-safety related. However, this Engineering Package (EP) will be classified as Quality Related to assure that good construction practices are followed and to assure added confidence during the design and installation to prevent mercury contamination of the condenser condensate, feedwater and the steam generators.

This EP covers modifications to the Main Condenser Pressure Monitoring System by adding one locally mounted, high precision, 35 inch range, vacuum mercury manometer per condenser and a 35 inch range barometer. These three instruments will be fitted with a 26 to 31 inch range Vernier scale to improve the reading precision to 1/100 of 1 inch. The improved accuracy will help in assessing when condenser cleaning is necessary.

A review of the additions implemented by this PCM 'was performed against the requirements of 10CFR50.59: As indicated in Section 3-0 of. this package, this EP does not involve an unreviewed safety question, nor does it require a revision to the Plant Technical Specification. Therefore, prior Commission approval is not required for implementation of this PCM.

SUPPLEMENT 1 This EP Revision incorporates the following:

a. Preventing the mercury from entering the condenser.

b. Verification of FSAR commitments for the use of mercury on site.

c. Impact of the use of mercury upon the NSSS equipment guarantees.

d. Special instructions for handling of mercury.

The original safety evaluation is not affected by this supplement.

PCM 061"287 SAPETY EVALUATION With respect to Title 10 of the Code of Pederal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question:

consequences (i) if the probability of occurence or the of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility" for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report, Section 10.4.1, "Main Condenser," is not increased since the mercury manometers and the barometer do not perform any safety function. In addition, these instruments are not essential for the safety of the plant and are not connected to any plant safety related systems.

ii) The possibility of an accident or malfunction of a different type other than any evaluated previously in the Safety Analysis Report, Section 10.4.1, "Main Condenser," is not created since the mercury manometers and the barometers are neither required for safe

,shutdown nor for mitigating the consequences of an accident.

iii) The margin of safety as defined in the bases for any Technical Specifications is 'not affected by this Engineering Package since the components involved in this modification are not included in the bases of. any Technical Specification and they do not change the original operational capability of the equipment.

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

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

PCM 062-287 ANNUNCIATOR NUISANCE ALARMS ABSTRACT The Engineering Package {EP) includes engineering and design necessary to correct annunciator nuisance alarms requiring set point and logic modification as well as alarm circuit deletions. By implementing this EP, these circuits will be consistent with the NUREG 0700 "Guidelines for Control Room Design Review" "Dark Annunciator" concept which allows for alternately flashing annunciators in the alarm state only. Under normal operating conditions no annunciators will be'lluminated.

This EP is classified as Nuclear Safety Related since it involves modifications of Nuclear, Safety Related circuits, necessary to correct these nuisance alarms. The safety evaluation has determined that this EP does not constitute an unreviewed safety question'nd does not require a'change in the Plant Technical Specifications. This PCH can be implemented without prior Commission approval.

This EP has no impact on plant safety or operation.

S PCM 062-287 SAFETY EVALUATION

'I Vith respect to Title l0 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 (11) if the .

possibility for an accident or malfunction of a different type than any evaluated previously 1n the safety analysis report may be created; or if (iii) the margin of safety as defined in the basis for any Technical Specification is reduced.

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

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the annunciators serve no controlling functions. Electrical separation is provided between redundant safety related wiring and components and annunciator logic which is separated to protect control'unctions from being affected by annunciation circuit failure. The Safety Related circuit modifications do not affect the purpose, function or operation of control circuits.

(ii) There is no possibility for an accident or malfunction of a d1fferent type than any previously evaluated since no changes have been made to the operational design of any control circuits or associated systems. The modified annunciator windows do not perform any Safety Related functions and do not'odify the control functions of any Safety Related circuit.

(iii) .The margin of safety as defined in the bases of any technical specification is not reduced since the modified annunciator alarms perform non- nuclear safety related functions and are not included in the bases of any technical spec1fication. The Safety Related circuits which were modified have been analyzed, and it has been determined that there is no effect on control circuit set points or response times prescribed by Technical Specifications.

The modified logic of some annunciator alarms is interfaced w1th the control logic of nuclear safety related equipment, therefore, this EP is classified Nuclear Safety Related.

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

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

PCM 063-287 NEW FUEL CRANE INTERLOCK ADDITION ABSTRACT This Engineering Package (EP) modifies circuits in and adds components to the New Fuel Crane in the Fuel Handling Building to provide impzovements as follows:

Install photoelectric sensor elements, control relays, and reflective tape as an interlock system to restrict fuel crane movement in order to prevent damage from collision between the fuel crane and observation plat ozm.

This EP is classified, as Qralit~elated since it provides for modifications to equipment not required to shut down the plant oz to mitigate the consequences of a Design Basis Accident but which is used to handle new nuclear fuel. The safety evaluation has shown that the implementation of this EP does not constitute an unzeviewed safety question nor would implementat on affect plant Technical Specifications. Thus, Commission approval is not required prior to implementation.

This EP has no impact on plant safety or operation.

SAPET7 EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50,59, a proposed change shall be deemed to involve an unzeviewed safety question: (i) if. the probability of occurrence or the consequences of an accident or malfunction of equipment important to

,safety. previously evaluated in the Safety Analysis Report may be increased; or. (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

PCH 063-287 The modifications included in this Engineering Package, which consist of photoelectric travel limit interlocks in the New Fuel Crane control circuits, do not involve an unreviewed safety question because:

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated are not increased since this EP provides for increased protection of the New Fuel Crane and structures in the Fuel Handling Building by reducing the potential for damage due to mishandling, and since no anticipated mode of interlock failure will affect equipment required to shut down the plant or 'to mitigate the consequences of an accident. The modifications do not change the designed function of the crane.

ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated. This EP does not modify the intended operation of the New Fuel Crane. The addition of the interlocks does not introduce the potential for new accidents because no anticipated mode of interlock failure will affect equipment required to shut down the plant or to mitigate the consequences of an accident, and because the new interlocks provide further restriction of movement but do not otherwise change the operating characteristics of the New Fuel Crane.

iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification, since no anticipated, mode of interlock failure changes any parameter referenced in the Technical Specifications.

This EP modifies equipment that is not Nuclear Safety-Related.

Howeverj since the equipment is used for handling new nuclear fuelj and since mishandling could result in fuel damage, this EP is classified as Quality Related.

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

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

The foregoing constitutes, per 3.0CFR50.59(b)j the written safety evaluation which .provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.

PCM 068-287 MFIV TERMINAL STRIP REPLACEMENT ABSTRACT This Engineering Package provides for the replacement of terminal strips in the terminal boxes of the four Main Feedwater Isolation Valves. The existing terminal strips have experienced a recurring problem with loose connections which causes unreliable valve operation. The replacement terminal strips are already in use in the steam trestle area and have been qualified by Environmental Qualification -Documentation Package 17.1, drawing number 2998-A-451-17.1.

This Engineering Package is classified as Nuclear Safety Related due to the classification and safety functions of the Main Feedwater Isolation Valves. Implementation of this PCM does not involve an .unreviewed safety question or a change to the Plant Technical Specifications. It may be implemented without prior Commission approval.

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

PCM 068-287 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unrevieyed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if the possibility for an accident or malfunction of a different type. than any evaluated previously in the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.

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

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since this EP" provides for the replacement of existing MFIV terminal strips with the more dependable Buchanan Type NQB112 terminal blocks.

(ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated since this EP does not affect the existing design philosophy of the MFIV control system.

(iii) This modification does not reduce the margin of safety as defined in the bases for the technical specifications since it improves the mechanical integrity of the MFIV control circuit.

Since this EP affects equipment that is identified as nuclear safety related in the PSL-2 Final Safety Analysis Report, subsection 10.4.7.1, this package is considered Nuclear Safety Related.

Although this EP involves equipment on the Essential Equipment List, it does not affect safe reactor shutdown or alternate shutdown. There are no other changes to equipment which involves 10CFR50 Appendix "R" fire protection (See Attachment 7.1). 'hus, the proposed design of this package is in compliance with the applicable codes and FSAR requirements for fire protection equipment.

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

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

implementation of this EP is not required.

PCM 069-287 RELOCATION OF THE SBVF HEATER CONTROL PANELS ABSTRACT The Shield Building Vent System (SBVS) maintains a negative pressure inside the annulus 'and filters for removal of fission products following a ZOCA.

Thus the SBVS prevents containment leakage from flowing directly from the annular space, through the Shield Building Structure, to the atmosphere. The SBVS is actuated automatically by a Containment'solation Actuation Signal or a high radiation signal from the Fuel Handling Building.

The Engineering Package (EP) covers the relocation of the Shield Building Ventilation Fan (SBVF) Heater Control Panels to a mild environment to allow for a reduction in EQ mainte'nance requirements.

The SBVF Heater Control Panel is part of the Shield Building Vent System and is classified as Nuclear Safety Related. Since this modification only covers relocation of the Heater Control* Panel (HCP) with no component changes to the panel, the same classification applies, Plant safety and operation are not affected by this change.

l The safety evaluation of this package indicates that the relocation of the HCP does not involve an unreviewed safety question, and does not require a change in the Plant Technical Specifications. Therefore, NRC notification is not required prior to implementation of this EP.

PCM 069-287 SAFETY EVALUATION Pith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed changed shall be deemed to involve an unreviewed safety question; (1) if the probabili.ty 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'n accident or malfunction of a d1fferent type than .any evaluated previously in the safety analysis if report may be created; or (iii) the margin of safety as defined in

,the bases for any technical specification is reduced.

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report will not be increased by this modification since there are no component changes associated with the relocation of the control panel.

Although the flame:test requirements for cables (IEEE 383-1974) were not addressed in the Action Test Report No. 17414 as required by St Lucie Unit 2 FSAR Section 8.3.1.1.4, the cables are in a dedicated conduit from end to end. Therefore, the operation of equ1pment described in the technical specification is not affected.

(11) The possibility for an acc1dent or malfunction of a different type than any evaluated previously in the safety analysis report will not be created by this modification since there is no change in system operation and the new location (a reeinforced concrete wall) is more rig1d than the location for which the CP was originally qualified (the side of the Shiel d Building exhaust fan).

(iii) The margin of safety as defined in the bases for any techn1cal specification is not reduced by this modification since the relocation of the equipment does not alter any circu1ts, and the relocation to a mild environment will reduce maintenance requirements.

The Shield Building Vent- System 1s Class IE (Electrical) and is Nuclear Safety Related, therefore, the Engineering Package (EP) is Nuclear Safety Related.

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

PCM 070-287 REPLACEMENT OF FISCHER AND PORTER CONTROLLERS ABSTRACT This Engineering Package (EP) covers the replacement of the now obsolete Fischer 6 Porter controllers with the currently manufactured and functionally equivalent Fischer & Porter controllers. The controllers are used to maintain the level and pressure parameters in the pressurizer within the required limits during the normal plant operation.

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

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

This EP has no impact on plant safety or operation.

PCM 070-287 SAFETY EVALUATION Vith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety-question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a for an accident or malfuntion of a different type than any 'ossibility evaluated previously in the Safety. Analysis Report may be created, or if (iii) the margin of safety as defined in the bases for any technical specification is reduced.

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

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

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

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

The new controllers replace the obsolete controllers on Class 1E main control board, therefore, this EP is classified Qality Related.'he implementation of this EP does not require a change to the Plant Technical Specifications, nor does it create an unreviewed safety question.

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

PCM 077-287 ERDADS/SAS UPGRADE ABSTRACT This Engineering Package (EP) provides for modifications in Control Room equipment to upgrade the Emergency Response'Data Acquisition and Display System (ERDADS), which is also known as the Safety Assessment System (SAS) and includes Safety Parameter Display System (SPDS) equipment. This EP will improve the performance and display capabilities of the existing system and will include new display (RTs and keyboards and a new color hardcopier.

The Engineering Package is classified as Quality Related since the SAS system is a computer based data processing and display system which assists Control Room personnel in evaluating the safety status of the plant and since the modifications in the Control Room involve installation of'equipment in RTGB-204. Implementation of this PCM does not involve an unreviewed safety question or a change to the Plant Technical Specifications. It can be implemented without prior Commission approval.

