L-89-316, Changes,Tests & Experiments Made as Allowed by 10CFR50.59 for Period of Jul 1988 - June 1989

Changes,Tests & Experiments Made as Allowed by 10CFR50.59 for Period of Jul 1988 - June 1989

ML20246L741
Person / Time
Site:	Turkey Point
Issue date:	06/30/1989
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-89-316, NUDOCS 8909070034
Download: ML20246L741 (184)

Text

. . _ - . _ - - .

i e n- -.

., - {

P.O. Box 14000, Juno Beach, FL 33408-0420 )

f- @ FPL I

l E -

4 l

L: SEPTEMBER .i.1989- -j L-89-316~ -l 10 CFR 50.59

.l l

i o

l U. S. : Nuclear Regulatory Commission L- Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: Turkey Point Unit 3 and 4 -l l Docket Nos. 50-250 and 50-251

~I 10 CFR 50.59 ReDort Florida Power & . Light Company's Report on " Changes, Tests and )

Experiments Made Without Prior Commission Approval" for the period j '

July 1, 1988 through June 30, 1989 is attached.

Very truly.yours, j C. O. Woody  ;

Acting Senior Vice President - Nuclear I COW /TCG/gp l Attachment j cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC ]

Senior Resident Inspector, USNRC, Turkey Point Plant '

)

l 1

l 1

til l i

gg*1888A$$8l$go

1. R  !

f an FPL Group company

e<

4 b

l,

i. i 16 TURKEY POINT PLANT UNITS 3'AND 4 DOCKET NUMBERS 50-250 AND 50-251 CHANGES, TESTS, AND EXPERIMENTS ,

MADE AS ALLOWED BY 10CFR50.59 FOR THE' PERIOD OF JULY 1, 1988 THROUGH JUNE 30, 1989 1

f;7  ;

L.

's ~!

j m j h

i i

INTRODUCTION-This report is submitted in.accordance with 10CFR50.59(b), which-requires:that:

i) changes in the facility as described in the SAR ii) changes in procedures as described in.the SAR, and lii) tests and experiments not described in the SAR which are conducted without prior Commission approval be reported to the Commission at least-annually. This report is intended to meet this requirement for the period of July 1, 1988 through June 30, 1989.

This report is divided into five (5) sections: the first, changes to the facility as described in the SAR performed by a Plant Change / Modification (PC/M) ; the second, changes to the facility or '.

procedures as described in the SAR not performed by a PC/M and tests and experiments not described in the SAR;' the third, a summary of the Unit 4 cycle 12 reload evaluation; the fourth, a list of. Power Operated Relief Valve (PORV) actuations, which is submitted in accordance with FPL's commitment to comply with the requirements of Item IIK.3.3 of NUREG 0737; the fifth, a summary of_the findings of the Unit 4 Steam Generator. tube inspection.

2

4 l

o >

Table of Contents SECTION 1 . . . . . . . . . . .. ... .- . . . . . .. . . . . . 12 l s83-139 . . . . . . . . . .o. . . . . . .. . . . . . . .. . 13 6/30/89; HALON SUPPRESSION FOR APPENDIX R MODIFICATIONS83-146 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 14 5/11/89; PENETRATION SEALS FOR APPENDIX R MODIFICATIONS83-148' . . . . . . . . . . . . . . .. . . .. . . . . . . . . 15 5/10/89; RACEWAY. PROTECTION FOR APPENDIX R MODIFICATIONS

.83-149 . . . . . . . . . . . . . . . .. . . . . . . . . . . 16 5/10/89; RACEWAY PROTECTION FOR APPENDIX R MODIFICATIONS83-150 . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 5/10/89; RACEWAY PROTECTION FOR APPENDIX R MODIFICATIONS83-154 . . . . .. .- . . . . . . . . . . . . . . . . . . . . 18 '.

5/5/89; ALTERNATE SHUTDOWN CAPABILITY - FOR APPENDIX R-MODIFICATIONS83-155 . . . . . . . . . . . . .. . . . . . . . . . . . . . . 20

.5/26/89; ALTERNATE SHUTDOWN CAPABILITY FOR APPENDIX R MODIFICATIONS83-200 . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3/25/89; INSTALLATION OF REACTOR EX-CORE . NEUTRON FLUX MONITORING SYSTEM 84-017 . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5/4/89; RHR VALVES CIRCUIT MODIFICATION UNIT 4 84-026 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3/1/89; CONTROL ROOM HABITABILITY HVAC MODIFICATION 84-070 . . . . . . . . . . . . . . . . . . . . . . .. . . . 25 1/28/89; POST ACCIDENT SAMPLING SYSTEM, LONG TERM MODIFICATIONS85-015 . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 5/5/89; DEDICATED COMMUNICATION SYSTEM FOR ALTERNATE SHUTDOWN ,85-074 . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 l

3/17/89; COMPUTER ROOM FIRE DETECTION E 85-081 . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 5/26/89; HYDROGEN LINE MODIFICATION 3

l l

k 85-122 . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 5/5/89; SAFE SHUTDOWN MODIFICATIONS - APPENDIX R 85-123 . . . .. . . . . . . . . . . . . . . . . . . . . . . . 32 5/5/89; SAFE SHUTDOWN MODIFICATIONS FOR APPENDIX R 85-142 . . .. . . . . . . . . . . . . . . . . . . . . . ' . . 34 10/11/88; FUEL TRANSFER SYSTEM MANIPULATOR CRANE DUAL CABLE MODIFICATIONS85-148 . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 10/1/88; SPENT FUEL POOL COOLING SYSTEM-SEISMIC UPGRADE 85-194 . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 12/9/88; BREATHING AIR COMPRESSOR TO FILL SCBA TANKS86-033 . . . . . . . . . . . . . . . . . . . . . . . . . . 38 4/22/89; SGWL CONTAINMENT ISOLATION VALVE REPLACEMENT 86-059 . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 ,

5/5/89; SEALING OF MANHOLIS FOR CABLE FIRE PROTECTION 86-130 . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 4/30/89; RCP OIL COLLECTION SYSTEM MODIFICATION / ANALYSIS86-182 . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 3/27/89; SI PUMP MINIMUM FLOW RECIRCULATION VALVE ACTUATION REPLACEMENT 86-185 . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 3/22/89; ANNUNCIATION IN MAIN CONTROL ROOM ON LOSS OF EDG CONTROL POWER 86-186 . . . . . . . . . . . . . . . . . . .. . . . . . . . 45 5/16/89; ADDITION OF STRAINERS AND DRAIN TRAPS TO INSTRUMENT AIR PIPING 86-191 . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 1/26/89; REPLACEMENT OF CONTROL SWITCHES CV-3-2913 AND CV-3-3722 86-195 . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 8/23/88; ADDITION OF CONTINUOUS TUBE CLEANING CAPABILITY TO THE CCW HEAT EXCHANGERS86-208 . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 10/1/88; IST GAUGE INSTALLATION FOR THE SPENT FUEL PIT COOLING PUMPS86-213 . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 5/15/89; EMERGENCY LIGHTING FOR APPENDIX R 4

f?

1 l

86-226 -

51 11/08/88; TURKEY POINT UNIT 4 ADDITION.0F NON-SAFETY l RELATED:MCC 4H (4B21)87-026'. . . - . .. . . . . - . . . . . . . ... . . .. .. .;. . . . .. 53-5/6/89; COMPONENT COOLING WATER SYSTEM NORMAL CONTAINMENT COOLER TUBE BUNDLE REPLACEMENT.87-034 . .- . . - . . .. . . . .. . . . . . . . . . . . . . . . . 54

, 5/15/89; DELETION OF CRDM COOLER FANS AUTO EDG LOADING 87-100 ' . - . . . . . . . . . . .. . . . . . .. . . ' . . . ..-

55 5/15/89; REACTOR CAVITY SEAL REPLACEMENT 87-212 57 2/10/89;.EDG ENHANCEMENT - SITE PREPARATION 87-310 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 - ~

