Text

Changes,Tests & Experiments Made W/O Prior Commission Approval,For Jul 1986 - June 1987
	ML20237J335
Person / Time
Site:	Turkey Point
Issue date:	06/30/1987
From:	Woody C; FLORIDA POWER & LIGHT CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	L-87-365, NUDOCS 8709030560
	Download: ML20237J335 (177)
	v • d • e

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I TURKEY POINT PLANT UNITS 3 AND 4 DOCKET NUMBERS 50-250 AND 50-251 CHANGES, TESTS, AND EXPERIMENTS MADE WITHOUT PRIOR COMMISSION APPROVAL ;

1 FOR PERIOD o

JULY 1, 1986 THROUGH JUNE 30, 1987 IN COMPLIANCE WITH TITLE 10, SECTION 50.59(b)

CODE OF FEDERAL REGULATIONS l

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INTRODUCTION )

This report is submitted in accordance with 10 CFR 50.59 (b), which requires that reports of:

i) changes in the facility as described in the FSAR ii) changes in the procedures as described in the FSAR, and fii) tests and experiments not described in the FSAR which are conducted without prior commission approval be reported to the Commission at least annually. This report is intended to meet this require-ment for the period July 1,1986 through June 30, 1987.

This report is divided into three sections; the first, Plant Change /Modifica-tions, covering changes in the facility as described in the FaAR; the second, Procedure Changes covering changes in the procedures as described in the FSAR; and the third, Tests and Expe riments , covering tests and experiments not described in the FSAR.

Appendix A to this report is a 'ist of safety and power operated relief valve actuations, which is submitted ii accordance with FPL's commitment to comply s with the requirements of Item IIK.3.3 of NUREG 0737. This report covers the period from July 1,1986 to June 30, 1987.

Appendix B to this report is a summary of the findings of the Steam Generator tube inspection performed on Unit 3 during the report period from July 1,1986 through June 30, 1987.

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TITLE 10, SECTION 50.59 CFR REPORT i) COMPLETED PC/M LIST JULY 1,.1986 THRU JUNE 30, 1987 1 PC/M TITLE UNIT TURNED OVER DATE ' j 85-86 AFW CV INSTRUMENT AIR FILTER MODIFICATION 4 7/26/86 85-108 NUCLEAR ADMIN. BLOG. POWER SUPPLY 3&4 3/15/86 85-171 INSTALLATION OF CHAIN OPERATOR ON AFW 4 7/26/86.

ISOLATION VALVES

'85-196 DIESEL GENERATOR SKID TANK SOLEN 0ID VALVE. 3&4 8/5/86 BYPASS LINE ADDITION 86-62 NORMAL CONTAINMENT COOLING FAN MODIF. 4 8/1/86.

86-94 EttERGENCY DIESEL GENERATOR "B" ROOM VENT 4 7/27/86 FAN 4V34 POWER SUPPLY.

86-95 PROVIDE NEW POWER FEED FOR NON-VITAL SECTION 3 7/24/86 l OF MCC 3A & 30 86-041 MODIF. OF MCC "D" AUTO TRANSFER 3 7/25/86 ,

.78-1028 STEAM GENERATOR BLOWDOWN REC 0VERY SYSTEM 4 7/17/86 o 84-02 MODIF. TO COMPLY WITH REG. GUIDE 1.97 REV. 3 4 9/15/86 RE0VIREMENTS TO PROVIDE QUALIFIED LIMIT i SWITCHES i 83-63 IMPROVED FLOOR DRAIN FOR CONTAINMENT SPRAY 3 12/12/85 PUMP ROOM ,83-114 REACTOR CAVITY FILTERS - LEAD SHIELDING 4 4/4/86 86-03 MSIV NITROGEN SUPPLY ADDITION 4 8/7/86 J I

86-64 4.160KV FUSE HOLDER SUBSTITUTION 3&4 8/7/86 86-70 REPAIR DAMAGED T/C CABLES AT TE-4-1418 4 6/24/86

& 1421 86-71 ICW BASKET STRAINER REPLACEMENT 4 6/24/16 82-83 ADD BACK-UP TO RELAYING IN SWITCHYARD 3&4 7/24/85 86-44 ICW BASKET STRAINER BELZONA LINING 4 12/1/86 87 CABLE REPLACEMENT FOR MSIV SOLEN 0ID 4 12/31/86 85-152 NIS INPUT TO TURBINE RUNBACK - REINSTATE 4 12/18/86 l 1/4 CONFIGURATION l 85-85 AFW CV INSTRUMENT AIR FILTRATION MODIF. 3 9/30/86 83/139 HALON SUPPRESION SYSTEM FOR APPENDIX R 3&4 8/20/86

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TITLE 10, SECTION 50.59 CFR REPORT i) COMPLFTED PC/M LIST JULY 1, 1986 YHRU JUNE 30, 1987 i PC/M TITLE UNIT TURNED OVER DATE 85-56 REPLACEMENT OF INTAKE WALKWAY 3&4 10/31/86 84-210 TURBINE RUNBACK MODIFICATION 4 8/19/86 85-151 ICW PIPE & STRAINER INSPECTION CLEANING 4 8/27/86 85-124 REMOVAL OF SPENT RESIN PIPE IN LAUNDRY ROOM 3&4 10/31/86 86-22 ICW HEADER ISOLATION VALVE REPLACEMENT 4 9/8/86 84-124 AFW FLOW TRANSMITTER REPLACEMENT 4 3/9/87 i 84-158 EMERGENCY DIESEL GENERATOR "48" CFD 4 12/17/85 RELAY REPLACEMENT

'83-141 FIRE BARRIERS FOR APPENDIX R 3 5/21/86-83-145 FIRE DAMPERS FOR APPENDIX R 3&4 8/14/86 84-20 POST ACCIDENT CONTAINMENT AIR SAMPLING 3&4 4/25/86 SYSTEM FLOW TRANSMITTER 86-067 TURBINE AUX. BLOCKING 0F AUTO LOAD ON 4 3/9/87 DIESEL GENERATORS 86-05 MSIV NITROGEN SUPPLY ADDITION (INTERIM) 3 6/9/86 i 85-130 AFW DISCHARGE FLOW CONTROL VALVE UPGRADE 3 3/9/87 85-149' SPENT FUEL PIT AIR INLET DAMPER REPLACEMENT 3 3/9/87 82-36 SPENT FUEL PIT LEVEL INDICATOR AND ALARM 4 5/15/86 84-167 DECONTAMINATION SHOWER FACILITY 3&4 3/26/87 i 86-158 THR0WBOLT REPLACEMENT ON INTAKE STRUCTURE 3&4 11/18/86 BAY HATCHWAY 86-166 FUEL TRANSFER SYSTEM CABLE DRIVE MODIF. 3 3/17/87 DISASSEMBLY OF PRESENT SYSTEM 86-207 IN SERVICE TESTING GAUGE INSTALLATION 3 3/19/87 FOR SPENT FUEL PIT COOLING PUMPS86-159 INTAKE STRUCTURE WOOD GRATING LATCHES 3&4 11/8/86 86-124 REPLACEMENT HIGH RANGE GAMMA RADIATION 4 11/26/86 READOUT MODULE 83-140 FIRC DETECTION FOR APPENDIX "R" MODIFICATION 3&4 12/1/86 84-124 AFW FLOW TRANSMITTER REPLACEMENT 4 3/9/87 82-311 AFW TURBINE STEAM SUPPLY STOP/ CHECK VALVE 3 3/13/87 85-143 BREAKER / FUSE C0 ORDINATION MODIF. 3&4 6/13/87 A2:PCM-LOG PAGE 2 1

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t TITLE 10, SECTION 50.59 CFR REPORT i) COMPLETED PC/M LIST JULY 1, 1986 THRU JUNE 30, 1987 PC/M TITLE UNIT TURNCO OVER DATE 85-60 MAIN TRANSFORMER FAN COOLER UPGRADE 3 6/11/87 84-171 MODIFICATION TO ALLEVIATE SHORTAGE OF 3 2/2/87 COMPARTMENTS ON MCC "38" j 86-236 REPLACEMENT OF TREATED WATER PUMP SEALS 3&4 4/4/87 l 85-197 RELOCATE INSTRUMENT AIR SUPPLY VALVES 3 9/26/86 40-4-098 & 40-3-641 85-170 INSTALLATION OF AFW VALVE ACCESS PLATFORM 3 1/15/87 ;

86-67 TURBINE AUXILIARIES-BLOCKING OF AUT0-LOADING 4 3/9/87 ON DIESEL GENERATOR .87-086' PERSONNEL AIRLOCK EQUALIZATION VALVE 4 3/28/87 i REPLACEMENT 87-095 "4A" ICW PUMP ANCHOR BOLT REPLACEMENT 4 5/14/87 f 121 UPENDER LEVELING DEVICE MODIF. 3 3/30/87 )86-029 AFW LOCAL INDICATION UNDER THE MAIN 3&4 3/28/87 {

FEED WATER PLATFORM 86-021 "C" AUX. FEEDPUMP REPLACEMENT IMPELLER 3&4 4/13/87 86-212 ENVIRONMENTAL QUALIFICATION LIST REVISION 3&4 2/12/87 86-103 ENVin0NMENTAL QUALIFICATION LIST REVISION 3&4 2/12/87 ,

85-65 G. E. SAM RELAY MODIFICATION, PC CARD 4 1/15/87 REPLACEMENT ;83-153 CABLE REROUTING - APPENDIX R MODIFICATION 3&4 12/18/86 87-23 MAIN STEAM HYDRAULIC SNUBBER REPLACEMENT 4 6/19/87 86-96 NEW POWER FEED TO NON-VITAL SECTION OF 4 9/8/86 i MCC "4A" 86-60 COMPUTER ROOM TEMPERATURE INDICATION 3&4 2/20/87 i 85-071 SPENT FUEL PIT BUILDING WALL 3 4/15/87 JOINT REPAIR ,

i 85-35 REPLACEMENT OF PYC0 RTD'S 3 6/19/87 I 87-97 INSTALLATION OF UNDER VOLTAGE TRIP 3 6/27/87 CIRCUITRY FOR TUR8INE AND POLAR CRANE BREAKERS l

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TITLE 10, SECTION 50.59 CFR REPORT l

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f) COMPLETED PC/M LIST L JULY 1, 1986 THRU JUNE 30, 1987 PC/M - TITLE UNIT TURNED OVER DATE i L

87-52 GENERATOR NEUTRAL GROUNDING TRANSFORMER 3 6/9/87 REPLACEMENT

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87-53 GENERATOR NEUTRAL GROUNDING TRANSFORMER 4 6/9/87 REPLACEMENT 87 INSTALLATION OF. UNDERV0LTAGE TRIP DEVICE 4 6/23/87 FOR POLAR CRANE BREAKERS80-117 UPGRADE AFW SUCTION, DISCHARGE, STEAM SUPPLY 3&4 11/15/85 84-111 A/C UNIT FOR PASS CONTROL PANEL 3&4 10/31/86 86-15 REPLACED TELEDYNE-FARRIS COMPONENT C0OLING 4 10/13/86 WATER RELIEF VALVES l 86-68 REMOVAL OF CCW PIPING TO THE PRIMARY SHIELD 4 10/31/86 COOLERS87-169 MODIFICATION TO COMPONENT COOLING WATER '4 6/19/87 SYSTEM 81-59 WATER TREATMENT PLANT FINAL EFFLUENT 3&4 2/17/87 8 CONDUCTIVITY TRIP 83-209 MSR FOUR TUBE PASS MODIFICATION 4 11/18/86 85-131 AUXILIARY DISCHARGE FCV UPGRADE 4 9/17/86 85-133 flSR MODERNIZATION 4 11/18/86 86-31 AFW PUMP CONTROL PANEL WIRING MODIFICATION 3&4 12/10/86 87-99 ICW/CCW GASKET STRAINER REPLACEMENT 3 6/26/87 87-156 ICW BASKET STRAINER ISOLATION VALVE 3 6/16/87 REPLACEMENT-SHAFT /0PERATOR ADAPTER

~86-80 SAFETY INJECTION ACCUMULATOR MAKE 3 6/26/87 LJ 'iEn0ER SEISMIC REPLACEMENT 86-76 DIESEL GEN "B" FRE00ENCY METER REPLACEMENT 3&4 6/3/87 86-96 ROOT VALVE # 4-20-698 SUBSTITUTION 4 7/8/86 85-10 ADDITION OF FW CONTROL VALVE DIRECT POSITION 4 2/18/87 INDICATION 84-209 REINSTATEMENT OF POWER MISMATCH WITHOUT R00 4 8/16/86 j WITHDRAWAL 1

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TITLE 10, SECTION 50.59 CFR REPORT i

1) COMPLETED PC/M LIST JULY 1, 1986 THRU JUNE 30, 1987 PC/M -

TITLE UNIT TURNED OVER DATE s86-162 REMOVAL OF CCW PIPING TO THE PRIMARY SHIELD 3 5/15/87 COOLERS87-126 ACCUMULATOR SAFETY INJECTION TEST 4 7/13/87 !

-LINE SOLEN 0ID VALVE REPLACEMENT

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87-160 BAILEY TEMP. TRANSMITTER REPLACEMENT FOR 3 7/13/87 TURBINE PLANT COOLING WATER 87-037 ICW PUMP FOUNDATION REPAIR ANCHOR BOLT 3 6/9/87 I

REPLACEMENT 87-102 REACTOR VESSEL HEAD INSULATION-REFLECTIVE 4 6/9/87 REPLACEMENT ,87-101 REACTOR VESSEL HEAD INSULATION-PERMANENT. 4 6/9/87 REPLACEMENT 87-177 CONTAINMENT SPRAY RESTRICTING ORIFICE 4 6/11/87 l 86-100 NIS SOURCE RANGE PRE-AMPIFIERS 4 10/31/86 l 86-184 RPI INVERTER REGULATOR TRANSFORMER REPLACEMENT 4 5/30/87 83-50 MASONRY WALL MODIFICATION 3&4 2/12/87 l 87-210 RPLACMENT SUPPORTS 4-SIH-42 AND 4-PRWH-11 4 6/12/87 l 84-16 RHR ISOLATION VALVE CIRCUIT MODIFICATION .3 4/13/87 l 85-182 Li!EMICAL ADDITION LINES SUPPORT REPAIR 3 10/29/86 f

85-139 REMOVE VALVES 4-524/525 AND PIPING 4 9/5/86 85-181 ROMOVE INACTIVE NITR0 GEN BLANKET LINE 4 10/3/86 i 84-11 MODIFICATION TO PRESSURIZER SPRAY SYSTEM (I.C.) 3 7/3/85 85-141 FUEL TRANSFER SYSTEM MANIPULATOR CRANE DUAL 3 6/29/87 i CABLE MODIFICATION 86-121 CONTROL POWER FUSE REPLACEMENT 4 5/30/87 86-107 MCC CONTROL POWER FUSE REPLACEMENT 3 5/30/87 l 86-026 4KV SWITCHGEAR BREAKER ELEVATING 4 5/30/87 f MECHANISM REBUILDING 85-11 MODIFICATION OF 4160V BREAKER "HH" SWITCHES 3 6/3/87 1 85-149 SPENT FUEL PIT INLET DAMPER REPLACEMENT 3 5/7/87 !

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q PLANT CHA %E/ MODIFICATION 85-086 PC/M CLASSIFICATION: NS UNIT: 4 TURNED OVER DATE: 07/26/86

SUMMARY

'0 ATE: 09/03/86 REVISION: 0 !

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I AUX 1LIARY FEEDWATER CV INSTRUMENT AIR FILTRATION MODIFICATION Summary: l This modification provided for the installation of new filters in the instru-

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ment air (nitrogen backed) supply line to each unit 4 Auxiliary Feedwater Control. Valve as shown on the drawing listed on Attachment 1. The installa-

. tion of the filters provides better quality air to the Control Valve position-ers and actuators.

The new filters were installed upstream of the instrument air-nitrogen suppli !

ti e-i n connections in the instrument air supply line. The filters are provided with isolation valves and bypass lines with a valve. for ease .of maintenance without isolating the Control Valve air supply. A new anchor, and, !

supports as required per 5177-PS-21 must be installed to isolate the . filter assembly from the downstream check valve. ;

This isolation is required to i ensure the functional and structural integrity of the system. All new filters, valves, and piping of the safety related portion upstream of the anchor '

3 need not be supported to seismic Category I requirements. The new piping will be 1/2-inch galvanized steel, Schedule 40. ._

The new filters are Parker Hannifin Standard Airline Filters designed to separate dirt particles and water. Sizing for this installation was deter- !

mined by the maxirrum flow capacity required for operation of the positioners on the Auxliliary Feedwater Control Valves. Using the flow rates required with 100 psi instrument air and 60 psi instrument air as the pressure range to operate the positioner, (13 scfm and 8.5 scfm respectively), and a required 2 to 5 psi pressure drop across the filter for efficient operation, and the particle . size filtration needed, in this case 5 micron, a filter size selec-tion is made form a chart supplied by the filter manuf acturter. Therefore, <

the filter size selected is specific for this application, and has the concur-rence of the filter manufacturer. This analysis is documented in Calculation M06-0424-01.

The method of operation of the air filter is such that pressurized air (instrument air at 100 psig) flows through a louvred deflector and is directed into a swirling pattern. Liquids and large dirt particles are thrown against the inside wall of the see-through polycarbonate bowl by cyclonic action and fall into a quiet zone below the lower baffle. This lower baffle maintains the quiet zone to prevent turbulent air from returning liquids and solids into the air stream. The instrument air, now f ree of liquids and dirt particles, pass through a filter element sized to remove particles down to 5 microns for this application.

Clean air then flows through the outlet port. Liquids are discharged from the bowl by the automatic drain valve. The Parker Hannifin nodel number for this following: installation is 04F158 which translates to the L

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'PC/M 8b'-066 Page'2 i l

04F - Minature Series 1 - 1/4" NPTF Port Size b - Automatic Drain i B - 5'Hicron Filtering Element Safety Evaluation:

These modifications provide for'the installation of new filters in the instru- :

ment air supply. line to each Unit 4 Auxiliary Feedwater Flow Control ' Valve. 1 The air filters, their associated isolation and bypass valves and piping will y not be installed to the requirements of seismic Category 1. The seismic i boundary anchor assures the structural integrity of the piping and components j downstream of the anchor. . The modifications provided by the PC/M include passive components whose function will not' be impaired by any design basis accident described in the FSAR. j

These modifications do not introduce new safety related equipment which .could I be affected by fire t or add new combustibles which could invalidate the fire # ;

Zone Heat Loading Analysis previously submitted to FPL per Bechtel letter SFB- i 1741 dated April 24, 1985. In addition, these modifications to not adversely l af fect any existing or proposed fire protection features of-the plant. 'There-fore, these modifications do not affect the Turkey. Point Fire Protection Program. i These modifications are not inside containment, are not attached to block walls, do not involve safety related snubbers and do not impact spent fuel ;

pool cooling operations of the plant. !

ho special ALARA considerations are required because the modifications are to be carried out in the areas outside of Radiation Control Area.

1 The modification does not involve the addition of electrical cable or any changes to existing raceways. The final modifications accomplished by this PC/M do not affect the flooding analysis as described in the NRC Safety Evalu-ation Report dated September 4, 1979, because they do not introduce a new source of flooding, modify the existing flood mitigating features, or install ;

or modify any safety related components which could be affected by flooding.

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

The probability of occurrence of an accident previously evaluated in the FSAR will not be increased because the modifications do not alter the function of Ins +rument Air or Nitrogen Backup Supply to AFW Control Valves.

The consequences of an accident previously evaluated in the FSAR will not be increased because the added tubing, filters and valves will not affect the performance of safety features.

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' The changes which are associated with plant safety features are minor 1 l in nature and do not change the function of the plant safety !

? features. Therefore, there is no possibility that an accident may be ]

i created that is different than any already evaluated in the FSAR. J This probability of occurrence of an equipmentimalfunction important to saf ety which have already been evaluated in the FSAR, . Will' not . be increased due to the installation of filters and isolation valves <

because this change does : not adversely af fect any equipment important L

to safety.

L The consequences of equipment malfunction important to safety which have already been evaluated in the FSAR will not be increased because the performance design basis has nt been changed from that described in the FSAR.

The modification to the AFW Control Valve Instrument Air Supply System ,

identified in this PC/M will not change the inherent function or design '

-basis for the system. In addition, the in-line air filters will be ;

l- inspected and cleaned (if necessary) on a regular basis 'in accordance with approved system maintenance procedures, to prevent possible block- e age. Therefore, the possibility of a malfunction of equipment impor-tant to safety which is of a different type than previously evaluated in the FSAR will not be created.

This modification provides cleaner air to the AFW Control Valve actua-tor and thus does not reduce the margin of safety as defined in the bases for any Technical Specification. l Based on the above, these modifications do not constitute an unreviewed safety question.

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-PLANT CHANGE / MODIFICATION 85-108 PC/M CLASSIFICATION: NNS )

UNIT: 3 and 4

j. TURNED OVER DATE: 3/15/86 ,

SUMMARY

DATE: 9/02/86 REVISION: 0 NUCLEAR ADMINISTRATION BUILDING POWER SUPPLY Summary:

This PC/M installed a power supply from the existing Florida City Feeder 'to the new fluclear Admt listration Building.

Safety Evaluation: ,

This PC/M does not adversely affect the operation of any nuclear safety related equipment as the Florida City Feeder and the Nuclear Administration Building do not perform a nuclear safety function and all work associated with-the installation of this PC/M is outside the plant perimeter fence. There- .,

fore, this PC/M is classified as non-nuclear safety related and ~does not involve an unreviewed safety question.

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l' PLANT CHANGE / MODIFICATION _85-171

, PC/M CLASSIFICATION: NS-V UNIT: 4' )

TURNED OVER DATE: _ 07/26/86 '

SUMMARY

DATE: _ 09/02/86 REVISION: __ U 1hSTALLAT10N Of CHAIN OPERATORS ON AUXILIARY FEEDW Summary:

i This modification provided for the addition of chain operators to AFW manual isolation valves which are currently inaccessible.

reviewed with respect to the reference documents, This PC/M has been

, l considered during the review.All applicable design verification elements of E 1S afety Evaluation:

o This modification manual installs isolation valves that chain operators for operation of inaccessible AFW operating conditions. require local operator action under off-normal pipe and no 'modifications stress analysis and the design of the pipe supports has are required.

addition analysis of the remains valid chainwheel to the (Refer to Attachment 7). valve and has concluded tha the seismic This modification is not inside containment, does not involve safety related snubbers, does not involve block walls, does not impact the spent fuel cooli operations of the plant, does not affect the Radioactive Waste Treatmen' changes to existing raceways. System of the plant, and does not involve th The modification guidances provided has been reviewed for ALARA requirements based upon the in Criteria for ALARA Evaluation per FPL letter JPE-PTP0-64-1239 and is acceptable.

. The analysismodificatf as or accomplished described in the NRC by this PC/M does not Safety Evaluation Report, affect the floodini September 4, dated k

1979, because the modification does not introduce a new source of i t

related components which could be affected by flooding. flo!

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Based on the preceeding, the following conclusions can be made: l The addition of chain operators does not change the inherent function ;

o of the AFW Sy s t em. The pipe and pipe support design have been l evaluated- with respect to this change and the additional weight is i' acceptable and no pipe support changes are . required. Therefore, the probability of occurrence of an accident previously evaluated in the FSAR will not be increased.

The consequences of an accident previously evaluated in the FSAR will I i

not increase because the modification does not affect the. operation of ,

any saf ety related system. {

i This modification does not change the function of any safety- related I system. ' Therefore,- there is no possibility that an accident may be created that is a different type than any previously evaluated in the J FSAR.

j This modification does not affect the probability of occurrence of any equipment malfunction important to safety previously evaluated in the ;

FSAR, because it does not adversely affect the inherent function,# '

. operation, or availability of equipment important to safety. .

Due to this modification, consequences of equipment malfunction l important to safety is not changed from one which is evaluated in the

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FSAR because this modification does not change the design basis l described in the F5AR for any safety related system. ;

I These modifications do not change the inherent function or design basis of the systems related to safety; therefore, the possibility of a !'

malfunction of equipment important to safety which is of a different type than previously evaluated in the FSAR will not .be created.

This modification does not change the operation of any system related ,

to safety, therefore, this modification does not reduce the margin of l safety as defined in the bases for any Technical Specification.

Based on the above, this modification does not constitute an unreviewed safety j question.

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l PLANT CHANGE / MODIFICATION 196 PC/M CLASSIFICATION: NS UNIT: 3&4

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TURNED OVER DATE: 08/05/86

SUMMARY

DATE: 09/02/86 I RE:ISION: 0 1

l DIESEL GENERATOR SKID TANK SOLEN 010 VALVE BYPASS LINE ADDITION Summary:

Add bypass lines with manual isolation valves for Solenoid Valves SV-3-3522 and SV-4-3522 in the respective fuel oil supply line to the Emerge.ncy Diesel Generator Skid Tanks. The installation of the bypass line provides an alter-nate fill path should the solenoid valve fail closed as a result 'of loss of power or a spurious signal to this solenoid valve. This modification is necessary to support the Appendix R Safe Shutdown System Analysis.

Safety _ Evaluation: )

i This modification provides bypass lines for Valves SV-3-3522 and SV-4-3522, fuel oil supply lines to the EDG Skid Tanks, to support licensing comitments >

associated with 10CFH50 Appendix R requirements. The new pipe will be seismi-cally supported in accordance with project requirements for small pipe. Since l the seismic integrity of the entire line, including this modification, will be ;

evaluated and ensured under the program to walkdown and evaluate small piping l and tubing, this approach is acceptable. !

This modification is not inside containment, does not involve safety related snubbers, does not involve block walls, does not impact the spent fuel cooling i operations of the plant, does not affect the Radioactive Waste Treatment System of the plant, and does not involve the addition of electrical cable or changes to existing raceways.

The modification has been revi ewed for ALARA requirements based upon the guidances provided. in Criteria for ALARA Evaluation per FPL letter JPE-PTP0-04-1239 and is acceptable. I The mo<11 fication accomplished by this PC/M does not affect the floooing 4 analysis as described in the NRC Safety Evaluation Report, dated l- September 4, 1979, because the modification does not introduce a new source of

i. flooding, modify the existing flood mitigating features, or install any safety

i. related components which could be affected by flooding.

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PC/M 85-1% i Page 2 I l

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

lhe probability of -occurrence of an accident previously evaluated in the FSAR will not be increased because the modification to the fuel ,

line pressure boundary is being performed to the same standards of the i original pressure boundary and does not alter the function of the i Emergency Diesel Generator System.

The consequences of an accident previously evaluated in the FSAR will '

not increase because the added piping will not ' affect the performance !

of safety features.

The changes are minor in nature and do not change the inherent fun'. tion of any plant safety features system. Administrative procedures shall ensure 'that the bypass valve is kept closed during normal operation.

In the event that the . bypass valve is lef t open, the " Diesel Engine l Trouble" alarm will sound when the fuel level in the- Skid Tank reaches 4" from the top. Therefore, there is no possibility that an accident may be created that is dif ferent than any already evaluated in the I FSAR. ,

This modification does not affect the probability' of occurrence of .any equipment malfunction important to safety previously evaluated in the FSAR, because it does not adversely affect any equipment important ' to safety.

L Due to this modification, consequences of equipment malfunction i important to safety is not changed from one which is evaluated in the-FSAR because this modification does not change the design basis j described in the FSAR for any safety related system.

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The modification to the Emergency Diesel Generator System identified it this PC/M will not change the inherent function or design basis for the system. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than previously evalu-ateo in the FSAR will not be created.

1 This modification does not reduce Lhe margin of safety as defined in }

the bases for any Technical Specification. It enhances the availabil-ity of a safety related system in the event of a fire.

Based on the above, these modifications do not constitute an unreviewed safety question. !