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

SUPPLEMENT 1 This EP revision provides for modifications in the Control Room in preparation for implementing an upgrade to the ERDADS/SAS equipment. Included in this work are installation of conduit and cable, relocation of existing ERDADS/SAS equipment, and installation of mounting hardware to allow future installation of ERDADS/SAS equipment.

The Engineering Package is classified as Quality Related since SAS is a computer based data processing and display system which assists'ontrol Room personnel in evaluating the safety status of the plant. Implementation of this PCM does not involve an unreviewed safety question or a.change to the Plant Technical Specifications. 'It can be implemented without prior Commission approval.

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

PCM 077-287 SAFETY EVAL'UATION Mith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in 'the Safety Analysis Report may be increased: or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

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

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated are not increased since the isolation of the ERDADS/SAS equipment will not be existing'nput modified and will maintain the same level of protection for safety-related equipment.

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

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

This EP does not modify equipment which is nuclear safety-related.

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

This EP has no effect on cables or components necessary for safe shutdown of the plant, or on equipment on the Essential Equipment List. Changes to equipment and structures involving 10CFR50 Appendix "R" fire protection requirements have been addressed.= (See Attachment 7.1). Thus, the proposed design is in compliance with applicable requirements for fire protection.

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

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

PCM 079-287

\

EXTRACTION STEAM PIPE AND PITTING MATERIAL UPGRADE ABSTRACT This design'package provides details and instructions to replace degraded carbon steel piping and fittings in the extraction 'steam systems with chromium-molybdenum alloys on. an "as-needed" basis. The extent of the replacement required for each situation will be based on ultrasonic inspection data to be reviewed by Power Plant Engineering during the l987 refueling outage. The required replacement will be reported to construction, and details of each replacement will be added to the package via the CRN process.

This PC/M also provides details'to replace two specific sections of extraction steam piping. These sections are identified for replacement since they are similar in 'terms of design and operating .conditions, to the section of Unit I extraction steam piping which failed during 1986. Theoretical erosion/corr'osion rates indicate that ANSI B 3l.l requirements for minimum wall thickness may be

'iolated during the next one to two power cycles.

This PC/M is classified as "Non-Nuclear Safety Related" since it affects only nonseismic, Quality Group D piping in Non-Nuclear Safety Related Systems.

~ Based on the failure modes analysis and 10 CFR 50.59 review, this modification does not impact any safety related equipment and is.not relied upon for any accident prevention or mitigation. Thus it does not constitute an unreviewed safety question. Since there are no unreviewed safety questions, and since no changes to technical specifications are involved, this PC/M may be implemented without prior NRC approval.

PCM 079-287 SAFETY EVALUATION .

The Unit 2. Extraction Steam System is a Non-Nuclear Safety Related System and as such is not required to fur ction during any existing analyze'd .

accident scenario. Therefore, modifications to these pipes affect only Non-Nuclear Safety Related, Quality Group 0 equipment.

The modification is 'a material upgrade only. The new material has been shown, in the Design Analysis, to meet all design requirements of the previous material.

Postulated failures of the extraction steam line would have no impact on safe shutdown of the plant or safety related systems. The extrac.ion steam lines are not used to prevent postulated accidents, mitigate the consequences of such accidents, maintain safe shutdown conditionsf or adequately store spent fuel.

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

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

Failure of an extraction steam line is not considered as an accident initiating eve'nt or considered in determining the probability of an accident. Also, since this design change does not alter or affect equipment used to mitigate accidents, the probability of malfunction of equipment important to safety remains unchanged.

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

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

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

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

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

PCM 083-287 I

MISCELLANEOUS ICW SYSTEM MODIFICATIONS ABSTRACT This engineering package enables minor modifications to be made to the'Intake Cooling Water (ICW) system resulting from disassembly, inspection, repair and reassembly during the 1987'refueling outage. Those mo4ifications that meet the criteria established by this design, package. shall be initiated via the Change Request/Notice form and 4ispositioned by engineering. Those modifications which do not meet the criteria established by this design package shall be.

implemente4 under separate design packages. Those modifications to the essential portion of the ICW System are classified as nuclear safety related, therefore the PC./h< is classified as safety related. hkodifications to the non-essential portion of the ICW System are classified as non-nuclear safety related unless the failure mode analysis determines an interaction with equipment important to safety. If so, quality requirements will be applied and the modification classified as Quality Related. The changes proposed in this design package are acceptable from the standpoint of nuclear safety, do not involve an unreviewed safety question, do not require a change to the Technical Specifications and do not require prior NRC approval prior to implementation.

SAFETY EVALUATION The modifications to the essential portion of the ICW system described in the project scope are classified as nuclear safety-related because the failure of the modified component in conjunction with the ~orst case single failure as analyzed per FSAR Table 9.2.2 would result in the inability of the ICW system to achieve its design basis safety function. Historically, the types of modifications to the ICW System resulting from- the disassembly and reassembly of the piping system for inspection and repair have been:

1. Modifications to pipe vent and drain lines (e.g., replacement of corroded material).

2. Modifications to support/restraints (e.g., documentation of weld symbols required to reassemble S/R's, excessive gap at S/R base plates, replacement of corroded material).

3. Weld repair to ICW pipe (e.g., documentation of pipe welds).

4. Pipe flange bolting material changes or bolt torque valve documentation.

As described in the design bases, these nuclear safety-related modifications shall be made in accor4ance with the design code requirements for Safety Class 3 pipe and pipe-components and for Seismic Class I support/restraints.

PCM 083-287 In accordance with the requirements specified in the design bases, each modification to the non-nuclear safety-related portion of the ICW system shall have a failure mode evaluation performed to determine if there are any. interactions with safety-related equipment or functions. Since the mon-nuclear safety related portion of the ICW system is not relied upon for any accident prevention or mitigation, failures which are determined to not impact., the function of the 'nuclear safety-related portion of the ICW system are acceptable with regard to nuclear safety. No Quality Related requirements will be applied to the design of these modifications.

However, if a modification to the non-nuclear safety-related portion of the ICW system is determined by the failure mode evaluation to interact with Nuclear Safety Related equipment', Quality Related requirements will be applied to the design of these modifications.

Based on the above, it can be demonstrated that an unreviewed safety question as defined by IOCFR50.59 does not exist.

The probability of occurrence or the consequences of a Design Basis Accident (DBA) evaluated in the FSAR is not increased because no DBA's deal with specific ICW component failures. The modifications restore the ICW system and original design condition and ensure its safety function will be performed.

The probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased because the modifications proposed by this design package are to passive. components only and they will be designed/implemented in accordance with safety class/FSAR requirements. The FSAR does not evaluate passive 'component failures.

iii) The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created because the modifications permitted by this design package do not alter the ICW system function or mode of operation, The FSAR evaluation of the ICW system envelopes the failure of the described modif ied corn ponents.

iv) The margin of safety as defined in the basis for a technical specification is not reduced. The modifications permitted by this design package have been reviewed and found acceptable. No changes to the design basis, function, or mode of operation of the ICW system is proposed IOCFR50.59 allows changes to a facility as described in the FSAR if an unreviewed safety question does not exist and if a change to the Technical Specifications is not required. As shown in the preceeding sections, the change proposed by this design package does not involve an unreviewed saf ety question because each concern posed by 10CFR50.59 that pertains to an unreviewed safety question can be positively answered since the PC/M returns the ICW system to its design condition and no Technical Specification change is required.

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

PCM 086-287 CONDENSER OUTLET TUBESHEET AND WATERBOX COATINGS ABSTRACT This engineering package addresses the addition of an epoxy coating to the condenser outlet tubesheets and waterboxes. This modification will enhance the corrosion resistance of the tubesheets and waterboxes and allow reduction of the cathodic protection system potentials and current densities.

The condensers and the circulating water system are classified as non-nuclear safety related. A safety evaluation and .failure mode evaluation has determined 'that the modification addressed in this engineering package does not consti.tute an unreviewed safety question as defined in 10 CFR 50.59. Furthermore, the addition of a protective coating to the condenser outlet tubesheets and waterboxes does not require a change to the plant Technical Specifications.

PCM 086-287 SAFETY EVALVATIOH As noted in FSAR Sections 9.2.1 and 10.4.1, the condensers and circulating water system perform no nuclear safety related function.

A failure mode valuation of the proposed condenser outlet tube sheet and water',ox coatings has determined there is no potential for interaction with equipment or functions important to nuclear safety. Accord';gly, the modification addressed by this engineering package is class. fied as non nuclear safety related.

Based on the abo:e evaluation and information supplied in the design analysis, it has 3een demonstrated that an unreviewed safety question as defined by 10 CFR 50.59 does not exist.

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

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

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

Ho new accidents or malfunctions associated with the failure of the condenser outlet tube sheet and waterbox coatings have

.been created.

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

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

In conclusion, the modification proposed in this engineering package is acceptable from the standpoint of nuclear safety, does not involve an un r eviewed safety question and does not require a change to any Technical Specifications. Accordingly, NRC approval prior to implementation is not required.

i PCM, 089-287 REMOTE REACTOR VESSEL LEVEL INDICATION ABSTRACT This Engineering Package (EP) is for the modification of the Remote Reactor Vessel Level Indicator. This modification will provide more reliable level indication during refueling and reduce personnel "radiation exposure since it replaces a temporary system which required more attention for operation.

The modifications considered in this EP are on the Reactor Coolant System. %he connections are'esignated as nuclear safety related and seismically qualified because they are within the Reactor Coolant Pressure Boundary and therefore this modification is classified as safety relate'de The instrument side of the system downstream of the piping isolation .valve is non-safety, seismic design. The safety evaluation has shown that this EP does not constitute an unreviewed safety question and prior NRC approval- is not required for implementation.

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

PCM 089-287 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question, '(i) if the probability of occurrence or the consequences of an accident or malfunction 'of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in 'the safety analysis report may be created; or if (iii) the margin of safety as defined in the bases for any Technical Specification is reduced'he modifications included in this Engineering Package are for the Reactor Vessel water level indicator installation involving piping, t'ubing, valves and orifices and differential pressure transmitters, all connected between the RCS. and the Pressurizer.-

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

i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to,safety previously evaluated in the safety analysis report is not increased since this modification provides 'a means whereby an accurate reactor vessel water level can be readily determined during refueling.

During power operation this system is isolated from the RCS. The portions of this modification within the normal RCS pressure boundary have been designed to the original requirements of the RCS pressure boundary, ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated because the modification provides double isolation valving which will isolate the system from'the RCS during power operation.

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

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

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

PCM 091-287 REACTOR HEAD TORUS RING MODIFICATION ABSTRACT This engineering package covers the modification of,the reactor head torus ring. The modification of the reactor head torus ring will enable the ring to be used as an air distribution header for the stud tensioner tuggers.

The reactor vessel head lifting rig assembly including the torus ring is non-safety" related. The lifting rig is operated near safety-related equipment including the reactor vessel. Failure of the torus ring during operation of the lifting rig could potentially damage fuel or nearby safety-related equipment. Seismic design criteria as well as the requirements of NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants", are applicable to the subject modification. Due to the factors mentioned above, quality-related requirements are applied to this design.

A safety evaluation of, this modification has been performed in accordance with 10CFR50.59. This evaluation indicates that implementation of this Engineering Package does not involve an unreviewed safety question. Furthermore, the implementation of this modification does not require a change to the plant Technical Specifications and has no detrimental effect on plant safety and operation. Therefore, prior NRC approval for impiementation of this modification is not required.

PCM 091-287 SAFETY EYALUATION 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 oJ the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

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

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

(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Updated Safety Analysis Report will not be created by this modification because the added nozzles are welded and made as part of the pipe assembly which does not perform a safety-related function.

(iii) The margin of safety as defined in the bases for any technical specification is not reduced since the modification does not require any revision to any technical specifications.

The reactor vessel head lifting rig assembly including the torus ring is non-safety related, The lifting rig is operated near safety-related equipment including the reactor vessel. Failure of the torus ring during operation of the lift rig could potentially damage fuel or nearby safety-related equipment. Seismic design criteria as well as the requirements of NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants", are applicable to this modification. Due.to the factors mentioned above, quality-related requirements are applied to this design.