6/23/89; EMERGENCY RESPONSE DATA ACQUISITION & DISPLAY' SYSTEM UPGRADE 87-317 . - . . . . . . . . . . . . . . . . . . . . . . . . . - . - . 59 11/08/88; PERMANENT INSTALLATION OF WHOLE BODY COUNTER 87-320 . . . .. . .- . . . . . . . . . . . . . . . . . . .. 60 7/21/88; CCW HEAT EXCHANGER CHANNEL HEAD REPLACEMENT 88-069 . . - . . . . . . . . . . . . . . . . . . . . . . . . . . 61 5/4/89; AFW AND MAIN STEAM SYSTEM TRAP AND DRAIN DRAWING UPDATE 88-078 . . .. . . .. . . . . . . . . . . . . .. . . . . . . . 62 2/11/89; RESIDUAL HEAT REMOVAL PUMPS MECHANICAL SEALS AND SEALS COOLER REPLACEMENT 88-131 . . . . . . . . . . . . . . . . . . . . . . .. . . . . 63 8/11/88; CONTAINMENT SPRAY PUMP PULL OUT ASSEMBLY REPLACEMENT 88-160 . . .. . . . . . . . . . . . . . . . . . . . . . .. . 64 7/2/88; RHR/ SAMPLING VALVE UPGRADE 88-172 . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 10/15/88; CONTAINMENT PURGE VALVE ACTUATOR VENTING MODIFICATION 88-193 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 4/21/89; DETECTION SYSTEM FOR MONITORING RCS LEAKS IN THE REACTOR HEAD AREA 5

l l

l-88-197 . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 4/21/89; INADVERTENT AC7"TATION OF PRMS RELAYS R-11 AND R-12 88-212 . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 4/26/89; PRESSURIZER LEVEL INDICATOR DRAWING DISCREPANCY, I LI-459B i 88-245 . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 4/21/89; MSIV AIR ACCUMULATOR SYSTEM 88-263 . . . . . . . . . . . . .. . . . . . . . . . . . . . 72 1/25/89; COMPONENT COOLING WATER HEAT EXCHANGER 88-282 . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 4/30/89; REPLACEMENT OF SGBD FLOW INSTRUMENTATION ' AND CHANGE OF SAFETY CLASSIFICATION Q-BASIS88-287 . . .. . . . . . . . . . . . . . . . . . . . . . . . 75 '

5/22/89; AFW FCV STEM TRAVEL LIMIT STOPS88-290 . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 9/27/88; CHANGING OF METEOROLOGICAL TOWER DATA FROM INSTANTANEOUS TO 15 MINUTE AVERAGED DATA 88-310 . . . . . . . . . . . . . . . . . . . . . . . . . . . 77 8/27/89; REMOVAL OF PAHM-3-008B AND PAHM-3-009 FROM THE SYSTEM 88-311 . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 8/27/89; REMOVAL OF PAHM-4-008B FROM THE SYSTEM 88-336 . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 3/1/89; REPLACEMENT OF PCV-4-118 AND PCV-4-119 ON THE HYDROGEN AND NITROGEN LINES TO THE CVCS VOLUME CONTROL TANK (VCT)88-359 . . . . . . . . . . . . . . . . . . . . . . . . . . . 80 4/24/89; SI ACCUMULATOR NITROGEN PRESSURE REGULATOR PCV-4-846 REPLACEMENT 88-404 . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 2/27/89; DEMINERALIZED CONNECTIONS FOR THE COMPONENT COOLING WATER SYSTEM 88-427 . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 12/10/88; PRESSURIZER PORV AIR SUPPLY TUBING ENHANCEMENT 88-442 . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 2/2/89; DEMINERALIZED CONNECTIONS FOR THE COMPONENT COOLING WATER SYSTEM 6

88-448 . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 1/25/89; REPLACEMENT OF TUBING, FITTINGS AND ISOLATION VALVES ASSOCIATED WITH ICW PRESSURE INDICATORS88-453 . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 12/7/88; UNIT 3 DRAWING DISCREPANCIES88-521 . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 4/30/89; UNIT 3 AND 4 DRAWING UPDATE FOR SYSTEM 023 EMERGENCY DIESEL GENERATOR 88-527 . . . . . . . . . . . . . . . . . . . . . . . . . . . 87 4/27/89; RESOLUTION OF DRAWING CHANGES ASSOCIATED WITH 5610-T-E-4503 88-528 . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 5/1/89; DRAWING UPDATE 5610-T-E-4065, SHEET 1 AND SHEET 2, LUBE WATER AND CIRCULATING WATER SYSTEMS88-530 . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 4/27/89; BREAKER LIST UPDATING FOR NON-CONFORMANCE REPORTS (NCRs) AND REQUEST FOR ENGINEERING ASSISTANCE (REAs)88-533 . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 4/28/89; UNIT 3 AND 4 DRAWING DISCREPANCIES ON DRAWING 5610-T-E-4510, SHEETS 1 & 2 88-534 . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 4/28/89; DRAWING DISCREPANCIES ON 5610-T-E-4534 SHEETS 1 AND 2 - CONTAINMENT VENTILATION SYSTEM 88-535 . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 3/2/89; PRESSURIZER PORV AIR AND NITROGEN SUPPLY TUBING ENHANCEMENT 88-536 . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 4/1/89; SAFETY INJECTION ACCUMULATOR LEVEL TRANSMITTER REPLACEMENT 88-584 . . . . . . . . . . . . . . . . . . . . . . . . . . 94 4/29/89; DRAWING REVISION FOR LIQUID WASL., DISPOSAL SYSTEMS OPERATING DIAGRAM 88-586 . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 5/3/89; CONTROL ROOM HVAC T-E DRAWING UPDATE 88-597 . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 4/27/89; DRAWING DISCREPANCIES ON DW6 5610-T-E-4063, SHEET 1 7

L _ _. _ __ _. _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - . _ _ _ _ _ . _ _ _ . _ - _ _ . _ _ _ _ _ _ _ _

-_ 7 _

. .a ,88-606' ... . . . . . . . . .. . . . . . . . . . . . . . - . ... '97

~5/1/89; DRAWING REVISION'FOR' CONDENSATE SYSTEM (5610-T-o >

_E-4062 SH'1)89-029 . . .: .< . . . . . . . . . . . .. . . . . . . . . . ... .

98 5/1/89; SPENT FUEL PIT.. (SFP) LEAK RECOVERY SYSTEM AS-

' BUILT VALVE TAGS I89-09L .. .. . . . . . .. . . . . . . . .. . . . - . . . . . . . 99 4/27/89; PRIMARY WATER, AND DEMINERALIZED WATER SYSTEM DRAWING UPDATE, 5610-T-E-4531 89-100. 100 5/1/89; MAIN STEAM DRAWING UPDATE OF DRAWING 5610-T-E--

4061, SHEET 1 OF 4-89-105 . . . . . . . . . . . . . . . . . . . . . . . . . . . 101 5/17/89; DRAWING UPDATE FOR . THE - CHEMICAL AND VOLUME -

CONTROL' SYSTEM 89-127' . . . .. . . . . . . . . . . . . . . . . . . . . . 102 4/25/89; UNIT 4 RESIDUAL HEAT REMOVAL (RHR) SYSTEM NO.

50' AND SAFETY INJECTION (SI) SYSTEM NO.' 62 DRAWING UPDATE FOR DRAWING 5610-T-E-4510, SHEET 1 AND 2

~89-139 .. . .. . . . . . . . . - . . . . . . . . . . . . . . 103 4/30/89; DRAWING DISCREPANCIES ON DRAWING 5610-T-E-4512, .

SHEETS 1 AND 2, " COMPONENT COOLING WATER SYSTEM"89-264 . . . . . . . . . . . . . . . . . . . . . . . . . . 104 4/12/89; REPLACEMENT OF YARWAY CONTAINMENT ISOLATION STOP VALVES89-373 . .. ... . . . . . . . . . . . . . . . . - - . . . . . . . 105 6/5/89; BONNET EQUALIZING LINES FOR MOV-4-750 AND MOV 751 89-396 . . . . . . . . . . . . . . . . . . . . . . . . . . 106 6/16/89; RCS RTD MANIFOLD FLOWRATE CHANGE SECTION 2 . . . . . . . . . . . . . . . . . . . . . . . . . 107 JPE-LR-87-038 Revision 0 . . . . . . . . . . . . . . . . . 108 INTERCONNECTING 125 VDC BUSSES 3B AND 4A WHILE BATTERY 3B(4A) IS REMOVED FOR TESTING AND THE UNITS ARE AT FULL POWER.

JPES-E-87-2484 Revision 0 . . . . . . . . . . . . . . . . . 109 THE INDIVIDUAL CELL EQUALIZING CHARGE AND ELECTROLYTE LEVEL CORRECTION 3A, 3B, 4A, AND 4B 8

JPE-PTN-SELJ-88-030 Revision 1 . .. . . . . .. . . .. .- 111

' ' ' BATTERIES c. 3 A, ) 3 B , 4A, AND 4B MINIMUM TERMINAL VOLTAGE l HIGHER THAN 105VDC JPE-PTN-SELS-88-031 Revision 0 ... . . . . . . . . .. . .- 113' REMOVAL ' OF AIRBORNE CONTAMINANTS USING THE EMERGENCY-CONTAINMENT FILTERS ~

.JPE-PTN-SEMS-88-039 Revision 1 . . . . . . .. . . . . . . 114 TEMPORARY SYSTEM ALTERATION TO FILTER FUEL OIL IN DIESEL l FUEL OIL STORAGE TANK JPE-PTN-SEMJ-88-040 Revision 0 . . . . . . . . . .. .. . 115

'LPT ROTOR PRE-INSTALLATION PERFORMANCE EVALUATION JPE-PTN-SEMS-88-041 Revision 1 .. .. . . . . . . . . . . . 116

. CHEMICAL CLEANING OF - THE TUBE ' SIDE OF THE COMPONENT COOLING WATER HEAT EXCHANGER .