.g PLANT CHANGE / MODIFICATION _86-%2 PC/M CLASSIFICATION: NS-UNIT: 4 TURNED OVER DATE: 08/01/86

SUMMARY

DATE: 09/03/86 REVISION: 0: I i

i NORMAL CONTAINMENT COOLING FAN MODIFICATIONS >

Summary:

i This design package provides modifications in the control circuits of the Normal Containment Cooling (NCC) Fans 4VlA, 4Vle, 4VIC and 4VID to prevent automatic loading of any of these fans on the Emergency Diesel Generator (EDG) on restoration of bus voltage during Loss of Offsite Power (LOOP). Should the bus voltage be lost the fans will trip and manual action will be required to restart the fans.

Currently, Fans 4VIA, 4VlB, and 4VIC are aligned to the EDGs via the vital section of Motor Control Centers 4807, 808 and 4805, respectively, and fan 4VID is aligned to the EDG via Load Center 4B02. The existing control circuit, '

automatically restarts all these fans which were running prior to Loss of '

Of fsite Power. This existing feature is being eliminated under this PC/M to reduce loading on the emergency diesel generators, gfetyEvaluation:

The existing design for Normal Containment Cooling Fans 4VIA, 4VlB, 4VIC and 4V10 includes a maintained contact control switch which allows the operating f ans to automatically start and load onto the en.ergency diesel generators in the event of a loss of offsite power. The modifications performed under this PC/M will eliminate the capability of automatic restart upon restoration of I bus voltage. Therefore, as required by 10 CFR 50.59, the following evaluation has been performed to determine if this modification constitutes an unreviewed saf ety question and requires prior NRC approval.

A proposed change, test or experiment 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 fnr any technical specification is reduced.

The probability of occurrence or the consequences of an accident or malfunc-tion of equipment important to safety previously evaluated in the FSAR is not increased by this PC/M for the following reasons:

The normal containment coolers are non-safety related and no credit is taken for their operation under loss of offsite power conditions to mitigate the ef fects of accidents considered in Chapter 14 of the FSAR.

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PC/M 86-062 Page'2 ,

' An analysis was perforemd to detereine the. temperature'inside contain- j ment without the normal containment coolers operating. This analysis ;

.(Reference Bechtel letter SFB-2507, dated May 2,1986) showed that the temperature would be maintained below 132*F for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided that I one cooler is started within one hour af ter the loss of offsite l

. power...This increased temperature over the normal maximum of 120*F !

was evaluated with respect to safe plant operation and determined to be acceptable for the following reasons:

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- Under accident' conditions, i.e., LOCA or MSLB, containment heat removal is accomplished via the emergency containment coolers and containment ' spray systems. The normal cc.1tainment coolers are i automatically de-energized under these conditions and perform no ,

active heat removal function. j An analysis performed by FPL Engineering (Reference JPE-IC-PTP 04-001, Revision 0 dated May 1986) determined that containment ;

temperature of 132*F on a non-accident unit would not adversely '!

affect equipment or instruments associated with engineered safe- 1 guards provided .that the operators can stabilize the plant after I reactor trip and maintain the conditions listed below:

Pressurizer pressure shall be maintained above 1800 psig '

through the alternate use of ' pressurizer- heaters an charging pumps. Charging pumps are needed 30 minutes out of every !

hour, to maintain the unit in hot standby. Du ring the remaining - 30 minutes at least 3 banks of pressurizer heaters of 50 kw each should be operated to make up for heat losses during the period without heater operation.

RCS pressure should be maintained less than 2258 psig to avoid automatic PORV or safety valve operation.

Containment Pressure should be maintained less than 3.0 psig.

Steam Generators pressure should be maintained less than 50 psig between SGs.

The St eam Generator pressure / steam valve feedback effect prevents the coincidence of SI actuation signals in the_SI logic for high steam flow and low SG pressure (600 psig) on ;

low Tavg (531*F) . However, operators should cut back on AFW to the SGs to assist in preventing a low Tavg below 531*F.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time limit specified in the anlaysis was considered appro-priate since at that time it is expected that sufficient diesel gener-ator load capability would exist and/or alternate means of containment cooling (i .e., use of purge system) would be possible. In addition ,

the calculation is consistent with the Appendix R safe shutdown i analysis which assumes loss of offsite power for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . l

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Page 3 l

The' operability of saf ety related t lectrical equipment and non-saf ety ,

related electrical equipment whose f ailure could prevent unsatisf ac- '

tory accomplishment of safety function is ensured since these compo-nents are cualified to much higher temperatures in accordance with the requirements of 10 CFR 50.49.

All new components and the existing relays used for this modification which are an integral part of, or 1nterface with an existing safety related system are qualified for their intended application and seis-mically installed, therefore, the integrity and reliability of existing safety related systems will not be degraded.

4 All new cables utilized for this modification are qualified for their intended application.and have been evaluated for ampacity and voltage drop ' and determined to be acceptable. In addition, all new cables will be routed in raceway installed seismically thereby precluding any adverse seismic or seismic II/I conditions.

An evaluation (Reference Bechtel letter SFB-2672 dated May 31, 1986)

' was performed to address the ability of the normal containment coolers, a and their associated discharge dampers to function at elevated temper-atures. The results of this evaluation show that operation of the coolers will not be adversely affected at a containment temperature of 132'F. The associated discharge dampers have been walked down and ,

verified on Unit 4 to be suitable for use in an ambient temperature of !

130"F. These dameprs are suitable for use since they will open at I hour, prior to the containment temperature reaching 130'F.

As specified in Attachment 5, appropriate procedural controls will be implemented by FPL to ensure that manual operation of the normal i containment coolers under loss of offsite power conditions will be l performed in a manner such that containment temperatures will not I exceed 130 F.

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

I previously evaluated in the FSAR is not created as a result of de-energizing j the normal containment coolers under loss of offsite power conditions since no new f ailure modes are introduced due to the resulting temperature increase.

An analysis perforemd by FPL Engineering, as identified above has demonstrated that equipment and instrumentation associated with safeguards actuation will not be affected by the temperature increase. Other safety and non-safety j related components required for safe plant operation are qualified for higher temperatures in accordance with the requirements of 10 CFR 50.49.

The margin of safety as defined in the basis for the Technical Specifications is not reduced since neither operation of the normal containment coolers or containment temperature are governed by Technical Specification ' limits and the normal maximum containment temperature of 120 F would only be exceeded during periods of of f-normal operation, i .e., loss of offsite power.

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.PC/M 86-062" -

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1 Based on the preceding - this modification does not cortstitute an unres ewed safety question'and prior NRC. approval is not required.

NOTE: An_ additional engineering evaluation is required to demonstrate -

that an increase- in containment temperature above the normal !

maximum of 120'F will not jeopardize the safe shutdown: capability i of a non-accident unit under loss .of offsite power conditions.

The results of this analysis will be incorporated:into_the safety j

l l- evaluation as necessary, ' prior to issuance of the PC/M. for l review / approval by the PNSC.

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'(i PLANT CHANGE / MODIFICATION 86-094 PC/M CLASSIFICATION: NS UNIT: 4 i

TURNED OVER DATE: 07/27/86

SUMMARY

DATE: 09/03/86 REVISION: 0 EMERGENCY DIESEL GENERATOR - B ROOM VENT FAN 4V34 POWER SUPPL Summary:

This PC/M changed the power and control supply. source to the EDG-B Room Vent Fan' formerfrom MCC 4AisVital section to MCC 4B Vital section.Control Power Trans-change. (CPT) size also increased from 50 VA to 100 VA as part of this The vent f an operation logic is not changed with this modification. i The PC/M.

change in- the EDG loading, as a result of this PC/M, is addressed in the The separation of raceways for the new cables to meet the Appendix R separation criteria is also addressed in the PC/M. All applicable design verification elements of EDPI 3.16-10, Exhibit H were considered during this review, o

Safety Evaluation:

The modifications specified by this PC/M package are necessary to ensure the operates. of power and control supply for EDG-B Room Vent Fan when EDG-B availability The Vent EDb-B Fanwhich system Motorhas as been a component is not Q-listed and is associated with the identified .

Shetdown systems in Appendix R analysis. as Safe Shutdown and Alternate Safe The vent fan motor should be added to the tion is FPL Q-list and important to the classified as of operation Important EDG-B.to Saf ety (ITS) since its opera-supplements and The Vent Fan when operating theref ore, important thatchanges the pattern of air flow inside EDG-8 Room. It is, should be ensured when EDG-B operates.the power and control supply to the Vent Fan M This PC/M package aligns the Room Vent Motor to EDG-B via MCC 4B vital section and as such ensures the power and control supply to the Vent Fan Motor when EDG-8 operates. The present func-tional Except control logic of the Vent Fan has not been modified by this PC/M. !

for the relocation of power and control supply, the design maintains the existing automatic / manual control features of the Vent Fan.

The modifications provided by this PC/M are associated with installation of relocated Motor Starter with larger size CPT and wiring in the MCCs. All work associated performed under with MCCs is considered safety related and, therefoe, will be appropriate Quality Assurance Programs. The CPT purchased will match the existing device / equipment quality level.

All cables associated with the modifications are environmentally qualified and all conduit supports will be designed and installed seismically.

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PC/M 86-094 L Pagt 2 L :

Construction ' activities associated with this PC/M require no unusual tech-niques or equipment , and will be accomplished in a controlled manner under existing procedures.

None of the work associated with this PC/M is in the Containment Building.

All work associated with this PC/M is in low radiation areas and, therefore, ALARA criteria is not applicable.

Penetrations through fire barriers and/or associated barrier seals, wherever needed, will be sealed with three-hour rated seals to maintain the integrity' of the fire barriers in accordance with the design documents. Conduits pene-

- l trating a fire barrier-are also sealed internally, as required, in accordance with Drawing 5610-A-182.

No work associated with this PC/M changes the ECCS heat sink analysis, nor affects radioactive waste treatment, safety related snubbers, spent _ fuel pit <

cooling systems,.or effluent monitoring system.

Cable routing for this PC/M has been reviewed in accordance with the require-ments of 10CFR50, Appendix R, and found acceptable. ,

Equipment or cable _ associated with this PC/M will not be attached to or installed in the proximity of any block walls which have not been previously analyzed to preclude their failure and subsequent damage to adjacent safety related equipment.

This modification has been reviewed with respect to the NRC Safety Evaluation Report, " Susceptibility of Safety Related System to Flooding from Failure of :

Non-Category 1 Systems", dated September 4,1979, and is acceptable for the !

following reasons:

identifiec modifications will not create new more limiting sources of {

flooding, or adversely affect the design or operation of any flood mitigating features.

Wiring added by this PC/M are located in existing equipment and as such, this PC/M does not add new equipnent which may be suceptible to <

flooding.

As required by 10 CFR 50.59, the following evaluation. has been performed to determine if this modification constitutes an unreviewed safety question and requires prior NRC approval . A proposed change, test or experiment 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 s different type than any evaluated previously in the safety analysis report may d be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.

PC/M B6-094 Page 3 I

i lhe of tion probability.

equipment of occurrence or the consequences of anf accident or malfu -

increased by this PC/M for the following reasons:important to safety previou The control logic of the vent fan has not been changed. -

The relocation 'of the power and control supply enhances the availabil-ity of the vent fan when EDG-B is operating.

i All new their components intended utilized for this modification are qualified for application.

i and voltage drop and found acceptable. Cables have been evaluated for ampa ;

i r, All new cables will be routed raceway installed . seismically thereby precluding any adverse setsmic or seismic II/I conditions.

evaluated previously in the safety analysis report is not '

Dieselacceptable, found generator loaoing has been evaluated for this modification and s

The functional control logic for the vent fan has not been modified.

This modification does not change the performance capability of the vent fan.

The margin of safety as defined in the basis for the Technical Specifications is not cal reduced since Specification limits.-the operation of the vent fan is not governed by Techni-safety question and prior NRC approval is not required. Base l

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PLANT CHANGE / MODIFICATION 86-095 PC/P CLASSIFICATION: NS UN1 3 TURN ' OVER DATE: 07-24

SUMMARY

DATE: 09-03-86 REVISION: 0 l

' PROVIDE NEW POWER FEED FOR NON-VITAL SECTION OF MCC 3A AND D Summary:

The intent of the design implementation package is to resolve a concern in the

. existing design that a single failure of the tie breaker between vital and t non-vital .,u s s es of Motor Control Centers (MCC) 3A and D can disable both Channel A and 8 diesels in providing emergency power.

{

This design modification provided a redundant means of isolating non-vital loads from the vital power source on a loss of offsite power (LOOP) or on a LOOP together with Safety injection Actuation Signal (SI).

This was achieved by physically separating the non-vital bus from the vital' bus at MCCs 3A and D and feeding the non-vital Busses directly from the sane safety related Load Centers 3A and 0 which provides the normal source of power to the vital busses of MCCs 3A and D. The feeder breaker at the load center and the tie breaker converted to the incoming breaker at the MCCs will be in series, to failure thereby trip.. . providing the necessary redunancy for the single breaker Since the trip logic for the tie breaker (now the incoming breaker to non- !

vitalaffected.

not bus) has not been changed the previously reviewed design' philosophy is Un' the other hand, the original design is improved by tripping the MCC feeder breaker at the load centers without any intentional time delay f or the design basis events mentioned above, thus quickly isolating the non-vital loads from the vital source of power. i The raceways and cables are installed by the PC/M 86-093. The installation I; meets all the criteria required for safety related equipment installation, I including seismic, Appendix R, blockwall, flooding, and nuclear qualification requirements. {

The new breakers added at Load Centers 3A and 3D are existing '

nuclear qualified spares. The modification to the bus work at MCCs 3A and D will not violate the original safety qualification for the equipment. This is based on confirmation by the equipment vendor who will provide documentation to substantiate this determination.

The design verification has considered the elements presented in exhibit H of s the EDPI 3.16-10 and a check has been made to verify compliance to the applic-able Project Standards and licensing commitments.

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' PC/M ~ B6-095 i Pa ge 2 '

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Safety Evaluation:

Per the existing plant design, the non-vital section of PCC 3A (D) is fed from the vital section1of MCC 3A (D) through a tie breaker, and is automatically shed . af ter a time delay on a loss of of fsite power (see Section 2.0) by '

tripping ~of the tie breaker. By implementation of this PC/M, the power feed for HCC 3A (D) non-vital bus is relocated form the vital bus of MCC 3A (D) to the source Load Center 3A (3D), and the existing tie breakers will be used as a the incoming breaker at the MCC non-vital bus Nv3A (NVD). NV3A (NVD) will be shed on loss of load center bus voltage, with no intentional time delay, by independent- trip circuits to the new load center Feeder Breaker 30103 (30411) and the MCC incoming Feeder Breaker (30535 (0832). ,Therefore, as required'by lDCFR50.69, the following evaluation has been performed f to determine if this modification constitutes an unreviewed safety question and requires prior NRC approval.

in accordance with 10CFR50.59 a proposed change, test or experiment shr il be deemed . to involve an unreviewed safety question (1) if the probabil ty of l occurrence or the consequences .of an accident or malfunction of equipment ;

important to safety previously evaluated in the safety analysis -report may be, .

i ncreased; ; or (ii) if a possibility for an accident or malfunction of a i different type than-any evaluated previously in the safety analysis report may be created; of (iii) if the rargin of safety as defined in the bases for any 4

technical specification is reduced. j To evaluate the above three conditions, every change resulting from this PC/M, as categorized below, has been investigated. j d

Electrical Distribution System Operating Characteristics 1

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System Response to LOOP /SI Re-energization of Non-Vital MCC NV3A (NVD) Loads Du ri ng Plant Recovery ,

L Appendix R Compliance 'l Effects of New Component Installation ,

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D PLANT CHANGE / MODIFICATION 86-041 PC/M CLASSIFICATION: SR UNIT: 3/4 TURNED OVER DATE: 7/25/86

SUMMARY

DATE: 10/20/86 REVISION: 0 Modification of MCC "D" Automatic Transfer

~Summaty:

PC/M 86-041 provides for. modification of the Automatic Transfer Circuit of MCC "D".

This evaluates single failures due to (1) Failure of Emergency Diesel Generator "B", (2)

Failure of Battery 3B/ Sequencer 3B, or (3) Failure of Breaker 3AB20. The circuit design prior to PC/M 86-041 modifications did not ensure an automatic transfer in case of failure of Battery 3D/ Sequencer 3B or failure of Breaker 3AB20 for certain conditions. This change request provides circult modifications for the above failures for various plant conditions.

Automatic transfer of MCC "D" on loss of offsite power with concurrent safety injection actuation signal on both Units is also addressed.

o Safety Evaluation:

This modification eliminates single failure susceptibilities in the MCC "D" automatic transfer ~ scheme associated with scenarios lis.ted tu Table 1. The probability of L occurrence of an accident previously evaluated in the FSAR is not increased because l~ the devices being added are environmentally and seismically qualified and seismically installed, effects on the existing cabinets have been evaluated, and no other safety related system functions or operations are jeopardized.

The devices and cable added by this PC/M are nuclear qualified. This modification does not alter the operation or function of any other plant safety system. By eliniinating single failure susceptibilities in the MCC "D" automatic transfer circuit, the availability of the MCC is increased. The existing transfer logic is retained. Therefore, the c.:. equences of an accident previously evaluated in the PS AR will not be increased.

The relays in the parallel logic added by this PC/M are applied in a fail-safe configu ration. Both relays on two load centers must drop out and remain dropped f out for 40 seconds to initiate an MCC "D" transfer. The existing permissives must also be satisfied to effect the transfer. Therefore, the probability of occurance of an equipment malfunction important to safety previously evaluated in the FSAR is not increased by this modification.

Sheet 2 of 2 PC/M 86-041 Modification of MCC "D" Automatic Transfer This modification improves the reliability of the MCC "D" transfer scheme and th.e availability of MCC "D" and the emergency containment coolers by eliminating single failure susceptibilities. Should Battery 3B fall in conjunction with a loss of offsite power on Unit 3 or Breaker 3AB20 fails to close without safety injection signal ' present, implementation of this modification ensures. that power will be available to MCC "D". This modification does not alter the operation or function of any other safety related systems. The EDG loading is not changed for the single EDG cace where one EDG has failed. However, for the two EDG cases wnere a single failure other than loss of an EDG has occured certain loads will not be stripped when MCC "D" transfers to its alternate supply. Since the auto loading of the EDG is within' the Technical Specification limitation, the consequences of an equipment malfunction important to safety previously evaluated in the FSAR is not increased by this modification.

The relays in the parallel logic added by this PC/M are applied in a fall-safe configuration. Both relays on two load centers must drop out and remain droppd out for 40 seconds to initiate and MCC "D" transfer. There are no interfaces -

with non-safety systems. The devices added are seismically and environmentally qualified and seismically installed and the effects on the existing cabinets have been evaluated. The cable is environmentajljg quallfled and seismically installed.

The loading on the emergency d!ssel generators is not increased. This modification does not alter the operation or function of any other safety related system. The existing MCC "D" transfer logic is retained, and this modification does not alter the operation or function of any other safety related system. MCC "D" and emergency containment cooler availability is improved. Therefore, the possibility of an accident of a different type than any previously evaluated in the FSAR is not created.

This modification does not reduce the integrity, . operation or function of any other safety related system addressed in the Technical Specifications. This modification improves the availability of MCC "D". Therefore, the margin of i safety. as defined in the basis for any plant Technical Specification is not decreased.

Based on the preceding, this modifiestion to the MCC "D" automatic transfer scheme does not constitute an unreviewed safety question.

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~ PLANT CHANGE / MODIFICATION 78-102B PC/M CLASSIFICATION: NS

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UNIT: 4 TURNED OVER DATE: 07-17-86

SUMMARY

DATE: 11-06-86 REVISION: 1 STEAM GENERATOR BLOWDOWN RECOVERY SYSTEM Sununary:

1 The Steam required Generator Blowdown Recovery System was designed to maintain the chemistry of the steam generator and recovery rate of 1% blowdown of maximum feedwater flow.for maximum water and heat 1% was designed to be discharged into the discharge canal. Blowdown in excess'of o

Safety Evaluation:

tion of equipment important to the safety of the plant, pre in the FSAR, has not been increased.

- or malfunction different than those previously evaluated.There is no Therefore, possibility of an it can be concluded that this PC/M does not pose any unreviewed safety questions.

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l PLANT CHANGE / MODIFICATION 84-02 PC/M CLASSIFICATION: NS UNIT: 4 )

TURNED OVER DATE: 09/15/86

SUMMARY

DATE: 11/12/86 REVISION: 0 MODIFICATION TO COMPLY WITH R.G. 1.97, REY. 3 REQUIREMENTS TO PROVIDE QUALIFIED LIMIT SWITCHES Summary: !

This modification consisted of replacing existing safety related non-qualified l'c.it switches for the Reactor Drain Tank (RCDT) and Component Cooling Water

'(CCW) containment isolatif on valves with fully qualified and documented Namco Series EA 180 limit switches.

Safety Evaluation: l This modification does not degrade the system or equipment.as follows:

1. The qualified limit switches will extend the environmental and olectrical integrity of the existing switches.

a. No system characteristics will be changed and the probability of occurrence of an accident would be no greater,

b. The consequences of an accident previously analyzed in Chapter 14.0 would.not be altered. ;

c. There is no potential to jeopardize the operation of other safety related systems.

d. The consequences of equipment malfunctions are no more severe than previously evaluated in FSAR Chapter 14.

2. The qualified limit switches do not decrease the design margins of the system, change the operation function of conditions, or affect other safety related equipment.

a. This change would not create the possibility of an accident not considered in FSAR Chapter 14.

b. The replacement limit switches would not create the possibility of malfunction of equipment not considered in Chapter 14 of the FSAR.

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c. This modification will not decrease any margin of safety discussed in any technical specification. ]

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~ PLANT. CHANGE / MODIFICATION 83-63 PC/M CLASSIFICATION:~NNSR UNIT: 3 ,

TURNED OVER DATE: ?2/12/85 SUMPARY DATE: 2/9/87 i REVISION: 0 4

i improved Floor Drain for the Containment Spray Pump Room i

Summary: This change modified the floor drain in the Containment Spray Punp Rootn to allow for better drainage. The drain piping was rerouted to provide a !

nore direct flow path to the drain header.

1 Safety Evaluation: This modification improved the containment spray pump- room '

to-function as intended. . The change was not nuclear safety - related. This I modification did not increase the possibility of. occurrence or the ;

consequences of an accident or malfunction of equipment previously evaluated in the FSAR, did not create the possibility for an accident or malfunction of 1 a different type than any evaluated previously in the FSAR and did not reduce' j the rargin of safety as defined in the basis for any technical specification. '

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s u PLANT CHANGE / MODIFICATION 83-114 PC/M CLASSIFICATION: NNS l-UNIT: 4 4/4/86 l, TURNED OVER DATE:

SU!HARY DATE: 2/16/87 REVISION: 0

-Unit 4 Reactor Cavity Filters - Lead Shielding Summary: Lead shielding and its associated supports were added inside the containment structure at the 58 foot elevation. The added shielding decreases

.the radiation dose rates in the surrounding area. .

Safety Evaluation: The shielding system does not' perform a nuclear safety related function. Since it is installed. in proximity to safety injection -

piping, it is built to withstand the maxinum earthquake loading used for the design of Turkey Points Units 3 and 4 seismic category I s t ruc' Jres in

.accordance with the FSAR. This modification does not increase- the possibility of occurence or the consequences of an accident or malfunction of equipment s

previously evaluated in the FSAR, does not create the possibility of. an accident or malfunction of a different type than any previously evaluated in the FSAR and does not reduce the margin at safety as defined in the basis for any technical specification.

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PLANT CHANGE /M00!FICATION 86-003 PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER DATE: 8/07/86

SUMMARY

DATE: 2/20/87 REVISION:

MSIV NITROGEN SUPPLY ADDITION FOR UNIT 4 (INTERIM)

Summary: ,

The PC/M provided an interim system to supply nitrogen to the 3 Main Steam Isolation Valves. This system manually actuated and serves as a backup to the existing Instrument Air Supply, to ensure that the MSIV's could be maintained closed in the event of a small steam leak downstream of the MSIV's, and the Instrument Air System is inoperable.

This system is designed in accordance with respect to FSAR paramenters and NRC concerns on valve closure.

The interim system. does not, however, address the specific requirements to achieve a second MSIV closure, single failure and other operability concerns;*

these will be addressed in the final (and permanent) Nitrogen Backup System Design.

Safety Evaluation:

This temporary change is the first step to upgrade the MSIV's to meet the FSAR closure requirements of five second closure with no steam flow and loss of instrument air. The operating limits on the instrument. air system have been specified, which permits power operation of the plant. The change does not affect the operability of the Instrument Air System with respect to closing the MSIV'S.

It is noted that the Nitrogen Supply System operability must be verified when i the system is placed in service (i.e. when air pressure < 66 psig and MSIV l will not close) in order to ensure that the MSIV's can be closed in the event 1 of a postulated accident.

i This modification is nuclear safety related with no unreviewed safety question l since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the FSAR.

This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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>3 PLANT CHANGE /M00lFICATION 86-064 PC/M CLASSIFICATION: NSR-

. (PWO) UNIT: 3&4 TURNED OVER DATE: 8/07/86 ,

SUMMARY

DATE: 2/23/87 REVISION: 0 U l

(PWO )

TITLE: 4.16K V FUSE HOLDER SUBSTITLITION Susanary:

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l 'This CPWO provided for replacement '4.16 KV SWGR fuse holders.

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Safety Evaluation:

l-A- Sa fety Evaluation has been performed which. approves the use of the replacement fuse holders. The evaluation resulted in the following

-determination. ,

lThis modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has -

not increased, nor was the possibility of an equipment malfunction / accident important to safety 'previously evaluated in the FSAR.

This modification will not decrease the margin of . safety as defined in the ,

bases _of any Technical Specification. .

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PLANT CHANGE /M00!FICATION 86-70' PC/M CLASS!FICt. TION: NNSR UNIT: 4 TURNED OVER DATE: 6/24/86 i SUPMARY DATE: 2/26/87 PC/M REVISION: 0 )

TITLE: REPAIR DAMAGED T/C CABLES AT TE-4-1418 8 TE-4-1421 Summary: Ables 4T-2000/00PS-TE-4-1418/1 for temperature element TE-4-1418 and repairable at the condulet as documented by NCR 109-86.

This modification consisted of installing a terminal box in the existing !

conduit run 4A422 and providing new thermocouple cables from the terminal box to temperature elements TE-4-1418 and TE-4-1421. This terminal box and new cables will replace the existing damaged cables as documented by NCR the condensate pump suction header piping. l Safety Evaluation: ,

This modification includes the installation of anew terminal box (TB4927)-and routing new thermocouple cables from the terminal box to TE-4-1418 and TE 1421 in order to replace the existing damaged cables. The entire modification is non-safety related consistent with the existing installation, This modification is non nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an '

equipment malfunction / accident important to safety previously evaluated in the FSAR. This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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PLANT CHANGE /M00!FICATION 86-71 PC/M CLASSIFICATION: NSR I . UNIT: 4 TURNED OVER DATE: 6/24/86

SUMMARY

DATE: 2/20/87 REVISION:

ICW BASKET STRAINER REPLACEMENT Summary: .

l This CPWO requested the replacement of the 4A ICW/CCW Rac ket Strainer due to

.the extensive corosion of the existing strainer. The new strainer was fabricated to ASME Sect VIII Di v.1 1983E0-1985 Summer Add.; the existing strainer was built to ANSI 831.1 requirements. The original strainer was

! built by Zurn Industries, and the new strainer was built by Zurn's new owner Hayward Industrial Products to the same dimensional specification, and essentially the same strainer body material. The new strainer was coated with Belzona E-C Barrier ceramic material, while the original strainer was coated with coal-tar epoxy.

Safety Evaluation: l All Dimensions and materials were provided on a one-to-one basis for the new strainer, and the new coatingis considered to be an improvement. The maximum specified nozzle loads for the strainer were found to be acceptable.

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluatad in the FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the FSAR. This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

i PLANT CHANGE /M001FICATION 82-83 PC/M CLASSIFICATION: NNS/0A/0C UNIT: 3&4 TURNED OVER DATE: 7/24/85

SUMMARY

DATE: 2/2d/87 PC/M REVISION: 0 TITLE: ADO BACK-UP TO RELAYING IN SWYD.