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

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

PCM 092-287 REPLACEMENT OF SAFETY RELATED BATTERIES 2A AND 2B This Engineering Package covers the modifications to the Safety Related Station, Batteries 2A and 2B which are part of the 125V DC Distribution System.

This Engineering Package will provide the engineering and design details required to implement the replacement of the existing batteries with new batteries. The existing batteries are showing signs of degradation (the battery acid is contacting the copper posts). The new batteries will also have an increased spare design margin (capacity) of 15X over the existing batteries, which were installed in the early 80s, for future load growth capability.

The station batteries, which are part of the 125V DC system, are classified as Class 1E, are seismically qualified and perform a safety related function.

This EP will be classified as Nuclear Safety Related.

This EP does not constitute an unreviewed safety question since the modifications described above were reviewed in accordance with 10CER50.59. and were determined to have no adverse impact on plant operations or safety related equipment.

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

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

SUPPLEMENT 1 Supplement 1 removes the holdpoints that were established (environmental and seismic reports received), adds and revises calculations and adds vendor drawings/manuals to PfDRAC system. The original safety evaluation is not affected by this supplement.

SUPPLPfENT 2 Supplement 2 incorporates CRN's ("as found" field. dimensions, seismic qualification note, and substitution of cables), revision to drawing revised vendor "F4 and Seismic Report" and additional calculations.

list, The original safety evaluation is not affected by this supplement.

PCM 092-287 SAFETy EVALUATION liith 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 acc'ident or malfunction of equipment important .to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced'his Engineering Package provides the engineering and design details required to implement the repl'acement of the existing batteries with new batteries. The existing batteries are showing signs of degradation which could reduce the capacity of the battery cells.

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

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

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

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

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

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

PCM 092-287 The implementation of this PCM does not require a change to the plant Technical Specifications.

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

PCM 096-287 PRESSURIZER INSTRUMENT NOZZLE REPLACEMENT hBSTRhCT This Engineering Package provides the design for the replacement of the four,, one inch, steam space pressurizer instrument nozzles. The existing nozzles have been fabricated from a heat . of Inconel 600 that has been shown to be susceptible to a form of intergranular stress corrosion cracking (IGSCC). hn analysis of the different environments experienced by the nozzles fabricated from this heat has determined that the nozzles with the greatest potential for development of IGSCC are those located in the pressurizer steam space. The replacement nozzles are identical in form, fit and function to the original nozzles with the exception that specific parameters for the Inconel 600 material are more closely controlled to significantly reduce susceptibility to IGSCC.

This Engineering Package is classified as nuclear safety related since it replaces the steam space instrument nozzles which are attached to a safety related component, the pressurizer, and are part of the reactor coolant boundary. The safety evaluation has shown that this Engineering Package does not constitute an unreviewed safety question nor does it require a technical specification change.

Therefore, prior NRC approval is not required for implementation of this PC/M.

This Engineering Package has no adverse impact on nuclear plant safety and/or operation.

Revision 1 This revision removes the construction hold point for material approval, changed the minimum heat treatment temperature from 1800 to 1750 degrees Fahrenheit in order to obtain the minimum hSME Section II yield strength requirements for the SB-166 portion of the replacement nozzle, and modified the hLhRA/implementation statements to incorporate plant comments. The safety evaluation has shown that this revision to the Engineering Package does not constitute an unreviewed safety question nor does it require a t chnical specification change. Therefore, prior NRC approval is not required for implementation of this PC/M.

PCM 096-287 SAFETY EVhLUhTION 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 accident or malfunction of equipment important to safety previously an evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or if (iii) the margin of safety as defined in the basis for any technical specification is reduced.

This engineering package replaces the four steam space pressurizer instrument nozzles with identical nozzles in design, dimensions, weightf and hSME Section II material specifications. This modification is considered to be safety related and does not involve an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the replacement of the nozzles will not impact the operation of the pressurizer, affect

~

downstream instrumentation or affect the parameters measured by such instrumentation..

There is no possibility for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operational design of the pressurizer and.

the new nozzles are equivalent in design.

(iii) This modification does not change the margin of safety as defined in the bases for any Technical Specification because the pressurizer nozzles are not included in any Technical Specification bases.

Implementation of this P/CM does not require a change to the plant Technical Specifications.

The foregoing constitutes, per 1QCFR50.59(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 P/CM is not required.

PCM 102-287 RCA PROTECTIVE CLOTHING BINS ABSTRACT This engineering package is being issued in response to a request from Plant Mechanical Maintenance. This package will provide the engineering documentation required for the installation of protective clothing bins in the Reactor Auxiliary Building {RAB) and the Fuel Handling Building (FHB). The bins are being installed to provide convenient locations for distribution of protective clothing, and to replace mobile carts currently used.

The protective clothing bins do not perform or affect any safety related function. However, this PC/M is classified Quality Related to provide the Q.C.

inspections necessary to ensure the location and installation of the bins are in accordance with the provisions of this engineering package: Quality Related requirements are applied to this modification.

Th'C/M is does not constitute an unreviewed safety question. The entation of this PC/M does not require a change to plant technica 1 specifications. This modification does not affect plant operations or sa ~et e y.

Based on the above, implementation of this PC/M does not require prior NRC approval.

ECM 102-287 SAFETY EVALUATION Safet Anal sis With respect to title 10 of the Code of Federal Regulations, Part,50.59, a proposed change shall be deemed to involve an unreviewed safety question:

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

The protective clothing bins do,not perform or affect any safety related system or function. However, this PC/M is classified as quality related to ensure Q.C. inspection of the installation.=

Consequently, the storage bins and support structures have been analyzed for the design basis conditions specified in the FSAR and Quality Related design requirements have been implemented, thus assuring the integrity of the installation.

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

(i) The probability of occurrence or the consequences of an accident or malfunction of, equipment important to safety previously evaluated is not increased since this modification will have no effect on equipment required to shut down the plant and monitor the plant in a safe shutdown condition.

(ii) -There is no possibility for an accident or malfunction of a different type ,than any previously evaluated since the protective clothing bins perform no safety function and no changes have been made to any operational design. Failure of the support structures could not occur since the modification has been designed for the design basi's conditions, (iii) This modification does not change the margin of safety as defined in the basis for any technical specification since installation of the protective clothing bins does not effect the basis for any technical specification. The implementation of this PC/M does not require a change to plant technical specifications.

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

PCM 103-287 TSI THRUST BEARING PROBE RELOCATION ABSTRACT This Engineering Package (EP) is for the relocation of the Turbine Generator thrust bearing probes from their existing location to the original Westinghouse probe location. These probes function as position detectors,.that is, to monitor the shifting of the rotor with respect to the thrust bearing.

This EP is classifed as non-safety related since these probes neither perform any safety function 'nor do they interact with safety related equipment. The safety evaluation has determined that this EP does not constitute an unrevi'awed safety question and implementation of this EP does not require a change to the Technical Specification. Therefore, prior NRC, notificati'on for implementation of this EP is not required.

This EP has no impact on plant safety and operation.

PCM 103-287 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction o'f a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Techn1cal Specification is reduced.

This EP is for the relocation and replacement of thrust bearing Bently Nevada probes, with similar shorter length probes, from their existing locat1on to the original Westinghouse probe location. The'modification implemented via this EP neither performs any safety function nor does it interact with safety related equipment, therefore this package is classified as nonnuclear safety related.

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

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The Turbine Supervisory Instrumentation are not,used in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.

(11) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created'he components involved 1n th1s mod1f1cation have no safety related function and no changes have been made to the operational design of the system.

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

The implementat1on 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 provided the bases that this change does not involve an unreviewed safety question or a change to the Plant Technical Specifications and prior Commission approval for the implementation of the PCM is not required.

PCM 111-287 MISCELLANEOUS SNUBBER MODIPICATIONS ABSTRACT Thi En ineering Package (EP) provides the engineering and design '

information for typical modifications to snubbers which may q f th I service Inspection findings. The anticipated typical modifications are expected to be replacement of the ex s g bb ith nubber on a one-for-one basis, of equivalent capacity of the same or a different style (e.g., replacement oof a Pacific Scientific snubber with an equivalent Anchor/Darling snubber).

This EP has been classified as Safety Related because the snubber modifications may affect nuclear safety related piping systems. The safety eva evaluation ua on has a determined that this EP does not, involve an unreviewed safety question, and implementation tion oof tthee EP does not require a c h ange too Plant Technical Specifications. Therefore, prior NRC approval is not required for implementation of this EP.

Modifications other than the typical ones shall be reviewed individually d i if the involve an unreviewed safety question as defined by 10CFR 50.59 or if they will affect any Technical S peecifications.

Documentation of these reviews shall be included in revisions to this EP.

Modifications performed under this EP shall be documented via Notices (CRNs) and/or revisions to this EP.

Change'equest A final revision may be issued, implementation of o if deemed necessary, t s PCM to include the following: a summary of all this after the th e modifications included in the pro)ect ct scope, sco e, aaffected ec e Support/Restraint Mark Numbers, documents, affected drawingsp TEDB and changes related to other sections of this EP.

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

PCM 111-287 SAFETY EVALUATION-With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed

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

This EP provides typical modifications to snubbers which may be found necessary during the Inservice Inspection of snubbers. These typical modifications are limited to the replacement of snubbers or components with snubbers and components of equivalent load rating. Since modifications may affect Nuclear Safety Related piping systems, these'typical this EP is classified as Nuclear Safety Related.

This EP has been determined not to involve an unreviewed safety question, based on the following:

(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased since the restraints for the piping will remain functionally identical to the existing configuration. In addition, since the restraint configurations are not changed, all ,previous analyses and conclusions are still valid.

The possibility for an accident or malfunction of a different type than any evaluated previously in a safety analysis report is not created because the system remains functionally identical to the configuration depicted in the existing stress analysis of record. Also, the affected restraints have been qualified to the same code requirements as those they replace.

(iii) The margin of safety as defined in the bases for any Technical Specification is not affected by this modification because the replacement components utilized perform the same restraining function as those they replace.

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

PCM 114-287 LOAD CENTER XFRMR VALVE PACKING MODIFICATIONS ABSTRACT: Load Center Transformers (2B2, 2A5, 2B5) radiator shut-off valves (Tranter Valves) have leaked silicone fluid. This change documents modifications performed to the Tranter Valves'36 Total) stem packing to prevent the leakage.

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

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

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

s A test or experiment not described in the.FSAR?

A change to the plant technical specifications?

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

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

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

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

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

Yes . No May the possibility of a malfunction of equipinent important to safety different than any already evaluated in th'e FSAR be created?

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

PCM 115-287 INSTRUMENT AIR DRYER AFTERFILTER ISOLATION VALVES AND BYPASS LINE This Engineering Package covers installation of isolation valves and a full flow bypass line with a valve on the Instrument Air (IA) dryer afterfilter. These modifications will provide for isolating the afterfilter while maintaining IA System operation. Provision for isolation capabilities is required to facilitate installation of the IA upgrade modification (PC/M 051-286), during plant operation.

This modification covers equipment located in the IA System which is- classified as Non-Nuclear Safety Related. Based on the failure modes evaluation and 10CFR 50.59 review, this modification does not impact any safety related equipment or functions>is not relied upon for any accident prevention or mitigation, and does not impact plant safety.

This EP does not constitute an unreviewed safety question and is correctly classified as Non-Nuclear Safety Related. In addition, this modification does not require changes to the Technical Specifications. 'Implementation of this modification, therefore, does not require prior NRC approval.

The subject modification govides for installation of inlet and outlet isolation valves and a full flow bypass line with a vtave to the IA System afterfilter. As defined in Section 9.3 of the FSAR, this system is considered Non-Nuclear Safety Related, Quality Croup D and is not required to perform a safety function. These modifications are therefore considered Non-Nuclear Safety Related. Based on the failure modes evaluation, as provided in the Design Analysis, failure of the IA System has no effect on Nuclear Safety.