JPE-PTN-SEEJ-88-042 Revision 0 . . . . . . . . . . . . . .. 117 DE-ENERGIZATION OF UNIT 4 4160 VOLT SAFETY RELATED BUSES JPE-PTN-SEEJ-88-047 Revision 0 . . . . - . . . . . . . . . . . 119

- BACKFEEO OF POWER THROUGH MAIN AND AUXILIARY TRANSFORMERS

JPN-PTN-SENJ-88-052 Revision 3 . . . . . . . . . . . . . . . 120-CONTAINMENT' BULK AMBIENT TEMPERATURES JPN-PTN-SELS-88-053 Revision 0 . . . . . . . . . . , . . . . 121 SPENT FUEL POOL' LEVEL REDUCTION JPN-PTN-SECS-88-060 Revision 0 -. . . . . . . . . . . . . . . 122 TEMPORARY . REMOVAL OF STEAM GENERATOR THRUST BEAM AND FLOOR GRATING JPN-PTN-SEMJ-88-067 Revision-0 . . -. . . . . . . . . . . . . 123 EMERGENCY DIESEL GENERATOR FIVE STARTING ATTEMPT TEST JPN-PTN-SELS-88-070 Revision 0 . . . . . . . . . . . . . . . 124 REACTOR CAVITY FILTRATION SYSTEM OPERATION JPN-PTN-SEMS-88-081 Revision 0 . . . . . . . . . . . . . . . 125 FREEZE ' SEAL SAFETY EVALUATION FOR REPAIR OF VALVE 3-312A AND 3-312B JPN-PTN-SEMJ-88-087 Revision 0 .. . . . . . . . . . . . . . 126 THE DELETION OF BACKFLOW DAMPERS IN THE CONTROL ROOM VENTILATION SYSTEM JPN-PTN-SEEJ-88-088 Revision 0 . . . . . . . . . . . . . . . 127 125 V DC CONTROL POWER FOR LOAD CENTER 3B41 FROM DC PANEL 3D01 9

JPN-PTN-SEES-89-002 Revision 0 . . . . . . . . . . . . . . . 128 UNITS 3 & 4 TEMPORARY SYSTEM ALTERATION FOR DISABLING OF LOCAL INDICATION FOR PRMS R-18 JPN-PTN-SEMS-89-008 Revision 0 . . . . . . . . . . . . . . . 129 TEMPORARY SYSTEM ALTERATION FOR CLEANING OF THE TURBINE LUBE OIL JPN-PTN-SEMS-89-010 Revision 0 . . . . . . . . . . . . . . . 130 DELETION OF BACKSEATING OF SAFETY INJECTION / CONTAINMENT SPRAY VALVES INSIDE CONTAINMENT JPN-PTN-SENJ-89-016 Revision 0 . . . . . . . . . . . . . . . 131 EMERGENCY DIESEL GENERATOR LOADING EVALUATION UPDATE JPN-PTN-SEMJ-89-023 Revision 0 . . . . . . . . . . . . . . . 132 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C JPN-PTN-SEMJ-89-031 Revision 0 . . . . . . . . . . . . . . . 134 TOP-514 VARIABLE BACK PRESSURE PERFORMANCE TEST JPN-PTN-SEMJ-89-035 Revision 0 . . . . . . . . . . . . . . . 135 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C JPN-PTN-SEMJ-89-036 Revision 0 . . . . . . . . . . . . . . . 137 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C JPN-PTN-SEMJ-89-038 Revision 0 . . . . . . . . . . . . . . . 139 RCP OIL COLLECTION SYSTEM JPN-PTN-SEMJ-89-041 Revision 0 . . . . . . . . . . . . . . . 140 RCP OIL COLLECTION SYSTEM JPN-PTN-SEMJ-89-043 Revision 0 . . . . . . . . . . . . . . . 142 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C JPN-PTN-SEMJ-89-044 Revision 0 . . . . . . . . . . . . . . . 143 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C JPN-PTN-SEMJ-89-045 Revision 0 . . . . . . . . . . . . . . . 145 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C JPN-PTN-SEMJ-89-047 Revision 0 . . . . . . . . . . . . . . . 147 COORDINATED LITHIUM / BORON CORRELATION FOR BORON CONCENTRnIIONS BETWEEN 1200 AND 2000 PPM l

10 l

l 1

JPN-PTN-SEMJ-89-055 Revision 0 . . . . . . . . . . . . . . . 148 CONTROLLED GAS DECAY TANK RELEASES USING ALTERNATE RADIATION MONITORS JPN-PTN-SEMS-89-059 Revision 0 . . . . . . . . . . . . . . . 149 EVALUATION FOR MOV-4-869 MATERIAL CHANGE JPN-PTN-SEMS-89-060 Revision 0 . . . . . . . . . . . . . . . 150 EVALUATION FOR MOV-3-869 MATERIAL CHANGE JPN-PTN-SEIJ-89-061 Revision 0 . . . . . . . . . . . . . . . 151 TSA INSTRUMENTATION LOOP MODIFICATIONS JPN-PTN-SENJ-89-064 Revision 0 . . . . . . . . . . . . . . . 153 INCREASED PRESSURIZER PRESSURE UNCERTAINTY JPN-PTN-SEMJ-89-067 Revision 0 . . . . . . . . . .. . . . . 154 CHANGE TO THE ADMINISTRATIVE TEMPERATURE LIMI'1S ON HEATUP AND COOLDOWN RATES TO CORRECT AN FSAR ERROR ".

SECL 88-530 . . . . . . . . . . . . . . . . . . . . . . . . . 155 ULTRASONIC INSPECTION OF NUCLEAR FUEL SECL-88-647 Revision 1 . . . . . . . . . . . . . . . . . . . 156 ALTERNATE BORATION PATH DURING THE REPAIR OF CVCS VALVE 3-268 4-EOP-ES-1.3 . . . . . . . . . . . . . . . . . . . . . . . . 157 TRANSFER TO COLD LEG RECIRCULATION TP-523 . . . . . . . . . . . . . . . . . . . . . . . . . . . 158 EMERGENCY DIESEL GENERATOR FUEL OIL DUPLEX FILTER FOULING AND CONFIGURATION TEST SECTION 3 . . . . . . . . . . . . . . . . . . . . . . . . . . 159 UNIT 4 CYCLE 12 CORE LOAD . . . . . . . . . . . . . . . . . . 160  ;

SECTION 4 . . . . . . . . . . . . . . . . . . . . . . . . . . 162 3 Unit 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . 163 Unit 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . 166 SECTION 5 . . . . . . . . . . . . . . . . . . . . . . . . . . 168 i

4 I

l 11 j l

_______________-_____-_-__a

SECTION 1-Changes to the facility as described in the SAR performed by a PC/M-12

y ,4g .m. r-- -

7 - - - - - -

4 o

2 PLhMT CHANGE / MODIFICATION 83-139- 1 ,

H 1

PC/M. CLASSIFICATION: QR UNIT: 3 & 4- ,

TURNED OVER.DATE: 5/30/89 j HALON SUPPRESSION FOR APPENDIE R 50 DEIFICATIONS Summaryt .

This., modification provides. for the installation of halon... j suppression systems and associated work in the Cable Spreading Room l (Fire -Zone - 98) , including the Cable Chase (Fire Zone - 132)', at  !

j Elevation-30'-O and_the. Invertor Rooms (Fire Zones 108A and 108B) atf ElevationJ 42 '-0. The'halon suppression systems provided are -!

a u t o m a t i c ,..L t o t a l f l o o d i n g t y p e . The systems are - designed to j provide a: concentration of 6.0 to 6.5 percent, by volume, within 10 seconds of actuation,Eand to maintain that concentration for a l

minimum of 30 minutes. .;

I l

safety-Evaluation: 1 The halon suppression system piping is installed to Seismic II/I requirements. These modifications are not inside containment, and '!

do not' affect radioactive-waste; treatment systems, do not involve j safety.related snubbers. The actuation of.the halon system could i not create a malfunction in equipment. The discharge nozzles are I placed such that there is no direct impingement of the halon against. the sensitive electrical equipment that could cause

' disruption in service.. The location. of the nozzles will also

, prevent the:halon.from significantly lowering the temperature of L electrical equipment and prevent the equipment from being exposed to turbulent discharge'from the nozzle. These modifications will not adversely affect the operation or availability of safety.

.related equipment.

i I

13 I

I

PLANT CHANGE / MODIFICATION 83-146 l

PC/M CLASSIFICATION: NNSR l UNIT: 3 & 4 TURNED OVER DATE: 5/11/89

- i PENETRATION SEALS FOR APPENDIX R MODIFICATIONS Summary:

The scope of this modification covers the installation of fire stop -

penetration seals to maintain the integrity of identified fire '

barriers necessary to meet the requirements of Section III.G of Appendix R to 10CRF50. The scope of work includes installation of three-hour fire rated penetration seals (or seals with lesser rating as specified on design drawings) and other related work in floors, walls, and ceilings which have been identified as fire barriers.