Summary: ,

The PC/M provided the design for installation of a solid state secondary relay-system for station and Davis 240 KV lines protection, and the installation of additional breaker failure protection for the 240 KV generator breakers.

Safety Evaluation:

The PC/M is 'non-nuclear safety related as the additional relaying does not perfrom a safety related function and is so located that it cannot affect any safety related equipment.

No unreviewed safety question exists since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety ;

previously evaluated in the FSAR. This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

PLANT CHANGE / MODIFICATION 86-044 PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER DATE: 12/01/86 i

SUMMARY

DATE: 2/20/87 REVISION: 0 i

UNIT 4 ICW BASKET STRAINER BELZONA LINING Summaty:

The original strainers were coated with coal tar epoxy, which degraded significantly over the years, which contributed to the extensive corrosion of 2 the strainer body. The new strainer is coated with Belzona E-C barrier ceramic material, and should provide greater protection due to it's superior adhesive strength, bonding strength, tensile strengt.h and toughness. The Belzona lining has demonstrated these characteristics elsewhere in 'the ICW system and other saltwater service applications. ,

Safety Evaluation:

The substitution of coal tar epoxy with Relzona provides greater protection, and the probability of the Belzona material disbonding from the strainer is no greater than that of the coal tar epoxy.

This modification is nuclear safety related with no unreviewed safety question ,

since the probability / consequences of an accident previously evaluated in the '

FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the FSAR.

This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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1 PLANT CHANGE / MODIFICATION 86-087 PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER DATE: 12/31/86

SUMMARY

DATE: 2/20/87 REVISION: O CABLE REPLACEMENT FOR MSIV SOLENOID FOR UNIT 4 Summary: ,

This PC/M provided the replacement of damaged centrol cables to the solenoids for Main Steam Isolation valves POV-4-2604 and POV-4-2605, (for damage description see NCR's #799-86 & 800-86). The original cables are no longer manufactured, so this PC/M provides a replacement cable with the same characteristics and Dwg. size as the original cable. The new cable is installed in a seismically supported conduit (cable #C24) and connect cable T84028 to the MSIV solenoids for the aforenientioned valves.

Safety Evaluation:

The replacement cable is of the same size and serve the same loads, are l qualified for the working enviornment and are seismically supported. No interfacing system loads.are affected by the new cables.

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR .has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the FSAR.

This modification will not decrease the margin of safety as defined in -the bases of any Technical Specification, t

PLANT CHANGE /M001FICATION 85-152 PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER 0 ATE: 12/18/86-

SUMMARY

OATE: 02/24/87 REVISION:

MIS INPUT TO TURBINE RUNBACK - REINSTATEMENT OF 1/4 CONFIGURATION _

Summary: . .

PC/M 85-103 provided an interim solution to the problem of NIS caused runbacks by changing the Unit 4 logic for initiating a turbine runback caused by a negative flux rate input (NIS signal) from a 1/4 to a 2/4 configuration. The PC/M stated that the " permanent solution will be forthcoming through implementation of PCM 84-211 during the next Unit 4 schedulded outage."

Because PC/M 84-211 assumes the previous 1/4 configuration, a new PCM,85-152, is required to change the affected portion of the system back to a status identical to the one before implementation of PC/M 85-103.

The task of this design package is therefore, to reinstate the system to its' original 1/4 logic for NIS initiated turbine runbacks, as a necessary precondition to implementation of PC/M 84-211.

Safety Evaluation:

This charge does not constitute an unreviewed safety question for the following:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not changed relative to this change. The logic for NIS initiated runbacks is changed back to its original 1/4 configuration. Therefore, the change would not prevent a runback from occuring to a single dropped rod.

2. The russibility for an accident or malfunction of a different type than has oeen proviously evaluated in the FSAR is not created due to the re-instatement of the logic to its original 1/4 state.

3. The margin of safety as defined in the bases for the technical specifications has not been decreased.

Based on the above discussion, we have concluded that there is no unreviewed safety issue associated with this modification. There is no change in the technical specifications. Therefore, the proposed change in the turnbine runback logic on NIS signal has been demonstrated to'be acceptable.

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1 PLANT CHANGE /M00!FICATION 85-085 PC/M CLASSIFICATION: NS !

. UNIT: 3 TURNED OVER DATE: 09/30/86 q

SUMMARY

DATE: 03/03/87 REVISION: 0

, AUXILIARY FEEDWATER CV INSTRUMENT AIR FILTRATION N00!FICATION 1

Summary:

This modification provided for the installation of new filters in the instru-ment air (nitrogen backed) supply line to each unit 3 Auxiliary Feedwater Control Valve as shown on the drawing listed on Attachment 1. The installa-tion of the filters provides better quality air to the Control Valve position-ers and actuators.

The new filt,ers were Installed upstream of the instrument air-nitrogen supply tie-in connections 10 the instrument air supply line. The filters are provided with isolation valves and bypass lines with a valve for ease of maintenance without isolating the Control Valve air supply. A new anchor, and s supports as required per 5177-PS-21 were installed to isolate the filter assembly from the downstream check valve. This isolation was required to ensure the functional and structural integrity of the safety related portion

of tha system. The new piping is 1/2-inch galvanized steel, Schedule 40.

The new filters are Parker Hannifin Standard Airline Filters designed to separate dirt particles and water. Sizing for this installation was deter-mined by the maximum flow capacity required for operation of the positioners on the Auxiliary Feedwater Control Valves. Using the flow rates required with 100 psi instrument air and 60 psi instrument air as the pressure range to operate the positioner, (13 scfm and 8.5 scfm respectively), and a required 2 to 5 psi pressure drop across the filter for efficient operation, and the particle size filtration needed. in this case 5 micron, a filter size selec-tion was made from a chart supplied by the filter manufacturer. Therefore, the filter size selected is specific for this application, and has the concur-ren'ce of the filt<tr manufacturer. This analysis is documented in Calculation M08-0424-01.

The method of operation of the air filter is such that pressurized air (instrument air at 100 psig) flows through a louvred deflector and is directed into a swirling pattern. Liquids and large dirt particles are thrown against the inside wall of the see-through polycarbonate bowl by cyclonic action and fall into a quiet zone below the lower baffle. This lower baffle maintains the quiet zone to prevent turbulent air from returning liquids and solids into

the air stream. The instrument air, now free of liquids and dirt particles, passes through a filter element sized to remove particles down to 5 microns. Clean air then flows through the outlet port. Liquids are discharged from the bowl by the automatic drain v0lve. The Parker Hannifin model number for this installation is 04F158 which translates to the following:

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04F - Minature Series L 1 - 1/4" NPTF Port Size 5 - Automatic Drain I B - 5 Micron Filtering Element I

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Safety Evaluation:-

These modifications provide for the installation 'of new filters in the instru- I ment air supply :line to each ~ Unit 4 Auxiliary Feedwater Flow Control . Valve. l

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.The air filters, their associated isolation and bypass valves and piping were j not installed to.the requirements of seismic Category I. The seismic boundary

. anchor -assures the structural integrity of the piping and components downstream ' of the anchor. . .The modifications provided by the PC/M. include passive components whose function wfil not be impaired by any design basis accident described in the FSAR.

These modifications di'd not introduce new safety related equipment which could be affected by . fire or add new combustibles which could invalidate the Fire, Zone Heat loading Analysis previously submitted to FPL per Bechtel letter SFB-1741 dated April 24, 1985. In addition, these modifications did not adversely i affect any existing or propose ~d fire protection features of the plant. There-fore, these modifications do not affect the Turkey Point Fire Protection -1 Program. -

These modifications are not inside containment, are not attached to block walls, do not involve safety related. snubbers and do not impact spent fuel ]

pool cooling operations of the plant.

No'special- ALARA considerations were required because the modifications were carried out in the areas outside of ' Radiation Control Area. l The modification did not involve the addition of electrical cable or any changes to existing raceways. ~ The final modifications accomplished by this i PC/M: did not affect the flooding analysis as described in the NRC Safety Evaluation Report dated September 4, 1979, because they do not introduce a new soorce of flooding, modify the existing ~ flood mitigating features, or install or modify any safety related components which could be affected by flooding.

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

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The probability of occurrence of an accident 'previously evaluated in the FSAR was not increased because the modifications do not alter the -'

function of Instrument Air or Nitrogen Backup Supply to AFW Control 1 Valves.

The consequences of an accident previously evaluated in the FSAR was I not increased because the added tubing, filters and valves do not affect the performance of safety features. I i_______________.___.__

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' 1 The changes which are associated with plant safety features are minor i in nature and do not change the function of the plant safety t features.. Therefore, there .is no possibility that an accident may be i created that is different than any already evaluated in the FSAR. l The probability ef occurrence of an equipment malfunction important to safety which had already been evaluated in the FSAR, will not be !

increased due to the installation of filters and isolation valves )

because this chance did not adversely affect any equipment important to (

l safety.

The consequences of equipment malfunction important to safety which had i

already been evaluated in the FSAR was not increased because the l l performance design basis was not changed from that described in the l FSAR. l i

I The modification to the AFW Control Valve Instrument Air Supply System i identified in this PC/M did not change the inherent function or design basis for the tystem. In addition, the in-line air filters will be ;

inspected end cleaned (if necessary) on a regular basis in accordance s. )

with approved system maintenance procedures, to prevent possible block-age. Therefore, the possibility of 'a malfunction of equipment impor-tant to safety which is of a different type than 'previously evaluated in the FSAR has not be created. '

i This modification provides cleaner air to the AFW Control Valve actua- J tor and thus does not reduce the margin of safety as defined in the l bases for any Technical Specification. i Based on the above, these modifications do not constitute an unreviewed safety question. i 1

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4 PLANT CHANGE /N00!FICATION 83-139 PC/M CLASSIFICATION: NNS-0A/0C )

l UNIT: 3&4 l TURNED OVER OATE: 08/20/86

SUMMARY

DATE: 03/05/87 )

REVISION: 0 HALON SUPPRESSION SYSTEM FOR APPENDIX R MODIFICATIONS l

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Summary:

The halon suppression systems are provided in response to licensing comitments to satisfy 10 CFR 50 Appendix R Section III.G requirements as described in the FPL Fire Protection Review report.

This modification provides for the installation of halon suppression systems in the Cable spreading Room (Fire Zone 98 ) and the Inverter Rooms (Fire Zones 108A end 1088). The halon suppression systems provided are autJmatic, total flooding type utilizing Halon 1301. The systems are designed to provide a concentration of 6.0 to 6.5 percent, by volume, within 10 seconds of actuation, and to maintain that concentration for a minimuim of 3') minutes.#

The concentration is based on industry standards and consideration for potential hazards to personnel.

The system consists of cross-zoned ionization detectors for actuation, local control panel, piping distribution system, and halon main and reserve supply for each area protected.

Safety Evaluation:

The probability of occurrence of an accident previously evaluated in the FSAR is not increased because these modifications do not change the function or arrangement of safety related features.

The halon suppression system is installed to Seismic II/I requirements and the discharge of halon will not thermally affect any sensitive electrical equipment. The halon storage cylinders do not need a missile shield above them because the bottles are only pressurized to 360 psi, the bottle mou':ted control heads are connected to a substantial flexible hose, and the franJng and header piping arrangement would impede control head travel.

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

FSAR: ,

All modifications are seismically installed. Therefore, the modification does not affect the consequences of any accident previously evaluated. In fact, the halon suppressica system will mitigate the consequences of a fire in the area.

With respect to the probability of malfunction of equipment important to safety previously evaluated in the FSAR:

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The modification is not associated with equipment important to safety !

previously evaluated in the FSAR and the modifications will not impact safety i related equipment since the modification additions are seismically supported and thermally will not impact sensitive electrical equipment. The halon suppression system in fact adds a margin of safety by reducing the threat of l exposure fires on equipment important to safety. l l

With respect to_ the consequences of malfunction of equipment important to safety previously evaluated in the FSAR:

The consequences of equipment malfunction important to safety which have already been evaluateo in the FSAR is not increased because the suppression system is seismically supported.

With respect to the possibility of an accident of a differenc type than any i analyzed in the FSAR" There is no possibility that an accident or malfunction of equipment important to safety *may be created which is of a different type than any already evaluated in the FSAR because these modifications are not associated with safety related systems'and are seismically supported.

o With respect to the possibility of a malfunction of a different type than any analyzed in the FSAR:

The modifications installing the halon suppression systems do not provide any

interaction with any safety related equipment and the systems are seismically installed. Therefore, the possibility of a malfunction of a different type than any analyzed in Chapter 14 of the FSAR is not created.

With respect to the margin of safety as defined in the basis for any Technical Specification: i These modifications relate to the Technical Specifications dealing with fire protection. The margi .s of safety, as defined in the associated bases, are not reduced because these modifications do not prevent the safety features from performing their intended safety functions. In fact these modifications tend to increase the margin of safety by decreasing the probability that performance of safety related features will be hindered by fire.

Based on the preceding these modifications do not constitute an unreviewed safety question. %

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I' PLANT CHANGE / MODIFICATION _85-56

PC/M CLASSIFICATION: NNSR UNIT: 3&4 TURNE0'OVER DATE: ___10/31/86

SUMMARY

DATE: 2/27/87 REVISION: 0 REPLACEENT OF INTAXE WALKWAY Summary:

Thisintake the PC/M replaced Canal. Thethe w Intake Access Walkway at the south retaining wall of slabs which were a one .alkway. was replaced with epoxy coated pre stressed for one replacement of the old slabs. The same anchoring methods eastern-most slab. were used for the new walkway with the exception of the !

The anchorage was approved per FCN. #607.

handrail (FCN #509).was replaced with a minor change to nounting to make it removableT Safety Evaluation: -

'i Thefor one ' Intake Access Walkway one replacement walkway.is non-nuclear safety related and was replaced wi Sufficient care was taken during the canal was not altered from its existing configuration. This change did not construction increase the possibility of occurrence or the malfunction of equiement previously evaluated in the FSAR, .did not create theco possibility for an accident or malfunction of a different type than any defined in the basis for any technical specification.previously eval j

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PLANT CHANGE / MOO!FICATION 84-210 PC/M CLASSIFICATION: NS

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, TURNED OVER DATE: 08-19-86

SUMMARY

DATE: 09-09-86 REVISION: 0 TURB. T RUNBACK M00!FICATIONS Susunary:

This modification increases the availability / operability of the plant by enabling operatina- to remove an unreliable input from the turbine runback logic and still i, mit the use of automatic rod control. The following modi-fications were performed:

Reconnected the bank selector switch auto contacts which were disconnected by PC/M 83-88.

Multiplied the Rod-on-' Bottom signals in the Rod Position Indication (RPI) rack to provide separate RPI signals into the turbine runback initiatjng logic., two Modified the governor runback and load limit runback logics so that either an RPI or a one-out-of-four Nuclear Instrumentation System (NIS) flux rate, signal (whec selected) will initiate both the turbine governor and load limit runback.t.

Installed a foer position, key-locked turbine runback selector switch on ,

the control console.

Disconnected the contacts on the defeat switch for the RPI input to the turbine runback logic and the control room annunciator.

Combined annunciator windows 81-7 with 82-7 and located the new alarm on window 81-7, and Modified the load limit runback logic so that a steam generator feedwater pump breaker trip with turbine first stage pressure above 60 percent load will automatically initiate a turbine runback.

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Provided an alarm via the annunciator system to indicate when the new selector switch is ' out of the normal (RPI) position or when the logic i matrix for the' RPI portion of the selector switch fails to actuate. !

l Safety Evaluation:

Some of the primary circuits that pnovide signals to Turbina Runback logic are Nuclear Safety Related. However, actual circuitry that initiates the runback logic is not safety related. There is no unreviewed safety question since NIS/RPI signal selection for turbine runback logic initiation wa's part of original design and no devices installed by this PC/M penetrate pressure '

boundary or affect any piping system analyses. None of the equipment will be installed adjacent to any block wall, and no equipment by this PC/M shall be installed inside the containment. It does not involve a significant increase in the probability or consequences of an accident previously considered and does not involve a significant decrease in safety margin.

PLANT CHANGE / MODIFICATION 85-124 PC/M CLASSIFICATION: NNS j UNIT: 3/4 TURNE0 OVER DATE:- 10/31/86 ,

SUMMARY

DATE: 02/26/87 REVISION: 0-l, l: . .

l: . REMOVAL OF SPENT RESIN PIPE IN LADERY ROOM Susuary:

This modification removed. unused portions of the resin - transfer ' lines.

Modification cut and capped'thoes portions of the piping system that remained in the auxiliary building and. physically removed thoes . portions no ' longer ;

needed. The benefits of the modification included lower dose rates in the i laundry room.

Safety Evaluation: .

i This modification is non-nuclear safety related with no unreviewed safety' i question. since the probability / consequences of an accident proviously evaluated in the FSAR has not increased, . nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in.the ..

! FSAR.

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This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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PLANT CHANGE / MOO!FICATION 86-22 PC/M CLASS!FICATION: NSR

. UNIT: 4 1

TURNED OVER DATE: 9-8-86

SUMMARY

OATE: 4-13-87

REVISION: 0 I l

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Susunary: -

l This PC/M replaced ICW. Isolation valves 4-50-307 through 4-50-310 with valves I of like kind valve 4-50-307 exhibited gear leakage. Internr. t inspection found that valve seat of valves 4-50-308 thru 4-50-310 had lost their resiliency and t may also be leaking.

Safety Evaluation:

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in thes l FSAR has not increased, nor was- the possibility (f an equipment {

malfunction / accident important to safety previously evaluated in the FSAR. l This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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l PLANT CHANGE / MODIFICATION 84-124 PC/M CLASSIFICATION: SR

'l UNIT: 4 TURNEDdvERDATE: 3/9/87 ;

SUMMARY

OATE: 4/8/87 REVISION: 0 UNIT 4 AUXILIARY FEEDWATER FLOW TRANSMITTER REPLACEENT l

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S usuna ry : 1 I

This -PC/M replaced Unit 4's Barton ITT-752 AFW flow transmitters with Rosemount 1153 flow transmitters. Transmitters replaced were FT-4-1401A & 8, FT-4-1457A & 8 and FT-4-1458A & B.

I Safety Svaluation: I This PC/M is safety related; the new transmitters are fully qualified. The, probability of occurrence, or the consequences of a design basis accident, or malfunction of equipment important to safety, as previously evaluated in the FSAR, will not be increased.

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PLANT CHANGE /M00! FICA.T!0N 84-158 PC/M CLASSIFICATION: NSR

, UNIT: 4 TURNEO OVER DATE: 12-17-85

SUMMARY

DATE: 04-13-87 REVISION: 0 EDG "48" CFD RELAY REPLACEMENT Sosunary:

This PC/M improves the fragility level of the 0/G differential circuit by reducing - the probability of relay trip due to mechanical vibration. This modification is accomplished solely by replacing the existing differentia l relays and cases, while implementing no internal or external wiring changes in the diesal generator control panel. This then precludes any new type of interaction with other safety related equipment. Therefore, this PC/M is nuclear safety related but does not involve an unreviewed safety ' question.

Safety Evaluation: #

This modification is nuclear safety related with no unreviewed safety question

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since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment mal function / accident important to safety previously evaluated in the FSAR.

This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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Pt. ANT CHANGE /M00lFICATION 83-141 PC/M CLASSIFICATf0N: NS UNIT: 3 TURNED OVER DATE: 5-21-86

SUMMARY

DATE: 04-13-87 REVISION: 0 FIRE BARRIERS FOR APPENDIX R M00!FICATIONS h=a ry :

The scoce of this PCM covers the installation and upgrading of Unit 3 fire barriers in accordance with the requirements of Section III G of Appendix R to 10CFR50. . The purpose of these fire barriers is to control or restrict the spread of fire from one fire area to another as identified in the Appendix R fire Protection Review Report. The type of modifications will include stairwell enclosures, addition of. new fire barriers, upgrading of existing fire barriers and addition of curbs around electrical manholes.

Safety Evaluation: ,

1. The probability of occurrence of an accident previously evaluated in the FSAR is not increased because the fire barriers, and. portions thereof, are designed for all applicable loads, including seismic, and the new barriers will not affect the function or operating conditions of safety related equipment.

2. A fire is not postulated to occur simultaneously with a design basis accident. However, since the barriers limit the threat of simultaneous fire exposure to systems, or portions thereof, which are redundant in the performance of safe shutdown functions, the consequences of an accident have not been increased.

3. There is no possibility that an accident will be created which is of a different type than any already evaluated in the FSAR because the only effect of the fire barrier installations is to enhance previously established fire area boundaries and does not diminish the quality of inter-spacial relationships of these areas for which credit may be taken for HVAC. In addition, the barriers have been designed seismically to preclude any interaction with safety related equipment.

4 The probability of occurrence of equipment malfunctions important to safety which have already been evaluated in the FSAR will not be increased because the intent of Appendix R and the design of these barriers precludes a common-threat exposure fire to safety features required for safe shutdown. The modifications under this PCM have no affect on the 4 equipment important to safety.

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5. The consequences of an equipment malfunction important to safety which has already been evaluated 'in the FSAR are not altered by the addition of these fire barriers. Furthermore, since safe shutdown equipment in one area is protected from an exposure fire.in 'another area where operability of the fedundant safety function equipment may be compromised, the consequences of equipment malfunction has not been increased.

6. There is no possibility that a malfunction of equipment important to ,

safety may be crerated which i s of a dif 4 rent type than any already .c 1 evaluated in the FSAR because the threat of exposure fires, component-for- '

component, remains unchanged by this modification, o

7. This modification relates to Technical Specification 4.15 with regard to -y fire barriers. This change does not reduce the margin of safety as R defined in the associated bases for limiting conditions for operation.

Based on the preceding, this modification does not constitute an unreviewed j safety question.

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y; it. ANT CHANGE / MODIFICATION'83-145 PC/M CLASSIFICATION: NS UNIT: 3 & 4>

TURNE0 OVER DATE: 8-14-86

SUMMARY

OATE: 4/2/87 REVISION: 0 FIRE DAW ERS FOR APPEN0!X R MODIFICATIONS Summary: .

These modifications provide ~ for the installation of seismically qualified fire dampers ~in'the fire barrier penetrations of the Auxiliary Building ventilation system and the Control Building ventilation system as shown on Drawings 5610-M-85/83 145 and 5610-M-91/83-145. The installation of these fire dampers is required to ensure the integrity of three hour fire rated barriers separating redundant train: of safe shutdown equipment, and is~ necessary to meet Appendix R fire protection commitments defined in the Turkey Point Fire Protection

' [, Review Report. These modifications also provide' for the installation of b.

isolation dampers used for the halon suppression system in the Control ,

building ventilation system ductwork. The halon suppression system, along with wiring ~ assoicated with damper actuation, is provided for under PCM 83-139. These modifications also provide for the necessary non-safety related ductwork modifications required for installation of the fire and- isolation

, dampers.'

t Safety Evaluation:

The portability of occurrence of an accident previously evaluated in the ,FSAR is not increased because these changes do not alter the design basis of the f acility with respect to any system or component required for plant safety.

1 The consequences of an accident previously evaluated in the FSAR will not be incrWased because c 'L '

The af fected conduit, air lines, and piping of systems and components v which are safety related that have to be rerouted due to the damper 1 instal'lation have been evaluated. These changes are minor and will

< not affect the design basis or function of the systems.

The affected system and cdmponents which are not reqaired for safe shutdown would not affect the performance of safety features during installation and as a result of operation or misoperation of the fire dampers. In addition, the change 'does not alter the basis of the ventilation systems operation as defined in the FSAR There . is no possibil k ty that an accident will be created which is of a 3 different type than anf already evaluated in the FSAR because T -

Modifications to safety relatea systems will only be accomplished in 4 ~

, accordance with aoplicable Techef:a1 Specification requirements and g< x when plant condit0cns are as specifigd in the PCM, and i

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The changes which are not associated with .any plant safety fsatures do not create any corditions'that could be associated with or be more limiting than any accidents defined in the FSAR.

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Therefore, there is no possibility that an accident may be created that is of .

a different type than any already evaluated in the FSAR. J The probability of occurrence of equipment malfunctions important to Jafety ;

which have already been evaluated in the FSAR will not be increased as. a rasult of the installation of fire dampers. Additional assurance is provided for fire damper modifications being implemented under qualified Quality 1 Assurance and Quality Control Programs. {

The consequences of an equipment malfunction important to safety which have already been evaluated in the FSAR will not be increased because the performance design basis has not been changed from that described in the FSAR.

The codifications to the ventilation system and any othr, existing system identifin in this PCM will not change the inherent function or design basis for the systems. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any already evaluated in the FSAR, will not be created.

Portions of thase modifications which are associated with safety related or fire portcetion systems shall be accomplished in accordance with Technical '

Specifico. ion requirements. Therefore, there is no reduction in the margin of safety as defined in the bases for any Technical Specification.

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PLANT CHAEE/ MODIFICATION 84-T1,., PC/M CLASSIFICATION: _ NNS-0AOC UNIT: 3&4 TURNED OVER DATE: 04-25-86

SUMMARY

-DATE: 04-13-87 REVISION: O POST ACCIDENT CONT. (PAC) AIR SAM LI E SYST. Fi.0W. TR. j hry:

The PAC provides calibratio'n readout during normal plant operation to verify the air sampling system line is open and available.

Safety Evaluation:

This PC/M is not safety related, it adds flow indication to the existing PAC air sampling line. Therefore, the probability of an accident previously evaluated in the FSAR is unchanged. Since it is not connected to any safety related system the consequences of an accident previously evaluated in the ,

FSAR are no greater.

l PLANT CHANGE / MODIFICATION 86-067 PC/M CLASSIFICATION: NSR UNIT: 4 i

TURNED OVER DATE: 3/9/87

SUMMARY

DATE: 4/2/87 REVISION: 0 i

i TURBINE AUXILIARIES - BLOCKING OF AUTO-LOADING ON DIESEL GENERATORS Summary:

This PC/M modified the control circuits of the Turning Gear Oil Pump, Bearing 011 Lift Pimp and Turning Gear Motor to prevent auto loading on the E00 on a loss of offsite power; and allow automatic starting only when offsite power is available. This was done to achieve better load management during operation of the EDG'S.

o Safety Evaluation:

This PC/M is nuclear safety related since it involves 4A & 48 sequencers. An unreviewed safety question is not involved since the turning gear system is non-class 1E and the ability to manually start the equipment is maintained.

Additionally the operability of the EDG is increased by reducing the total load automatically sequenced on following a Loop. This modification does not reduce the integrity, operation, or function of any safety related system addressed in the Tech. Specs.

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[ PLANT-CHANGE /M00!FICATION 86-005 PC/M CLASSIFICATION: NSR l- UNIT: 3 TURNED OVER DATE: 06-09-86 i

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SUMMARY

0 ATE: 04-13-87 REVISION: 0 l

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1 l MSIV NITROGEN SUPP13 ADDITION FOR UNIT 3 (INTERIM) j

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l Summary:

The PC/M provides an interim system to supply nitrogen to the 3 Main Steam Isolation Valves. This system is manually actuated and serves as a backup to the existing Instrument Air Supply, to ensure that the MSIV's could be >

maintained closed in the event of a small steam break downstream of the MSIV's i and the Instrument Air System is inoperable. t This sy. stem is designed in accordance with respect to FSAR parameters and NRC j concerns on valve closure. l The interim system does not, however, address the specific requirements to l achieve a second MSIV closure, single failure and other operability concerns; '

these will be addressed in the final (and permanent) Nitrogen Backup System, !

Desgin.

_ Safety Evaluation:

This temporary change is the first step to upgrade the MSIV's to meet the FSAR closure requirements of five second closure with no steam flow and loss of Instrument Ai r. The operating limits on the In3trument Air System have been specified, which permits power operation of the plant. The change does not ;

affect the operability of the Instrument Air System with respect to closing the MSIV's. l It is noted that the Nitrogen Supply System operability must be verified when the system is placed in service (i.e. when air press <66 psig and MSIV will not close) in order to ensure that the MSIV'$ can be closed in the event of a postualted accident.

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the FSAR.

This modification will not decrease the margin of safety as defined in the bases of any Technical Specification'."