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

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

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

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

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

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

This change has no effect on any existing Technical Specifications.

In conclusion, this modification is acceptable from the standpoint of nuclear safety since it does not involve an unreviewed safety question, as defined by 10 CFR 50.59, and does not require changes to the Technical Specifications. Implementation of this modification does not require prior NRC approval.

PCM 117-287 CONDENSER EXPANSION JOINT IMPINGEMENT PLATE MODIFICATION ABSTRACT: The existing impingement plate design is inadequate for satisfactory long-term performance. Welded attachments on the plates have continuously failed, causing plates to fall on the damage condenser tubes. The new plate design will involve no welding and will aU.ow plate installation which will prevent any further failure.

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

TYPE OF CHANGE Yes No iQ A change to the plant as described in the FSAR'i Yes No~~ A change to procedures as described in the FSAR?

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

Yes .

No ~r' change to the plant technical speciiications?

EFFECT OF CHANGE Yes No ~Y Vill the probability of the FSAR be increased?

an accident previously evaluated in Yes No Will the consequences of an accident'previously evaluated

~r'es in the FSAR be increased?

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

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

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

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

Yes No~r Vill the margin of safety as defined in the bases to any technical specification be reduced?

PCM 118-287 v

CABLE SUPPORT STRUCTURE CONNECTION MODIFICATION ABSTRACT: Reinstallation of the reactor bead cable support structure at the end of each refueling outage requires the replacement of connection bolts. These bolts,. specified on drawing 2998-B-791,

.SH.11, are unique and must be special ordered., This modification will specify standard connection bolts which are readily available. This modification does not involve an unreviewed safety question and no technical specification changes are required.

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

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

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

Yes No~ A test or experiment not describedin the FSAF?

Yes .. No~ A change to the plant technical specifications?

EFFECT OF CHANGE Yei No ~ Will the probability of an accident previously evaluated in the FSAR be increased?

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

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

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

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

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

Yes No~ Will the mar'gin of safety as defined in the bases to any technical specification be reduced?

PCM 120-287 GROUTING OF MASONRY BLOCK WALLS ABS'IR ACT In the course of preparing the Fire Protection Appendix of the Unit 2 FSRR, a concern was raised as to whether certain masonry block walls assumed to be 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barriers are actually grout filled. A safety evaluation was performed (Reference 6.5) which established that, if these walls are in fact not filled with grout and therefore not providing the full 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of fire protection, the plant still maintains its ability to achieve safe shutdown.

This safety evaluation recommended that an inspection of these walls be performed to establish their's-built condition. Such an inspection was recently performed and concluded that the walls are not fully grouted.

This Engineering Package (EP) provides the details/requirements for pressure grouting the voids in block walls 127, 128, 129A, 129B, and 137. The remaining walls will be addressed in a future revision to this EP.

This modification does not involve an unreviewed safety question, has no effect on plant safety and operation, and does not involve a change to Technical Specification. Upon completion of this modification, the any'lant

. action in Technical Specification 3/4.6.12 will no longer be required for the walls modified. This EP is classified Quality Related since all of the walls involved are seismically designed and required per 10 CFR 50 Appendix R to be fire barriers.

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

When a concern was raised that the walls modified by this EP might not be fully grouted, a report (Reference 6.5) was written to evaluate the safety implications if the walls were found to not be if fully grouted. This report demonstrated that, an ungrouted condition was confirmed, no unreviewed safety questions exists and continued operation of the plant is justified. This EP provides the details/requirements for pressure grouting the walls so that they are in conformance with the design bases established in Subsection 3.11.2 of the St Lucie Unit 2 FSAR Appendix 9.5A; consequently, this modification cannot give rise to an unreviewed safety question.

Although the walls do not perform a safety-related function, this EP is classified Quality Related, since failure of the walls could damage safety-related equipment.

PCM 120-287 Baaed on the above,'he following provides, the justification that an exist:

The 'f unreviewed safety question does not probability occurrence or the consequences accident or malfunction of equipment important to safety-of an previously evaluated is not increased; since these walls are seismically designed; no accidents due to structural failure are postulated. The only other type of accident potentially associated with these walls involves damage that could occur if the walls fail to provide three hours of fire protection.

The JCO discussed above, however, demonstrated that no single fire event could impair the plant's ability to achieve safe shutdown. Consequently, there - are no .accidents or malfunctions of equipment important to safety previously postulated which involve these block walls.

There is no possibility for an accident or malfu'nction of a different type than any evaluated previously, since the modification provides the walls with a three hour fire rating while the design ensures that the seismic integrity of the walls is maintained.

iii This modification does not change the margin of safety as defined in the bases for any Technical Specification. The basis .for Technical Specification 3/4.6.12 indicates that fire barriers ensure that fire damage will be limited and the ibili f a single fire event involving more than wi'll one fi a ea minimized.

prior to'etection and extinguishment w The referenced JCG indicated that the curr bee ent it s u& tion in combination with compensatory measures, does not violate this basis. When the walls are fully grouted, barriers will be fully operational, eliminating the need for the said compensatory measures.

The implementation of this PC/M does not require a change to plant technical specifications.

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

PCM 121-287 STEAM GENERATOR TUBE PLUGGING/CE DESIGN PLUGS

~BS'~lCT This 'PCM docu'ments Engineering review and concurrence for the use of Combustion Engineering expanded type plugs in the St. Lucie Unit 2 steam generators; This PCM also provides the information necessary to as-build affected documents.

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

Based upon a failure mode evaluation and 10 CFR 50.59 review, this modification does not involve an unreviewed safety question nor require changes to the technical specifications. Therefore, prior NRC approval is not required for implementation of this modification. The modification has no adverse affect, on plant safety or operability.

PCM 121-287 modification involves documenting the maintenance practice of plugging steam generator tubes. Steam generator tubes are nuclear safety related, therefore this engineering package is classified as nuclear safety related. 'Ihe PC/M pravides engineering concurrence for the use of the Gcxtibustion Engineering expanded tube plug design (previously utilized on the Unit 1 steam generators), the required 50.59 review of the modification, and the information requu~ for update of affected donments.'0 CPR 50.59 allows a change to a nuclear facility without prior NRC if appraval if an unreviewed safety question does nat exist and ch ages to Technical Specifications are nat involved. 'Ihe followirg argunents demonstrate that an unreviewed safety question does nat exist relative to this modification:

not increased since this modification does nat decrease the design margin of the'CS pressure bound-uy (the tube plugs meet or exceed all design requireaents for ASME Section III, Class 1 canponents) .

'Ihe consequences of a previously postxQ.ated design basis accident are nat made mare severe for the same reasons given in (i) arxl since no existing accident mitigation equipnent or sy.~ns are altm~ by this modification.

iii) 'Ihe possibility of an accident of a different type than previcasly,aCkhessed in the FSAR does nat exist since no new systems or equi~t are intnduced by this modification.

Failure of a tube plug would be no more severe than a steam generator tube rupture, a previously evaluated condition.

'Iherefore, no new accidents are created.

iv) The margin of safety as defined in the basis for any technical specification is nat reduced since the total number of tubes plugged in the steam generators following this modification is less than asauned in the Cycle Four Reload Analysis.

Since the abave arcpments demonstrate than an. unreviewed safety question does not exist, and since a revision to the Technical Specifications is not requuM, the addition of the CmkMstion-Ergineering tube plugs to the Unit 1 steam generators does not recpdre prior NRC approval.

PCM 122-287 2A/B SPARE STEAM GENERATOR INSTRUMENT NOZZLES CLOSURE MODIFICATION ABSTRACT: The exi'sting blank flange connections have degraded, resulting in steam leaks. In order to prevent further leakage, and since: the nozzles are no longer required, the flanges will be removed and welded caps will be installed. No unreviewed safety questions exist as defined by 10 CRF 50.59, and no Technical Specifications are affected by this modification. Therefore, prior Commission approval is not required.

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

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

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

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

Yes . No X A change to the plant technical specifications?

EFFECT OF CHANGE Yes No Will the probability oi an accident previously evaluated.in...,.

the FSAR be increased?

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

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

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

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

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

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

PCM 126-287 INCORE INSTRUMENTATION (ICI) PLATE MODIFICATION ABSTRACT This Engineering Package covers repair modifications to the Incore Instrumentation (ICI) Plate Assembly to correct local deformations in the plate. The major features of this package include the repair process and the acceptability of the repaired ICI plate for continued service. Included is information pertaining to the repair tooling, repair process procedures, repaired plate acceptance criteria and an assessment of the safety impact of continued use of the repaired ICI plate. '

Work performed under this Engineering Package has been classified as Quality Related. This classification was selected because the repair work discussed herein was conducted over the reactor vessel and its internal components. Further, during operation the ICI plate has the potential to interact with safety related components. The ICI plate assembly itself, however, is a non-safety related component since it neither prevents nor mitigates the consequences of accidents. Changes in the ICI plate

-configuration resulting from the repair will have no effect on its design function.

Based on a Failure Modes and Effects Analysis (FMEA) and 10CFR50.59 review, the repair modifications implemented by this engineering package are acceptable from the standpoint of nuclear safety as it does not involve an'nreviewed safety question and does not change plant Technical Specifications. As such, prior Nuclear Regulatory Commission approval is not mandatory to implement this Engineering Package.

Su lement l This supplement revises only the design interface record of'he package to add two signatures. This supplement does not affect the conclusions of the safety evaluation that the 'hange does not involve an unreviewed safety question and does not change plant Technical Specifications.

0168L

PCM 126-287 SAFETY EVALUATION This Engineering Package has considered the safety related consequences of the repair process proposed to straighten the deformed Zncore Instrumentation (ZCI) Plate Assembly at St. Lucie Unit 2 and its subsequent return to service. The package has been classified Quality Related due to the location of the repair (over the reactor vessel) and the potential for interaction with safety related components. The ZCI plate assembly itself, however,unreviewed performs no safety related safety question or function. No Technical Specification change was identified as a result of these considerations. As such, it been concluded that the, repair of the plate and has its return to service can proceed as planned.

In order to straighten the deformed ICI plate, it was necessary to develop repair tooling which could be used to reverse the load path on the plate that originally caused the deformation.

This tooling took the form o f various pieces o f hardware which could be used to grip or push on the plate and apply the necessary loads. The loads applied were aimed at either straightening the plate in a global sense by pushing down at its center or by loading the locally deformed areas to bring them back into proper alignment. Zn practice, only the center push tool had to be employed. All jacking load paths for the plate bending process were through the ZCI plate itself and the UGS .lift rig and did not involve any other plant structures or reactor vessel internal components. Straightening of the QCI plate was actually accomplished through two"center push operations. Each operation was accomplished over a series of controlled increments.

Zn making its determination regarding the safety aspects of this process, C-E considered containment integrity, shutdown cooling system operation, fuel damage, impact on, plant structures, loose'articles due to grinding operations , heavy loads, fire hazards and the acceptability of repair hardware materials. The details of this evaluation were transmitted (L-MPS-87-033) to FPL prior to the repair and are included with this package in Attachment 7.

After straightening the ZCZ plate, it was discovered that one of the T-brackets which

PCH 126-287 supports incore detector guide tube runs interfered with the outside wall of an adjacent Control Element Assembly (CEA) guide tube shroud during, the lowering of the ICI plate. The interference was sufficient to prevent the ICI plate from seating completely.

A plan to reduce the T-bracket deformation and eliminate the interference was developed. The corrective action called for a suitably rated chain-fall (or come-along) to be attached to the T-bracket near its top and to a. bracket mounted on the southwest corner of the "A" steam generator biological shield wall. FPL determined that this bracket was capable of withstanding a load substantially greater than the maximum 5000 pound load limit for the T-bracket corrective procedure.

To faciliate bending of the T-bracket, a relief slot was cut in the vertical leg of the bracket.

The'slot allowed the T-bracket to undergo a one-time local plastic deformation when tensioned with the come-along. In this manner, the T-bracket was pulled into a more upright position eliminating the interference.