Safety Evaluations These fire stop penetration seals are non-nuclear safety related, arti n:Et inside containment, do not involve safety related snubbers, do not affect spent fuel cooling operations, but do enhance the separation of equipment required for safe shutdown. The structural integrity of affected block walls will not be impacted by these modifications. These modifications do not adversely alter the function or arrangement of safety related features of the plant, including the HVAC's performance.

14

PIANT CHANGE / MODIFICATION 83-148 PC/M CLASSIFICATION: NNSR UNIT: 3 TURNED OVER'DATE: 5/10/89 RACEWAY PROTECTION FOR APPENDIX R MODIFICATIONS guaynary:

This package provided for fireproofing of raceways carrying safe

' shutdown cable belonging to one channel when cables of redundant channels are routed in the~ same fire zone. Material used for.

fireproofing is approved 'for such application and has the required -

rating. ~ Method of application is consistent with supplier's recommendations.

Safety Evaluation:

The raceway protection installation .is non-nuclear safety related, does not involve. safety related snubbers'or blockwalls, does not affect spent. fuel cooling operations, but does provide and maintain the fire protection integrity needed for the separation of redundant safenhutdown equipment. The fireproofing material has been approved ; for _ installation in nuclear plant facilities by American Nuclear . Insurers and is installed in a nunder of plants accepted for operational licensing by the . Nuclear Regulatory Commission.

The addition of the fire protection material has been decigned and will - be installed to withstand a Design Basis Safe Shutdown Earthquake.

An evaluation of all the power cables enclosed with fire protection was completed and all the protected power cables were within the allowable ampacity including the TSI derating for 3-hour protection for both the protected cables and their load.

15

PLANT CHANGE / MODIFICATION 83-149 PC/M CLASSIFICATION: . NNSR UNIT: 4 TURNED OVER DATE: 5/10/89 RACEWAY PROTTCTION FOR APPENDIX R MODIFICATIONS Summary:

This package provided for the fireproofing of raceways carrying ,

safe shutdown cables. Material used for fireproofing is approved for such application and has the required rating. Method of application is consistent with supplier's recommendations.

Safety Evaluation:

The raceway protection installation is non-nuclear safety related, does not involve safety related snubbers or blockwalls, does not affect spent fuel cooling operations, but does provide and maintain the fire protection integrity needed for the separation of redundant safe shutdown equipment. The fireproofing material has been approved for installation in nuclear plant facilities by American Nuclear Insurers and is installed in a number of plants approved for operating licenses by the Nuclear Regulatory Commission.

The addition of the fire protection to the electrical raceways and/or supports will result in an increase in the loading of the supports. This additional load has been approved to be added to the existing supports, and where required, additional supports will be added. The supports will be protected, as required, in areas where 3-hour rated protection is required.

Also, the fire protect';on material has been designed and will be installed to withstand a Design Basis Safe Shutdown Earthquake.

An evaluation of all the power cables enclosed with fire protection was completed and all the protected power cables were within the allowable ampacity including the Thermal Science Inc. (TSI) derating for 3-hour protection.

l 16 l

l E . _ . _ _ _ _ _ _ _ _ . _. _ _ _ _ _ _ _ _ _ ._ _ _ _ _ ._______________________o

m

]

1 i

PLANT CHANGE / MODIFICATION 83-150  !

PC/M CLASSIFICATION: NNBR l UNIT: 3 & 4 TURNED OVER DATE: 5/10/89 RACEWAY PROTECTION FOR APPENDIX R MODIFICATIONS UNITS 3 &4 Summary:

This package provided for fireproofing of raceways carrying safe .

shutdown cables. Material ur,2d for fireproofing is approved for -

such application and has the required rating. Method of application is consistent with supplier's recommendations.

Safety Evaluation:

The raceway protection installation is non-nuclear safety related, does not involve safety related snubbers or blockwalls, does not affect spent fuel cooling operations, but does provide and maintain the fire protection integrity needed for the separation of redundant safe shutdown equipment. The fireproofing material has been approved for installation in nuclear plant facilities by American Nuclear Insurers and is installed in a number of plants approved for operating licenses by the Nuclear Regulatory Commission.

The addition of the fire protection to the electrical raceways and/or supports will result in an increase in the loading of the supports. This additional load has been approved to be added to the existing supports, and where required, additional supports will be added. The supports will be protected, as required, in areas where 3-hour rated protection is required.

Also, the fire protection material has been designed and will be installed to withstand a Design Basis Safe Shutdown Earthquake.

An evaluation of all the power cables enclosed with fire protection was completed and all the protected power cables were within the allowable ampacity including the Thermal Science Inc. (TSI) derating for 3-hour protection.

17

PLANT CHANGE / MODIFICATION 83-154 PC/M CLASSIFICATION: SR UNIT: 3 TURNED OVER DATE: 5/5/89 ALTERNATE SHUTDOWN CAPABILITY FOR APPENDIX R MODIFT_QATIONS Summary:

This design package provides Alternate Shutdown Capability, with adequate instrumentation and controls to bring the unit to a hot standby condition, and maintain' it, in the event a fire in the ;

Control -Room,- the Cable Spreading Room, or the North-South Breezeway necessitates evacuation of the Control Room. Additional instrumentation and controls to achieve and maintain hot standby are provided on the Alternate Shutdown Panel, C264 and supplemented by manual functions at local' stations for achieving cold shutdown.

Safety Evaluation:

Electrical channel separation is maintained for this modification.

No single transfer switch or new device added by this package is utilized for redundant safety functions. Therefore, no single failure will result in a loss of redundant safety functions. Also, isolation by mechanically actuated dry contacts and transformer coupled converters assure the alternate shutdown controls are independent of the three postulated fire areas and, 'therefore, are isolated from the effect of the postulated fire.

Isolation of pneumatic operated components by block solenoid valves i

added by this package assure the alternate shutdown controls are l i* dependent of the three postulated fire areas and, therefore, are L isolated from the effects of the postulated fire. Current to pneumatic (I/P) converters provided for atmospheric dump valve operation are non-safety grade components. However, during normal plant operation the atmospheric dump valve's block solenoid is de-y energized thereby isolating the I/P converter from the system.

These solenoid valves are nuclear qualified, seismically installed, and normally de-energized, as are all the new solenoid valves installed by this package and as such do not degrade the existing safety related system.

l l

18

n: ;

2

' Administrative . and : operational controls will be established to provide" assurance ofl.. proper-component positions for normal plant and alternate shutdown modes of operation.

.. The . design modifications which interface ' with. existing . safety L

N .related systems does'not-' change the function or~ reduce.the-design-

. basesof.the existing systems. All new. components,-cables,.and

. instrument.: tubing.are seismically installed.

19

f PLANT CHANGE / MODIFICATION 83-155 1

PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 5/26/89 {

ALTERNATE SHUTDOWN CAPABILITY FOR APPENDIX R MODIFICATIONS Summary:

This design package provides Alternate Shutdown Capability, with adequate instrumentation and~ controls to bring the unit to a-hot ',

standby : condition, and maintain it, in the . event a fire in the Control Room, the Cable Spreading Room, or the North-South Breezeway necessitates evacuation of the Control Room. Additional-instrumentation and controls to achieve and maintain-hot standby are provided on the Alternate Shutdown Panel, 4C264 and supplemented by manual functions at local stations for achieving cold shutdown.

Safety Evaluation:

Electrical channel separation is maintained for this modification.

No single transfer switch or new device added by this package is utilized for redundant safety functions. Therefore, no single failure will result in a loss of redundant safety functions.

Isolation by mechanically actuated dry contacts and transformer coupled converters assure the alternate shutdown controls are independent of the three postulated fire areas and, therefore, are isolated from the effects of the postulated fire.

Isolation of pneumatic operated components by block solenoid valves added.by this package assure the alternate shutdown controls are independent of the three postulated fire areas and, therefore, are isolated from the effects of the postulated fire. Current to pneumatic (I/P) converters provided for atmospheric dump valve operation are non-safety grade components. . However, during normal plant' operation the atmospheric dump valve's block solenoid is de-j energized thereby isolating the I/P converter from the system.

o -

These solenoid valves are nuclear qualified, seismically installed, I and normally de-energized, as are all the new solenoid valves installed by this package and as such do not degrade the existing safety related system.

l' l

20 ll

7-

.i e'

. Administrative- 'and operationalicontrols ' will:- be established to provide assuranceiof. proper-component positions ~for normal plant

~

sand alternate shutdown modes of operation.

2 The' design' modifications which ~ interface with existing safety'

.,, related systems,does.not' change'the. function of. reduce the design

+ bases 1of the existing systems. All new-components, cables, and

. instrument tubing are. seismically'.-installed.

't g s*  :

1 e

21 l.

L t- - - - - - - - - - _ _ _ - _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

r 7 i

i PLANT CHANGE / MODIFICATION 83-200 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 3/25/89 INSTALLATION OF REACTOR EX-CORE NEUTRON FLUX MONITORING SYSTEM Summaryl.