PLANT CHANGE / MODIFICATION 85-130 PC/M CLASSIFICATION NSR j

. , . UNIT: 3 IMPLEMENTED: 3/9/87 )

SUMMARY

DATE: _3/11/87_ l s

-REVISION: 0' l i

AFW DISCHARGE FLOW CONTROL VALVE UPGRADE 4

Summary:

This PC/M modifies the AFW flow control valve trim to provide better controlabjlity at the present design and operating conditions. The existing valves have experienced ,

" oscillations" due to the valve operating outside the desirable controllability range ,

of the trim. !

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S&;ety Evaluations l This modification is consistent with all app 1l cable design requirements for the AFW system and will resolve the flow oscillation problem. Appropriate testing criteria is provided for verifying the acceptability of the modifications. )

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.Therefore, this modification does "not involve an un' reviewed safety question and is considered acceptable. ~ l l

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PUUIT CWWEE/M00!FICAT10N 85-149. PC/M CLASSIFICAT10N8 0VALITY RELATED-NNSR UNIT: 3 L TURNED. 0VER DATE: 3/9/87 SIM MAY DATE: s/11/87 REVISION: 0 SPENT FUEL P0OL AIR INLET DAltPER REPLACEMENT Susanary:

1 The engineering package provided for the replacement of the two existing air inlet dampers, integral roughing fi.iters and damper actuators of the Spent Fuel Pool Area Ventilation System. The replacement dampers are Pathway ]

Parallel Multi-Blade dampers provided with integral filters and motorized {

damper actuators. -

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i The original dampers and integral roughing filters exhibited corrosion j around the supporting frames. Examination revealed that the supporting; a j frames and dampers were only partially operable and the filter housing #: j was beyond repair. J

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1 Safety Evaluation:

The dampers, damper moun ti nos , damper actuators and associated raceways have been designed as Qua11+ Related in accordance with FSAR Appendix SA l requirements, to prevent interaction with safety related equipment or i function.

Tne installation of the replacement dampers will be performed in a. controlled manner under existing, approved plant procedures. The SFP Air Inlet Dampers do not interface with safety related equipment or perform safe shutdown functions, and do not adversely affect system operation. Therefore, the consequence of a malfunction of equipment important to safety previously !

evaluated in the FSAR will not be incrc. ed.

Based on the above evaluation and information supplied by design ~ analysis, it can be concluded that the modification specified in this PC/M does not ,

require a chan.ge to any Technical Specification nor does it constitute '

an unreviewed safety question.

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l PUWT CHANGE / MODIFICATION 82-036 PC/MCLASSIFICAT!bNi NNSR-0A/0C UNITt 4 )

TURNED.0VER DATE: 5 / 1 5 / 815 )

SUPMARY DATE: 5/11/87 )

REVISION: 0 1

SPENT FUEL PIT LEVEL INDICATOR AND ALARM a

Summary:

PC/M 82-36 addressed the installation of the new "SFP Level Alarm System".

The System consists of a level transmitter and level indicator supported by structural steel members attached to the spent fuel pool floor. Other equipment consists of a receiver support, control panel, and inputs to the control room. j The f to thesystem provides high and continuous low level alarms. S,FP level indication locally with inputs, L

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Safety Evaluation:

The SFP level alarm system is not safety related per the Power Plant Engineering I&C Section', and failure of any of the systems structural {

i supports will not adversely affect any safety related systems or components. 1 Tne SFP level indicator does not perform a nuclear safety related function. l The probability of occurrence or the consequences of a design basis accident l or malfunction of equipment important to the safety of the plant, previously I evaluated in the FSAR, has not been increased. There is no possibility l' of accident or malfunction different than those previously evaluated.

Therefore, it can te concluded that this PC/M does not pose any unreviewed safety questions. j i

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PLANT. CHANGE / MODIFICATION 84-167 PC/M CLASSIFICATION: NNSR UNIT: 3&4 l l

TURNED V0'ER DATE: 3/26/87 l

SUMMARY

DATE: 5/11/87 REVISION: 0

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3/4 DEC,0KTAMINATION SHOWER FACILITY I

1 Summary:

A shower facility was added next to the access and dress facility near the auxiliary building. The facility provides means of showering personnel for the purpose of decontamination. The area has filtered ventilation and sump which prevents the contaminant from escaping to the environment. '

Safety Evaluation: .

This shower facility does not'providt a safety function or provide protection for safety related systems or equipment. The building is located away from ;

all safety related structure systems and components.. This modification does,:

not increase the possibility of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the FSAR, does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR and does not reduce the margin of safety as defined in the basis for any technical specification dc/F2 O

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PLANT CHANGE /M00!FICATION 86-158 PC/M CLASSIFICATION: NNSR i i

UNIT: 3/4 ,!

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TURNED'0VER DATE: 11/18/86

SUMMARY

DATE: 5/11/87 r REVISION: 0 ;

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. ' i UNIT 3/4 THROWBOLT REPLACEMENT ON INTAKE STRUCTURE BAYS HATCHWAY L j I Summaryl l U-Bolts and padlocks were installed to replace the throwbolts on the intake l hatchcovers east of the travelling screens. This modification was necessary ]

to ensure security from infiltrators. i Safety Evaluation: j 1 This modification replaced the throwbolts on the intake hatchways with U-bolts :

l and padlocks. There was na impact on the intake concrete or hatch struc- )

ture. Therefore this modification did not increase the possibility of occurre ;

ence or the consequences of an accident or malfunction of equipment previously',

evaluated in the FSAR, did not create the possibility of an accident of mal-function of a different type than any evaluated previously in the FSAR and did l not reduce the margin of safety as defined in the bisis for any technical l specification. l i

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PLANT CHANGE /N001FICATION 86-166. PC/M CLAS$1FICATION: OUALITY RELATED, NNSR

>- UNIT: 3 l

TURNED OVER DATE: 3/17/87

SUMMARY

DATE: 5/11/87 REY!SION: 0 i

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l, FUEL TRANSFER SYSTEM CABLE DRIVF f40DIFICdTinN - Df9a99FMRlY nF PRESENT SYSTEM Suenary *., ,

This Engineering Package directed the disassembly of the Unit 3 air motor i drive fuel transfer system. The new cable drive system was installed under !

PC/M 85-53, Fuel Transfer System, Cable Drive fiodification -

New System Installation, which replaced the old system.

Implementation of this modification, together with PC/M 85-53 improved- l the reliability of the fuel transfer system and, therefore, reduced. . 1 maintenance activities, which ' occasionally affect the critical path for'I "

refueling outages. !

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Safety Evaluation-The disassembly of the Unit 3 fuel transfer air operated motors and chain drive systems does not ilave any direct effects on nuclear safety since .

the fuel transfer system is not safety related. It could, however, l indirectly affect the function of the new transfer system components, and l of the transfer and refueling canal liner plates, if these components are l damaged during the disassembly effort. As discussed in the Design Analysis, l QC inspections of these components will be required af ter the completion I of the disassembly effort to insure that no damage has occurred.

Based on the above, the probability of occurrence or the consequences of a design basis accident or malfunction of equipment important to the safety of the plant has not been increased. There is no possibility of an accident or malfunction diTferenc than those previously evaluated in the FSAR. ,

Also, the margin of safety as defined in the Plant Technical Specifications l has not been reduced. Therefora, it can be concluded that these -

modifications to the fuel transfer system do not pose an unreviewed safety question pursuant to 10 CFR 50.59.

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PLANT CWWEE/M00!FICATION 86-2n7 PC/N CLASSIFICATION: 00AL f TY RELATE 1

UNIT: 3 TURNED DVER DATE: _3/19/87

SUMMARY

DATE: 5/11/87 REVISION: 0 t

IST GAUGE INSTALLATION FOR THE $ PENT FUEL PIT COOL!tlG PUMPS Suasary:

The engineering package installed instrumentation for Inservice Testing of the Unit 3 Spent Fuel Pit Cooling Pumps. Specifically it installed a flow element and flow gauge for flow measurement and pump suction and discharge gauges for pressure measurement. This was repaired because these pumps nave been added to the Turkey Point Inservice Testing Program.

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Safety Evaluation:

o The flow element added by this engineering package is classified as Quality Related for seismic con's iderations. The flow and pressure gauges are flot fluclear Safety Related. None of these instruments perform a safety function.

The work associated with this modification is Quality Related. This modification does not constitute an unreviewed safety question.

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PLANT CHANGE / MODIFICATION 86-159 PC/M CLASSIFICATION: NNSR UNIT: 3&4 TURNED'0VER DATE: 11-8-86 t

SUMMARY

DATE: 5-11-87 REVISION: 0 j 3/4 INTAKE STRUCTURE WOOD GRATING LATCHES Summary:

This modification installed locking devices on the east ends of the wood grating covers just east of the travelling sreens at the 16' elevation. The covers were required to be lockable for security reasons.

Safety Evaluation:

This change installed a locking device to the' existing wood grating covers.

l The installation did not affect the structural integrity of either the con-l crete slab of wood grating. Therefore, this modification did not increase the ;

probability of occurrence or the consequences of an accident or malfunction of , ,

equipmentpreviouslyevaluatedintheFSAR,anddidnotcreatethepossibilit/

of an accident or malfunction of a different tyoe than any evaluated previous- '

ly in the FSAR and did not reduce the margin of safety as defined in the basis for any technical specifications.

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PLANT CHANGE /M00!FICATION 86-124 _ PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER DATE: 11/26/86

SUMMARY

DATE: 7/1/87 REVISION: 0 4

REPLACE OF HIGH RANGE G#MA RADIATION READOUT MODULE Summary:

The PCM provided for design ~ modification to the Containment high range gamma radiation readout modules for Ra T-4 6311A and B for Unit 4, located in the '

control room on vertical panels 4g1, 82, these readout modules were not capable of full scale deflection (10 R/HR) as required for the Channel check and channel fuction test outlines. The modifications to ghe reagout modules provided the capability of operating withia t'ne range of 10 to 10 R/HR. The modifications were internal to the equipment and do not effect original qualifications to the inherent function, or design basis cf the system.

Safety Evaluation:

The entire modification was nuclear safety related. The changes were very I i

' minor and did not coastitute an unreviewed safety question and is considered acceptable. ,

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Pt. ANT CHANGE / MODIFICATION 83-140 PC/M CLASSIFICATION: NNSR UNIT: 384 TURNED OVER DATE: 12/01/86

SUMMARY

DATE: 7/1/87 REVISION: 0 FIRE DETECTION FOR APPENDIX R MODIFICATIONS 4

Sumary:

This PC/M installed a new low voltage fire detection system in vital areas of the plant to meet the requirement of appendix R.

Safety Evaluation:

This PC/M is not safety related and has a minimum interface with S.R.

equipment also was implemented in a controlled manner. It will improve the reliability of the system, therefore, the probability of the occurrence of an accident previously evaluated in the FSAR will not increase. The consequences '

of an accident previously evaluated in the FSAR will not increase.

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84-124 PC/M CLASSIFICATION: SR PLANT CHANGE / MODIFICATION UNIT: 4 TURNED OVER DATE: 3/9/87

SUMMARY

DATE: 4/8/87 REVISION: 0 i

UNIT 4 AUXILIARY FEEDWATER FLOW TRANSMITTER REPLACEMENT Summary:

This PC/M replaced Unit 4's Barton ITT-752 AFW flow transmitters with Rosemount 1153 flow transmitters, Transmitters replaced were FT-4-1401 A & B, FT-4-1457A & B and FT-4-1458A & B.

Safety Evaluation:

This PC/M is safety related; the new transmitters are fully qualified. The ,

probability of occurrence, or the consequerices of a design basis accident, or' malfunction of equipment important to safety, as previously evaluated in the

.FSAR, will not be increased. ,

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!; PUWT CHANGE / MODIFICATION 82-311 PC/M CLASSIFICATION: SR l 3

p VNIT: 3 TURNED OVER DATE: 3/13/87

SUMMARY

DATE: 7/1/87 REVISION: 0 AUXILIARYFEEDWATERTURBTfESTEAMSUPPLYSTOP/CHECKVALVE.

Summary:

This modification consistec in replacing existing Walworth valves for the auxiliary feedwater pump valves with ones from Pacific Valve Company. The original valve manufacturer could not provide replacement valves that met current required specifications.

Safety Evaluation:

This modification is nuclear safety related with no unreviewed safety question since valves with original specifications were installed, with no change in i' probability of malfunction / accident previously analyzed in the FSAR. The '

margin of safety was not decreased as previously analyzed in the.FSAR.

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PLANT CHANGE / MODIFICATION 85-143 PC/M CLASSIFICATION: NSR UNIT: 3&4 i

TURrlEO OVER DATE: 6/13/87

SUMMARY

DATE: 7/1/87-REVISION: 0 _ _ _

l BREAKER / FUSE COORDINATION MODIFICATIONS Summary:

, This PC/M changed the power supplies to the 120 V AC vital subpanels from a vital panel breaker to the vital panel main breaker. It replaced the 480V-SWGR BXRS. 3(4)0112, 3(4)0206, 3(4)0306, 3(4)0407, 30406 & 40311 trip settings from 1000 amps. to.800 amps.

Safety Evaluation:

This PC/M is safety related. It does not involve an unreviewed safety question since the reliability of the 120V AC vital panels a subpanels and .

i 480 V 1aad center breakers does not change or affect the design basis function' of the axisting plant systems.

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' PLANT CHANGE / MODIFICATION ' 85-60 PC/M C1. ASS!FICATION: 'NNS l

1 l- UNIT: 3 TURNED OVER DATE: 6-11-87

SUMMARY

DATE: '7-1-87 l REVISION: 0 i

MAIN TRANSFORMERS FAN COOLER UPGRADE l

Sunnary:

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-This PC/M added 10 fans for additional cooling capacity to the main I transformer, j Safety Evaluation:

This PC/M is not safety related. This modification involves the one for one replacement of original fan cooler units with improved fan cooler units, and does not af fect any S.R. equipment. Therefore, no unreviewed safety question is involved. o i

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I PLANT CHANGE /M00!FICATION 84-171 PC/M CLASSIFICATION: SR UNIT: 3 TURNED 0VER DATE: 2/2/87

SUMMARY

DATE: 7/1/87 i

REVISION: 0 1

.l MODIFICATIONS TO ALLEVIATE SHORTAGE OF COMPARTMENTS ON MCC 38 i

l Summary:

This PCM relocated the feeder from MCC 38 (3806) to load center 38 (3802) for normal containment cooler fari 38.

Safety Evaluation:

This PC/M is safety related, this change does not alter the reliability of the normal containment cooling fan 38, the probability of occurrence of. an accident previously evaluated in the FSAR is not increased. Also changes do l not constitute an unresolved safety question. , l I

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I Pl. ANT CHANGE /M00!FICATION 86-236 PC/M CLASS!FICATION: NNSR l

. UNIT: 3&4 !

TURNED OVER DATE: 4/4/87 l

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SUMMARY

DATE: 7/1/87 l l

REVISION: 0 -]

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l REPLACEMENT OF TREATED WATER PUMP SEALS I

Summary:

1 The treated water pump (P-18A AND P-188) packing was replaced with mechanical split seals to eminate the excessive leakage and required maintenance. {

Safety Evaluation:

This change was made to non-safety related components and did not alter or i impact any safety related components or systems. No unreviewed safety ;

question existed as a result of this modification.

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-I PLANT CHANGE / MODIFICATION 85-197 PC/M CLASSIFICATION: NNSR UNIT: 3 TURNED OVER DATE: 9/26/86 l

SUMMARY

DATE: 7/1/87 REVISION: 0 RELOCATION OF INSTRUMENT AIR SUPPLY val'iCS 40-4-098 AND 40-3-641.

Summary:

This modification relocate'd valves 40-4-098 and 40-3-641, which supply instrument air to the Unit 3 MFW and AFW flow control valves. This relocation was needed to make these valves more accessible.

Safety Evaluation:

This modification was evaluated and it was concluded that no unreviewed safety question existed as a result of' the modification, primarily because the only changes were the physical location of the valves, and the new locatforis were, more accessible to the operators.

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PLANT CHANGE / MODIFICATION 85-170 PC/M CLASSIFICATION: SR UNIT: 3 TURNE0 OVER DATE: 1/15/87

SUMMARY

DATE: 7/1/87 REVISION: 0 ,

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INSTALLATION OF AFW VALVE ACCESS PLATFORMS Summary:

This PCM provided two new . access platforms with their associated component lighting for the auxiliary feedwater flow control valves (CV-3-2831, 2832, and 2833) and manual isolation valves (3-141, 241, 341; AFPD-3-006, 007 and 008) ,

which were not easily accessible. l Safety Evaluation:

The platforms do not perform safety related functions and are not associated

- with safety related systems. However, they are in the vicinity of safety related equipment. Therefore , they have been designed for seismic II/I considerations, hurricane, and tornado wind loads. In addition, they are, protected from externally generated missiles due to their location under Unit ,

3 feedwater platform. the design neither creates a new condition nor altered an existing condition in the plant which has not been a'nalyzed in the FSAR.

Therefore, the design did not create an unreviewed safety question.

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PC/M CLASSIFICATION: NSR PLANT CHANGE / MODIFICATION 86-067 UNIT: 4 TURNED OVER DATE: 3/9/87

SUMMARY

DATE: 4/2/87 REVISION: 0

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TURBINE AUXILIARIES - BLOCKING OF AUTO-LOADING ON DIESEL GENERATORS Summary:

This PC/M modified the control circuits of the Turning Gear Oil Pump, 8 earing Oil Lift Pump and Turning Gear Motor to prevent auto loading on the EDG on a loss of offsite power; and allow automatic starting only when offsite power is available. This was done to achieve better load management during operation of the EDG'S, o

Safety Evaluation:

This PC/M is nuclear safety related since it involves 4A & 4B sequencers. An unreviewed safety question is not involved since the turning gear system is non-class 1E and the ability to manually start the equipment is maintained. '

Additionally the operability of the EDG is increased by reducing the total i load automatically sequenced on following a Loop. This modification does not reduce the integrity, operation, or function of any rafety related system addressed in the Tech. Specs. ,

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i PLANT CHANGE / MODIFICATION 87-086 i PC/M CLASSIFICATION: SR

- - 7 UNIT: 4 l

TURNED OVER DATE: 3/28/87 !

SUMMARY

DATE: 6/18/87 l

REVISION: 0 PERSONNEL AIRLOCK E00ALIZATION VALVE REPLACEMENT l Sumn ary: ,

The Unit 4 Personnel Hatch air lock equalization valve was replaced with a k valve determined to be identical with the exception of the valve stems. This replacement valve stem was modified to match the original valve stem prior to installation. After installation, the new valve was tested satisfactorily.

Safety Evaluation: ,

This modification is nuclear safety related. Thw replacement valve and linkage meet the original design requirements and function in the same manner ,

as the original. This modification did not increase the possibility of ]

occurrence or the consequences of an accident or malfunction of equipment .

previously evaluated in the FSAR, does - not create the possibility for an accident or malfuention of a different type than any evaluated proviously in the FSAR and does not reduce the margin of safety as defined in the basis for any technical specification. j q

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d PLANT CHANGE /M00!FICATION 87-095 PC/M CLASSIFICATION: SR ,

UNIT: 4 TURNEO OVER DATE: 5/14/87 l-

SUMMARY

DATE: 6-18-87 l REVISION: 0 I

4 A ICW PlMP ANCHOR BOLTS REPLACEMENT I

Summary: .

The Southeast and Southwe'st anchor bolts for the 4A ICW pump base were replaced. The bolts were replaced with through bolts of a larger diameter and a plate on the lower side, which provide greater anchorage than .he originally designed anchor bolts. In the process of changing the anchor i bolts, some of instrument tubing used to measure differential pressure of the travelling screens was routed to avoid interference with the through bolt.

-Safety Evaluation- l This anchor bolt replacement is Nuclear Safety Related since the ICW pumps are safety. related. The anchor bolts installed by this CPWO provide a greater ,

anchoring capacity than the originally designed and installed anchor bolts. I The new anchor bolts do not affect the operation of the ICW pump. The l instrument tubing to the travelling screen differential pressure indicator was j not affected by the rerouting of the tubing. Therefore this modification did )

not increase the possibility of occurrence of the consequences of an accident j or malfunction of equipment previously evaluated in the FSAR, does not create i the possibility for an accident or malfuention of a different type than any )

evaluated proviously in the FSAR and does not reduce the margin of safety as i defined in the basis for any technical specification.

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PLANT CHAME/ MODIFICATION 84-121 PC/M CLASSIFICATION: NS UNIT: 3 TURNED OVER DATE: 3/30/87

SUMMARY

DATE: 7/7/87 REyISION: 3 UPENDER LEVELING DEVICE MODIFICATION - UNIT 3 Summary:

Structural Modi fications were made to the Unit 3 Refulling Po o'. Upender Leveling Device Bracket Attachment including Bracket shim, Plate removal, Bracket modification, and the addition of stiffener plates to the existing Embed Plate.

Safety Evaluation-I This Modification does not involve safety related Snubwrs, does not change SFP operation, and does not affect Block Walls. It wil; also not cause or be !

adversely affected by flooding. This Modification does not alter the design '

bases or function of the_ upender leveling device.

This change does not constitute an unreviewed safety question. It does not change or increase the probability of an accident previously evaluated in the FSAR.

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' PLANT CHANGE / MODIFICATION 86-029 UNIT: 4 TURNED OYER DATE: 3/28/87

SUMMARY

DATE: 7/7/87 i REVISION: 0 AFW LOCAL INDICATION UNDER THE' MAIN FEE 0 WATER PLATFORM ,

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Summary:

This PCM provided for the installation of ficw indicators for AFW Trains I and 2, and Steam Generator wide range level indicators under the Main Feedwater Platform for use during manual operation of the AFW Train 2 Flow Control Valves for Unit '4, in case of a Control Room evacuation. In addition, existing flow indicators located on the Main Feedwater Platform were replaced ,

to maintain system compatibility and reliability.

Safety Evaluation: l o l The previous design requires that plant operations depend on radio communication when minually operating the Train 2 Fl ow Control Valves.

Operation of these valves is required during Control Room evacuation.

Installation of new flow and level indicators will help preclude possible errors during a Control Room evacuation and will increase overall reliability of manual system operation and operator action. The modification under this PCM does not constitute an unreviewed sa'aty question.

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i PLANT CHANGE / MODIFICATION 86-021 PC/M CLASSIFICATION: SR-UNIT: 3&4

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TURNED OVER DATE: 4/13/87

SUMMARY

DATE: 7/7/87 REVISION: 0 "C" AUXILIARY FEED PLMP REPLACEMENT IMPELLER Summary: -

This CPWO replaces the currently installed rotating element in the Auxiliary Feedwater Pump C with the spare rotating element. The impeller vanes of the spare rotating element have been underfiled in the vendor shop in order to increase the head of the Auxiliary Feedwater pump. The replacement rotating element is identical to the previously installed element except for the <

underfiled 10peller vanes.

Safety Evaluation:

The replacement of the pump impeller assembly (rotating element) does not* ,

involve an unresolved safety question because the probability of occurrence or l the consequences of a design basis accident or malfunction of equiphient important to safety previously evaluated in the FSAR is not increased, the possibility for an accident or malfunction of a different type than. evaluated previously in the FSAR is not created, and the margin of safety as defined in the basis for a Technical Specification is not reduced.

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l PLANT CHANGE /M00!FICATION 86-212 PC/M CLASSIFICATION: SR-UNIT: 3&4 TURNED OVER DATE: 2-12-87

SUMMARY

BATE: 7-7-87 REVISION: 0 ENVIRONMENTAL QUALIFICATION LIST REVISION i Summa ry:

The purpose of this PC/M was to correct EQ list deficiencies and update EQ document packages. No hardware change was involved.

Safety Evaluation: ,

The PC/M was. designated SR because the E0 list and associated Doc. Packs are SR. This PC/M did not make a physical change to the plant. No technical i specification change was required. The PC/M does not constitute an unreviewed (

safety question. , j i

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I PLANT CHANGE /M00!FICATION 86-103 PC/M CLASSIFICATION: SR UNIT: 3&4 TURNED _OVER DATE: 2/12/87

SUMMARY

DATE: 7/7/87 REVISION: 0 l'

ENVIR0 MENTAL 00ALIFICATION LIST REVISION Summary:

The PCM updated EQ List based upon the E0 update E0 Doc Packs and the !

associated review of PC/M'S for the purpose of identifying equipment and components within the scope of 10 CFR 50.49. The PCM only revised drawing 5610-E-1435 rev 1 (EQ LIST) and involved no plant Mod's. ,

l Safety Evaluation

I The PCM is S.R. but did not constitute an unreviewed safety question, did not require any changes to TECH. SPEC'S and therefore is considered acceptable.

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PLANT CHANGE /M00!FICATION '85-65 PC/M CLASSIFICATION: SR j

UNIT: 4 f

TURNED OVER DATE: 1/15/87 -l

SUMMARY

'0 ATE: 7/7/87 l REVISION: 0 l l

l G.E. SAM RELAY N00., P.C. CARD REPLACEMENT I l

l1 Summary: l l

This PC/M replaced a printed circuit card on existing SAM Relays to eliminate !

the possibility of incorrect timing due to initiating contact bounce.

Safety Evaluation: .

This PC/M.is. safety related and does not . involve an unreviewed safety question i because it involves the replacing of a P.C. with a slightly modified card. No .j external wiring changes were made and the relay's function remain unchanged. i

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PLANT CHANGE / MODIFICATION 83-153 PC/M CLASSIFICATION:. SR 1

UNIT: 3&4 i

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TURNED OVER DATE: 12/18/86 l

SUMMARY

.DATE: ,

7/7/87 f REVISION: 0 )

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f' CABLE REROUTING - APPENDIX R MOOS. UNITS 3 & 4.

Summary:

This PC/M re-routed cables' in conduits to ensure hot and/or cold shutdown capability and for 'which protection in place is not practical to meet section III G.2 of appendix R.

Safety Evaluation: i

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This PC/M is safety related, the probability of an accident previously evaluated in the FSAR will'not be increased as there are no additional changes other than the pulling and termination of new cables in new raceways which are '

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e PLANT CHANGE /M00lFICATION 87-023 PC/M CLASSIFICATION: OR ;

UNIT: 4

)

TURNED O'VER DATE: 6/19/87

SUMMARY

DATE: 7/7/87 REVISION: O l

MAIN STEAM HYDRAULIC SNUBBER REPLACEMENT I

Summary: .

PC/M # 87-023 provided for the replacement of certain hydraulic snubbers in i the Main Steam System for Turkey Point Unit 4. The affected supports for this

. snubber replacement effort were as follows: !

- MSHX-1 * - MSHX-7 ,

- MSHX-2 * - MSHX-8 I

- MSHX-4A * - MSHX-10 )

- MSHX-4B The replacement of these snubbers was necessitated by a history of problems 1 and poor performance of this particular type of hydraulic snu',ber. PC/M # l 87-023 replaced these hydraulic snubbers with Anchor /rarling mechanical I snubbers manufactured in accordance with ASME Section III, aubsection NF which were determined to be equal to or better than the original hydraulic j snubbers. The replacement mechanical snubbers prov.de maintenance free l operation, a capacity to test in place and a suitabili',y for operation in the !

vibration environment of the Main Steam piping.

Safety Evaluation:

The Hydraulic snubbers that were replaced by PC/M # 87-023 are classified as I Quality Related and as such are not addressed in the Turkey Point Technical l Specification, Sections 3.13 and 4.14. Furthermore, the current FSAR and !

Technical Specification specify snubber examination and test requirements only on safety related snubbers.

However, the replacement of the subject hydraulic snubbers in the Main Steam l System with mechanical snubbers that.a.re equal to or better than the existing snubbers does not affect nuclear safety. This modification is considered a one-for-one replacement and therefore does not impact on existing safety i related functions, create malfuncticas of a different type than previously !

evaluated in the safety analysis report or reduce the margin of safety for any technical specification.