To support the feasibility of this procedure to straighten the T-bracket; a test was performed on a similar structure at C-E's Windsor Test Facility on Nov. 6, 1987. A T<<bracket was deformed approximately 1/4 to 1/2 inch by applying a load of less than 3500 pounds. In the St. Lucie Unit 2 plant, the actual load was applied at an angle of approximately 35 above vertical because of access limitations in the work area. The C-E test demonstrated that the T-bracket could be deformed without affecting adjacent ICI plate structures (e.g., guide tube clusters) .,

The in plant procedure was performed with the ICI plate raised and supported by the UGS lift rig and with the ICI plate compressed by the center push tool applying a load of approximately 8000 pounds.

An analysis was performed to evaluate areas of potential concern:

a) Motion Within the Lift Rig Analysis showed that the vertical component of the bending force was less than the seating force at each leg of the lift rig. Therefore, lift off was not anticipated and tipping could not occur.

PCM 126-287 since the moment balance about a potential tipping point was stable.

b). Lift Rig Column Bending (Lift rig in bending and shear)

Analysis showed the bending stress to be below the minimum material yield strength of 30 Ksi. As such, column bending was not a concern. Combined cable side and vertical loading due to dead weight and center jacking stabilizing loads were combined. The total stress. in the rig leg column was < 8000 psi (i.e.,

lift combined bending and axial shear).

c) UGS Stability A moment balance about the base of the UGS showed that the system was stable and would not tip under the maximum applied 5000 pound pulling load.

d) Acceptability of Slotting T-Bracket The flow loads in the relatively isolated upper head region were reviewed and found to be below those required to adversely impact the slotted T-bracket.

The pull will not affect the ability of the T-bracket to perform its design function of supporting the instrumen't guide tubes in. any way. Further, the.

slotted and straightened bracket will not interfere with the function of any other components in the area of the ICI plate.

During the repair, it was observed that the surface of the ICI plate covered with metal chips in the localized area in which grinding of the locating pins took place'.

The chips were characterized by FPL divers as ranging from fine particles to slivers of between 1/8 - 1/4 inch in length (maximum) with a thickness of less than

'25 mils on average. The surface density of the chips was reported as approximately 10 - 15 chips per square inch in the grinding area falling off to sparse coverage in areas outside the grinding location.

If the chips enter the reactor vessel, the potential for interference with control rod motion and possibility of degraded core thermal margin and/or fuel damage can be postulated. 1n order for the small metal

PCM 126-287 chips to impede control rod mot'ion they would have to become wedged between the control element assembly (CEA) and the guide tube. The diametral gap between the stainless steel sleeve and the CEA is 54 mi'ls and the metal chips are less than 25. mils. Therefore, it is very unlikely that any chips that happened to fall into the guide tube could cause the CEA to jam.

Degraded core thermal margin could only occur if significant flow blockage were to occur as a result of a

the metal chips. Critical heat flux tests have been run in which flow blockages were simulated. Blockage of 11 out of 34 subchannels of a 5X5 test section showed little effect on bundle critical heat flux capability compared to similar tests without the blockage. Considering the distribution and sizes of chips observed on the ICI plate, it is concluded that is is not possible for a sufficient quantity of chips to agglomerate in-one local region of the core and block the inlet of more than 11 subchannels.

Therefore, no premature DNB due to subchannel inlet flow blockage is expected. It is conceivable that a small number of chips could become trapped at other spacer grids above the core inlet. Again, the total blockage that might occur would be extremely small for any one subchannel and would not be expected to adversely impact DNB margin.

The potential for fretting-induced fuel failure due to the chips entering the active fuel region, however, does exist. It is not likely., though because of their small size. Any fretting-induced fuel failures would occur gradually over time, and become apparent to the reactor operator in the form of higher coolant activity levels. Technical Specifications on coolant activity level preclude plant operation at levels which would pose an undue risk to the health and safety of the public. Therefore, with regard to.the issue of control rod motion and DNB performance and fuel damage, the metal chips found on the ICI plate do not create the possibility of an accident or impact the operation of equipment important to safety.

In addition to the repair process, this Engineering Package has considered the acceptability of the repaired ICI plate assembly for continued service.

Evaluation of the loads experienced by the plate during its deformation, during the repair (straightening)g process and during operation ensure that the plates design function were not. adversely impacted.

Using a finite element model of the ICI plate, the.

loads necessary'o cause the observed deformation were

PCM 126-287 determined. This information was used to evaluate 'the strain levels within the plate and its acceptability for repair. Results of'.this investigation indicated that the plate was not strained beyond limits which would preclude a repair process to straighten the deformed areas. This same finite element model was employed to determine the loads and their points of application to be applied in order. to straighten the ICI plate. The strain levels imposed during the straightening process were also evaluated to assure that plate integrity would not be compromised. Using the as-repaired dimensions of the ZCI plate, its acceptability for re-installation in the xeactor vessel was assessed along with its ability to caxry out its design function.

The safety evaluations discussed above were conducted to determine whether any unreviewed safety question or change to Technical Specifications was involved in the

'proposed ICI plate assembly repair process or in its return to service. The overall conclusions, which are elaboratored below, indicate that the use of the special tooling and procedures developed for this repair effort as well as the continued use of the repaired ZCI plate assembly does not involve an unreviewed safety question or require a change to the plant Technical Specifications.. Specifically, the safety evaluation conclusions are that the ICZ plate repair and continued use does not:

1) Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report; As indicated in Attachment 7, the repair process was conducted in such a fashion that containment integrity and shutdown cooling system operation were unaffected, therefore, the safety functions for both systems were not .impacted. Fuxthermore, examinations of the possibility of dropping tools into the reactor vessel or of impacting Heavy Loads considerations during the repair pxocess wexe conducted. The repair process was found to not introduce any additional loads to the UGS lift rig that exceed its design limits. Heavy Loads, therefore, were not impacted in any way by the repair. Furthermore, since the repair work was IC1'late

PCH 126-287 conducted above the UGS, which was, in its normally seated position, there was no increase in the probability of occurrence of fuel damage due to potentially dropped tooling.

The ICI plate and fine alignment pins are passive components and do not provide a safety function. The modification of the fine alignment pins could, however, increase the lateral movement of the ICI plate, resulting in some slight contact- of the ICI instrumentation thimbles and HJTCs. Even did occur, it if would not affect the function of these contact instruments. The potential for increased wear of any of these components was considered insignificant because flow loads in the isolated upper head region are small or non-existent and the ICI plate assembly is heavily damped. by 56 incore instruments and two HJTC probes. Instrumentation guide tube cluster engagement with the reactor vessel head also provides damped restraint. Therefore, the probability of unanalyzed equipment malfunction is not increased, nor is the probability of an occurrence or. consequences of an accident previously evaluated in the safety analysis report increased.

Evaluation of the load to be applied during the pull to eliminate intereference between the T-bracket and the CEA shroud were well within the capability of the UGS lift rig and "A" steam generator biological The modified T-bracket is capable of performing shield'all.

its design function and will not adversely impact any other components in the area, of the ICI plate; As such, the probability of occurrence or the consequences of accidents evaluated in the safety analysis report with respect to use of the the ICI plate are"unaffected.

lift rig or the function of As stated in Section 3.9.5.4.2 of the St. Lucie Unit 2 UFSAR, the ICI plate assembly by itself is not a safety-related component since its satisfactory performance does not prevent accidents nor mitigate'the consequences of accidents that could cause undue risk to the'ealth and safety of the publi.c. Nothing in the, repair process nor the continued use of the ICI plate has any impact on the UFSAR statement.

PCM 126-287

2) Create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report; In assessing the possibility of creating new accidents or malfunctions as a result of the repair process, examinations of pool contamination, fire hazards, loose parts and impacts on, crane handling design capability were conducted. The findings indicate that the grinding operations performed on the two ICI plate locating pins were executed in a controlled manner.

Particles, that could potentially have escaped the vacuuming operations would be of insufficient size to create the possibility of an accident or impact operation of any safety related equipment. Hydraulic fluid used in conjunction with the plate straightening tools has been found acceptable for unrestricted use at St. Lucie Unit 2. With respect to additional fire hazards being generated due to the repair process, only the short time use of a cutting torch (under water) was necessary to cut a slot in the T-bracket. This operation was controlled by FPL in accordance with their fire protection procedures. No fire hazards were, therefore, created. The size and mass of all repair tooling was examined and found to be well within the design loading capacity of the tool handling cranes. Protection against inadvertently losing tooling was ensured through use of lanyards and administrative controls.

The possible affects of the modification to the ICI plate fine alignment pins could have some slight effect on the potential for interaction between the incore instrumentation thimbles and HJTCs and their respective guide tubes. The slight contact that may potentially occur does not adverseley impact the ability of.these instruments to provide their design function. The ICI plate will not interfere with other reactor vessel internal structures in its repaired configuration.

Therefore, the possibility of an unanalyzed accident or malfunction is not created by this modification.

Evaluation of the load to be applied during the T-bracket pull indicated that the UGS lift rig will not tip nor will its columns deform. Similarly, the UGS itself is stable and cannot be tipped by the load to be applied. The integrity of the T-bracket is not compromised by the slot and, in its new orientation,

PCM 126-287 it does not i'nterfere with other design function of the T-bracket components. The in supporting the instrumentation guide tubes will, therefore, be met.

As stated above, the ICZ plate assembly does not have any design function related to the prevention or mitigation of accidents. The evaluations of the its as-repaired ICZ plate indicate, furthermore, that structural integrity is not compromised and that it can continue to provide its design" function. In the extreme, if a hypothetical failure of the plate (complete through plate crack) were assumed in any of the repaired areas, no adverse consequences have been identified by the Failure Nodes and Effects Analysis.

The CEA shrouds extending from the upper guide structure pass through the plate in numerous locations over its surface area and would act to hold any separated segments in place. Further, since the upper head region in which the ICZ plate resides is a relatively low flow area, the probability that any would adverse movement of the plate or segments thereof occur is also extremely low. Being a highly isolated region from the remainder of the reactor vessel internals will preclude the generation of any postulated loose parts from adversely affecting plant operation.

3) Reduce the margin of safety as defined in the basis of any technical specification; An examination of the repair process has not led to the identification of any Technical Specifications that will be impacted. In addition, the modifications to the ICI plate, locating pins and T-bracket during the repair do .not affect the Technical Specification margins for safety because these structures do not serve any safety related function. As.a result, the margin of safety in the bases in plant Technical Specifications will remain unchanged.

Based on the evaluations discussed above, C-E concludes that the proposed corrective actions do not involve an unreviewed safety question or a change to the plant's Technical Specifications. Further, Combustion Engineering has found no reason to preclude the repaired ZCZ plate assembly from being returned to service and being able to carry out .its design function for the remainder o the St. Lucie Unit 2 design life.

PCM 126-287 Attachment 7 L"MPS-87-033 Page 1 of 5 St. Lucie Unit 2 Incore Instrumentation Plate Repair Program Safety Evaluation Summary:

Combustion Engineering (C-E) has considered .the safety r elated consequences of the repair proposed to straighten the deformed incore instrumentation ( ICI) plate assembly at St. Lucie Unit 2.

In making its determination, C-E considered containment integrity, residual heat removal system operation, fuel damage, impact on plant structures, loose particles due to grinding operations, heavy loads, fire hazards and the acceptability of repair hardware materials. No unreviewed safety question nor technical specification change was identified as a result of these considerations. As such, C-E concludes that the repair can proceed as planned.

Discussion:

In order to straighten the deformed ICI plate it was necessary to develop repair tooling which could be used to reve se the. load path on the plate that. originally caused the defor.~ation. This tooling is in the form of various pieces of hardware which can be used to grip or push on the plate and apply the necessary loads.

The loads applied are aimed at either straightening the p'late in a global sense by pushing down at its center or by loading the locally deformed areas to bring them back into proper alignment.

All jacking load paths for the repair process are through the ICI plate itself and the upper guide structure (UGS) lift rig and do not involve any other plant structures or reactor vessel internal components.

An evaluation of the repair hardware or the repair orocess for any adverse safety related consequences was conduc ed ih the following areas:

Containment Inte rit Neither the repair process nor the results of the repair will impact containment integrity in any way. The equipment required is self contained and can be powered from sources inside containment normally used during refueling or plant maintenance operations. No outside containment support is required for the operation of repair hardware. Normal access hatches will be used only to bring in and remove equipment. No breaches of containment integrity, therefore, will be required before or during the repair.