This modification provided. installation of;a new fully qualified ex-core neutron flux monitoring system which provides neutron flux .

indication in the control room and on the alternate shutdown panel. -

In a d d i t i o n , -.t h i s system provides inputs to the control room annunciator system and the containment evacuation alarm system.

Safety Evaluation:

This system does not interface with the existing NSSS supplied nuclear instrumentation. All outputs from this new system which interface with non-safety related systems are isolated. The addition of this system does not alter or affect any other safety related system.

This new system does not provide any protective or control functions and does not interface with any reactor protection circuitry.

22

Y ,  :  ;

P ,

3. x 4

. PLANT CHANGE / MODIFICATION G4-017.

E c .- .

.PC/M CLASSIFICATION: NN8R-UNIT: 4 TURNED OVER DATE:- 5/4/89.

h y RER VALVES CIRCUIT MODIFICATION ITNIT 4 Summary:

i

- This package'provided the operator with annunciation for loss of

m . letdown to the Residual. Heat Removal'(RHR) System with the reactor in,heatup'or cooldown and.the overpressure.witigation System (OMS) -

in the: " low pressure" mode. .In addition, it;provided the operator.'

with a mechanism to restore the letdown-p.ath ijdickly, provided thes

- Reactor Coolant. System (RCS) pressure does not prohibit- it, thereby

. reducing"the potentialifor challenges to the OMS.

Safety Evaluations-The modifications addressed by this package involve changes in the control? circuits of standard motor operated valves (MOVs). These change.G will essentially allow choice of operation between that of

- a "stop" valve and a modulating or " throttle" valve, : provided automatic, constraints. permit it. Existing supervisory control.of z these valves remains intact. .When the automatic circuitry requires the valve L to . close, it will' close, and the action cannot be manually overridden. Neither can the."open" pressure permissive be overridden.. The only functional ' change is the ability to reverse ' closure of the valve, provided ' there is no automatic

~

closure" signal present, without having to. wait 160 seconds'until

~

.the valve is fully closed.-

The. changes of this. package are made, following standard engineering practices and methods of application for motor operated -. valves which are . to be L used as either stop or throttle valves. This simply means that valve action is either sealed in (stop valve)-or modulated by a control switch (throttle valve) depending on system h,_

operating conditions. Under either mode of operation, the valves-can,~and always could, fail open, fail closed or fail in some-intermediate position; however, an FSAR accident analysis has

. previously been conducted on this basis. No other syr ' ems will be affected by this modification.

23

- - -__.._____.____________m____ -

I' PLANT CHANGE / MODIFICATION 84-026 PC/M CLASSIFICATION: BR.

UNIT: 3 &4 TURNED OVER DATE: 3/1/89 CONTROL ROOM HABITABILITY HVAC MODIFICATION Summary:

This package enhances the Control Room HVAC system. The major equipment added to provide redundancy of active HVAC system .

equipment are emergency ventilation supply fans, ventilation inlet -

dampers, recirculation dampers and radiation detectors.

-safety Evaluation:

The permanent modifications are designed to enhance the redundance and reliability of the Control Room HVAC System. The addition of the redundant radiation monitors in the normal air intake enhances the reliability of the Control Room HVAC System by providing a redundant independent actuation signal for the emergency mode of operation. Components of the system were specified to applicable standards and quality level, ensuring qualification for the designed application and use, Approved and qualified Quality Assurance (QA) and Quality Control (QC) Programs, as applicable to the original Control Room HVAC System / Components, will be utilized for these modifications, All permanent equipment will be seismically installed and the effects on' block walls have been reviewed and found to be acceptable. The supply fans will be added to the emergency diesel generators; however, the Auxiliary Steam Condensate Pumps ares being moved to the non-vital side of their respective MCC busses. This will have no affect on the overall Emergency Diesel Generator loading.

24

I-p PLANT CHANGE / MODIFICATION- 84-070 PC/M CLASSIFICATION: NNSR UNIT: 3&4 TURNED OVER DATE: 1/28/89

' POST ACCIDENT SAMPLING SYSTEM, LONG TERM MODIFICATIONS Summary:

This modification is to improve the. reliability and accuracy of the .

• Post- Accident Sampling' System (PASS). This modification will replace an obsolete hydrogen analyzer with one of a current design, provide'for improved sample conditions for the analyzers, provide a new sample point for'the system, and provide. increased system reliability.

Safety Evaluation:

The newisample_ point for PASS is located down stream'of HCV-142, the safety-related boundary valve for the Residual Heat Removal (RHR) System. The additional sample flow path, backwash capability, sample cooling, drain capability, sample cask and pump protection are installed to the same desigst criteria as the original PASS. The valves eliminated are unused and not required for system operation. This modification does not affect the structural integrity of previously analyzed block walls or the The modifications covered will containment heat sink analysis.

have no effect on the operating conditions of any safely related system.

25

PLANT CHANGE / MODIFICATION 85-015 PC/M CLASSIFICATION: QR UNIT: 3 & 4 TURNED OVER DATE: 5/5/89

(

DEDICATED COMMUNICATION SYSTEM FOR ALTERNATE SHUTDOWN Summary:

The scope of work provides for the installation of a combined dedicated Alternate Shutdown Communication System for Units 3 and

4. This communication system is completely independent of the -

existing plant communication system and is provided to enable tre operators to coordinate the required manual operations and to monitor the status of the plant during Alternate Shutdown conditions.

Safety Evaluation:

The design of the system minimizes any interfaces with tha safety function of vital plant equipment. Although powered from c vital source via an isolation transformer and a primary side fuse, it is electrically independent of other safety related equipment and postulated electrical failure in this system will not result in the damage or loss of any safety related equipment. In addition, all conduit and equipment required for these modifications is seismically installed where failure of this equipment.or supports due to an earthquake could result in possible damage to existing or new safety related equipment.

An isolated section of the Alternate Shutdown Communication System is provided in the Control Room to enable the operator to coordinate the reestablishment of control and monitoring functions in the Control Room. This section is isolated by use of keylock switches and administratively controlled procedures. Therefore, the isolated section to the Control Room does not impair the Alternate Shutdown Communication System or any other safety system of the plant.

No work associated with this package changes the heat sink analysis, or affects the radioactive waste treatment system, tha spent fuel pit cocling system, the effluent monitoring systems or 26

s.

N snubbers. In addition, no work associated with this package alters -

- the radiation effluent monitoring systems capabilities; changes the discharge flow paths;' or changes the effluent quantities, types or temperatures. - The structural integrity of block walls will not'be af facted since ' field ; attachments to block.. walls are reviewed by Engineering' prior.to installation..

l l

l:

27 Q- _ .

W,T ':m., ,

4 '

.. 'gw l

m

[If. - b ,

. , - 'O

/ $

• 1 .

y' v

'i E , -k - 'PLhMT CHANGE / MODIFICATION 074 PC/M CLASSIFICATION: -QRL U

. UNITr. 3 & 4-TURNED OVER'DATE: ;3/17/89 o

, , COMPUTER ROOM FIRE DETECTION

-)

Summary:

~

The; Fire Detection System.provides fire det'ection of the Computer Room and the Control Room on one alarm circuit. The Fire Detection 1 System'sfintentEls to alert operators . in ' order to preclude 3 the

~possible damage to safety related equipment.- These modifications ;

~

- - will.. enhance:theioperation of the fire detection' system.

Safety Evaluation

~

There are nof electrical interfaces.: with safety related systems under this modification. Electrical failure in this system will'

'not result in:the damage or loss of any safety-related system._ In-

' addition,' . all conduit and equipment required for these t . modifications.will1be seismically installed. The conduit does not require fireproofing. However, conduits will be sealed internally, iwhere required. . All penetrations through fire barriers will be

' sealed with : grout to restore the' barrier to its ' original fire ratingLin accordance with normal construction practices.

Cable : required .for this modification is purchased to Class IE requirements. The cable has been used in a. low voltage circuit for monitoring the operation of-fire detectors. Current flow in the-circuit is of the order of milliamps and hence.ampacity is not~a _

?. .  : concern. As all new cabl'es will be routed in new conduit, this modification will have no affect on the ampacity of the existing J cables andiwill not adversely affect supports for the existing raceway. systems. ' This modification will not adversely affect block walls evaluated under I.E. Bulletin 80-11.

Construction activities required by these modifications require no unusual; techniques or equipment. 'All ~ work will be closely

coordinated with plant personnel and controlled in accordance with xthe approved plant procedures and Technical. Spec ii f cat ons. i As a result, these modifications will not decrease the reliability or availability of the fire protection system or other plant equipment during the implementation phase. This modification does not reduce

-the audible dispersion of the Plant Paging System.