PC/M 87-023 is provided with a written safety evaluation in accordance with 10 )

CFR 50.59. The safety evaluation concludes that the change performed under the PCM does not involve an unreviewed safety question, consequently prior commission approval for the implementation of the modification was not i required. I l

PC/M CLASSIFICATION: SR PLANT CHANGE / MODIFICATION 86-96 UNIT: 4 t

TURNED OVER DATE: 9-8-86

SUMMARY

DATE: 7-7-87 REVISION: 0 NEW POWER FEED TO NON-VITAL SECTION OF MCC 4A Summary:

PCM eliminated the consequences of a failure of MCC 4A non-vital to vital tie breaker to trip during EOG load sequencing.

Safety Evaluation:

Per the existing plant design, the non-vital section of MCC 4A is fed from the vital section of MCC 4A through a tie breaker, and is automatically shed after a time delay on a loss of offsite power by tripping of the tie breaker... By.

implementation of this PCM, the power feed for MCC 4A non-vital bus is (61o-/

cated from the vital bus of MCC 4A to the source Load Center 4A, and the existing tie breaker will be used as the incoming breaker at the MCC non-vital bus NV4A. NV4A will be shed on loss of load center bus voltage, with no intentional time delay, by independent trip circuits to the new load center Feeder Breaker 40103 and the MCC incoming Feeder Breaker 40535 is required by 10CFR50.59, an evaluation has been performed to determine if this modification constitutes an unreviewed safety question and requires prior NRC approval.

The results are as follows.

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment malfunction /-

accident important to safety previously evaluated in the FSAR. This modifica-tion will not decrease the margin of safety as defined in the bases of any Technical Specification.

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PLANT. CHANGE / MODIFICATION 86-060 PC/M CLASSIFICATION: NSR UNIT: 3&4 .l

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TURNED OVER DATE: 2-20-87

SUMMARY

DATE: 7-7-87 REVISION: O COMPUTER ROOM TEMPERATURE INDICATION Summary:-

This PCM adds two redundant, temperature indication loops to enable the control room operators to monitor the Computer room temperature. The operators need thf s information to determine when to restart the chillers and HVAC equipment <

following a loss of offsite power. Thus, the load management program for the diesels will be enhanced by not adding unnecessary KW. i Safety Evaluation:

The chillers' and air handling equipment associated with the. Gomputer Room do l not automatically reload onto the vital AC bus after SI and/or LOOP. Since

the Computer Room houses safety related equipment (ICCS cabinets) which have finite temperature requirements, the operator must manually restart the HVAC before reaching the temp. limit. This PC/M, therefore,'does not constitute an unreviewed safety question; it provides the operator with vital information to

. prevent violation of operating limits.

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PLAKT CHANGE / MODIFICATION 85-071 PC/M CLASSIFICATION: NNS-0A/0C UNIT: _

3 TURNED OVER DATE: 4-15-87

SUMMARY

DATE: 7-7-87 REVISION: 2 SPENT FUEL PIT BUILDING WALL JOINT REPAIR I

l Summary:

l This modification provided for the repair of expar-ion joints in the South l wall of the SFP Bldg., and in the South Door Column in the East wall. The l

existing material was replaced by an epoxidized polyurethane sealer. The new Joint material is a "Dymeric" Sealant, backed by a polyethylene Rod.

i Safety Evaluation:

l The new joint material does not perform a safety related function the I installation, on failure of this new material will not affect any safety related system. This does not pose an unreviewed safety question pursuant to ,

10CFR 50.59.

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s PLANT CHANGE / MODIFICATION 85-35 PC/M CLASSIFICATION. SR UNIT: 3 TURNEDdVERDATE: 6/19/87

SUMMARY

DATE: 7/7/87 REVISION: 0 REPLACEMENT OF PYC0 RTD'S Summary:

This PC/M replaces PYC0 RTO's used in the RCS and charcoal filters with CONAX RTO'S. The PYC0 RTD'S had reached the end of their qualified life thus necessitating replacement.

Safety Evaluation:

This PC/M did not change the function of the SR RCS RTO'S. The replacement CONAX RTO'S are fully qualified. No other SR equipment was affected by this PC/M. This PC/M did not constitute an unreviewed safety question. ,

PLANT CHANGE / MODIFICATION 87-97 PC/M CLASSIFICATION: SR

. UNIT: 3 TURNEO OVER DATE: 6/27/87 l-

SUMMARY

DATE: 7/13/87 REVISION: 0 INSTALLATION OF UV TRIP CIRCUITRY FOR TURBINE & POLAR CRANE BREAKFRL .

Summary:

This PC/M installed trip coils & associated circuitry in breakers for Turbine Crane & Polar Crane. These will trip the breakers when an undervoltage condition is sensed on the affected 480V bus. This is to ensure that the loads will be disconnected from the Emergency Diesel Generators following a loss of offsite Power.

Safety Evaluation: #

This PC/M is safety related. It does not involve .an unreviewed safety question since it does not add or change any equipment required to operate during an accident & it does not involve any Technical Specification.

Additionally, it improves the margin of safety of the E0G's by minimizing loads that could be auto connected .

PLANT CHANGE / MODIFICATION CPWO 87-52 PC/M CLASSIFICATION: NNSR UNIT: 3 TURNED 0VER DATE: 6/9/87

SUMMARY

DATE: 7/20/87 REVISION: 0-UNIT 3 GENERATOR NEUTRAL GROUNDING TRANSFORMER REPLACEENT Sunnargy This CPWO replaced the existing gerneator neutral grounding transformer which contained PCB's with one that did not contain PCB's. The replacements are manufactured to original physical & operational design requirements Safety Evaluation:

This CPWO is non-safety related because the transformer is part of. the main generator & is dated phase bus system which is non-safety related, and non-essential for safe shutdown. This change does not involve an unreview'e'd; safety question and no TECH SPEC changes are requires.

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PLANT CHANGE / MODIFICATION CPel0 87-53 PC/M CLASSIFICATION: NNSR c UNIT: 4 TURNED O'VER OATE: 6/9/87 '

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SUMMARY

DATE: 7/20/87' j REVISION: 0 l

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UNIT 4 GENERATOR NEUTRAL GROUNDING TRANSFORER REPLACEENT .

1 Sunnary:

This CPWO replaced the existing gerneator neutral grounding transformer which contained PCB's with one that did not contain PCB's. The replacements are 'j manufactured to original physical & operational design requirements i

Safety Evaluation:

This CPWO is non-safety related because the transformer is part of the main generator & is dated phase bus system which is non-safety related, and non-essential for safe shutdoen. This change does not involve an unreviewed ,

safety question and no TECH SPEC changes are requires.

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PC/M CLASSIFICATION: SR l PLANT CHANGE / MODIFICATION 98 I

UNIT: 4 l- .

! TURNED OVER DATE: 6/23/87

SUMMARY

DATE: 7/20/87 REVISION: 0 .

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INSTALLATION OF UNDERVOLTAGE TRIP DEVICE FOR P0LAR CRANE BREAKER i

Summary 1 This PC/M installed trip cofis 8 associated circuitry in breaker for Polar Crane. This will trip the breaker when an undervoltage condition is sensed on i the affected 4808 bus. This is to ensure that the Polar Crane will be disconnected from the Emergency Diesel Generator following a loss of offsite power.

Safety Evaluation:

This PC/M is safety related. It does not involve an unreviewed safety

' question since it does not add or change any equipment required to operate' during an accident & it does not involve any Technical Specifications.

Additionally, it improves the margin of safety of the EOG by minimizing loads that can be auto connected.

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PLANT CHANGE / MODIFICATION '80-117 PC/M CLASSIFICATION: NNS !

t UNIT: 3&4 1 TURNED OVER DATE: 11-15-85

SUMMARY

DATE: 7/29/87 )

REVISION: 3 l UPGRADE AUXILIARY FEE 0 WATER SUCTION, DISCHARGE AND STEAM SUPPLY PIPING l 1

Sunnary:

1 This modification consisted of adding redundant steam supplies to the AFW. !

trubines. The modification also replaced the auxiliary feedwater control {

valves and removed the following lines from the condensate storage tank .)

discharge line: condensate makeup / reject line, condensate recovery system 0 discharge line and condensate transfer pump line. l Safety Evaluation: .

1 This change -is safety related but does not involve an unreviewed safety '

question as this modification does not. affect, create or increase the'

probablity of occurrence of any accident / malfunction already addressed, or i new, in the FSAR. j l

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PLANT CHANGE /M00IFICAT!0N ,84-111-PC/M CLASSIFICATION: NNS = OA/0C

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UNIT: 3a4 i TURNED.0VER DATE: 10/31/86

SUMMARY

DATE: 07/29/87 'l REVISION: 0 l A/C UNIT FOR PASS CONTROL PANEL l

Summary:

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This PC/M provided. an air , conditioner for P.A.S.S.. control panel CZ14 in the 1 auxiliary building hallway outside of the P.A.S.S. room. The air conditioner J was added in order to prolong the life and increase the relaibility o f. l electrical components in the cabinet. j q

Safety Evaluation: -

The addition of the air conditioner to Cabinet C-214 does not increase the probability of occurrence of any accident previously evaluated in the FSAR, I nor will it affect the consequences of any. accident previously evaluated ir) J

-the FSAR. The P.A.S.S. Control Cabinet has no safety related function and the !

failure of the' air conditioner has no potential for interaction with safety related equipment. The modification will not affect ' the consequences of ]

I malfunction of. equipmernt important to safety previously evaluated in the FSAR, and the possibility of an accident of a different type than any analyzed in the. FSAR is not created by this PC/M. The margin of safety as defined in the basis for any Technical Specification is not affected by this PC/M.

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^ l PLANT-CHANGE / MODIFICATION CPWO 86-15 PC/M CLASSIFICATION: NSR UNIT: 4

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TURNED OVER DATE: 10-13-86

SUMMARY

DATE: 07-29-87 REVISION: 0 1

REPLACEMENT OF TELEDYNE-FARRIS CCW RELIEF VALVES I

Summary:

l CPWO 86-15 allowed obsolete, Teledyne-Farris model 1870, relief valves that failed periodic testing to be replaced by Teledyne-Farris .model 1850 valves. i The model 1850 is identical to the 1870 except for minor. internal changes. '

All specifications of the 1870 valve are met by the replacement.

Safety Evaluation:

This nuclear safety related CPWO provided for component cooling water system D' relief valve replacement. The . replacement relief valves are functionally- a identical to the original valves. Therefore, implementation of this CPWO did' 1 not involve an unreviewed safety question. I l

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PLANT CHANGE / MODIFICATION 86-068__ PC/M CLASSIFICATION: NSR UNIT: 4

. . TURNED.0VER DATE: 10/31/86

SUMMARY

DATE: 07/29/87 REVISION: 0 REMOVAL OF CCW PIPING TO THE PRIMARY SHIELD COOLERS r

Summary:

PCM 86-068 removed the Component Cooling Water Piping valves, instrumentation and associated hardware for the Primary Shield Coolers. The coolers were originally installed to maintain temperatures within the primary shield wall below 150';, however, the design of the shield wall did not take credit for cooling by ti.e primary shield coolers. Since the coolers were not required and were in need of maintenance it was decided. to abandon them by performing the work described in this PCM.

Safety Evaluation:

PCM 86-068 is considered nuclear safety related since it removes portions of' the component cooling water system associated with the primary shield coolers. This PCM does not adversely effect any other function of the component cooling water system. The primary shield coolers were not required for either normal plant operation or post accident recove ry. Therefore, implementation of this PCM did not involve an unreviewed safety question.

PLANT CHANGE / MODIFICATION 87-169 PC/M CLASSIFICATION: SAFETY RELATED UNIT: 4 6/19/87 TURNED OVER DATE:

SUMMARY

DATE: 7/29/87 REVISION: 0 l

l MODIFICATION TO COMPONENT COOLING WATFR SYSTEM Summary:

Plant Change / Modification 87-169 covers modifications to the pipe supports on the CCW lines to and from the Reactor Coolant Pump Thermal Barriers. These improvements were made to minimize the possibility of damage to the CCW lines if a thermal barrier were to fail. This modification was not a direct NRC commitment or requirement.

Safety Evaluation:

The section of Component Cooling Water Pipe that is b' eing mcdified by this PC/M is safety related. An unreviewed safety question is not involved with, these modifications. The changes in this PC/M are to pipe supports and do not change the system operation, create malfunctions of a different type than that (valuated in the safety analysis report or reduce the mar ~ gin of safety.

i Pt. ANT CHANGE /M00!FICATION 81-059 PC/M CLASSIFICATION: NNSR ;

UNIT: COMMON _,

TURNED OVER DATE: 2-17-87 _

SUMMARY

DATE: 7-30-87

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REVISION: 0 l WATER TREATMENT PLANT FINAL EFFLUENT CONDUCTIVITY TRIP Summary:

Under PC/M 81-059, an automatic isolation valve in the common Water Treatment Plant (WTP) discharge line to Units 3 & 4 was installed. The automatic isolation valve was provided with a full flow bypass with appropriate isolation valves to permit maintenance of the automatic isolation valves to permit maintenance of the automatic isolation valve. A handswitch ,with valve open and valve closed position indication lights, was provided in the WTP control panel to permit operators to remotely open and close the automatic isolation valve.

In addition to the above, a conductivity cell was provided upstream of the -

automatic isolation valve in order to monitor WTP final effluent, conductivity. The design is such that if effluent conductivity levels , as ser. sed by the newly installed conductivity cell, increase to an established setpoint then'the following ocur:

The automatic isolation valve closes The demineralized feed pumps in the WTP trip

.An annunciation in the WTP control panel annunciated Safety Evaluation:

PC/M 81-059 involved ohly changes to the Water Treatment Plant (WTP), a non-nuclear safety related system. This system does not function to achieve and maintain safe shutdown condition, or to safely store and cool spent fuel, or to prevent or mitigate accidents, which could result in potential of f-site radiological exposurer comparable to those cited in 10 CFR 100.11.

The change performed under PC/M 81-059 does not interface with any safety-related equipment nor is it located in the vicinity of any safety-related equipment. Therefore, failure of it's pipe supports would not adversely affect safety related systems, e or structures, and it can be concluded that PC/M 81-059 does not'quipment pose any unreviewed safety question.

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( ' PLANT CHANGE / MODIFICATION 83-209 PC/M CLASSIFICATION: NFSR UNIT: 4

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TURNED OVER DATE: 11-18-87

SUMMARY

DATE: 7/30/87 REVISION: 0

_MSR FOUR TUBE PASS MODIFICATION Summary: .

PC/M 83-209 provided for the installation of a Scavenging Steam Vent Condenser (SSVC) drain lines from each Moisture Separator Reheater (MSR) to a HP (No.6)

Feedwater Heater or to the Main Condenser. The modifications performed under this PC/M are related te modifications performed under PC/M 85-133 which converted the internals of each MSR from a two-pass tube arrangement to a four-pass tube arrangement for reheating cycle steam. The additional two passes produce excess condensate of the reheating steam which is removed from each MSR via an installed SSCV drain line.

The SSCV lines discharge directly into a HP Feedwater Heater (6A or 68) via an extraction steam line during normal operation and to the Main Condenser during start-up operation. The SSVC lines also provide a means of venting the MSR'S' to purge non-condensable gases.

All changes in steam ca condensate flowrates due to the above modifications were found to be acceptable and did not require changes to other existing piping and valves except to previously existing vent lines to the MSR'S.

These vent lines were cut and capped under PC/-M 83-209 since the SSVC drain lines provide the necessary vent path during startup from the MSR's to the condenser.

Safety Evaluation:

The modification performed by PC/M 83-209 is classified as Non-Nuclear Safety Related. The modification involves only Ouality Croup D secondary side system components such as the MSR's the HP Feedwater Heater and the Main Condenser.

The modification does a t involve safety related snubbers, safety related instrument lines or any other components important to safety.

The modification does not affect any limiting condition for operation per Turkey Point Technical Specifications.

The modification does not involve '"

the addition of electrical cable or any changes to existing receways.

The addition of the SSVC drain lines by PC/M 83-209 does not impact high energy line break analyzers already evaluated in the FSAR nor do they affect the flooding analysis as described in the NRC Safety Evaluation Report dated September 4, 1979. In addition, the installation of these lines does not create a new hazard to existing safety related systems or components.

PC/M 83-209 is provided with a written safety evaluation in accordance with 10 CFR 50.59 . The conclusion of the safety evaluation is that the modification does not involve unreviewed safety questions.

1 PLANT CHANGE /M00lFICATION 85-131 PC/M CLASSIFICATION: NSR 1

UNIT: 4 TURNED OVER DATE: 9/17/86 l

SUMMARY

DATE: 7/29/87 REVISION: 0 i l

1 AUXILIARY DISCHARGE FLOW CONTROL VALVE UPGRADE l

Summary:

This PC/M replaced the valve seats, retainers and plugs on the AFW Flow Control Valves CV-4-2816, 2817, 2818, 2831, 2832, and 2833. This change allows better operation at the 125 gpm automatic initiation setpoint. The i manual handwheels were also locked to restrict valve stem travel to a maximum of 85 percent of full open. This limits flow to less than 525 gpm in the l event a flow control valve fails open. I Safety Evaluation:

This modification improves performance of the system at the revised setpoint' of 125 gpm and limits flow if an FCV fails open. This modification does not change the operation of the AFW system and it does not increase the probability or consequences of- any accident previously analyzed or different f rom those analyzed.

Therefore this change does not create an unreviewed safety question. i l

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PLANT CHANGE / MODIFICATION 85-133 PC/M CLASSIFICATION: NNSR ,

UNIT: 4 l

. TURNED OVER DATE: 11-18-86 )

SUMMARY

DATE: 07/29/87 l REVISION: 0 l

1 MSR MODERNIZATION Sunnary:

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PC/M 85-133 provided a modification package for the Moisture Separator Rehaeters (MSR)s to increase MSR performance. The modification included the replacement of each MSR tube bundle and modifications to moisture separation equipment.

Each MSR tube bundle used for reheating cycle steam was replaced by a Westinghouse tube bundle design. This design increased the total number of tubes inside .each MSR, and changed the original two-pass tube arrangement to a four-pass tube arrangement. The excess condensate of the reheating steam that '

occurs as a result of the two additional passes is removed by a Scavenging, Steam Vent Condenser (SSVC) drain line that 'was installed under PC/M 83-209.

Modifications to other moisture separation equipment also enhanced overall MSR performance. These changes included the replacement of the original mesh-type moisture separators with cherron separators. Other modifications were also performed to optimize steam flow distribution and to prevent mositure entrainment within the moisture separator section of the MSR's.

PC/M 85-133 also provided for the installation of two Reheater Drain Tank drain line flow measuring instruments, test connection points and thermo-wells.

Safety Evaluation:

The modification performed by PC/M 85-133 is classified as Non-Nuclear Safety Related. The modification involved only Quality Group D components on the secondary side of the plant. These components include the four MSR's and the two Reheater Drain Tanks. The modification does not involve safety related snubbers, safety related instrument lines or any other components inportant to safety.

The modification does not affect the evaluation of any accident previously performed in the FSAR nor does it create or increase the possibility of any accident not already evaluated in the FSAR The modification does not affect any limiting condition for operation per Turkey Point Technical Specifications.

The modification does not affect the flooding analysis as described in the NRC Safety Evaluation Report dated September 4, 1979.

PC/M 85-133 is provided with a written safety evaluation in accordance with 10CFR50. 59. The conclusion of the safety evaluation is that the modification does not involve unreviewed safety questions.

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t PLANT CHANGE / MODIFICATION 86-31 PC/M CLASSIFICATION: NSR UNIT: 3&4

! TURNED OVER DATE: 12/10/86 L

SUMMARY

DATE: 7/29/87 I l

REVISION: 0 1

' AUXILIARY FEEDWATER PUMP CONTROL PANEL WIRING MODIFICATIONS Summary:

l This PCM provided wiring modifications to the Auxiliary Feedwater Pump Control Panels to correct a design deficiency which resulted in the .short-circuiting and arcing on remote selector switch contacts in the main control boards, t This modification corrected this deficiency by utilizing spare contacts on the i existing relays already installed within the control circuit. This also i' provided modifications to the AFW Pump Trip and Throttle Valve limit switch wiring connections to correct the current OPEN and CLOSED sequence of operations for the position indicating lights. Also the Trip and Throttle Valve positions instead of the Limitorque Motor Operator Positions will now be indicated on the main control board.

j Safety Evaluation:

The PCM enhanced operator interface with the AFW pump and .trubine control !

panel, and increased the reliability of the controls. None of these changes increased the probability or consequences of any previously or not previously analyzed accident, and therefore do not result in an unreviewed safety j

question.

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- PLAKT CHANGE / MODIFICATION 87-099 PC/M CLASSIFICATION: NSR .__

, ,, UNIT: 3 TURNED OVER DATE: 6/26/87 -

SUMMARY

DATE: 7/31/87 l REVISION: 0 ICW/CCW BASKET STRAINER REPLACEMENT l

l Summary:

This CPWO requested the rplacement of the Unit 3 ICW/CCW Basket Strainers due to the extensive corrosion of the existing strainers. The new strainers were -

f abricated to ASME Sect. VIII Div.1 1983ED-1985 Summer Add.; the existing l strainer was built to ANS! B31.1 requirements. The ' original strainers were j built by Zurn Industries, and the replacement strainers were built by Zurn's ]

new owner Hayward Industrial Products, to the same dimensional specifications, and essentially the same strainer body material. The strainers are coated with corroglass 200 epoxy material, while the original strainers were coated with coal-tar epoxy. The replacement strainers can also be equipped with a, I sacrificial zine anodes to provide additional corrosion protection.

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Safety Evaluation: ;

All- dimensions and materials were provided on a one-to-one basis for the replacement strainers, and the new coating is considered to be an i improvement. The maximum specified nozzle loads for the strainer were found !

to be acceptable. The addition of zinc anodes improves the service life of l

- the. strainers. -

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment ;

malfunction / accident important to safety previously evaluated in the FSAR. j This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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l PLANT CHANGE / MODIFICATION 87 156 PC/M CLASSIFICATION: NSR UNIT: 3 TURNED OVER DATE: 6/16/87 _

SUMMARY

DATE: 8/03/87 REVISION: 0 l

UNIT 3 ICW BASKET STRAINER ISOLATION VALVE REPLACEMENT -

SHAFT /0PERATOR l ADAPTER 1

Summary: ,

This CPWO was prepared to provide the necessary parts to adapt the new l replacement valves with the existing hand operators. Tite replaced valve shaf t is smaller than the original valve shaft, requiring .a modification to the Henry Pratt MOT-4 operator.

Safety Evaluation:

Tne valves are normally open and are closed for maintenance purposes only.,

These valves are not required to operate to perform any safety function and are safety related for pressure boundary concerns only. The change will not affect the valves operability as attested by the manufacturer, Henry Pratt Co.

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the FSAR.

This modification .will not decrease the margin of safety as defined in the bases of any Technical Specification, o.

PLANT CHANGE /M00IFICATION 86-076 PC/M CLASSIFICATION: NSR

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UNIT:. 3&4 TURNEDOYERDATE: 6/3/87

SUMMARY

DATE: 7/31/87 REVISION: 0

-l L ' DIESEL GENERATOR' "8" FREQUENCY METER REPLACEENT J

Summary:

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This modification ~ involved'the replacement' of an existing, obsolete frequency meter Lin . the Emergency- Diesel Generator B Control. Cabinet 4C12, with an i

upgraded, functionally. equivalent frequency meter: and removal of the assoicated frequency. impedor module also installed in Cabinet 4C12. The new-meter with ~self-contained tranducer has only two wiring terminals versus the old meter, which. had! three wiring terminals and required .the use of a --

frequency impedor module.

LThe Control . Cabinet 4C12 is listed in the Essential Equipment List, Lowever, this PCM does not impact its safe shutdown capability.. ,

. Safety Evaluation:

This. ~ modification involved .the replacement of -an existing frequency meter. in the Emergency Diesel Generator. B Control Cabinet 4C12, with an upgraded, functionally equivalent frequency meter. The existing frequency impedor'also installed in Cabinet 4C12 is to be removed as the new frequency meter does not require. the . use .of a frequency impedor. Al though the emergency diesel !

generator ; frequency meter does not perform a safety related function, this modification is nuclear safety related to ensure the circuit integrity of ;

existing safety related Control Cabinet 4C12.

This modification was not inside containment, does_not involve safety related snubbers, does not involve block ~ walls, does not.inpact the spent fuel cooling i operations of the plant and does not affect Radioactive Waste Treatment System of the plant.

The modification was reviewed for ALARA requirements based upon the guidance 4 provided in Criteria for ALARA Evaluation per FPL letter JPE-PTP0-84-1239.

The modification accomplished by this PCM did not affect the flooding analysis as described in the NRC Safety Evaluation Report, dated September 4,' 1979, because the modification does not . introduce a new source of safety related componets which could be affected by flooding.

Equipment or cables associated with this work were not attached to or in.

proximity of any , block walls which - have not been previously analyzed to preclude their failure and subsequent damage to adjacent safety related equipment. .No now cables required. The existing internal wire was used for connecting the new frequency meter.

The replacemht of the frequency meter did not adversely affect the seismic qualification of Panel 4C12 as no structural modifications to this panel are required as the new frequency meter utilizes the same mounting arrangement as the existing frequency meter and the meters are of approximately the same size and weight i

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

The replacement 'o'f- the existing l frequency . meter;in the Emergency ' Diesel Generator B Control Cabinet 4C12, with an upgraded functionally equivalent frequency meter did not_ change the design function of the Emergency Diesel Generator System. Therefore, the probability of occurrence of an accident previously evaluated in the FSAR will not be increased.

.The consequences of an accident previously evaluated in the FSAR was not increased because the basic function'of the Emergency Diesel.

Generator System-remains the same, and no other safety' related systems are adversely affected by this modification.

.This modification did not change the inherent function 'of any safety' related-systems.. Therefore, there was no possibility that an accident may be created which is . a different type than any already evaluated in the FSAR. ,

This modification was for the replacement of the obsolete frequency meter !

with an _ upgraded version of_ the frequency meter. Therefore, the !

probability of occurrence of e';uipment malfunctions important to safety i previously evaluated in the FSAR was not increased.

All work: associated with this modification was accomplished'in accordance with' approved procedures and final system design will be tested to ensure its proper function and operability. Therefore, consequences of equipment, malfunction important to safety previously evaluated in the FSAR was not

-increased.

This modification did not adversely affect the inherent function or design basis of the systems related to safety; therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the FSAR was not created.

This modification did not reduce the margin of safety as defined in the bases for any Technical Specification since this modification replaces the obsolete frequency meter with an upgraded model. However, strict adherence to the requirements of the Technical Specifications,- Section 3.7.0 shall be observed.

Based on the above, this modification did not constitute an unreviewed safety question and is considered acceptable.

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.PtANT CHANGE / MODIFICATION 86-80 PC/M CLASSIFICATION: HSR UNIT: 3 _ ,,,

TURNED OVER DATE: 06/2o/87

SUMMARY

DATE: 07/31/87 REVISION: 1 S.I. ACCUMU1.ATOR MAKEUP HEADER SEISMIC REPLACEMENT Summary: -

This modification qualifies the piping and support configuration of the Accumulator Makeup Header inside containment to the requirements of the FSAR Appendix 5A seismic design criteria by the installation or modification of supports per Specification 5177-M-56 and Computer Stress Analysis.

This modification is inside containment. It consists of new supports and modifications to some existing supports. The Structural steel required for these modifications will change the heat sink by approximately ten square feet of 1/8- inch thick steel. This is a negligible change to the 48,300 square, feet used in the ECCS heat sink analysis (Refer to Table 14.2.4-1 of the FSAR). Therefore, this modification does not alter the ECCS heat sink analysis. This modification does not involve safety' related snubbers, or block walls, does not impack the spent fuel cooling operations of the plant, does not i nvol'/e additions of electrical cable or changes to existing raceways, and does not compromise the Turkey Point Fire Protection program.

This modification resolves the safety concerns addressed in JPE-M-85-029 which evaluated the consequences of a post-LOCA break in the acn mulator fill line. The possibility of a break had increased when it was retermined that the pipe was not seismically supported.

Safety Evaluation:

This modification is nuclear safety related with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment :

malfunction / accident important to safety previously evaluated in the FSAR.

This modification will not decrease the margin of safety as defined in the bases of any Technical Specification.