PCN 126-287 Page 2 of 5 Shutdown Coolin S stem 0 eration The repair process does not impact the shutdown cooling system (SDC) operation. The SQC system will be in its normal refueling configuration throughout the repair process. The repair process will have no direct interaction with the SQC system. All repair actiors will be preformed in the refueling pool above the normal elevation of the ICI plate and above the reactor vessel.

Fuel Oama e There is no increased potential to damage fuel during the r'epair process. The size and mass of repair tools and equipment are within the envelope of tools and equipment normally used for work in this area. 'he repair work will all be conducted above the upper guide structure which is located so as to preclude any dropped equipment from reaching the core region. As such, the repair evolution will not result in any risk of damage to the fuel. Further considerations regarding the potential-for fuel damage are discussed in relation to heavy loads below.

Loose Particles Grindin Grinding of any ICI plate components as part of the repair evolution will be done in a controlled manner. A vacuum system will be used to minimize the potential for particles entering the refueling pool water and possibly making their way into the reactor vessel or primary coolant system. Any rancom particles not captured by the vacuum system would be of insufficient size or volume to create a safety hazard.

Impact of Tools on An Plant Structure The size and mass of the repair tools are within the capabilities of the cranes to be used in support of the repair process. All tools are hand-'pump oper ated metal working tools wh',ch operate independent of other plant systems. The recommended repair process does not involve any repair equipment coming into contact with any of the plant structures other than the ICI plate and the UGS lift rig. Tools will be administratively controlled and procedui es will include the use of lanyards to assure that no tools are inadvertently lost in the refueling pool. No adverse impact, therefore, is expected as a result of the repair process or any postulated tool or equipment failure.

Acce tabi lit of Material Bein Put in Pool Mate~

No material will be introduced into the refueling pool water that is not typically used at refueling outages. Repair tooling is fab~icated from carbon steel and stainless steel. The hydraulic

PCH 126-287 Page 3 of 5

fluid that will be used for the repair tooling has been evaluated and found to be acceptable for unrestricted use in C-E desigred

NSSSs. The potential for contamination of the water or for adverse reaction with reactor system components is, therefore, precluded.

Yeav Loads The Upper Guide Structure (UGS) lift rig was designed and load tested in accordance with the guidance of NUREG-612 and ANSI N14.6-1978. The repair process does not introduce any additioral loads from repair tooling which exceed the lift rig design limits. Abnormal load transfer through the lift rig structural components due to the repair process has been reviewed and determined not to adversely impact structural integrity. The recommended repair process will limit the applied loads to the ICI plate and the lift rig and will not result in any loads being applied to fuel, core support components or plant structures.

The potential consequences of dropping heavy loads into the open reactor vessel have been considered. The UGS wi-11 be-in its normally seated position and, therefore, heavy load drop is not an issue pertinent to'his repair. Only the ICI plate, which weighs less than 10% of the UGS, will be within i s normal range of raised positions for the repair evolution. She~id the ICI plate drop, the impact load would be taken up by ="e reactor vessel flange and not the internal components. A :rop such as this is conservatively bounded by reactor vessel read drop analyses which have been shown not to have any fuel damage consequences. The long slender ICI thimbles which extend down from the plate are guided in the UGS and fuel regions by surrounding tubes. These tubes would preclude damage to the fuel even in the unlikely event the ICI plate were to arop.

Fire Hazards The repair program does not involve the use of any =lammable substances or open flame. All of the repair tooli~g is prefabricated outside containment. The repair process, therefore, will not cause any fire hazards.

Conclusions:

A safety evaluation was conducted to determine whether any unreviewed safety question or change to technical specifications is involved in the proposed incore instrumentation plate assembly repair process. The overall conclusions, which are elaborated below, indicate that the use of the special tooling and procedures developed for this repair effort will not involve an unreviewed safety question or require a change to the plant technical specifications. Specifically, the safety evaluation conclusions are that the ICI plate repairs does not:

PCH 126-287 Page 4 of 5 r

Increase the probability of occurrence or the consequences of an accident or malfunction of

.equipment important to safety previously evaluated in the safety analysis report; As indicated previously, the repair will be conducted in such a fashion that containment integrity and shutdown cooling system operation will be unaffected, therefore the safety functions for both systems will not be impacted. Furthermore, examinations of the possibility of dropping tools into the reactor vessel or of impacting the Heavy Loads Analysis during the repair process have been conducted. The repair process was found to not introduce any additional loads to the UGS lift rig that exceed its design limits. The Heavy Loads analysis will, therefore,.not be impacted in any way by the ICI plate repair. Furthermore, since the repair wor k will be, conducted above the UGS, which will be in its normally seated posi tion, there is no increase in the probability of occurrence of fuel damage due to the dropped tooling.

Create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report:

In assessing the possibility of creating new accidents or malfunctions as a result of the repair process, examinations of pool contamination, fire hazards, loose parts and impacts on crane handling design capability were conducted. The findings indicate that the grinding operations of any ICI plate components will only be executed in a controlled manner that minimizes the potential for particles entering the pool or reactor coolant system. The vacuum system that will be used will ensure that any random particles that may escape will be of insufficient size to create any safety concern. Furthermore, the hydraulic fluid that will be used has been found acceptable for unrestricted use at St. Lucie 2. With respect to additional fire hazards being gener ated due to the repair process, it was found that no flammable substances or open flames will be used, No fire hazards will therefore be created: The size and mass of all repair tooling have been examined and found to be well within the design loading capacity of the tool handling cranes.

Protection against inadvertently losing tooling has been ensured through use of lanyards and administrative controls on all.

PCM 126-287 Page 5 of 5 3, Reduce the margin of safety as defined in the'asis of any technical specification; An examination of the repair process has not led to any technical specifications that will be impacted. As a result, the margin of safety for plant technical specifications will remain unchanged.

Based on the information set forth above, C-E has concluded .that it is acceptable to proceed with the incore instrumentation plate assembly repair as planned.

PCM 131-287 AS-BUILT CCW SUPPORT ABSTRACT: NCR 2 123 has been generated to resolve discrepant field conditions for component cooling system restraint Mark No; CC-2074-44. This Engineering Package documents the evaluation performed for the 'as found'ondition and provides a mechanism for permanent plant drawing update. This modification does not involve an unreviewed safety question and no technical specification changes are required.

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

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

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

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

Yes. No X.' A change .to the plant technical -specificatio'ns'P EFFECT OF CHANGE Yes No Will the probability of an accident previously evaluated'in the FSAR be increased?

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

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

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

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

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

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

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PCM 136-287 CHECK VALVE V27101 REPLACEMENT ABSTRACT: Existing check valve V27101 by Rockwell has been damaged and is to be replaced with an equivalent valve by Kerotest, on FPL PO 87630-90117.

No unreviewed safety questions exist as defined by 10 CFR 50.59, and no Technical Specifications are impacted by this modification. Therefore, prior

-commission approval is not required.

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

TYPE OF CHANGE Yes No V h change to the plant as described in the FSA'R?

Yes No < h change to procedures as described in the FSAR?

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

Yes No v A change to the plant technical specifications?

KFF~ OF CHANGE Yes No v 1L'ill the probability of an accident previously evaluated in the FSAR be increased.

Yes No v Wild the consequences of an accident previouslywvaluated =-

hi the FSAR be increased?

Yes No~ lhay the possibility ot an accident which is different than any already evaluated in the FSAR be created?

'4 Yes No v~  %'ill the probability of a malfunction of equipment important to safety previously'evaluated in the FSAR be incr e a se d?

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

=

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

'es No~  %'ill the margin of safety as defined in the bases to any technical specification be reduced'

PCM 137-287 RCS INSTRUMENT NOZZLE INSULATION TEMPORARY MODIFICATIONS ABSTRACT: Minor Insulation modifications are outlined in this package for five hot leg RTD locations and one pressrizer lover head nozzle location. The changes are required to support requirement of JPE-M-87-112, Revision 0, "RCS Instrument Nozzle Cracking-Justi,fication for Continued Operation."

t No unreviewed safety questions exist as defined by 10 CFR 50.59 ~ and no Technical Specifications are impacted by this modification. Therefore, p commission approval is not required. prior NUCLEAR SAFETY EVALUATION CHECKLIST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded

'y the design analyses 'is attached to the Design Equivalent Engineering Package. The answers below are suppor ted by this evaluation.

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

No~

Yes Yes Yes No No

~ A change to procedures as described in the FSAR?

A test or experiment not described in the FSAR?

A change to the "plant technical specifications?

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

Yes No 2 Will the consequences of an accident previously evaluated in the FSAR be increased?

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

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

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

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

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

PCM 138-287 SPLICE BOXES B2124,34,35 o ABSTRACT!

re lecin FD t e conduit boxes the bend radius of the Raychem This EP documents the acceptability of vith larger splice boxes to prevent violating splices contained vithin. This modification does not involve an unrevieved safety question and does not require a change to the Plant Technical Specifications. This is evidenced by the attached Nuclear Safety va uat on ec st.

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

. evalua tion.

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

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

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

Yes No x A change to the plant technical specifications?

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

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

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

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

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

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

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

PCM 139-287 REPLACEMENT OF I-FCV-25-7 AND 8'ACCUMULATOR CHECK VALVES ABSTRACT: Replacement of the check'valves for the Instrument Air supply to the accumulators for FCV 25-7,8 (Containment Vacuum Relief). The replacement valves are NUPRO SS-4CP2-1 pu'rchased on FPL PO C38610 98267.

No unreviewed safety questions exist as defined by 10 CFR'50.59, Technical Specifications are impacted by this modification. and no commission approval is not required. Therefore, prior NUCLEAR SAFETY EVALUATlON CHECKLIST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below are supported by this e valua tion.

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

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

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

Yes. No A change to the plant technical specifications?

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

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

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

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

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

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

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

PCM 150-287 CEA MG SETS LOCKOUT RELAY ABSTRACT: This change modifies drawings (see drawing list) to relay 52Y contact (16-17) as normally closed per vendor manual show lockout representation. No physical change is required, only correction of drawing. No unreviewed safety question or change to technical specification is involved.

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

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

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

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

Yes No~ A change to the plant technical specifications?

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

s Will the consequences of an accident previously evaluated Yes Yes No No

~ in the FSAR be increased?

May the possibility of an accident which is different than any already evaluated in the FSAR be created?

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

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

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

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

PCM 155-287 EXCORE START/UP AND CONTROL CHANNEL CIRCUIT MODIFICATION

'BSTRACT This engineering package covers modifications to the Excore. Start/up and Control Channel Linear Power Circuits. The ma)or feature of this package is the modification of the feedback loop on the linear power subchannel inputs to increase the channel gain. This modification will compensate for the lower values of leakage flux at the excore detectors which result from the current St. Lucie Unit 2 Fuel Management Program.

Because the Excore Start/up and Control Channels are non-safety related and since this modification does not impact any safety related systems, all work covered by this engineering package is classified as non-safety'elated.

Based on a 10CFR50.59 safety evaluation this modification does not affect plant safety or operation, nor does-it involve any unreviewed safety questions or require changes to the plant Technical Specifications. As such, prior NRC approval is not required to implement this engineering package.

PCM 155-287 SAFETY EVALUATION Mith respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be increased, or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created, or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

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

i. The change described herein does not increase the probability of occurrence or the consequences of an accident, or malfunction of equipment important to safety previously evaluated in the safety analysis report.

This engineering package has considered the safety related consequences of the modifications proposed to the excore start/up and control drawers. Because the excore start/up and control channels are non-safety related, all work covered by this engineering package is classified as non-safety related.

All of the modifications implemented by this package are confined to the linear amp and summer cards (LASI-2) which are located inside the excore start/up and cont~ol channel drawers. The chnages consist of modifying the first stage feedback loop on the two linear subchannels to accommodate the actual, and anticipated future flux levels as seen by the detectors. Section 10 describes the modification in detail. These changes do not impact the functionality of the drawer; There are no changes to any other equipment interfacing with the excore start/up and control channels.