28 j

t PLANT CHANGE / MODIFICATION 85-081 i

PC/M CLASSIFICATION: NNSR UNIT: 3 &4 TURNED OVER DATE: .5/26/89 p

NYDROGEN LINE MODIFICATION p

Summary:

This package provides modifications'for the Units 3 and 4 hydrogen supply headers to install Excess Flow Valves with manual bypass '-

valves. The Excess Flow Valves allow the hydrogen flow for normal operation of the Volume Control Tank (VCT) but close when-the flow increases beyond a specific.value such as following a guillotine break of the hydrogen line downstream of the valves.

Safety Evaluation:

The modification to the hydrogen line to the VCT does not alter the function of the hydrogen supply system. The changes which are associated with plant safety features are minor in nature and do not change the function of the plant safety features. The changes

.which, are not associated with any plant safety features do not create - any condition that could be associated with or be more limiting ~than any accident defined in the FSAR.

These modifications are not inside Containment, are not attached to block walls, do not involve safety related snubbers and do not impact spent fuel cooling operations of the plant. The modifications do.not involve the addition of electrical cable or any changes to existing raceways. The final modifications accomplished by this package do not affect the flooding analysis.

29

PLANT CHANGE / MODIFICATION 85-122 PC/M CLASSIFICATION: NNSR UNIT: 3 TURNED OVER DATE: '5/5/89-SAFE SHUTDOWN MODIFICATIONS - APPENDIX R SummarYt This package provided design changes to meet 10CFR50 Appendix'R,Section III.G,, " Fire Protection of Safe Shutdown capability",

requirements. This PC/M ensures that at least one train of the -

equipment described in this package will be, available for use, '

following a' single area fire, in bringing the fire affected unit to a controlled hot standby and subsequent cold shutdown condition.

Safety Evaluation:-

~

The design modifications do not change the functional operability of the existing safety related systems for plant safety or. normal' plant operation.

Elec" vical channel separation between redundant circuits. are maints.ned for this modification.

All new cables and raceways within the power block and turbine deck area are seismically installed.

All new components and cables which interface directly with existing safety related systems are nuclear qualified (environmentally and seismically) or at least the same quality levels as the existing components.

Administrative and operational controls were established to provide assurance of proper switch positions for normal plant and ' safe shutdown modes of operation.

No single device added by this package is' utilized for redundant safety functions. Therefore, no single failure will result in a loss of redundant safety functions.

30

g _

h'; <i t

\'i i~ c

-. i ',,

1 An engineering evaluation has been performed regarding the ampac'ity;

~

, Land. voltage: drop for--the'added and existing cables and was found .

i _p- acceptable.

s.

3 lt _

e f

l 34

-e

.h -

31

c

?

!4 _

' PLANT CHANGE / MODIFICATION 85-123 PC/M CLASSIFICATION: NNSR' UNIT: 4 TURNED OVER.DATE: 5/5/89

'8AFE SHUTDOWN MODIFICATIONS FOR APPENDIX R' Summary:

This package provided design changes to meet.10CFR50 Appendix R, Section III.G, " Fire Protection of Safe Shutdown Capability",

requirements. This PC/M ensures that at least one train of the -

-equipment described in this package will be available for use, ~

following a single area fire, in bringing the. fire affected unit to a controlled hot standby and subsequent cold shutdown condition.

Safety Evaluation:

The design modifications do not change the functional operability of the existing safety related systems for plant safety or normal plant operation.

Electrical channel separation between redundant circuits are maintained for this modification.

All new cables and raceways within the power block and ' turbine deck -

area are seismically installed.

All new components and cables which interface directly with existing ~ safety. related systems are nuclear qualified (environmentally and seismically) or at least the same quality levels'as the existing components.

Administrative and operational controls were established to provide assurance of proper switch positions for normal plant and safe shutdown modes of operation. -

No single device added by this package is' utilized for redundant safety functions. Therefore, no single failure will result in a loss of redundant safety functions.

32

!!L ,

• - E

+ $ .< b 3

An'engin'eering~ evaluation has been performed regarding the ampacity and voltage drop for.the added-and'. existing, cables and'was found acceptable.

b 4

i i

33

PLANT CHANGE / MODIFICATION 85-140.

PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: 10/11/88 FUEL TRANSFER SYSTEM MANIPULATOR CRANE DUAL CABLE MODIFICATIONS Summary:

This package covers modifications to the Unit 4 manipulator crane.

These 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. 3.) The installation of a load test fixture used

to load test the manipulator crane.

This package adds - redundancy to the cable hoist portion of the refueling machine. This enhancement will increase reliability, reduce maintenance-time, and incorporate the latest state of the art capabilities into the refueling system.

Safety Evaluation:

This modification enhances the safety of the manipulator crane load handling operations. The replacement of the single cable system with the dual cable system will prevent the dropping of a fuel assembly in the event of a single cable failure, thus adding redundancy, to the system. In addition, a mechanical backup overload safety switch was added which will automatically shut the hoist off when an overload condition is encountered. This was added as a backup to the load cell sensing device to function in the event of a malfunction of the load cell. All other system safety features and built-in interlocks are maintained or improved.

Stearns Catalytic provided their analysis based on a design criteria different from the FSAR. However, the comparison of the two criteria shows that although they are not identical, the stresses by the more conservative Stearns Catalytic analysis are within the FSAR design criteria allowable limits.

The electrical modifications consist of the replacement of a single cable hoist motor with a dual cable hoist motor, the replacement of mechanical geared limit switches, and replacement of the Dillon l

34 l

. load sensing device with a Sensotec device. The addition of the.

above equipment does not adversely affect any safety related equipment as the equipment is located on the manipulator crane and is fed from a non-safety related. power supply. All electrical equipanent installed by this package has been installed seismically to prevent interaction with any safety related equipment during a seismic event.

There are severa components provided by these modifications that do not have qualified protective coatings. This is unavoidable, because certain items, like the hoist and load cell, are supplied with standard manufacturer's coating and cannot be ordered with a qualified coating and cannot be~recoated at the site without the risk of being damaged. However, the majority of these components replace existing components that also have unqualified costings.

Therefore, the net change in unqualified coatings is insignificant and the overall impact if these coatings during Loss of Coolant Accident (LOCA) is considered acceptable. .

The load test fixture modification is not classified as safety related since the load test fixture does not perform or affect any safety related function. The class I refueling pool liner plate integrity is not affected by this modification. Failure of the load test fixture could not affect any safety-related system or function.

I 1'

1 35

p

_ PLANT CHANGE / MODIFICATION- 85-148 PC/M CLASSIFICATION: SR UNIT: 4 TURNED'OVER DATE: 10/1/88 SPENT FUEL POOL COOLING SYSTEM-SEIBMIC UPGRADE SummarYi This provides the design to upgrade the Spent ~ Fuel Pool (SFP)

Cooling System to' ensure-that the cooling function of the system-is not lost as a result of a seismic event.' Also included in this ;

package is the upgrading of the system . piping . to include pool boiling - (212

• F) as the operating temperature for ' piping stress analysis. purposes only.

Safety Evaluation:

The . modification ensures that the SFP Cooling System remains-functional during, and subsequent to, a seismic event. The modification:also ensures that the thermal effects of fuel pool boiling can be structurally accommodated by the SFP Cooling. System-piping.- The modification does not affect the integrity, function or operation of any other safety related system.

36

- _ _ - _ _ --_-- _ - - _ _ _ _ a

9 PLANT CHANGE / MODIFICATION 85-194 r -

PC/M CLASSIFICATION: NNSR

-UNIT: 3&4 TURNED OVER DATE: 12/9/88 l BREATHING AIR COMPRESSOR TO FILL SCBA TANKS Summaryi

- This package provided the plant with the capability to fill . the.

~

larger (4500 psi) Scott air packs. The package provided for a ,

compressor plus compressed air holding tanks which store air for -

filling up to "four Scott Air packs without having to start the-compressor. -The compressor package also includes the. equipment necessary'for the air to meet the various standards required for-breathing air.

Safety Evaluation:

'This system does not perform any safety related function or affect the operation of any safety related equipment. The modification specified in this package provides for the installation of electrical equipment or raceways, but the electrical equipment per this modification is not located in a harsh environment and does not perform any safety related function. The modification does not involve safety related snubbers. The modification per-this package i' does not affect the' spent fuel pool or its cooling function. None of the new supports per this modification are attached to a block wall. This modification is not inside containment, therefore, it does not affect the heat sink analysis.

The installation of the Self Contained Breathing Air (SCBA) system J does not affect the High Energy Break Analysis evaluated in the FSAR because this system does not perform any safety related function or interact with any safety related equipment / function in a manner where a high energy line break in this system could affect

.the operation of a safety related system.

' The installation of the SCBA system does not reduce this audible ]

dispersion of the Plant Paging System or affect any fire protection features.

37

.____________-_____O

PLANT CHANGE / MODIFICATION 86-033 PC/M CLASSIFICATION: BR UNIT: 4 TURNED OVER DATE: 4/22/89 SGWL CONTAINMENT ISOLATION VALVE REPLACEMENT Summary:

The existing Steam Generator Wet Iayup System (SGWL) Containment - '

Isolation Valves leak and cannot provide adequate isolation.