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Pt. ANT CHANGE /MODIFICA, TION 86-090 PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER DATE: 7/8/86

SUMMARY

DATE: 8/03/87 !

REVISION: 0 1 i

i ROOT VALVE NO 4-20-698 COMPONENT SUBSTITUTION Summary:

This CPWO replaced broken 3'/4" 1500 psi Rockwell globe valve 4-20-698 (FT 497 isolation) in the Feedwater system with an equivalent 3/4" 1500 psi globe.

valve.

Safety Evaluation:

This CPWO will not increase the probability of' occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR, nor will the possibility of an accident or a, malfunction of a different type than any evaluated previously in the FSAR be created. Additionally, the margli, of safety as defined in the basis for any technical specification is not reduce,1.

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7, L PLANT . CHANGE / MODIFICATION 85-010 PC/M CLASSIFICATION: NS UNIT: 4 TURNED OVER DATE: 2/18/87 L

SUMMARY

DATE: 8/03/87 REVISION: 0 ADDITION OF FEEDWATER CONTROL VALVES DIRECT POSITION INDICATION gmmary:

In order to provide direct position indication for the Main and By-Pass Feedwater Control Valves, FCV-4-478, 488, 498 and FCV-4-479, 489, 499, respectively, qualified limit switches (NAMC0 Type EA-180) and indicating lights were added.

'The indicating lights are located in the Control Room Control Console 4 directly above each valve's corresponding controller. A new safety related power supply has been provided for the indicating lights.

Conax electric conductor seal assemblies are also provided to maintain the environmental integrity of the limit switches, due to their being located in a, high energy line break area. l Safety Evaluation:

This modification consists of adding qualified limit switches and indicating lights for the feedwater control valves to meet the requir m ots of Regulatory Guide 1.97, Rev. 3. In the event of a single . failure of Train A, valve position can be determined by : 1) the main feedwater flow indicators, 2)

SPDS/SAS, or 3) visual inspection of valve.

The circuitry involved in this modification is for indication only and does not perform any control function. The load added to the safety related power supply by this addition is negligible.

This modification did not involve an unreviewed safety question because:

1.a With respect to the probabilit/ of occurrence of an accident previously evaluated in the FSAR: '

This addition provided valve position indication only; the existing i control functions were unchanged. All conduits have been seismically installed. All modifica.tions utilize qualified safety grade components and were done under plant OC procedures. Addition of new items adds negligible weight and had no effect on the equipment seismic response.

Therefore, the probability of occurrence of an accident previously evaluated in the FSAR is not greater, 1.b With respect to the consequences of an accident previously evaluated in the FSAR:

This modification meets the requirements of Regulatory Guide 1.97, Rev. 3. With the addition of qualified limit switches and indicating lights there is a greater assurance that the valves are in the correct position. The control of the valves was not affected by this f

change. Ad,dition of new items adds negligible weight and had no effect on the equipment seismic response.

Therefora the consequences .of an accident previously evaluated in the FSAR was niot increased.

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

All Conduits and equipment 'were seismically installed. The addition of the limit switches on the control valves was wired totally independent of the wiring for the solenoid valves controlling the feedwater valves. There were no inter ties with any other safety related system. Addition of new items adds negligible weight and had no effect on the equipment seismic response.

Therefore, the probability of malfuention of equipment important to safety is no greater.

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

The addition of qualified indicating lights provides for better operator information. The wiring for this modification had no interaction with any other safety related system. All the equipment was seismically installed.

Therefore, the consequences of rulfunction of equipment important to safety previously evaluated in the FSAR were not increased.

4 2.a With respect to the possibility of an accident of a ;;fferent type than any analyzed in the FSAR:

The new conduits and equipment were seismically installed. This system provides valve position indication only. Existing valve function and control remained unchanged.

Therefore, the possibility of an accident of a different type than any analyzed in the FSAR was not created.

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

The circuitry involved in this modification is independent of the l control wiring for each valve. This new system is not connected to any other safety related system. The function of the valves was not changed.

Therefore, the possibility of malfunction of a different type than any analyzed in the FSAR WAS,not created.

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

This addition is not addressed in any Technical Specification.

Therefore, no margin of safety as discussed in the Technical Specifications has been decreased.

These modifications added terminal blocks and indicating lights to the control console. The negligible weight of the new equipment and the new cut-outs did not affect the seismic response of the console.

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1 The weight' of' the limit switches and brackets being added to each-valve are. negligible and did not affect the, seismic response.

All added conduit was seismically installed per ~5177.-E-302. l The modifications twere' not installed on or adjacent to any . " block"

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PLANT CHANGE / MODIFICATION 84-209 PC/M CLASSIFICATION: WS UNIT: 4 l TURNE0 OVER DATE: 08/16/86

SUMMARY

DATE: 08/06/87 REVISION: 0

_ REINSTATEMENT OF POWER MISMATCH WITHOUT AUTOMATIC ROD WITHDRAWAL Summary:

l The Automatic Rod Control System, with power mismatch circuitry was poten-l tially susceptible to undesirable control system operations induced by ari adverse environment (i.e., a steam line break inside containment could subject the excore detectors and cables to elevated temperatures which could cause rod withdrawal, if the rods were in the automatic mode prior to a reactor trip).

Power mismatch was disconnected from automatic rod control by PC/M 81-13. The possibility of the NIS System initiating a spurious low power signal without causing a reactor trip.on negative flux rate could have been eliminated by the removal of automatic rod withdrawal circuit. Because the rod insertion circuit was also eliminated, it is deemed necessary to reinstate automatic rod insertion control circuit. When operating in automatic mode, the automatic rod insertion would occur, if nuclear instrumentation system detects a high

power signal (OT-0T).

Safety Evaluation:

This change does not involve an unreviewed safety question, because the modf-fication reinstates the power mismatch circuit associated with automatic rod insertion only. The probability of occurrence of uncontrolled rod cluster control assembly (RCCA) withdrawal is not made more likely, since this modifi-cation affects the rod insertion circuitry only and all the rod withdrawal circuitry will be disconnected. The power mismatch' circuitry was provided as part of the original NSSS package. This modification only reinstalls the automatic rod insertion ci rcuitry to its original state and removes the circuitry associated with automatic rod withdrawal, since this modification does not add a control system that did not exist. Hence the probability of occurrence of an accident previously evaluated in the FSAR, or consequences of a accident, or probability of malfunction of equipment important to safety, or consequences of the malfunction of equipment important to safety previously etaluated in the FSAR has not changed. Since the modification reinstates a system that was provided in origine.1 NSSS package ano does not adversely affect any safety system or introduce any possibility of e accident of a different type than any analyzed in the FSAR, the control rod insertion limits will not be changed for Technical Specification 3.2 and the margin of safety as defined in the basis for Technical Specifications will not be reduced.

No device penetrates any pressure boundary or affects any existing piping stress analysis. No equipment shall be added to containment, so there is no effect on heat sink of containment. No cables are bein~ g added, so there is no effect on raceways and no requirements for conduits and supprot. No additional load or modification is performed to the racks so no seismic evalu-ation is required.

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PLANT CHANGE /H00!FICATION 86-162 PC/M CLASSIFICATION: SR UNIT: 3 TURNED OVER DATE: 5/15/87

SUMMARY

DATE: 08-03-87 REVISION: 0 I

, REMOVAL OF CCW PIPING TO THE PRIMARY SHIELD COOLERS Summary:

PCM 86-162 removed the component Cooling Water Piping Valves instrumentation and associated hardward for the Primary Shield Coolers. The coolers were originally installed to maintain temperatures within the primary shield wall below 150' F; however, the design of the shield wall did not take credit for cooling by the primary shield coolers. Since the coolers were not required I and were in need of maintenance it was decided to abandon them by performing the work described in this PCM.

Safety Evaluation:

PCM 86-162 is considered nuclear safety related since .it removes portions of the component cooling water system associated with the primary shield coolers. This PCM does not adversely effect any other function of the component cooling water system. The primary shield coolers wern't required for either normal plant operation or post accident recovery. Th erefore, implementation of this PCM did not involve an unreviewed safety question.

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n PLANT CHANGE / MODIFICATION 87-126 PC/M CLASSIFICATION: SR l' -

UNIT: 4 TURNED OVER DATE: 7/13/87

SUMMARY

DATE: 8/03/87 REVISION: 0 ACCUMULATOR SI.' TEST LINE SOLENOID VALVE REPLACE?'ENT Summary:

This CPWO replaced the following solenoid valves SV-850 B, SV-850-0 and SV- ,

850-F. l Safety Evaluation:

i This CPWO did not change the function of these solenoid valves. The ;

replacement solenoid valves are qualified in accordance with IEEE 323 and IEEE 344. In addition seismic supports were added for these solenoid valves. This modification did not involve an unreviewed safety question.

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' PLANT CHANGE / MODIFICATION 87-160 PC/M CLASSIFICATION: NNSR UNIT: 3 TURNE0.0VER DATE: 7/13/87

SUMMARY

DATE: 8/03/87 REVISION: 0

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BAILEY TEMPERATURE TRANSMITTER REPLACEMENT FOR TPCW Summary:

1 This CPWO replaced TE-1432 and associated transmitter TT-2201. This CPWO also added a thermowell in which the new temperature element without compromising the TPCW heat exchanger pressure boundary.

Safety Evaluation:

This CPWO did not change the function of this temperature loop. The TPCW system is non safety related and non seismic. The thermowell does not degrade the TPCW piping classification. This modification did not constitute an unreviewed safety question. ,

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PLANT CHANGE / MODIFICATION 87-037 PC/M CLASSIFICATION: SR

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UNIT: 3 TURNED OVER DATE: 6/9/87

SUMMARY

DATE: 7/25/87 REVISION: 0 ICW PlMP FOUNDATION REPAIR ANCHOR BGLT REPLACEMENT Summary:

This CPWO provided for replacement of Anchor Bolts a pump bases for Unit 3 ICW i Pump Foundations and repair' of an Intake screen backwash pump basket strainer drain pipe, and ICW pump grounding cable. The drain pipe and grounding of the anchor bolts.

The replacement bolts -are larger and longer to provide increased anchorage capacity for the ICW pumps. This CPWO improved the reliability of the ICW pumps.

Safety Evaluation: j This CPWO -is considered safety related since it is associated with the intake #

cooling water pump support. This CPWO does not adversely affect any function of the ICW system. It does not increase the probability of an accident or the possibility of'an accident, different than that in the FSAR.

The consequences of an accident are not changed from those described in the FSAR.. The Margin of safety is not reduced from that currently defined in the Technical Specifications.

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I PLANT CHANGE / MODIFICATION 87-102 PC/M CLASSIFICATION: QUALITY RELATED

-CPWO UNIT: 4

. TURNED OVER DATE: 06/09/87

SUMMARY

DATE: 8/4/87 REVISION: 0 ;

REACTOR VESSEL HEAD INSULATION-REFLECTIVE REPLACEMENT Summary:

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. I Boric acid accumulation from the Conoseal Leak within nine reflective insulation panels has - reduced the thermal performance of the insulation. l Cleaning of the insulation has proven ineffective and replacement was deemed '

prudent. .This CPWO installed replacement insulation panels fabricated of all !

stainless steel interior and exterior ' plates with stainless steel foil filler. This design is considered an upgrade due to the absence of the aluminum foil filler originally used. The additional weight- of the replacement panels (160 lb/ panel) has been evaluated and the increased load are acceptable. The new panels are fabricated to the original dimensions, l

-therefore the installation process is unchanged.

. , f Safety Evaluation:

CPWO 87-102 is considered quality related since it attaches to the reactor vessel. This CPWO does not adversely affect the r? actor vessel or associated components. The replacement insulation meets the criteria of specification 67449 and is therefore acceptable from a thermal performance standpoint. ,

Therefore, implementation nf this CPWO did not involve an unreviewed safety ;

question.

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PLANT CHANGE / MODIFICATION 87-101 PC/M CLASSIFICATION: QR UNIT: 4 t

. TURNED OVER DATE: 6/9/87

SUMMARY

DATE: 8/4/87

, REVISION: 0 t

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1 REACTOR VESSEL HEAD INSULATIO -PERMANENT REPLACEMENT Summa ry: _ .

l As a result of damage from, the leaking Unit 4 Conoseal, it was necessary to replace the " Permanent" reactor vessel insulation. The original insulation a consisted of block type unibestos (or equal) with asbestos cement filler, asbestos tape coating and Ih " layer of "one cote" cement over tape. The replacement material consists of two 9/f layers. of B & W Kaowool, Fiberfrax (

cloth coating with /g 1 thick coating of fiberfrax cement and waterproof coating of GE SM-2010 Silicon Release Emulsion. The replacement materials conform to the requirements or Reg. Guide 1.36 and have an equivalent or better performance as compared to the original insulation (eg. thermal performance, materials etc). ,

Safety Evaluation:

CPWO 87-101 is considered Quality Related since it is applied to the Reactor Vessel. This CPWO does not adversely affect the Reactor Vessel Head or appurtenances. Application or the Reactor Vessel Head Permanent insulation does not represent an unreviewed safety question.

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l PLANT CHANGE / MODIFICATION 87-177 PC/M CLASS!FICATION: SR UNIT: 4 TURNED OVER DATE: 06/11/87

SUMMARY

DATE: 08/05/87 j REVISION: 0 CONTAINMENT SPRAY RESTRICTING ORIFICE Summary:

This PC/M added a restricting orifice on the Containment Spray Pump discharge .

flange, to reduce the spray pumps injection flow rate to an acceptable level, l with respect to NPSH requirements and accident analysis assumptions. The PC/M also: affected several Emergency Operating Procedures to ensure pump cavitation is avoided, and avoid violating the Emergency Diesel Generator Loading; and a setpoint change was made to the RWST low level alarm to i preclude pump cavitation.

This PC/M evaluated all _ POST-LOCA operating modes to determine all pump flows and required flows.

Safety Evaluation:

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The probability of an occurrence of an accident or malfunction of equipment previously evaluated is not increased, since the Containment Spray System's only function is to mitigate accidents. The consequences of an accident will not change as diesel loading requirements are still acceptable. ,

i' Due to this systems function as accident mitigation and the minimal change in system configuration, the possibility of an accident of a different type than those currently evaluated in the FSAR is not increased. j With respect to malfunction of a different type than those previously analyzed in the FSAR, the changes in stress load, support loading, and diesel generator loading are all within the allowables and therfore preclude the increase in probability of failure or malfunction of any equipment important to safety /

For the reasons stated above, the margin of safety as currently defined in I the Technical Specifications is not reduced due to this PC/M. !

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PLANT CHANGE /M00!FICATION 86-100 PC/M CLASSIFICATION: NSR UNIT: 4 10/31/86 TURNED OVER DATE:

SUMMARY

DATE: 8/17/87 REVISION: 0 MIS SOURCE RANGE PREAMP REPLACEENT

.t Summary:

This CPN0 replaced both Unit 4 Source Range preamplifiers (N-31 & N-32) with new Westinghouse preamplifiers. The new preamplifiers have better noise !

rejection characteristics. This change was part of an overall effort to reduce noise on the excore channels.

Safety Evaluation: i The source range preamplifiers are nuclear safety related because these channels are required by Technical Specifications during refueling operations and plant startup. The replacement preamplifiers are fully qualified i n, :

accordance with IEEE 323 8 344 The function of the preamplifier was not

, changed (this was simply a one for one part replacement). This CPWO did not constitute an unreviewed safety question.

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PLANT CHANGE /M00!FICATION 86-184 PC/M CLASSIFICAT!ON: ITS UNIT: 4 TURNED 0VER OATE:

- 5/30/87

SUMMARY

DATE: 8/17/87 REVISION: 0 RPI INVERTER REGULATOR TRANSFORMER REPLACEMENT Sununary:

This package covers the replacement of the failed Solotron Line Voltage Regulato r.

Safety Evaluation:

The replacment regulator has the same nominal voltage and KVA ratings. This change does not involve an unreviewed safety' question because the probability of occurrence or the consequences of a design basis accident are not increased. The malfunction of safety related equipment previously evaluated in the FSAR is not increased. The Possibility for an accident or malfunction, of a different type than evaluated previously in the FSAR is not created. The margin of safety as defined in the bases for a Technical Specification is not reduced.

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PLANT CHANGE / MODIFICATION 83-50 PC/M CLASSIFICATION: NSR UNIT: 3a4 2/12/87 TURNED OVER DATE:

SUMMARY

DATE: 8/17/87 REVISION: O CR-2 MASONARY WALL M00!FICATIONS - UNITS 3 & 4 Summary:

An inspection of masonry walls in the Auxiliary Building, Control Building, i Turbine .Buf1 ding, Diesel Generator Building and Stream Generator Feed Pump Enclosure indicated these walls were erected without rebar and/or grout. As these walls supported safety related components, these walls were modified to be able to carry the design loads. i i

Safety Evaluation:

This modification improved the walls load bearing strengths by adding robar and grout or as required rebuilding the wall. This modification does not increase the possiblity of occurrence or the consequences of an accident or, malfunction of equipment previously evaluated in the FSAR , .does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR, and does not reduce the margin of safety as ,

defined in the basis for any technical specification. !

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PLANT CHANGE /MODIFIC.h lqN 87-210(CPWO) PC/M CLASSIFICAThJ: NSR UNIT: 4 TURNED OVER DATE: 6/12/87

SUMMARY

DATE: 8/17/87 REVISION: 0 REPLACEMENT OF SUPPORTS 4-SIH-42 AND 4-PRWH-11 Summa ry:

Two Pipe Supports 4-S!H-42 and 4-PRWH-11 on the line from the Refueling Water Storage Tank to the Charging Pump Suction were replaced with new supports designed to provide the same vertical and lateral restraint. The supports were located within 10" of the old supports. The old supports were considered inadequate due to corrosion.

Safety Evaluation:

This line and supports are nuclear safety related. The design of the supports provide the same vertical and lateral support and are located at the same, design location. The supports are mounted on the trench wall vice floor. The wall has been analyzed to be an adequate mounting location. This modification does not increase the poss'ibility of occurance or the consequences of an accident or malfunction of equipment previously evaluated in the FSAR, does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR and does not reduce the margin of safety as defined in the basis for any Technical Specification.

PLANT CHANGE /M00!FICATION 84-16 PC/M CLASSIFICAT!0N: NNS UNIT: 3 TURNED OVER DATE: 4/13/87

SUMMARY

DATE: 8/17/87 ,,

REVISION: 0 RHR ISOLATION VALVE CIRCUIT MODIFICATION Su. wiry:

i The change installs an ov.erride pushbutton on VP8 to allow the operator to open MOV750, M0V751 should they close on a momentary high pressure spike when a on RHR. These pushbuttons allow the valve to be opened before it completes- l the complete closure cycle thus preventing a loss of RHR letdown and i subsequent challenges to the OMS System.

Safety Evaluation:

This change is nuclear safety related because it affects the RHR system )

isolation function. It does not present an unreviewed safety question because, 1 it does not change RHR isolation, but only allows the valves to be opened i rapidly after the isolation signal has cleared.

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PLAKT CHANGE /M00!FICATION 85-182 PC/M CLASS!FICATION: NSR UNIT: 3 TURNED OVER DATE: 10/29/86

SUMMARY

DATE: 08/17/87 REVISION: 0 CHEMICAL ADDITION LINES SUPPORT REPAIRS Summary:

The chemical addition lines to the main feedwater piping were found to have inadequate supports for seismic Category 1. This modification provided new supports, new anchors and modified existing supports to adequately support-the chemical addition line.

Safety Evaluation:

l This modification improved the line supports to meet Seismic Category 1 criteria. The line is nuclear safety related. This modification does not increase the possibility of occurrence or the consequences of an accident or' malfunction of equipment previously evaluated in the FSAR, does not create the possibility for an- accident or malfunction of a different type than any evaluated previously in the FSAR and does not reduce the margin of safety as >

defined in the basis for any Technical Specification.

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PLANT CHANGE / MODIFICATION 85-139_ PC/M CLASSIFICATION: NSR UNIT: 4 TURNE0 OVER OATE: 09/05/86

SUMMARY

DATE: 08/17/87 REVISION: 0 l

l REMOVAL OF VALVES 4-524, 4-625 AND ASSOCIATED PIPING (PRESSURIZER MINI-SPRAY) 1 Summary:

l On Turkey Point Unit 4, pressurizer spray mini-bypass lines and valves exist (Ref. PC/M 82-161) for the two relocated spray valves (with new bodies).

Additionally, two mini-bypa:s lines and valves use to exist for the abandoned original plant spray valve bodies. The two original installation bypass lines which are no longer in service were a high source of radiation. Their function is now performed by the newer valves installed under PC/M 82-161.

This PC/M removed the original 3/4" mini-spray bypass lines and valves provided for the valve bodies abandoned per PC/M 82-161. These mini-spray bypass lines were located above the Containment 54' elevation and were a high, source of radiation. Their removal greatly improved ALARA concerns in the immediate vicinity and eliminated the need to shield these valves and piping.

This removal did not affect the capability of the new mini-spray bypass lines (currently at 14' elevation).

Safety Evaluation:

This design removed equipment whose function is no longer required. The removal was to further limit radiation to amounts as low as reasonably achievable. The equipment that superseded the function of the equipment removed by this PC/M received the appropriate level of safety analysis via PC/M 82-161. The removal of equipment in this PC/M has undergone sufficient review to assure that all current analysis (stress analysis or the 4" pipe) or design practices (provision of sufficient minispray flow) impacted by the equipment has not been detrimentally affected by its removal For the above reasons, the probability or consequences of an accident previously evaluated in the FSAR have not increased. Furthermore, the probability or consequence of malfyn.ction of equipment important to safety remain unchanged, and the possibility of an accident of a different type than any previously analyzed has not been created. Finally, the margin of safety as defined in the Technical Specifications is not affected by this equipment removal.

PLANT CHANGE / MODIFICATION 85-181 PC/M CLASSIFICATION: NSR UNIT: 4 TURNED OVER DATE: 10-3-86

SUMMARY

DATE: 8/17/87

, REVISION: 0 REMOVAL OF INACTIVE NITR06EN BLANKET LINE Summary:

This FCM removed an existing plant nitrogen system supply line connected to the Train 1 Auxiliary Feedwater Piping of Unit 4. This line was used during

, construction for blanketing portions of the secondary side of the plant with nitrogen. The .line was no longer in service.

Safety Evaluation:

This modification is consistent with all applicable design requirements for the Auxiliary _Feedwater System and affects no other systems. The QC requirements imposed are sufficient to ensure that the modification is mades I

accurately and adequately.

i Therefore, this modification does not involve an unreviewed safety question and.is considered acceptable.

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PLANT CHANGE /MO0!FICATION 84-11 PC/M CLASSIFICATl0N: NSR UNIT: 3 TURNED OVER DATE: 7/3/85

SUMMARY

DATE: 8/17/87 REVISION: 0 M00!FICATION TO PRESSURIZER SPRAY SYSTEM - FUNCTIONALITY (I.C.)

Su m er All work in this PC/M was . completed under Change Request #4 to PC/M 81-146.

This PC/M was for the transfer of documentation only in order to clarify the Engineer of Record for PC/M 81-146. The work was support modifications to the Pressurizer Spray System Functionality (Inside Containment) required for compliance with NRC IE Bulletin 79-14.

Safety Evaluation:

The modifications and analyses ensured that the design criteria of the, original piping system design documents were met. Therefore,.the probability or consequences of an accident previously evaluated in the FSAR have not increased. Futhermore, the probability or consequence of malfunction of equipment important to safety remain unchanged, and the possibility of an accident of a different type than any previously analyzed has not been created. Finally, the margin of safety as defined in the Technical S'ecifications was not affected by this PC/M.

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t PLANT CHANGE / MODIFICATION 85-141 PC/M CLASSIFICATION: OR UNIT: 3 l TURNED OVER DATE: 6/29/87

SUMMARY

DATE: 8/17/87 REVISION: 0 l

FUEL TRANSFER SYSTEM MANIPULATOR CRANE DUAL CABLE MODIFICATION _

Susunary:

This PC/M covers modifications to the Unit 3 manipulator crane. These j modifications consist of:

1. Upgrading the manipulator crane from a single cable hoist to a dual cable hoist.

2. The installation of a new hoist load indicator system.

The installation of a load test fixture used to load test the main hoist

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on the manipulator crane. ( An existing structure will be used to load test the manipulator crane's auxiliary hoist.)

This PC/M added redundancy to the cablo hoist portion of the refueling # '

machine. This enhancement increased reliability, will reduce maintenance time, and incorporated the latest state of the art- capabilities into the refueling system.

Safety Evaluation:

The installation of the manipulator crane modifications does not have any direct effect on nuclear safety since the manipulator crane is not safety related. This modification could indirectly affect nuclear safety since a crane. accident could result in a design basis accident. However, the modifications associated with this PCM include changes which reduce the possibility of potential accidents. i I

An unreviewed safety question, as defined by 10CFR 50.59 does not exist and no )

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PLANT CHANGE /M00lFICATION 86-121 PC/M CLASSIFICATION: SR UNIT: 4 5/30/87 TURNED OVER DATE:

SUMMARY

DATE: 8/17/87 REVISION: O CONTROL POWER FUSE REPLACEMENT Sumary:

l This Package covers' the replacement of Gould Shawmut Control Power Fuse Type l A2Y 3 AMP Type 2 with Type A6Y 3 AMP Type 2.

Safety Evaluation:

This change does not involve an unreviewed safety question because the probability of occurrence or the consequences of a desigr. basis accident are f not increased. The malfunction of safety related equipment previously l evaluated in the FSAR is not increased. The possibility for an accident or malfunction of a different type than evaluated previously in the FSAR is not, l

created. Also the margin of safety as defined in the basis for a Technical Specification is not reduced. The preceeding is correct because all electrical specifications of the replacement fuse are identical to the old fuse. The voltage rating of the replacement fuse is higher than the old fuse.

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I PLANT CHANGE / MODIFICATION 86-107 PC/M CLASSIFICATION: SR UNIT: 3 TURNED OVER DATE: 5/30/87

SUMMARY

OATE: 8/17/87 /

REVISION: 0 MCC CONTROL POWER FUSE REPLACEMENT Sununary:

- This package covers the replacement of Gould Shawmut Control Power fuse Type A2Y 3 AMP Type 2 with Type A6Y 3 AMP Type 2. This control Power Fuse is used on the Emergency Containment Filter 38 Safety Evaluation:

This change does not involve an unreviewed safety q'uestion because the probability of occurrence or the consequence of a design basis accident are not increased. The malfunction of safety related equipment, previously evlauated in the FSAR is not increased. The possibility for an accidentj or, malfunction of a different type than evaluatedpreviously in the FSAR is not; created. The margin' of safety as defifsd in the basis for a Technical Specification is not' reduced. !The proceeding i s' correct because the replacement is equal or better !Wnen cof@a' red to the old fuse in all r categories. ,e . ,

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PLANT CHANGE / MODIFICATION 86-026 PC/M CLASSIFICATION: SR !

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1 UNIT: 4 l TURNED OVER DATE: S/30/87

SUMMARY

DATE: 8/17/87 REVISION: 0 i

4KV SWITCHGEAR BREAKER ELEVATING MECHANISM REBUILDING Summary:

This CPWO involves the rebuilding and reinstallation of the "E" shaped right and left side elevating mechanisms on the 4KV Switchgear Startup Transformer Breaker 4AA05; and the Unit Auxiliary Transformer Breaker 4AA02.

Safety Evaluation:

The newly installed parts do not perform any load bearing function except' 1 during elevating the breaker assembly. Since the rebuilding of this elevating mechanism does not affect the seismic capabilities of this "E" shaped bracket to support the breaker assembly there will be no detrimental effect upon the, capabilities of this part_to perform its design function. Therefore it can be said that: This change does not involve an unreviewed safety question because the probability of occurrence or the consequences of a design basis accident are not increased. The malfunction of safety related equipment previously evaluated in the FSAR is not increased. The possibility for an accident or malfunction of a different type than evaluated previously in the FSAR is not created. The margin of safety as defined in the bases for a Technical )

Specification is not reduced.