These changes were reviewed to determine the impact on the existing seismic and environmental requirements with no negativ'e findings.

PCM 155-287 Based on the above, the modification'is confined only to the excore start/up and control drawer, and has no impact on existing

/

analysis or design basis. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not increased.

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

The changes will not result in any new functional circuitry being added to the equipment, therefore, the procedures for performing maintenance and calibration for the drawer are essentially identical, and there is no impact on personnel performing maintenance on this .

equipment.

Based on the above, the change has no effect on existing setpoints, or system operation, and therefore, does not create the possiblity for an accident or malfunction of a different type than evaluated previously.

The change does not reduce the margin of safety as defined in the basis for any technical specification. The change does not result in an increase in the surveillance requirements. In addition, no operational parameter or technical specifications are impacted by the changes in this Engineering Package, therefore, no change to the technical specifications are requi~ed.

Because the Excore Start/up and Control Channels are non-safety related and since this modification does not impact any safety related systems, all work covered by this engineering package is classified as non-safety related.

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

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

PCM 027-288 ROSEMOUNT XMTR FT9021-Correct a model number discrepancy between installed hardware and affected documents for Rosemount instrument FT 9021. This is in response to RFD-009-87 and NCR 2-047. It is a document change only. It does not affect system function or qualification. It does not require a Tech. Spec. change and it does not involve an unreviewed safety question.

NUCLEAR SAFETY EVALUATION CHECKLIST e

The written evaluation of the proposed design change to demonstrate that the change does

.not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below are supported by this evaluation.

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

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

Yei No x A test or experiment not described in the FSAR?

Yes No x A change to tlie plant technical specifications?

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

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

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

Yes No X Will the probability of .a malfunction of equi'pment important to safety previously evaluated in the FSAR be increased?

Yes No X. Will the'onsequences of a m'alfunction of equipment important to safety previously evaluated in the FSAR be increased?

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

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

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PCM 044-288 EQ LIST REVISIONS SPARE PARTS ABSTRACT: The EQ List has an "X" in the SPEER column to designate the components/replacement parts which must be evaluated by engineering prior to ordering. Equipment included in this DEEP "X" in has been preapproved by engineering for purchase so the the SPEER -column is being deleted.

Tnis change does not involve a change to the Technical Specifications or an unrevi'et'ed safetv auestion as defined in 10CFR50.59 so prior 5'RC approval is not reauired for implementing this change. This change has no impact on plant safetv or operation since it is being implemented to assure that, procurement of replacement parts for EQ comoonents is being done in accordance with the FPL QA program requirements to assure compliance with the provisions of 10CFR50.49.

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

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

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

Yes Yo A test or experiment not described in the FSAR?

Yes Nlo x A change to the plant technical specifications?

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

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

Yes No x 4'ay the possibility of an accident which is different than any already evaluated in the FSAR be created?

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

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

Yes No iVay the possibility of a m'alfunc tion of equipment important to safety different than any already evaluated in the FSAR be created?

Yes No Will the margin of safety as defined in the bases to any technical speci fica t ion be reduced?

PCM 048-288 DRAWING/INSTRUMENT LIST CORRECTIONS REGARDING ICW PUMPS ABSTRACT: To correct a drawing error (CWD 2998-B-327 SH 882) for a Main Generator relay by Interchanging the contact designations and to correct designations. In the Instrument List and TEDB flow.

indicating switches used on Lubewater for the ICW pumps on the Instrument list and TEDB. An unreviewed safety question does not exist and there is no change to technical specification involved.

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

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

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

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

Yes No X A,change to. the plant techni0al specifiCations?

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

Yes No X Will the consequences of an accident previous!y evaluated in the FSAR be increased?

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

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

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

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

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

PCM 056-288 ICW & CW PUMP PACKING REPLACEMENT ABSTRACT: The existing CW'& ICW pump packing contains asbestos. The packing will be changed to an all graphite material. This DEEP provides for this change including guidelines for procurement.

\

No unreviewed safety questions exist as defined by 10 CFR 50.59, and no Technical Specifications are impacted by this modification. Therefore, prior commission approval is not required.

NUCLEAR SAFETY EVALVATION

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

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

Yes No A charge to procedures as describe'd in the FSAR?

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

Yes No A change to the plant technical specifications?

EFFECT OF CHANCE Yes No X V'ill the probability of an accident previously evaluated in the FSAR be increased?

Yes No 8 Will the consequences of an accident previously evaluated in the FSAR be increased?

May the possibility of an accident which is different than any already evaluated in the FSAR be created?

Yes No 0( %ill the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

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

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

'es No %ill the margin of safety as defined in the bases to any technical specification be reduced?

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PCM 080-288 NAMCO LIMITS SWITCHES FOR PCV-8801 THRU 5 ABSTRACT: Designate replacement limit switches for PCV-8801, 8802, 8803, 8804 and 8805. Revise affected drawings and documentation. The existing limit switches D2400X to be replaced with limit switches model EA170-11100.

No unreviewed safety questions exist as defined by 10 CFR 50.59, and no Technical Specifications are impacted by this modification. Therefore, prior commission approval is not required.

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

TYPE OF CHANGE Yes No

)C'es A change to the plant as described in the FSAR?

No A charge to procedures as described in the FSAR?

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

Yes No A change to the plant technical specifications?

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

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

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

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

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

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

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

PCM 096-288 DRAWING CLARIFICATION-ENGINEER) SAFEGUARDS CABINET ABSTRACT: This PC/M clarifies a,Unit 2 Control Wiring Diagram to more clearly show the appropriate terminal board numbers for the termination of certain wiring within the engineering safeguards cabinet. Improper interpretation of the terminal board designation by plant personnel has previously .contributed to a plant trip. This is merely a drawing clarification. An unreviewed safety question does not exist and there is no change to technical specifications.

'UCLEAR SAFETY EVALUATION CHECKLIST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the

. Design Equivalent Engineering Package, 'he answers below are supported by this e valua tion.

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

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

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

Yes. No~ A.change to, the plant technical specifiCations?

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

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

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

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

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

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

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

PCM 098-288 PDIS REPLACEMENT ABSTRACT: Replacement of failed PDS-2216, Barton Model No 288A with a Barton DPIS Model No 288A currently in stores. The replacement Barton 288A switch was procured as spare for Unit No 1 PDIS 02-1, which performs the same function in Unit 1 as the failed PDS-2216 in Unit No 2.

No unreviewed safety questions ezist as defined 'by 10 CFR 50.59, and no Technical Specifications are impacted by this modification. Therefore, prior commission approval is not required.

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

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

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

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

Yes No~ A change to the plant technical specifications?

Yes Yes No No~

~ EFFECT OF CHANCE i?ill the probability of the FSAR be increased?

an accident previously evaluated in Will the consequences of an accident previously evaluated in the FSAR be increased?

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

Yes No~ %ill the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

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

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

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

PCM 102-288 EMDRAC DRAWING 2998-738, REV 4 ABSTRACT The ASME Sect III Class 2 'check valves on drawing 2998-738 Rev 4 were never installed. Hydraulic operated gate valves were installed in place of these valves for the main feedwater system containment, isolation function. Since these check valves were subsequently sold, drawing 2998-738 shall be deleted and associated documentation updated accordingly. No technical specifications have been affected and there aie'no unreviewed safety questions.

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

Yes No ~

~

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

Yes Yes Yes No No No

~

~

A change to procedures as described in the FSAR?

A test or experiment not described in the FSAR?

A change to the plant technical specifications?

EFFECT OF CHANGE Yes No X, Will the probability of an accident previously evaluated in the FSAR be increased?

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

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

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

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

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

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

PCM'108-288 r

FUEL POOL PURIFICATION SYSTEM PUMPS MECHANICAL SEAL REPLACEMENT ABSTRACT: The existing mechanical seals utilized in the subject pumps are being replaced by a cartridge type mechanical seal. The change vill reduce the subject pumps'>> downtime required for seal replacement and is considered a maintenance enhancement. Also addressed is a clarification of the seal material specified for the subject pumps in the Unit 2 PSAR. This change does not affect any technical specification and there are no unreviewed safety questions.

NUCLEAR SAFETY EYALUATlON CHECKLiST The written evaluation of the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is a t tached to the Design Equivalent Engineering Package. The answers below are suppor ted by this evalua tion.

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

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

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

Yes No x A change to the, plant technical specifications?

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

Will the consequences ot an accident previously evaluated Yes No ~ in the FSAR be increased?

May the possibility of an accident. which is different than Yes Yo ~ any already evaluated in the FSAR be created?

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

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

May the possibility of a mal func tion of equipment important to safety different than any already evaluated in the FSAR be created?

Yes . No~ Will the margin of safety as defined in the bases to any technical specifica tion be reduced?

PCM 110-288 CONDENSATE RECOVERY SYSTEM PUMPS MECHANICAL SEAL REPLACEMENT ABSTRACT: The exi.sting mechanical seals utilized in the subject pumps are being replaced by a cartridge type mechanical seal. The change will reduce the subject pumps'owntime required for seal replacement and is considered a maintenance enhancement. 'This change does not affect any technical specifications and there are no unreviewed safety questions.

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

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

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

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

Yes No x A change to the plant technical specifications?

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

Yes No x Will the consequences of an accident prg viously evaluated Yes No ~ in the FSAR be increased?

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

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

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

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

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

PCM 112-288 TURBINE GLAND SEAL SYSTEM PUMPS MECHANICAL SEAL REPLACEMENT ABSTRACT: The existing mechanical seals utilized in the sub]ect pumps. are being replaced by a cartridge type mechanical seal. The change will reduce the subject pumps'owntime required for seal replacement and is considered a maintenance enhancement. This change does not affect any technical specifications and there are no unreviewed safety questions.

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

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

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

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

Yes No x A change to the plant technical specifications?

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

Will the consequences of an accident prf'viously evaluated Yes No ~ in the FSAR be increased?

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

Vill the probabili ty of a ma lfunc tion of equipment important to safety previously evaluated in the FSAR be increased?

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

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

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

PCM 149-288 DOCUMENTATION CORRECTIONS FOR PS-29 4~ 4 lj 4 ABSTRACT: Revise theSt Lucie Unit 2 Instrument List, and the Total Equipment Data Base to reflect as-built conditions for pressure switches PS-29-4, PS-29-1, and PS-29-4-2. Also revise EMDRAC 2998-13526.

1) The model number for pressure switch PS-29-4 appears incorrectly in the St Lucie Unit 2 Instrument List. The following change shall be made:

Model number B464 B-XNF shall be listed as B464BXNF.

(See Attachment 4, Sheet 1)

J

2) The model numbers for pressure switches PS-29-4, PS-29%-1, and PS-29-4-2 appear incorrectly in the Total Equipment Data Base. The following change shall be made:

The listing of the model number for each pressure switch shall change from B464-B-XNF to B464BXNF (Reference Ashcroft Bulletin 110, dated 4/80). (See Attachment 4, Sheet 1.)

3) The set points for pressure switches PS-29-4, PS-29"4-1, and PS-29-4-2 appear incorrectly in the St Lucie Unit 2 Instrument List. The following changes shall be made:

Incorrect Correct Ta Number Point Listin 'et Set Point Listin PS-29-4 HI ANN HI INT 6.5" WC 5.8" WC P S-29-4-1 HI INT ID ANN 5.8" WC 1.0" WC PS-29-4-2 LO ANN HI HI ANN 1.0" WC 6.5" WC (See Attachment 4, Sheet 1 )

4) The mounting location for pressure switches PS-29-4-1 and PS"29-4-2 shall change from IR 10-1B to IR 10-IA. (See Attachment 4, Sheet 1)

5) The information "QTY-2" and "TAG-PS-29-4-2" shall be deleted from EMDRAC Drawing ¹2998-13526. (See Attachment 4, Sheet 2)

No unreviewed safety questions exist as defined by 10 CFR 50.59, and no Technical Specifications are impacted by this modification. Therefore, prior commission approval is not required.