Therefore, spectacle flanges were added to the SGWL System piping by the origina'. issue of this package to serve as temporary Containment Isolation boundaries and new ISI code boundaries. This i package then replaces the leaking valves with valves of the same type. After replacement of the valves, the Containment Isolation Boundaries and ISI Code Boundaries are to be returned to their original location at the valve.

Safety Evaluation:

i The affected portions of piping shall be seismically supported in accordance with project requirements for small pipe. The affected ,

portions of small bore piping shall be supported in accordance with 1 5177-PS-21: therefore, the existing stress analysis for the Main Feedwater Piping and Steam Generator Blowdown piping are not j affected and no now stress analysis are required.

This modification is not inside contair. ment, does not involve safety related snubbers, does not involve block walls, does not impact spent fuel cooling operations, does not affect the Radioactive Waste Treatment System of the plant, and does not i involve the addition of electrical cable or changes to existing 4 raceways. l The Steam Generator Wet Layup System does not perform any safety related function and is not addressed in the FSAR or Technical Specifications except in regard to containment Isolation. The locations of the new spectacle flanges are consistent with the 1 locations of the existing valves. Therefore, this modification is  !

considered bounded by the existing design in considering a terminal l end High Energy Line Break. The modification does not affect the '

l function of any safety related system.

38  !

.S

'1,'

. The package then replaces the leaking Containment Isolation valves

~

with ' new valves . to the same . type. It also provides for : the relocation to the containment Isolation boundaries and the ISI Code Boundaries back to their original position at the. valves.' Thus, l .this' package' restores the safety related portion of the piping to I-

~its original configuration.

i.

39

PLANT CHANGE / MODIFICATION 86-059 PC/M CLASSIFICATION: QR UNIT: 3 &4 TURNED OVER DATE: 5/5/89 SEALING OF MANHOLES FOR CABLE FIRE PROTECTION Summary:

This package pro"4ded for modifications to all manholes containing .

safe shutdown Lad /or alternate shutdown cables. The manholes -

containing these particular cables were to be modified to seal their covers to prevent any accidentally spilled combustible liquid from entering and creating a fire hazard.

Safety Evaluation:

The modifications contained in this package provide a design that prevents a postulated spill of combustible liquids from entering into these manholes, therefore these cables are adequately protected from all credible fire hazards.

Modifications under this package do not affect radioactive waste treatment systems, plant effluent, or the flooding analysis, and do not involve any safety related pipe snubbers. Block walls are not affected by these modifications.

This package does not involve modifications in the containment, therefore, the heat sink analysis is not affected.

These modifications do not change the configuration or function of any safety related systems.

40 ,

- - _ - _ _ _ - _ _ i

{

l i

L ,

PLANT CHANGE / MODIFICATION '86-130 PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: 4/30/89 RCP OIL COLLECTION SYSTEM MODIFICATION / ANALYSIS Summary:

This engineering package covered the modification to the Reactor -

Coolant Pump (RCP) Oil Collection System piping. This modification "

installed flexible . hoses between the pumps and the hard piping which drains the oil to the oil collection tank. Installation of the flexible hoses provides free thermal. movement of the RCP and aids in removal of the RCP motors for repair or replacement. This package also provided for the seismic analysis and includes upgrade of the RCP Oil - Collection System to meet the requirements of 10CFR50 Appendix R Section III.0 as committed to by FPL per letter L-80-336, dated October 7, 1980.

Safety Evaluation:

The piping modification does not affect the performance design basis of any safety related system. Pipe supports were added to the modified piping system to ensure that the piping system will remain functional following a Safe Shutdown Earthquake.

The modification specified in this package does not involve installation of any electrical equipment or raceways. It does not

-involve safety related snubbers or affect spent fuel cooling system and is consistent with the requirement for fire protection as stipulated in 10CRF50 Appendix R. The new supports for this modification increase the inventory of the material inside the containment. However, the increase in the amount of the total inventory of material inside the containment is negligible as compared to the figures given in Table 14.3.2-3 of the FSAR.

Therefore, this modification does not adversely affect the heat sink analysis of the containment.

The RCP Oil Collection System in a gravity drain, ambient temperature system. Therefore, the high energy line break analysis, evaluated in the FSAR is not affected.

41

PLANT CHANGE / MODIFICATION 86-182 PC/M CLASSIFICATION: BR UNIT: 4 TURNED OVER DATE: 3/27/89 SI PUMP MINIMUM FLOW RECIRCULATION VALVE ACTUATION REPLACEMENT Summarv:

This engineering package provided design for the replacement of the .

actuators on Isolation Valves 4-856A and 4-856B for the Safety Injection (SI) pump minimum flow recirculation lines.

. Safety Evaluation:

This modification will replace the air operators with motor operators on the recirculation valves. A Failure mode and Effects Analysis has been performed to ensure that at least one valve will operate when required and will not fail in an unacceptable state (closed when recirculation to the pumps is required). The results of this evaluation have been used in the design of the electrical circuit to ensure that no loss of power or electrical fault would render both valves inoperable or cause them to fail closed at an undesirable time.

This modification is not inside Containment, does not involve

-safety related snubbers, does not involve block walls, does not impact Spent Fuel cooling operations, does not affect the Radioactive Waste Treatment System, and does not reduce the audible dispersion of the Plant Paging System.

Since this modification is not inside Containment a.d ,therefore, there is no adverse impact on the heat sink analysis. In addition, the modification does not affect the operation of the Safety Injection System. Therefore, it can be concluded that the modification does not affect the 10CFR50 Appendix K analysis.

This modification installs new cables fror, MCCs "4A" and "4B". All cables are routed in separate channelized raceways. This channelization provides the separation required for redundant safety related components. Cable ampacity and voltage drop calculations have been performed and found acceptable. New electrical components added by this modification have been 42

seismically qualified, as required. All new piping and associated supports have been designed and installed in accordance with the requirements of the FSAR, Appendix 5A.

This modification installs new safety related motor starters in the vital sections of existing Motor Control Centers (MCCs) "4A" and "451". The addition of these vital loads does not adversely affect Emergency Diesel Generator (EDG) loading due to the intermittent operational requirements of these MOV's. Additionally, MOVs 4-856A and 4-856B have control power removed at all times except when required to be operated which further eliminates any potential adverse effect on EDG loading.

i 43 I

PLANT CHANGE / MODIFICATION 86-185 l

l' PC/M CLASSIFICATION: SR UNIT: 3 & 4 TURNED OVER DATE: 3/22/89 ANNUNCIATION IN MAIN CONTROL ROOM Od LOS8 OF EDG CONTROL POWER Summary:

This package provided annunciation in the Main Control Room on loss -

of control power to Emergency Diesel Generator (EDG)-A and/or EDG B. This was accomplished.d by replacing auxiliary relay (174V/X) in both respective EDG Control Panels, since these relays did not have sufficient spare contacts to provide the new alarms. This package also replaced the non-resistored full voltage indicating lights at EDG-A. Engine Panel 3C13 and EDG-B Engine Panel with resistored indicating lights.

Safety Evaluation:

There are no cable ampacity or voltage drop concerns associated with this modification since no new cables or additional loads are added by this package. The new relays and lights require less power.

The replacement components are qualified for the intended applications in the existing equipment and these components will not degrade the existing equipment. This modification does not change the performance capabilities of EDG system. This modification does not impact the design function of any safe shutdown components.

44

PLANT CHANGE / MODIFICATION 86-186 PC/M CLASSIFICATION: NNER UNIT: 3&4 TURNED OVER DATE: 5/16/89 ADDITION OF STRAINERS AND DRAIN TRAPS TO INSTRUMENT AIR PIPING g ummary:

This engineering package provided for the installation- of additional strainers, drain traps, and associated valves to the -

Instrument Air piping. These additions increased the reliability ~

of the Instrument Air System. Strainers were installed upstream of existing drain traps to minimize the possibilities of drain trap plugging and increase the effectiveness of the existing drain traps. Three additional drain traps were installed at newly identified low points to remove the condensed water from the air piping and improve the Instrument Air quality.

Safety Evaluation:

The strainers and drain traps are not installed in instrument air supply paths; therefore, malfunction or improper operation of any of the strainers or traps does not result in c limitation on operation of the Instrument Air System.

The Instrument Air System is not relied upon fer accident prevention or mitigation. The addition of strainers and traps will not cause a failure to equipment or functions the Instrument Air System serves. The failure of these components does not have any adverse interaction with any safety related equipment or function.

There are no safety related systems or equipment located in the area affected by this modification, therefore, there is no possibility of interaction with safety related systems or equipment i and there is no need for seismically qualified supports.

The strainers and drain traps do not affect the operation of any safety related components and are not required for safe shutdown or accident mitigation.

No work associated with this installation, changes to the heat sink analysis, or snubber program, or affects the radioactive waste treatment, spent fuel pit cooling, or effluent monitoring systems.