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PLANT CHANGE / MODIFICATION 85-11 PC/M CLASSIFICATION: SR L . UNIT: 3 o

TURNED OVER DATE: 6/3/87

SUMMARY

DATE: 8/17/87 REVISION: 0-I MODIFICATION OF 4160 V BREAKER HH SWITCHES i

Summary.

This PC/M changed the 4160 V Bkr. Position Switches "52 HH" for a better designed switch and added wiring so that the breaker position white !

indications light monitors the action of the "52HH" switches.

I Safety Evaluation:

This PC/M 'is. safety related. The new switches and wiring enhance the 4160 V Bkr. - operation. The probability of occurrence of 'an accident previously evaluated in the FSAR is not increased, and an unreviewed safety question .is not involved.- ,

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PLANT CHANGE / MODIFICATION 85-149 .PC/M CLASSIFICATION: OR' UNIT: 3 TURNED OVER DATE: 5-7-87

SUMMARY

DATE: 6-18-87 REVISION: 0 SFP AIR INLET DAMPER REPLACEMENT 1

Summary:

This PC/M replaced the Unit.3 Spent Fuel Pit Room Fresh Air Inlet Dampers.

Safety Evaluation:

This PC/M is quality related in accordance with FSAR Appendix 5A requirements with no unreviewed safety question since the probability / consequences of an accident previously evaluated in the FSAR has not increased, nor was the possibility of an equipment malfunction / accident important to safety previously evaluated in the 'FSAR. This modification will not decrease the.

margin of safety as defined in the bases of any Technical Specification. ,

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. 11) PROCEDURE CHANGES The following procedures were changed, reviewed, and approved and reissued during the reporting period. The precedure changes are as described below and only those procedure changes constituting changes in the procedures as described in the Final Safety Analysis Report (FSAR) are reported. Minor or routine procedure changes not ,

affecting procedures as described in the FSAR are not reported. j

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1) A review of the manufacturer's recommendations for cooling water temperature i for the diesel driven fire pump determined that the criteria in the procedure did not match the manufacturer's recommendations. The fellowing procedure was changed to reflect the manufacturer's recommendations:

a) OP 15524: Fire Protection Pumps and Power Supplies - Periodic Test Gafety Evaluation Summary:

The proposed change only changes cooling water temperature limits to those specified by the pump manufacturer. This change does not alter the pump operation or change che pumps capability to perform its intended function as specified by the manufacturer. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment ;

important to safety. No new accident or malfunction of a different type will ;

be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: July 3, 1986

2) The existing procedure for fire stop and cable tray fireproofing was revised to identify specific fire areas that are required to have cables coated, and the criteria for applying and maintaining the coatings in compliance with !

Regulatory requirements. In addition various portions were removed that were no longer applicable. The following procedure was revised:

a) MP 0725: Fire Stop and Cable Tray Fireproofing (New title is Application and Maintenance of Flame Retardant Cable Coatings (Flamemastic 77))

Safety Evaluation Summary:

The evaluation serves to clarify the commitments that have been made in regards to the use of cable coatings or IEEE-383 qualified cable. As such, the safety of the units will not be reduced if qualified or coated cables are provided in only the areas specified by detailed fire hazt.rd analysis reports. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor..will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis j for any Technical Specification is decreased. I Date of Change: July 3, 1986 1

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3) Spacial Tests ware parformsd to determine the proper. component cooling water system flow balance fot Unit 4. The results of these tests were used to revise the following procedures:

a) 4-OP-030- Component Cooling Water System j b) 0-ADH-205 Administrative Control of Valves, Locks, and Switches )

c) AP 0103.19 Monthly Verification of Safety Related Systems Flowpaths d) AP 0103.32 Reactor Cold Shutdown Conditions Safety Evaluation Summary The system alignment established by the special test will serve to ensure that the system is capable of satisfying design basis heat removal requirements during 1 accident conditions. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety ralated equipment important to safety. No new accident or malfunction of a different type will l be created and no margin of safety as defined in the basis for any Technical l Specification is decreased.

Drie of Changes: July 24, 1986

4) A study of emergency diesel generator (EDG) loading determined that a potential existed to exceed the FSAR and emergency operating procedure loading limits for the EDGs. Based on this plant changes were made along with revising the following procedures:

a) 3(4)-E0P-ES-0.0 Rediagnosis ;

b) 3(4)-EOP-ES-1.3 Transfer to Cold Leg Recirculation !

c) 3(4)-E0P-ES-1.4 Transfer to Hot Leg Recirculation j d) 3(4)-EOP-ES-3.1 Post-SGTR Cooldown Using Backfill e) 3(4)-EOP-ES-3.2 Post-SGTR Cooldown Using Blowdown f) 3(4)-E0P-ES-3.3 Post-SGTR Cooldown Using Steam Dump g) 3(4)-E0P-ECA-0.0 Loss of All AC Power l h) 3(4)-EOP-ECA-3.2 SGTR With Loss of Reactor Coolant-Saturated Recovery l Recovery Desired j i) 3(4)-EOP-ECA-3.3 SGTR Without Pressurizer Pressure Control l j) 3(4)-E0P-F-0 Critical Safety Function Status Trees k) OP-057 Containment Normal Ventilation and Cooling System ;

1) OP-0205.2 Reactor Shutdown - Hot Standby to Cold Shutdown ;

Condition i m) OP-10304.6 Computer Room Chilled Water System - Operating Instructions n) H:NPO-3.3 Nuclear Plant Operator i o) H:NPO-4.3 Nuclear Plant Operator Safety Evaluation Summary Implementation of the EDG loading evaluation assumptions demonstrates that the i FS AR load evaluation remains bounding,'and that the EDGs meet the loading '

requirements set forth in the PTPN Technical Specifications. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the s basis for any Technical Specification is decreased.

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Date of Changes: July ,27, 1987. l

5) A study of emergency diesel generator (EDG) loading determined that a potential existed to exceed the FSAR and emergency operating procedure loading limits for the EDGs. Based on this plant changes were made along with revising the following procedures:

a) 3(4)-EOP-E-0 Reactor Trip or Safety Injection b) 3(4)-EOP-ES-0.1 Reactor Trip Response l c) 3(4)-EOP-ES-0.2 Natural Circulation Cooldown I d) 3(4)-EOP-ES-0.3 Natural Circulation Cwoldown With Steam Void in Vessel (With RVLMS (QSPDS))

e) 3(4)-EOP-ES-0.4 Natural Circulation Cooldown With Steam Void in Vessel (Without RVLMS) f) 3(4)-EOP-E-1 Loss of Reactor or Secondary Coolant g) 3(4)-EOP-ES-1.1 SI Termination ,

h) 3(4)-EOP-ES-1.2 Post LOCA Cooldown and Depressurization

1) 3(4)-EOP-E-3 Steam Generator Tube Rupture j) 3(4)-E0P-ECA-0.1 Loss of All AC Power Recovery Without SI Required k) 3(4)-EOP-ECA-0.2 Loss of All AC Power Recovery With SI Required

1) 3(4)-EOP-ECA-2.1 Uncontrolled Depressurization of All Steam Generators m) 3(4)-EOP-ECA-3.1 SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired n) 3(4)-0N0P-004 Loss of Offsite Power o) 3(4)-0P-006 480 Volt Switchgear System p) OP 9104.1 Main Transformer - Periodic Tests q) ONOP 9108.1 Main Transformer - Malfunction r) 3(4)-0P-007 480 Volt Motor Control Centers Safety Evaluation Summary Implementation of the EDG loading evaluation assumptions demonstrates that the FSAR load evaluation remains bounding, and that the EDGs meet the loading requirements set forth in the PTPN Technical Specifications. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Changes: July 28, 1987.

6) During a review of the 4160 volt bus lockout schemes for both units, two types of devices were identified, whose failure could adversely affect the ability of safety related equipment to perform their intended safety function. Based on this the following procedures were revised:

a) 3(4)-EOP-ES-0.l Reactor Trip Response b) 3(4)-EOP-E-0 Reactor Trip or Safety Injection Safety Evaluation Summary:

The addition of cautiona on ensuring that the battery chargers are energized within 30 minutes will provide additional assurance that the requirements of the FSAR are met. Therefore, the proposed procedural change will 3

l not increase the probability or consequences of an accident analysed in the I FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin cf safety as defined in the basis for any Technical )

Specification is decreased.

Date of Changes: August 15, 1986 )

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7) A temporary procedure was written to provide instructions for running the B AFF pump to maintain the Unit 3 steam generator levels, so that maintenance can Fs performed on PI-3-1435. The following temporary procedure was developea: l l

a) TP 277 Use of AFW to Maintain SG Levels During Repair of !

PI-3-1435 Safaty Evaluation Summaary:

The TP allows operation of train 2 AFW to maintain plant power at equal to or less than 5 percent in accordance with the AFW system design. Therefore, the )

proposed procedural change will not increase the probability or consequences of I

) an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: August 24, 1986

8) The FSAR states that the mechnical overspeed trip setpoint has a value of 11%

above rated speed (1998 RPM). The folinwing procedure was revised with an on-the l spot change to provide an acceptance b.ad for this setpoint.

a) OP-8004.1 Turbine Generator - Overspeed Trip Test Safety Evaluation Summary Revising the overspeed setpoint to less that 11% of rated speed is a more conservative setting with regard to protection of the turbine. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: September 6, 1986

9) In order to incorporate the requirements of Technical Specification Amendment 118 (Fecility Operating License DPR-31 for Unit 3) and Ametiment 112 (Facility Operating License DPR-41 for Unit 4) which incorporated a new Technical Specification for the Standby Steam Generator Feedwater System, the following procedures were revised:

a) OP-0202.1 Reactor Startup - Cold Shutdown to "ot Standby Conditions Date of Change: July 17, 1986 b) AP-0190.16 Scheduling and Surveillance of Periodic Tests and Checks 4

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R quired by Technical Specifications !

Date of Change: September 26, 1986 l c) AP-0103.7 Reports Required by Technical Specifications and 10 CFR d) ONOP-0208.9 Annunciator List - Panel G - Miscellaneous e) OP-16001.2 Technical Specification Surveillance Requirements for l Core Refueling j f) 0-GP-018 Demineralized Water System i g) 0-0P-074.1 Standby Steam Generator Feedwater System h) 0-0N0F-074.1 Standby Steam Generator Feedwater System Operation With Loss of Offsite Power and Loss of Auxiliary Feedwater l i) 0-OSP-200.1 Schedule of Plant Check and Surveillance j j) 0-0SP-074.3 Standby Steam Generator Feedwater Pumps Availability Test I k) 0-OSP-074.4 Standby Steam Generator Feedwater Pumps / Cranking Diesels Test

1) NPO-4:2 Nuclear Plant Operator Logsheet l

Date of Changes: October 2, 1986 l l

m) AP 0103.12 Notification or Significant Events to NRC Dage of Change: October 9, 1986 j

10) Procedure changes were made to incorporate the operation of diesel afr compressors i as described in temporary procedures (TP) 250 and 251 into existing plant procedures. These diesel air compressors were installed for the emergency diesel genert'or loading evaluation. The following procedures were changed.

a) 3(c.)-0P-013 Instrument Air System b) 0-ONOP-013 Loss of Instrument Air Safety Evaluation Sunnary Implementation of the EDG loading evaluation assumptions demonstrates that the .

FSAR load evaluation remains bnunding, and that the EDGs meet the loading {

requirements set forth in the PTPN Technical Specifications. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor,-will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction j of a different type will be created and no margin of safety as defined in the l basis for any Technical Specification is decreased.

Date of Changes: July 28, 1987.

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11) Due to a change in a breaker number, the following procedures were revised:

a) 3(4)-OP-005 4160 Volt Buses A and B l b) 0-0N0P-074.1 Standby Fdedwater System Operation With Loss of Offsite l Power and Loss of Auxiliary Feedwater c) ON0P-0208.8 Annunciator List - Panel F - Electrical d) ONOP-0206.16 Annunciator List - Panel J - Auxiliary Electrical Power e) ONOP 9408.2 Energizing 4KV Buses Using the Cranking Diesels Bus Tie ,

j Lines or Startup Transformer from the Opposite Unit l f) ONOP-9308.1 Startup Transformer - Malfunction l l 5

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The changes consisted of administrative changes only and did not affect the operation of the breaker. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no macgin of safety as defined in the basis for any Technical Specification is decreased.

Date of Changes: October 7, 1986 L2) The FSAR states that the mechnical overspeed trip setpoint has a value of 11%

above rated speed (1998 RPM). The following procedure was revised to permanently incorporate a prior on-the-spot-change approved on September 6, 1986.

a) OP-8004.1 Turbine Generator - Overspeed Trip Test Safety Evaluation Summary Revising the overspeed setpoint to less that il% of rated speed is a more conservative setting with regard to protection of the turbine. Therefore, the proposed procedural enange will not increase the probability or consequences of an accident analyzed in the 7SAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: October 14, 1986 L3) In order to load test the 4A bactery da the vital DC system a temporary procedure was written to perform this evolution. The following temporary procedure was written:

a) TP 286 Unit 4 - Removal and Return to Service 4A DC Battery for Load Test Safety Evaluation Summary The evaluation concludes that during the test the 3B and 4A DC busses will be interconnected to keep all four instrument channels functional. The plant will be operating under a DC system LCO per Technical Specification 3.7.2 during the plant condition, while battery 3B and battery chargers 33 and 4S feed both DC busses 3B and 4A. Neither the basic design nor die permanent configuration of any equipment are affected by this change. The interconnection of DC busses 3B and 4A will be governed by the requirements of Technical Specifications for a battery out of service. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR not, will it impact the functioning of any safety related equipment important to safety.

No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: November 25, 1986 6 l 1

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a) TP 290 Unit 3 - Removal and Return to Service 3A DC Battery for Load Test Safety Evaluation Summary l The evaluation concludes that the temporary battery can be used to replace the 3A '

battery to feed bus 3A with a DC systes LCO invoked during the service testing of i

the 3A battery. 'Therefore, the proposed procedural change will not increase the i

probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety.

No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is J decreased.

Date of Caange: November 28, 1986 l

15) A review of the design of the control room ventilation system discovered a the potential for loss of the control room ventilation system that could prevent the system from performing its intended function. Based on this the .

following procedures were developed: l a) TP 291 Loss of Control Room Ventilation System (CRVS) Air Conditioning Safety Evaluation Summary The interim actions taken ensure that sufficient control room cooling and atmosphere cleanup capability exists for accident scenarios described in the FSAR. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it ;

impact the functioning of any safety related equipment important to safety.

No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: December 18, 1986

16) In order to incorporate the requirements of Technical Specification Amendment 120 (Facility Operating License DPR-31 for Unit 3) and Amendment 114 (Fncility Operating License DPR-41 for Unit 4) which incorporated a revision to the emergency diesel generator 18 month preventative maintenance requirements which changed them to coincide with the Unit 3 refueling outage schedule. The following procedures were revised:

a) OP-4304.3 Emergency Diesel Generator - Eight Hour Full Load Test 1 and Load Rejection b) OP 16001.2 Technical Specification Surveillance Requirements For Core Refueling '

c) 0-OSP-200.1 Schedule of Plant Checks and Surveillance i d) 0-OSP-200.2- Plant Startup Surveillance e) 3(4)-OSP-203 Engineered Safeguards Integrated Test 7

Date of Chznges: January 7, 1987 17.) A review of the design'of the control room ventilation system discovered a the potential for loss of the control room ventilation system that could prevent the system from performing its intended function. This evaluation was revised to allow operation of the heating feature of the control room heating /

air conditioning system. Based on this the following procedure was revised: ;

a) TP 291 Loss of Control Room Ventilation System (CRVS) Air Conditioning Safety Evaluation Summary The interim actions taken ensure that sufficient control room cooling and i atmosphere cleanup capability exists for accident scenarios described in the FSAR. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it 1 impact ' the functioning of any safety related equipment important to safety.

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No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: January 29, 1987

18) In order to load test the 4B battery in the vital DC system a temporary procedure was' written to perform this evolution. The following temporary procedure was written:

a) TP 310 Units 3 and 4 - Removal anc Return to Service 4B DC Battery for Load Test Safety Evaluation Summary The evaluation concludes that the temporary battery can be used to replace the 4B l

, battery to feed bus 4B with a DC system LCO invoked during the service testing of ,

the 3A battery. Therefore, the proposed procedural change will not increase the !

probability or consequences of an accident analyzed in the FSAR nor, will it ,

I impact the functioning of any safety related equipment important to safety.

No new accident or malfunction of a different type will be created and no l margin of safety as defined in the basis for any Technical Specification is decreased. l Date of Change: March 10, 1987 ;

19) In order to lower the Unit 3 spent fuel pit level for maintenance activities, the following temporary procedure was developed:

a) TP 313 Unit 3 3FP Level Reduction for Maintenance Safety Evaluation Summary: <. .

During the reduction of SFP level, no fuel movement shall occur, no crane operation with loads shall be conducted, and the temperature of the SFP water shall not be allowed to exceed 180 degrees Fahrenheit. By following these restrictions, the proposed temporary procedure will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new 8

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accidant or malfunction of a different typa will ba created and no margin of safety as defined in the, basis for any Technical Specification is decreased.

Date of Change: March 17, 19G7

'20) A temporary procedure was written to provide instructions for the recirculation and draining of Demineralized Water Storage Tank. This was required after routine

- samples of the tank contained very low levels of cobalt 58 and cobalt 60. The tank was isolated and analysis of the tank indicated that the concentrations were acceptable for direct release to the atmosphere. These instructions include all necessary prerequisites and guidelines for the tank to be recirculated for sampling requirements and for the gravity draining to the discharge canal. The procedure also ensures that the tank is isolated from external sources of liquids during the draining process and during non-recirculation modes of operation.

a) TP 334 DWST Recirculation and Release to Discharge Canal Safety Evaluation Susumary:

The flushing and discharge of' the contents of the DWST to the cooling canal l neither involves nor is associated with any accident scenario described in the FSAR. The TP incorporates the requirements of Technical Specification 3.9 for radioactive liquid releases. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defincd in the basis for any Technical Specification is decreased.

Date of Change: April 27, 1987

21) In order to incorporate the requirements of Technical Specification Amendment 123 (Facility Operating License DPR-31 for Unit 3) and Amendment 116 (Facility Operating License DPR-41 for Unit 4) which incorporated a revision to the reporting requirements of section 6.0. The following procedures were revised. l a) AP-Oln3.7 Reports Required By Technical Specifications and 10 CFR b) ONOP-2608.2 Chemical and Volume Control System Malfunction of -

Boron Concentration Control System c) MP-15537.3 Surveillance of Penetration Fire Barriers d) ON0P-15538 Fire and Smoke Detection System and Fixed Fire 1 Protection Equipment / Systems - Operating Instructions )

Date of Changes: May 5, 1987

22) An evaluation of the switchover to cold leg recirculation during a LOCA determined that a 10 minute interruption of flow could not be supported using the latest analytical assumptions. The fdllowing procedures were revised, a) 3(4)-EOP-ES-1.3 Transfer to Cold Leg Recirculation b) 3(4)-EOP-ES-1.4 Transfer to Hot Leg Recirculation i

Safety Evaluation Summary The revisions have been made to maintain continuous SI flow to the RCS following i

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larga breaks (unless spray cooling for the ecergancy containment filters is required or where RHR f, low indicator FI-605 exhibits erratic behavior) and to minimize the time of interruption of SI flow during switchover to recirculation for small breaks. Therefore, the proposed procedural change will not increase the probability or. consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment impo rtant to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Changes: May 26, 1987

! 23) Due to problems with nitrogen inleakage through the boric acid transfer pumps, peocedure changes were made to have the nitrogen supply normally isolated to the pumps. The following procedures were revised:

a) ONOP-2608.2 Chemical and Volume Control System Malfunction of Bdron Concentration Control System j b) 0-0P-065.3 Nitrogen Gas Supply System l c) 0-0SP-201.2 SNPO Daily Logs d) 0-ADM-205 Administrative Control of Valves, Locks and Switches 1

Safety Evaluation Summary l These changes do not introduce any conditions that are beyond the design capabilities of the boric acid transfer pumps or represent a condition that is a outside the design basis for the boric acid transfer pumps. Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety !

related equipment important to safety. No new accident or malfunction of a j different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased. j j

Date of Changes: June 10, 1987 I

24) In order to incorporate the requirements of Technical Specification Amendment 124 (Facility Operating License DPR-31 for Unit 3) and Amendment 118 (Facility '

Operating License DPR-41 for Unit 4) which incorporated a revision to the requirements for the auxiliary feedwater system and the condensate storage tanks. !

These changes include adding the modes of applicability and additional action I statements. The following procedures were revised as a result of this amendment.

a) 3(4)-0P-018.1 Condensar.e Storage Tank (CST) b) 3(4)-0P-075 Auxiliary Feedwater System c) 3(4)-0SP-075.1 Auxiliary Feedwater Train 1 Operability Verification d) 3(4)-0SP-075.2 Auxiliary Feedwater Train 2 Operability Verification e) 3(4)-OSP-075.3 AFW Nitrogen Backup System Low Pressure Alarm Setpoint and Leakrate Verification f) 3(4)-0SP-075.6 Auxiliary Feedwater Train 1 Inservice Test g) 3(4)-OSP-075.7 Auxiliary Feedwater Train 2 Inservice Test h) 0-0SP-075.9 AFW Overspeed Test Date of Changes: June 18, 1987 i) 3(4)-OSP-201.1 RCO Daily Logs 10

Date of Changes: July 1,.1987

25) An evaluation of the HVAC system for the DC equipment / inverter rooms discovered setnarios which could result in a loss of the HVAC function. The loss of the HVAC could result in elevated temperatures in the affected~ rooms and could affect the operability of equipment in these rooms. The following procedures were !

I developed or revised.

a) 3(4)-ON0P-004 Loss of Offsite Power !

b) TP-347 DC Equipment and Inverter Rooms Supplemental Cooling Monitoring and Standby Conditions ;

c) TP-348 DC Equipment and Inverter Rooms Continuous Supplemental j Cooling d) TP-349 DC Equipment and Inverter Rooms Supplemental Cooling i During Normal Conditions e) TP-350 DC Equipment and Inverter Rooms Supplemental Cooling l During Fire Conditions f) TP-351 DC Equipment and Inverter Rooms Supplemental Cooling '

During LOOP Conditions

.g) TP-352 DC Equipment and Inverter Rooms Supplemental Cooling ;

During LOOP Conditions with LOCA Date of Review: June 5, L987 h) TP-346 DP-412A Transfer Switch Inspection )

i) TP-355 DC Equipment / Inverter Rooms HVAC Maintenance ,

Date of Review: June 16, 1987 l l

l ,j) TP-356 CRDM M/G Set 4A/4B Load Sharing k) TP-357 CRDM M/G Set 3A/3B Load Sharing j l

Date of Review: June 18, 1987 i

Safety Evaluation Summary The implementation of the interim actions will ensure that temperatures remain within electrical equipment design temperatures. Therefore, the proposed procedural changes will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety I related equipment important to safety. No new accident or malfunction of a j d

different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

l I Date of Changes: June 5, 1987 l

26) In order to inspect cable splices in the electrical penetrations, installed fire protection material (Flamemastic) was'r'emoved from selected cables for Unit 4. An 3' evaluation was written to suppo'rt the temporary removal of the Flamemastic from the affected cable trays in the Unit 4 containment penetration areas until the '

next refueling outage when the Flamemastic will be reapplied. Based on this evaluation, the following procedure was revised.

a) ON0P-15538 Fire and Smoke Detection System and Fixed Fire Protection 11 )

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Equipment /Systens - Operating Instructions !

Safety Evaluation Summary Removal of Flamemastic does not increase the probability of occurrence of a ,

malfunctin of equipment important to safety since Flamemastic is considered a i passive fire . protection feature. With the Flamemastic removed, compensatory ,

measures as incorporated in the above procedure change will be implemented. The I implementation of these compensatory easures together withthe existing available

fire protection features provides a high degree of assurance that at least one train of redundant safe shutdown cables will remain free of fire damage. l Therefore, the proposed procedural change will not increase the probability or consequences of an accident analyzed in the FSAR nor, will it impact the functioning of any safety related equipment important to safety. No new accident or malfunction of a different type will be created and no margin of safety as defined in the basis for any Technical Specification is decreased.

Date of Change: June 5, 1987 i

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i (iii) TESTS AND EXPERIMENTS This section contains the results and conclusions for special tests that were completed during the reporting period. Special tests still in progress at the .)

end of the reporting period are also described. l SPECIAL TEST NLMBER TITLE UNIT (S) 86-20 "B" Auxiliary Feedwater Pump - Performance 3 and 4 Test 1 86-24 Main Feedwater Check Valve Cycling Test 3 and 4 86-26 Containtrent Spray Recirculation Piping 4 Flushing and Pressure Drop j Unit 3 TPCW H )

86-28 x Performance Test to Evaluate 3 Effectiveness of Chemical Dispersion Technique 86-29 Unit 4 TPCW H x Performance Test to Evaluate 4 i Effectiveness of Chemical Dispersion Techniques 86-30 Condensor Cathodic Protection Special Test - 3 and 4 Reduce Cathodic Protection Contributions to , '

Hydriding 86-31 Waste Gas System Leak Test 3 and 4 86-32 Unit 3 VCT Purge Special Test 3 86-35 Unit 4 VCT Purge Special Test 4 :

87-01 System Generator Blowdown Control Valves 3 and 4 Delayed Actuation Verification Test i

87-03 WTP, DWST, and Condensate Dissolved 4 0xygen Monitoring Program 87-08 Condensor Tubing Cathodic Protection 3 Monitoring System l 87-11 Safety In,iection System M0V Differential 3 and 4 Pressure Stroke Testing Valves:

1 M0V-3-843 A/B and M,0V-878 A/B 87-13 Condensor Tubing Strain Measurement Test 3 l

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SPECIAL TEST 86-20 "B" AUXILIARY FEEDW5TER PU W - PERFORMANCE TEST Performed: August 7, 1986 ,

BACKGROUND INFORMATION:

This special test was conducted to study the performance of the "B" Auxiliary '

Feedwater Pump after the installation of the underfiled pump impellor.

_TI:ST RESULTS:

The study concluded that the under-filed impellor replacement improved the pump perfomance to above the design requirement.

SAFETY EVALUATION:

In Section 9.7 of this Special Test, the "B" AFW pump recirculation line back to the Condensate Storage Tank will be closed. The procedure alerts the Reactor Control Operator to maintain flow through the pump at all times during this I;ortion of the test to prevent the pump from dead heading. In addition, the Turbine Operator shall be standing by the recirculation valve 20-277 to open ,

it in the event of an AFW AUTO START signal. These precautions will ensure the safe operation of the AFW pump.

Sections 9.9 and 9.10 of this Special Test require the Terbine Operator to use the governer overspeed trip device to increase AFW. turbine speed to 6000

! 15 rpm and 6100 2 15 rpm respectively. The electrical overspeed trip at 6200

- t 50 rpm and the mechanical overspeed trip at 6500 50 rpm are both operable and will provide turbine protection if required. f Section 9.11 obtains data to evaluate what effect the Turbine Lube Oil Cooler I line coming from the AFW pump second stage has on pump performance. The valve manipulation in this section shall be independently verified to ensure that ]

i adequate cooling is maintained to the pump and turbine lube oil and that no service water recirculates back to the Condensate Storage Tank. j This Special Test does not involve an unreviewed safety question, nor does it increase the probability or severity of an accident either previously or not previously analyzed in the FSAR. It does not increase the probability of an actise or passive failure of any AFW equipment important to the function j of the AFW System. Similarly, it does not increase the severity of the 4 consequences of an AFW equipment malfunction.

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SPECIAL TEST 86-24 MIN FEEDWATER CHECK VALVE CYCLING TEST BACKGROUND INFORMTION:

This information is a special test to determine if the check valve shows any indication of binding of the rockshaft when rotated.

l TEST RESULTS:

Test was run 7/21/86 through 7/23/86 on Units 3 & 4. Data sent to Site Engineering August 22, 1986.