PCM 149-288 NUCLEAR SAFETY EVALUATION CHECKLIST The written evaluation of.the proposed design change to demonstrate that the change does not alter the plants design basis and is bounded by the design analyses is attached to the Design Equivalent Engineering Package. The answers below are supported by this evaluation.

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

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

A test or experiment not described in the FSAR? ~

A change to the plant technical specifications?

Yes Yes No No~

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

Will the consequences of an accident previously evaluated in the FSAR be increased?

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

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

Yes No X WiU the consequences of a malfunction of to safety previously evaluated in the pSAR be equipment'mportant increased?

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

~

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

PCM '227-984 TURBINE GANTRY CRANE SEPARATION REQUIREMENTS INTRODUCTION This PCM Supplement provides, restrictions on the clear distance to be normally maintained between the Units 1 and 2 turbine gantry cranes and loading combination restrictions under which the cranes may be operated regardless of separation.

DESCRIPTION The St Lucie Units 1 and 2 turbine gantry cranes share a common runway. In order to prevent overstressing of the Turbine Building structures, it is necessary to either maintain a specified distance between the cranes or restrict the loads allowed on each crane.

This PCM Supplement reiterates the separation requirement originally provided by Supplement 0 of this PCM. Where this separation is maintained, the cranes may be loaded to design capacity simultaneously.

'I This PCM Supplement further provides load tables indicating maximum lifting capacities of each crane in con)unction with various loads on the other crane, when normal separation between the cranes cannot be maintained. The load tables envelop all possible crane loading conditions and lo'cations for each building bay.

The allowable loads do not distinguish between loading on the main and auxiliary hooks, but represent the total load on both hooks.

PCM, 227-984 SAFETY ANALYSIS With respect to Title 10 of the 'Code of Federal Regulations

=50.59, a proposed change shall be deemed to involve an unrev'.

safety question: (i) if the probability of occurrence or consequences of an accident or malfunction of equipment importan:

safety previously evaluated in the Safety Analysis Report may increased; or (ii) if a possibility for an accident or malfunct-of a different type than any evaluated previously in the Safe Analysis Report may be created; or (iii) if the margin of safety defined in the basis for any Technical Specification is reduced.

This PCM Supplement imposes load and proximity restrictions on the Turbine Building gantry cranes. The Turbine Building" is a Non-Seismic Category I structure and contains no safety-related equipment. Therefore, this PCM Supplement does not increase the probability of occurrence or the consequences of an accident. or malfunction of equipment important to safety previously evaluated in the FSAR.

The load and proximity= restrictions provided ensure that the cranes are operated without exceeding the design capacity of the turbine building structure. Therefore, this PCM Supplement does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR. This PCM Supplement does not-involve any change to the St Lucie Unit 1 or 2 Technical Specifications.

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; and prior Commission approval is therefore not required for the implementation of this PCM Supplement.

PCM 087-985 HYPOCHLORITE CELL FLUSH SYSTEM This Engineering Design Package (EDP)'. modifies the existing portable cell flush cart. The cart will be permanently mounted on a

'all

'ypochlorite concrete pad and piping and electrical connections will be made permanent.

This to a EDP is classified as non-safety related since it is a modification non-safety .related system. The safety evaluation has shown that this EDP does not constitute any unreviewed safety question, nor does require a Technical Specification change. Therefore, prior NRC approval it is not required for implementation of this EDP.

PCM 087-985 Safet Evaluation With respect to Title 10 of the Code of Federal'egulations, Part 50 '9, 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 r'eport may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any technical specification is reduced.

The Sodium Hypochlorite'ystem is a chemical treatment systemMor the water in the Circulating Water and Intake Cooling Water Systems and does not perform any safety related function. Accordingly, all components of the system are classified non-safety class, Quali.ty Group D, and non-Class 1E.

Chlorine, in the form of Sodium Hypochlorite is used to control biological fouling in the Circulating Water System by use of a Hypochlorite Generating System, serving both St Lucie Units 1 and 2.

This modification does not involve an unreviewed safety question because:

i) The probability of occurrence or the consequences of an accident o" malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The Hypochlorite Cell Flush System equipment and associated piping and power supply are not used in any safety analysis for accidents or malfunction. of equipment. This system is non-safety related and will have no effect on equipment vital to plant safety.

ii) The possibility for an accident or malfunction of a different type than any 'evaluated previously in the safety analysis report is not created. The components involved in this modification have no safety related function. No changes have been made to the operational design of the hypochlorite system.

iii) The margin of safety as defined in the bases- for any Technical Specification is not affected by this PCM, 'since the components involved in this modification are not directly included in the bases of any Technical Specification. Failure of 'this system will be identified by instrumentation used to detect effulent chlorine content. This system can be restored to its operable status prior to unacceptable levels of slime accumulation.

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

'Technical Specification.

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

'PCM 020-986 F

INTAKE CANAL DREDGING AND SLOPE RESTORATION ABSTRACT Several areas of the intake canal have been sub]ected to continuous erosion and sedimentation. Re'cent inspections of the areas indicate that'he

'eterioration is due to, various factors contributing to different extents; among these probable causes are canal currents, tidal action, and rainfall.

This PC/M provides restoration of canal embankments, the installation of new protection against erosion, and the removal of sedimentation east of the A1A bridge.

This PC/M does not involve an unreviewed safety question and has no effect on plant safety. The intake canal is not considered safetymelated and the work be done will restoring the canal to its original design profile.

'o be

7CM 020-986 ShPETY EVhLUhTION With respect to Title 10 of the Code of Federal Regulations, Pait 50.59, a proposed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occurrence malfunction of importarit or to consequences of an accident or equipment safety previously evaluated in the Safety Analysis Report may be increased; (ii) if a possibility for an accident or malfunction of a different type may be crested; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

This PC/M restores Intake Canal embankments and does not involve an unreviewed safety question for the following reasons:

i The probability of occurrence or the consequences accident or malfunction of equipment important to safety of an previously evaluated is not increased since none'f the work will be performed in . the proximity of safetymelated equipment; most of the work will be done outside the plant security fence. Pai.lure of this system will not prevent the safe and orderly shutdown of the plant. The Emergency Cooling Water System, through Big Mud Creek, can provide adequa'te makeup and has been considered.

ii There is no possibility for an accident or malfunction of,a different type than any evaluated previously 'since the Intake Canal is nonmafety 'related, and this modification cannot affect any safet~elated system. 'his modification will, in fact, increase the reliability and decrease the probability of an accident by restoring 'he canal to its original configuration.

iii This modification does not change the margin of safety as defined in the basis for any Technical Specification. This PC/M may be performed in any plant mode of operation.

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

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

PCM 039-986 BLOWDOWN BUILDING RADIATION MONITORING SYSTEM ABSTRACT This engineering package provides for the replacement of the effluent gaseous portion of the presently installed Nuclear Measurement Corporation (NMC) Blowdown Building Ventilation Airborne Radiation Monitoring System with a spare General Atomics (GA) Technologies Airborne Radiation Monitor.

The sensitivity levels of the GA Airborne Radiation monitor will be equal to or greater than the presently installed system and the outputs will be duplicated.

The Blowdown Building. Airborne Radiation monitor will be used to record and annunciate the gros's airborne trends in the Blowdown Building Ventilation System and the amount of radioactive releases to the atmosphere. Although this system provides no safety related function and the monitor will be physically located inside the Steam Generator Blowdown building, this PC/M is classified as +ality Related, since the monitors are used to assess the Blowdown Building's contribution of airborne radiation to the total airborne radiation effluent at the site.

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

PCM 039-986 Safety Evaluation With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (I) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) If the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.

The probability of occurence as the consequences of an accident or malfunction of equipment previously evaluated in the Safety Analysis Report is not increased by this PC/M. These modifications to the blowdown building radiation monitoring system will duplicate the outputs of the replaced system and reuse the existing Isokinetic probe and output recorder.

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

a. The new equipment mounting will be seismically analyzed for additional loading in accordance with St Lucie Design Criteria Manual, Section I.

b, The new radiation monitor will be located in the blowdown building, which is considered to be a mild environment.

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

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

PCM 111-986 SIMULATOR TRAINING FACILITY PIPING TIE-INS ABSTRACT This engineering package is being issued to cover the addition of the St. Lucie Simulator Training Facility fire protection, service water and sanitary piping tie-ins. No aspect of this project will add to, modify or otherwise affect any plant safety related system. Fire system modifications associated with this project that tie into the plant fire loop up to the first isolation valve shall be classified as "Quality Related" QA/QC required. The remainder of the fire protection, all service water and all sanitary piping shall be Non-Nuclear Safety Related.

The addition of the Simulator Training Facility fire protection, service water and.

sanitary tie-in lines do not pose any unreviewed safety questions nor involve any changes to plant Technical Specifications.

PCM 111-986 SAFETY EVALVATION Related The St. Lucie plant fire protection loop is defined as a QuaHty of this modification that tie into the fire loop up to system. Those portions QuaHty and including the first isolation valve have been designated >s Related and conform with the requirements of the original fire .oop. The remainder of the fire protection piping added by this modificatio", has been designated Non-Nuclear Safety Related Quality Group D. Those portions of the modification providing service water and'These sanitary piping and tie-ins are classified as Non-Nuclear Safety Related. components tie into the existing plant service water and sanitary systems which are also classified as Yon-Nuclear Safety Related QuaHty Group D.

A failure mode anlaysis was perfoimed on the Non-Nuclear Safety Related portions of the modification. Based on this analys~ failure of the service water and sanitary piping or components and those portions of the fiie main downstream of the first isolation valve will not inhibit the operation of any safety related equipment or components. These materiali are located remote from any safety related equipment or components aad as such cannot fall on a hit such components. Failure of the service vater line will cause loss of service water to the simulator building. Failure of the sanitary Hne wiH inhibit the use of the Simulator Building saritary system. Failure of the downstream fire main piping will not inhibit the functional capabilities of the fire loop since the post indicator valve, located upstream of these portions of the system provides adequate isolation capabilities to ensui e functional integrity of the fire loop.

~

Those portions of the modification providing fire protection piping tie-ins to the first isolation valve can affect the functional capabilities of the fire loop and therefore can affect fire protection capabilities for Safety Related equipment and components. As addressed in the Design Analysis, these portions of the modification have been designed and construction requirements have been specified to comply with the necessary Quality Related requirements. Since the equipment affected by this modification is not considered by the FSAR in determining the probability of accidents or possible types of accidents or in the evaluation of the consequences of accidents, it can be concluded that the probabiHty of occurrence of accidents previously addressed in the FSAR is unchanged and the possibility of new accidents not considered in the FSAR has not been created.

Therefore, the potential failure mode of this system and degree of

'rotection provided to nuclear safety related .equipment remains unchanged.

Based on this information, it can be demonstrated that an unreviewed safety question as defined by 10CFR50.59 does not exist since the consequences of all analyzed accidents remains unchanged. Additionally, with respect to Nuclear Safety, no new accidents or'malfunctions are introduced as a result of this modification. Finally, the margin of safety as defined in the Technical Specifications has not been reduced nor have changes to the Technical Specifications been requii e*

In conclusion, this modification is acceptable from the standpoint of nuclear safety since it does not involve an unreviewed safety question nor require changes to the Technical Specific'ations. Thus implementation of this modification does not require prior NRC approvaL

PCM 106-988 STEAM GENERATOR BLOWDOWN TREATMENT FACIIITY SYSTEM PUMPS MECHANICAL SEAL REPLACEMENT ABSTRACT: The existing mechanical seals utilized in the subject pumps are being replaced by cartridge type mechanical seal. The -change vill reduce pumps'owntime required for seal replacement and is considered a maintenance enhancement. This change does not affect any technical specifications and there are no unrevieved safety .questions.

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

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

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

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

Yes No x A change to the plant technical specifications?

EFFECT OF CHANGE Yes No x Will the probability of an accident previously evaluated in Yes Yes No No

~ x s

the FSAR be increased?

Will the consequences of an accident pn'viously evaluated in the FSAR be increased?

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

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

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

May the possib ility o f a mal func tion of equipment important to safety different than any already evaluated Yes . 'o~ in the FSAR be created?

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