SAFETY EVALUATION:

SPECIAL TEST OF MIN FEEDWATER CHECK VALVES ;

SAFETY ANALYSIS:

The special test has been evaluated with the following results:

1) The probability of occurrence or the consequences of a design basis accident or malfunction of equipment important' to safety previously evaluated in the FSAR has not been increased due to both units being in cold shutdown and the Tech. Spec. 3.4 requirement for two coolant loops being maintained.

2) The possibility of an accident or malfunction of a different type than evaluated previously in the FSAR has not been created as all ,

Tech. Spec. requirements for cold shutdown are being maintained.

Realignment of the feedwater system does not impact the unit as feedwater is not required and typically isolated at cold shutdown ;

with RHR in service. l

3) The margin of safety as defined in the basis for a Technical Specification has not been reduced as the Tech. Spec 3.4 requirement for two coolant loops is being maintained by one RHR pump and one RCP/SG pair ava11abe for service. There is no requirement that makeup j be provided to the steam generators during cold shutdown; however. l the test provides an extra margin of safety by providing an available )

makeup supply path during the test. ;

Therefore, the proposed special test does not involve an unreviewed i safety question. l EK/rlh/:026 l 1

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SPECIAL TEST 86-26 l

CdNTAINMENTSPRAYRECIRCULATIONPIPING FLUSHING & PRESSURE DROP 1

l Background Information: 1 1

This special test was performed in order to determine if there was any blockage in the containment spray recirculation piping for unit 4. This test was performed with both unit 3 and 4 in cold shutdown.

Test Results:-

This test was performed and the results were forwarded to engineering and the flow rates calculated from the pressure drop data matched design flow rates. . j I

Safety Evaluation: j j

'This test was written to be performed in modes 5 or 6 only. These modes do J not require any containment spray trains to be operable. Therefore, special j test 86-26 did not pose an unreviewed safety question. j o

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SPECIAL TEST 86-28 Unit 3 TPCW'HX Performtsee Test to Evaluate Effectiveness of Chemical Dispersion Technique i

Background Infonnation:

The special test was performed to give an indication of the effectiveness of adding a Calgon water treatment chemical to reduce HX tube fowling rates.

This Test was to compare to Unit 3 data. Completed 10-21-86 .

l Test Results:

The tests were inconclusive. Future evaluations will be performed.

Safety Evaluation:

The test was performed with adequate safeguards to assure that the ICW system functioned as normal. No changes were made which would adversely affect the systems operatinq capability. Normal line-up was maintained. Based on the above Special Test 86-28 did not pose an unreviewed safety question.

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Unit 4 TPCW HX Performance Test to Evaluate Effectiveness of Chemical Dispersion Technique i Background Information: ,

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The special test was performed to give an indication of the effectiveness of ]

I adding a Calgon water treatment chemical to reduce HX tube fowling rates, i This Test was'to compare to Unit 3 data. Completed 10-21-86 l

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l Test Results:

l The tests were inconclusive. Future evaluations will be performed.

Safety Evaluation: ,

i The test was performed with adequate safeguards to assure that the ICW system functioned as normal. No changes were made which would adversely affect the systems operating capability. Normal line-up was maintained. Based on the above Special Test 86-29 did not pose an unreviewed safety question. 1 i

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SPECIAL TEST 86-30 SUleULRY

- . l Background Information:

The purpose of this test is to reduce / eliminate the excessive hydriding of the titanium tubes in the condensers of Unit 3 and 4.

The test involves disconnecting some of the anodes in the condensers and reducing the rectifier outputs.

Test Results:

On Going

' Safety Evaluation:

The configuration for this test does not involve an unreviewed stifety question nor does it increase the probability of an accident because:

1. The Condenser Cathodic Protection System is non-safety related.

The design intent of the Cathodic Protection System remains the same.

This modification does not alter the operation or function of any other plant safety or non-safety systems. Therefore, the probability of malfunction of equipment important to safety has not been increased, and the consequences of malfunction of equipment to safety previously evaluated in the FSAR are not affected.

2. This test does not interface with any safety . equipment. This modification does not alter the operation or function of any safety related systems. Hence, the possibility of an accident of a different type than analyzed in the PSAR is not increased, and the possibility of a malfunction of a different type than analyzed in the FSAR has not been created.

3. This modification ~ does not affect the integrity, operation or function of any safety related system addressed in the Technical Specifications. Therefore, the margin of safety as defined in the basis for any plant specification is not decreased.

In conclusion, the subject special test does not constitute an unreviewed safety question. .

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SPECIAL TEST 86-31 WASTE GAS SYSTEM LEAX CHECX Background Information:

Subject test was completed on 12/86 l

Test Results:

The Test consisted of isolating the equipment that uses the Waste Gas System: ,

CVCS, Hold-up Tanks, Spent Resen Storage Tank, Reactor Coolant Drain Tanks, l Sample Rooms, Gas Stripped Boric Acid Evaporator and Volume Control Ta nk s . l The Waste Gas System was then filled with Np and He gases and it was leak checked using a H e detector. The system was aTso sectionalized using existing i system isolation valves, and pressure drop tests were taken for each section tested. l The Test proved certain theories to be erroneous and provided data useful in resolving some waste gas system problems. l l

l Safety Evaluation-l l

The test proceeded uneventfully and at no time during it's course was there j any circumstances that could have created an unreviewed safety question nor , ,

lowered factors of Safety of any Tech. Specs. !

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SPECIAL TEST 86-32 UNIT 3 VCT PURGE SPECIAL TEST Background Information:

l This special test was written and performed to provide a method to reduce the amount or waste gas system piping involved in the Volume Control Tank (VCT) purge. . The waste gas compressor discharge to cover gas cross tie was isolated to preclude VCT purge gases from entering the cover gas header. The gas flow path for this special test was the same as per plant procedures, from the VCT to the waste gas compressor and to the Gas Decay Tanks. A secondary purpose of this special test was to locate the general area of waste gas system leakage if any.

1 Test Results: i i

This special test was performed and some waste gas system leakage was located in the area of the waste gas compressors.

Safety Evaluation:

This special test was found not to envolve an unreviewed safety question ,

because the flow path used is the same as described in the FSAR. The cross ;

ties from the waste gas compresscr discharge to the cover gas header were isolated to minimize the amount of waste gas system piping exposed to VCT purge gases and had no detrimental effect on waste gas compressor operation, i l

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SPECIAL TEST 86-35 UNIT 4 VCT PURGE SPECIAL TEST Background Information:

s This special test was written and performed to provide a method to reduce the amount or . waste gas system piping involved in the Volume Control Tank (VCT) )

purge. The waste gas compressor discharge to cover gas cross tie was isolated to preclude VCT purge gases from entering the cover gas header. The gas flow path for this special test was the same as per plant procedures, from the VCT to the waste gas compressor and to the Gas Decay Tanks. A secondary purpose of this special test was to locate the general area of waste gas system leakage if any.

Test Results:

This special test was performed with no significant leachages detected from the waste gas system piping.

Safety Evaluation:

This special test was found not to envolve an unreviewed safety question , ,

because the flow path used is the same as described in the F3AR. The cross l ties from the waste gas compressor discharge to the cover gas header were !

isolated to minimize the amount of waste gas system piping exposed to VCT purge gases and had no detrimental effect on waste ga'. compressor operation.

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SPECIAL TEST 87-01 STEAM GENERATOR BLOWDOWN CONTROL VALVES DELAYED ACTUATION VERIFICATION TEST Background Information:

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The purpose of Special Test 87-01 is to verify the setpoints on the Agastat timer to Steam f Generator Blowdown Control Valves CV-6275A, CV-6275B and CV-6275C. Each Agastat j timer provides a delayed opening signal to its respective control valve following a placement of control switch (HS-6275A-1, HS-6275B-1, or HS-6275C-1) from the closed to the open l position. A stopwatch will be used by the Control Room personnel to measure the time l l Interval from the point a control switch is turned to the open position until the first )

indication of stem travel of the tested control valve is observed (i.e., when dual position .)

lights of the tested valve have liluminated on the control board). Any discrepancies between - '

the value of the setpoints indicated on the Agastat timers and the actual measured setpoints

. (i.e., the time intervals determined during the test) will be documented for incorporation into the system's maintenance package.

The test will also utilize field personnel stationed near the control valves. Close radio l communications will be maintained between the Control Room and field personnel durittg test evolutions. The field personnel will verify that the pressure in the header being tested attains the secondary side steam generator pressure (as indicated on a temporarily installed pressure gauge) prior to the actuation of the tested control valve.

Status:

The procedure to perform the special test is not yet completed, but is tentatively scheduled

, to be available by October 1,1987.

Test Results:

No testing has yet been performed.

Safety Evaluation:

The safety evaluation for Special Test 87-01 will be available following a completion of the written test procedure. ,,, l l

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SPECIAL TEST 87-03 WTP, DWST AND CONDENSATE DISSOLVED OXYGEN MONITORING PROGRAM Background Information:

.Special Test 87-03 monitors dissolved oxygen levels in the make-up water system to Unit

4. High levels of dissolved oxygen in make-up water can cause corrosion in secondary system piping. Juno Beach Nuclear Services, Westinghouse and INPO recommend less than 100 ppb of oxygen in the Demineralized Water Storage Tank (DWST). The Turkey Point Nuclear Units presently average about 2000 ppb of oxygen in their tanks.

This special test samples deoxygenated water at various points from the Water Treatment Plant (WTP) effluent to the discharge of the Condensate Pumps to Unit 4. The points sampled include the DWST, the Demineralized Water Degasifier, the Condenser Hotwells, the suction side of the Condensate Pumps, as well as, the WTP effluent and the Condensate Pump discharge. Dissolved oxygen values are obtained during normal operation of the WTP to the DWST with the Demineralized Water Degasifier in service and with it out of service.

Status:

The test is presently awaiting the site arrival and subsequent installation of the dissolved oxygen analyzers. Installation of all equipment is scheduled for completion by August 1,1987. The test is tentatively scheduled to be completed by the end of September,1987. #

Test Results:

Test data is scheduled to be taken in August, 1987 following the installation of all test l equipment.

1 Safety Evaluation:

The components and sytems which make up Special Test 87-03 do not interact with any nuclear safety-related systems. The WTP is a Quality Group C system which is defined in the FSAR as "important to operation but not essential to safe shutdown and isolation of the reactor or control of the release of substantial amount of radioactivity". The Condenser and Condensate Systems are Quality Group D systems while the DWST is described in the FSAR as an alternate, non-safety source of water for the Auxiliary Feedwater (AFW) pumps or the Condensate Storage Tanks (CSTs). The installation and operation of the test equipment does not affect the normal operation of these systems.

Any postulated failure of the test equipment or of these systems would include tank rupture, pipe breaks, sample ilne breaks and varve. failure. These types of accidents would not impact any equipment or system important to safety.

Based on the above, it can be said that the modifications performed under Special Test 87-03 are acceptable from the standpoint of nuclear safety. The modifications do not constitute an unreviewed safety question nor do they require a change to the Technical Specifications.

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(fISPf2ALTEST87-08b CONbENSERTUBINGCATHODICPROTECTIONMONITORINGSYSTEM ,.

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Background Inf'armatign,:_ g'

. The purpose of .the test is to obJain data on cathodic protection levels in

%the Unit 3 condenserfwaterboxes and. tube sheets. This data will be evaluated A

y ( ty General Engineerirjg. and usedf to obta.in safe operating levels of the cathodic 4 protection syst(g k N/uut causing excessive. hydrogen evolution. Excessive -l 1 yydrogen can be ibrorbed into the' condenser titanium tubes causing their I embrittlement and evonpul failure. 4 3 s

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F The cathodic protectihn monitoring system installed under .the special test '

3 consist of six refereAce electrodes mounted on each inlet and outlet tube sheet gy'. to the conden;er, connecting wires that will, be routed through an existing J,, penetration in M, -

data aquisition. 'each waterbog, a junction box and a personal computer for j

s 4 ;

Status: i '

e<NWith the exception of th.e personal computer, all equipment associated with l I

/ the monitoring system under Special Test 87-08 has been installed. The computer .l t 1 will be installed prior to a startup of the unit. The computer will be used l to obtaim the necessary data on cathodic protection levels during normal !

poweropeptiop

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fPMe has not been taken shce the system is not yet functional and the unit !

tispotonline.

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Safety Evaluations:

The condenser, condenser waterboxes and the monitoring system installed under I Special Test 87-08 perform no safety related functions. The condenser and its associated componenta he classified as Non-Nuclear Safety Related, Quality Group D and Non-Classi 1E. The installation, operation and any postulated {

f;guures of the installed monitoring system will not interfere nor interact {

with any safety related equipment, j Based on the above, it ca'n be stated that the modifications performed under l Special Test 87-08' are acceptable from the standpoint of nuclear safety. The d modifications do not constitute so unreviewed safety question, nor do they l j require a change to the Technical $pnifications. j 0

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SPECIAL TEST 87-11_ f SAFETYIN[ECTIONSYSTEMMOVOIFF. PRESS.STROKETESTING VALVES: MOV-3-843 A/B /ND MOV 878 A/B ;

Background Information: ,

Test completed, awaiting complete documentation package prior to transmittal to Document Control.

NRC I & E Bulletir, 85-03 item C (CTRAC # 86-0692-03)= requires demonstration of M0V . operability in HHSI systems required to be tested for operational-readiness in accordance with 10 CFR 50.55A.

The test cycled selected valves under tne maximum practical achievabl e .

differential pressure, with an SI pump operating at minimum flow conditicns and suciion from either the' RWST or RHR pumps, as required by. the design conditions established for each valve. !

2 2st Results:

l The selected valves cycled properly, without difficulty. The differential pressures achievdd for each valve are as follows:

i M0V-3-878, DP = 1505 psid i MOV-3-843, DP = 1585 psid ,

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Safety Evaluation: 1 This test aligned the system.in modes' described in the FSAR and cycled .alves to demonstrate operability under maximum achievable differential pressures.

Therefore the probability of occurance of an accident, or the seriousness of the consequences of an accident, or the probability of malfunction of equipment, or the seriousness of the consequences of malfunction of equipment is not increased.-

There were ,nc. har.1 Ware modifications required as part of this test.

Therefore, thcre is 'no . possibility for an accident of a. different type, as possibility -for a malfunction. of a different type than any evaluated previously.

The test was performed within Tech. Spec. limitations, therefore the margin of safety was preserved.

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l SPECIAL TEST 87-13

.J CONDENSER TUBING STRAIN MEASUREMENT TEST f j

Background Information:

The purpose of Special Test 87-13 is to measure tne strain of Unit 3 condenser titanium tubes following the unit's return to power after the 1987 refueling i outage. Data on actual tube strain levels during full power operation is required in order to develop a more accurate criteria for future tube plugging due to hydriding (hydrogen embrittlement) of the tubes.

The special test consist of strain gauge rosettes installed on the shell side i of six (6) condenser titanium tubes. Connecting cables to the gauges are mounted along the tube sheet to the 3BS outlet waterbox and encased in ARCOR 100% solids epoxy. The cables are then mounted within 1 " carbon steel, Schedule 40 piping installed along the west side and south side condenser walls, through an existing penetration and to a computer for data acquisition.

Status:

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Installation of the strain gauge test equipment has been completed. However, only 5 of 6 titanium tubes were installed with strain gauges due to a lack of unplugged ' tubes in one desired area. The installation and operability 6f 1 the strain gauge equipment was verified by the Technical Department on June j 25, 1987. Tube strain data will be taken during the startup of the unit. l l

l Test Results:

Test data has not yet been obtained but will become available following a startup i of the unit. >

Safety Evaluation:

The Unit 3 condenser is a Non-Nuclear Safety Related component and, as such, )

is not required to function during any existing analyzed accident scenarior.

Therefore, modifications to the condenser as performed under Special Test 87-13 affects only Non-Nuclear Safety Related, Quality Group D equipment.

- The chemical content of all materials or solutions used in preparing the test will not adversely affect any components or the secondary chemistry.

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The condenser wall penetration utilized during the test and capped following the completion of the test will meet all design criteria of existing penetrations to insure that the condenser pressure boundary is maintained. j i

.SPECIAL TEST 87-13 l CONDENSER TUBING STRAIN MEASUREMENT TEST Any ' postulated failures of the materials or equipment used in the test will have no impact on the safe shutdown on the plant, or on any safety related systems. Any materials associated with the test located inside the condenser which could be postulated to become dislodged would be caught in the condenser hotwell pump screen. None of these materials would be large enough to impact pump suction.

Additionally, postulated failures of the condenser would have no impact on the safe shutdown of the plant, or on'any safety related systems. .The condenser

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is not used to prevent postulated accidents,. mitigate the consequences of such '

accidents, maintain. safe shutdown conditions, or adequately store spent fuel.

Based on the above, it can be said that the modifications performed under Special Test 87-13 are acceptable from the standpoint of nuclear safety. The modifications do not consititute an unreviewed safety question nor do they require a change to the Technical Specifications.

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APPENDIX A ANNUAL REPORT OF SAFETY ')

AND RELIEF VALVE CHALLENGES '

By-letter dated. June 13, .1980 (L-80-186), Florida Power and Light stated the intent to comply with the requirements of Item IIK.3.3 of Enclosure 3 to the commission's letter of May 7,1980 (Five Additional TMI-2 Related Requirements for Operating Reactors).

The following is' a list of safety valve ?.nd power operated relief valve (PORV) j

. actuations for Turkey Point Units 3 and 4 from July 1,1986, to June 30,1987. J Unitj July 15, 1986 PORV 455C and 456 were cycled as per 3-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory. ,

July 15, 1986 PORV 455C and 456 were cycled open as per OP 0209.1, Valve Exercising Frocedure. PORV operation was

' satisfactory.

o July 24, 1986 PORV 455C and 456 were opened to provide a 2.2 square inch vent path while the reactor coolant system was depressurized as per Technical Specification 3.15.

Both PORV's opened satisfactorily.

July 29, 1986 PORV 455C and 456 were cycled as per 3-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

August 1, 1986 PORV 456 was cycled manually to test annunciator A 4/1, PORV/ Relief Valve open. PORV operation was satisfactory.

December 27, 1986 PORV 456 cycled open due to high reactor coolant system pressure during a turbine runback. The valve opened but would not close. Yhe associated block valve MOV-3-535 was closed.

PORV 456 was cycled during Maintenance work. P00V operation was not satisfactory due to dual position indication.

December 28, 1986 PORV 455C was cycled twice for maintenance. PORV operation was satisfactory.

December 28, 1986 PORV 455C was cycled as per 3-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

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Turkey Point Units 3 and 4

-Docket Nos. 50-250, 50-251 Report of Safety and Relief Valve Challenges Page 2 ,

December 29, 1986 PORV 655C was cycled as per Appendix 8 and P of OP 0209.1, Valve Exercising Procedure. PORV operation was satisfactory.

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December 29, 1986 PORV 455C was cycled twice while collasping the steam ]

bubble in the pressurizer. PORV operation was satis-factory.

1 January 1, 1987 PORV 455C was opened to provide a 2.2 square inch vent path while the reactor coolant system was depressur-ized as per Technical Specification 3.15. PORV opera-tion was satisfactory.

January 1, 1987 PORV 455C .and 456 were cycled as per 3-0SP-041.4, {

Overpressure Mitigating System Nitrogen Backup Leak and Functional Test. Both PORV's operated satisf ac-torily.

January 1,1987 PORV 455C was cycled as per 3-0SP-041.4, Overpressure Mitigating System Nitrogen Backup Leak and Functional Test after maintenance was performed to repair leaks found during previous test. PORV operation was satis-factory. s January 1,1987 PORV 455C and 456 were cycled as per 3-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

January 1,1987 PORV 456 was cycled open as per OP 0290.1, Valve Exercising Procedure. PORV operation was satisfactory.

January 2, 1987 PORV 456 was cycled as per Temporary Procedure (TP) 292, PCV-3-456 Closure Test. PORV operation was satisfactory. This test was done as part of past maintenance testing done for PORV 456 after it did not close on December 27, 1986.

January 4,1987 PORV was cycled open while Unit 3 was in hot standby as final operability check after maintenance completed work on the valve when it did not close on December 27, 1986. PORV operation was satisfactory.

March 7,1987 PORV 455C wa's cycled open due to a misaligned adjust-ment pot which indicated a full pressurizer spray demand. PORV operation was satisfactory.

March 8, 1987 PORV 455C and 456 were cycleo as per 3-0P-041.4, Overpressure Mitigating Sys :em. P0RV operation was satisfactory. 4 I

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' Report of Safety and Relief Valve Challenges Page 3

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March 9, 1987 PORV 456 cycled open during the performance of 3-OSP-

. - 050.8, RHR MOV's' 750, 751, 862 and 863 Interlock Test. The test simulated a high reactor coolant >

system pressure and provided no precautions to prevent ;

PORV operation when aligned in the low pressure mode of the Overpressure Mitigating System. Procedure change was completed to correct problem. PORV opera-tion was satisfactory.

March 16, 1987 PORV 455C and 456 were cycled as per OP 0209.1, Valve Exercising procedure. PORV operation was satisfactory, i

June 22, 1987 PORV 455C and 456 were cycled as per 3-0SP-041.4,. '

Overpressure Mitigating System Nitrogen Backup Leak i and Functional Test. PORV operation was satisfactory..

June 22, 1987 . PORV 455C and 456 were cycled as per 3-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

June 23, 1987 PORV 456 cycled open and did not close when power supply breaker tripped. Maintenance reset power ;

supply and PORV 456 closed. PORV operation was- satis- !

factory.

June 24, 1987 PORV 455C and 456 were cycled open as per 3-0P-041.4, # ,

Overpressure Mitigating System, .upon completion of !

repairs to TE-423A. PORV operation was satisfactory.

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-Report of Safety and Relief Valve Challenges Page 4 Unit 4 l July 5, 1986 PORV 455C and 456 were cycled as per 4-0P-041.4, Overpressure Mitigating System. PORV operation was satisf actory.

1 July 7, 1986 PORV 455C and 456 were cycled open as per 4-0SP-041.4, L Overpressure Mitigeting System Nitrogen Backup- Leak .)

and Functional. Test. ;

July 20, 1986' PORV 455" was cycled open to reduce reactor coolant system pressure. PORV operation was satisfactory.

July 27, 1986 PORV 455C and -456 were cycled oren to provide a 2.2 l square inch vent path while the reactor coolant system ,

was depressurized as per Technical Specification 4 l 3.15. PORV operation was satisfactory.

o August 1, 1986 PORV 455C and 456 were cycled open as per 4-0SP-041.4, Overpressure Mitigating System Nitrogen Backup Leak and Functional Test. PORV operation was satisf actory.

August 1,1986 PORV 456 cycled open while hanging in-plant equipment l clearance 4-86-8-002 for safeguards rack OR80. PORV i operation was satisfactory.

August 7, 1986 PORV 455C cycled open due to a high pressure condition in the reactor coolaat system. PORV operation was satisfactory.

August 12, 1986 PORV 455C and 456 cycled open as per OP 1300.2, l Operational Test of MOV *-535, 536 and PORV *-455C ,

456. PORV operation was satisfactory.

August 14, 1986 PORV 455C and 456 were cycled open as per 4-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

' August 15, 1986 POPV 455C an2 456 were cycled open as per 4-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

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i Turkey Point Units 3 and 4 Docket Hos. 50-250,.50-251 Report of Safety and Relief Valve Challenges Page 5 October 25, 1986 PORV 455C and 456 were cycled open as per 4-0P-041.4, Overpressure Mitigating System. PORV 456 operated satisfactory. PORV 455C was declared out of service due to dual position indication. PORV 455C was repaired and retested as per 4-0P.-041.4. PORV 455C operated satisfactory.

March 11, 1987 PORV 455C and 456 were cycled open as per OP 0209.1, Valve Exercising Procedure. PORV' operation was satis factory.

March 13, 1967 PORV 455C and 456 were cycled open as per 4-0P-041.4, Overpressure Mitigating System. PORV operation was satisfactory.

May 12, 1987 PORV 456 was cycled as . per 4-0P-041.4, Overpressure Mitigating System and 4-0SP-041.4, Overpressure Miti-gating System Nitrogen Backup Leak and Functional Test. PORV operation was satisfactory.

o May 13, 1987 PORV 455C was cycled as per 4-0P-041.4, Overpressure 4 Mitigating System and 4-0SP-041.4, Overpressure Miti-gating System Nitrogen Backup Leak and Functional Test. PORV operation was satisfactory.

May 22, 1987 PORV 455C cycled open due to an overpressure condition that resulted from PVC-4-145 not responding quickly to l a reactor coolant system pressure spike. PORV oper-l ation was satisfactory.

June 10, 1987 PORV 455C and 456 were cycled as per 4-0P-041.4, ,

Overpressure Mitigating System. PORV operation was satisfactory.

June 17, 1987 PORV 455C and 456 were cycled as per 4-0P-041.4, Overpressure Mitigating System. PORV 455C operated satisfactory. PORV 456 did not obtain proper indica-tion. Maintenance was performed on the valve and it was retested satisfactory.

r. .

n . .

I. )

i

(

APPENDIX B l

i ABSTRACT l

An inservice Multi Frequency Eddy Current Examination of Steam Generators

3A, #3B and #3C at the Turkey Point Nuclear Plant was performed during the period of June 8 through June 13, 1987.- The examination was conducted by the -

Juno Nuclear Staff - Material Codes & inspection group and supplemented by Florida Power & Light Compmy certified Eddy Current personnel and complimented .by contractor personnel. The examination program consisted of multi-frequency testing for detection of tube wall anomalies.

The data from the examiantion was recorded on magnetic tapes and also on computer diskettes. The recordings were evaluated by data analysts and the results recorded on computer print outs. An independent review of the data was also performed during the examination program.

The examinations were full length inspection of tubes, from the inlet' side of each of the three Steam Generators.

A tctal of 1029 tubes (324 in #3A,332 in #3B and 373 in #3C) were inspected. A

~

total of three tubes were plugged. Two were non-full length tubes and were plugged electively and the third had an imperfection d.pth in excess of the 40%

Technical Specification limit.

i i

l 1

l SDFl/062/l u_ _ _ --. -_ ;

l .

1 l.

4 L FORM NIS-99 OWNER $' DATA REPORT FOR E00Y CURRENT EIAMINAi!0N RESULTS l I As required by the provisions of the ASME Code Rules page 1 of 7 i

i . .. ____. __ .. __ .. .... .. .....___.... .. .. .. .. .......____...

t l 1

SUMMARY

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! EIAMINAi!0N DATES: JUNE 8,'1987 THRU JUNE 13, 1987 l 1 i l STEAM i TOTAL ! TOTAL 1 T0iAL i TOTAL i TOTAL l

? GENERATOR ! TUBES i IN0!CAil0NS i INDICAT!DNS : TU8ES PLU66ED AS .! TUBES i i NUM9ER ! , INSPECTED l ) OR TO 201 TO 391 i ) OR TO 401 TO 100% ! PREVENTIVE MAINT ! PLUB6ED l l 1 l l l 1 :

1 3E2104 1 324 : (16) 2 1 0 l 0 1 0 t l.

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! l 1 CERTIFICATION OF RECORD !

l l l We certify that the statements in this record are corrtet and the tubes inspected were tested full length l !

! in accordance with the requirements of Sectfod 11 of t$v ASME Code. I I l 1 FLORIDAPONERandLl6HTCOMPANY j i i l l (Organization) !

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L , P. O, BOX 14CM, JuMO BEACH, FL 33408-0420

h AUGUST 3 1 1987 L-87-365 U. S. Nuclear Regulatory Commission Atin: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: Turkey Point Units 3 and 4 ,

Docket Nos. 50-250 and 50-25l 10 CFR 50.59 Report Florida Power & Light's Report on " Changes, Tests and Experiments Made Without Prior Commission Approval" for the period July 1,1986 through June 30,1987 is

- attached.

Very truly yours, t ,

1 C 0. Woody roup Vice President Nucle'or Energy COW /SDF/gp Attachmen t cc: Dr. J. Nelson Grace, Regional Administrator, Region 11, USNRC Senior Resident inspector, USNRC, Turkey Point Plant l

TG47 an FPL Group company n.

'SDFl/061/1

L-87-365, Changes,Tests & Experiments Made W/O Prior Commission Approval,For Jul 1986 - June 1987

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