45 l

L - - - - - _ - _ _ - _ _ _ _ _

W ,

, d Y-I PLANT CHANGE / MODIFICATION 86-191 PC/M CLASSIFICATION: -QR

. UNIT: 3 TURNED OVER DATE: 1/26/89 RE21hCEMEMT_QF__GQMIRQL SWITCHES CV-3-2913 AND CV-3-3722 Summary 8

- This' package. replaces contro1' Switch CV-3-2913 at Control Console .

3C02 and. Control Switch CV-3-3722 at Vertical Control Board 3C04.

- Contrei Switch CV-3-2913 provides control of valves CV-3-2910,

- 2911, 2912-and 2 W for purge steam supply to Moisture Separator and. Reheaters l 3A, JB,.3C, and ' 3D. Co.ttrol Switch CV-3-3722 provides control' of Valves CV-3-3717 through 3723 for '. main . ste,am

- piping condensate drain to Condenser. These switches are'being. -

changed.to meet the . requirements of NUREG 0700 (Guidelines for Control._ Room Design Reviews) Section - 6. 4. 2.1, " Direction of Movement."

safety Evaluation:

The solenoid valves associated with the control valves do not perform any nuclear safety-functions, and Control Switches CV '

2913 and CV-3-3722 are not . associated Jith Class 1E circuits.

However, since these control switches are to be mounted in nuclear safety-related control panels in the. control Room, an evaluation-has been performed to ensure that these switches will 'not seismically interact with any safety related system, structure or component.

The modifications made the switch positions consistent with similar switches.in the Control Room and resolved human factor concerns on possible operator confusion and error due to inconsistent switch positions.

The modifications did not add any new equipment or components that could create new safety hazards of any nature and do not affect the

- plant safety or. operation.

46

PLANT CHANGE / MODIFICATION.86-195 PC/M CLASSIFICATION:. SR UNIT: 4 TURNED OVER DATE: 8/23/88 ARDII;LQE OF CONTINUOUS TUBE CLEANING CAPABILITY TO THE CCW HEAT EXCHANGERS Summary:

This package covers installation, testing, and evaluation of an on 1

. line mechanical tube cleaning system. The new cleaning systems will

. operate by introducing sponge rubber balls into the Intake-Cooling Water (ICW) supply line of each Component Cooling Water (CCW) heat

exchanger. The normal process flow will then force the balls through the heat exchanger tubes to maintain cleanliness. Screens in the discharge lines will collect the balls, a' centrifugal pump _

will recirculate the balls to the injection point, 'and a ball .

collector will allow addition or retrieval of th'e cleaning. balls..

Safety Evaluations The piping and components added by this package, as well as affected existing ICW piping, is seismically- supported in accordance with the requirements of the FSAR Appendix 5A for Class I system. All ICW pressure retaining components required for this modification were procured as safety related, seismically.

qualified. All components were procured to the requirements of ICW design basis conditions. This modification is not inside containment, does not involve safety related' snubbers, does not involve block walls, does not impact spent fuel cooling operations, does not affect tha Radioactive Waste Treatment System, does not create any new flooding concerns and does not adversely impact the Plant Paging System operation or audible dispersion.

The modification . involves the addition of new cables. All new cables utilized for this modification are qualified for their intended application and have been evaluated for ampacity and voltage drop and have been found to be acceptable. In addition, all new cables will be routed in seismically installed raceways to preclude any adverse seismic interactions.

I 47 i L

L

.__m,_.. _ __

I r

e "~l:

' The! addition of continuous tube cleaning to the CCW heat exchangers

~

does not' affect the design basis'of.the ICW system or of the CCW

. system as. described of the FSAR. Continuous. tube-cleaning serves

- no safety.related function, and failure of the components will not cause'a condition detrimental;to. nuclear safety.

l l

48

h' L

t:

j PLANT CHANGE /MODIFICATIOtt 86-208 p

PC/M CLASSIFICATION. SR UNIT: 4 TURNED OVER DATE: 10/1/88

-IST GAUGE INSTALLATION FOR THE SPENT FUEL PIT COOLING PUMPS Summary:

'This modification installed instrumentation for Inservice' Testing of the Unit 4 Spent Fuel Pit (SFP) Cooling Pumps because the pumps ,

have been added to the Turkey Point Inservice Testing, Program. -

Specifically, a flow element (annular type) with a local flow indicator was installed in the common suction line of the pumps and existing suction and discharge gauges for pressure measurement were upgraded with . gauges- having accuracies in accordance with the requirements of ASME XI, Subsection IWP (Inservice Testing of Pumps in Nuclear Power Plants).

Safety Evaluation:

The modification affects the Spent Fuel Cooling S;rstem. The instrumentation added is seismically -installed to .orevent its

-interaction with safety related equipment. The ancular flow element, flow indicator, and pressure gauges'added are not required for safe shutdown or accident mitigation. Also a calculation was performed which postulated the worst case accident resulting from the failure of this. equipment. The accident resulted in an one inch . hole in- the SFP Cooling Pump's suction piping. The calculation supports the fact that should the damage occur, the SFP

. cooling system will not be. prevented from performing its normal function.

49

f PLANT CHANGE / MODIFICATION 86-213 pC/M CLASSIFICATION: QR UNIT: 3-E 4-TURNED OVER DATE: 5/15/89 EMERGENCY LIGHTING FOR APPENDIX R Summary:

This modification will provide the necessary illumination'to meet the requirements of 10CFR50, Appendix R, Section III.J plus additional lighting facilities for operator convenience. . This '.

engineering package also provides layout drawings depicting Appendix R equipment and lights.

Safety Evaluation:

This addition and relocation of emergency lighting fixtures will enhance the operators ability to access equipment necessary for the safe shutdown of the plant, and will be seismically supported, as applicable,_to preclude interaction with safety related equipment.

The fixtures do not interface with safety related equipment or

. perform safe shutdown functions. The equipment'added is suitable for use in the environment in which it is located. The implementation of this package does not change the functional or operational requirements of the existing safety related systems for-safe shutdown or normal plant operation ~.

50

PLANT CHANGE / MODIFICATION 86-226 PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: 11/08/88 TURKEY POINT UNIT 4 ADDITION OF NON-SAFETY RELATED MCC 4H (4B21)

Summary:

The scope of this package consisted of adding a new Motor Control .

Center, MCC 4H, to provide 480 volts power feeds to the Continuous -

Tube Cleaning (CTC) system to be installed for the CCW heat exchangers. The installation of a new MCC was required since there were no sparc breakers or available space for new breakers in existing MCCs in the vicinity of CCW heat exchanger CTC units.

Safety Evaluation:

1he added MCC is powered from a non-safety related source of power and will be utilized to power only non-safety related loads.

Failure of the MCC or any component therein will not affect any safety related system or function. All systems and structures including the relocated reach rods, added by this engineering package, have been designed and installed to protect existing nuclear safety related equipment and systems from the potential effects of adverse interactions. No work associated with this engineering pcckage is inside containment and hence does not affect heat sink analysis, nor does it affect the radioactive waste treatment system, spent fuel cooling system or safety related snubbers. Equipment or cable associated with this engineering package will not be attached to or be in proximity of any block walls.

All new cables utilized for this modification are suitable 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 seismically installed raceways. The MCC anchorages to the supporting steel have been designed for the applicable loads in accordance with FSAR, Appendix SA to preclude any adverse interactions with safety related structures, systems or components.

51

g- . -

o b '- The addition of the'MCC requires the' relocation of: the reach rods for ccw Jsystem'. valves 787A. and 4-787B. The relocation:of~the I. . - reach : rods .does :, not affect the capability ' to' operate either~of .

f- . these valves post-accident.

s 4

4 4

52

PLANT CHANGE / MODIFICATION 87-026 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 5/6/89 COMPONENT COOLING WATER SYSTEM NORMAL CONTAINMENT COOLER TUBE BUNDLE REPLACEMENT Summary:

This package provides details for the replacement of the tube bundles in the. Normal Containment Coolers (NCCs) in the Component -

Cooling Water (CCW) System. Also, this package provides for the ~

replacement of the sheaves on the NCC fan motors in ordsr to increase the fan speed and maintain air flows essentially the same as before.

Safety Evaluation:

The existing component tube bundles and fan sheaves are being replaced by tube bundles and fan sheaves of similar design and materials. No changes have been made to the operational design of the system. This will-have no adverse'effect on the equipment.

Existing. fan motor feeder cables have been evaluated and are acceptable for the. increased motor current.

The tube bundles involved in this modification are slightly larger but similar in design to'the existing components.- Likewise, the new fan sheaves are also different size but similar in design to the existing components. No changes have been made in the operational design of the system. The increased electrical load due to the increased CCW flow is below the Emergency Diesel Generator (EDG) maximum allowable.

l l

53

PLANT' CHANGE / MODIFICATION 87-034 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 5/15/89 DELETION OF CRDM COOLER FANS AUTO EDG LOADING Summary:

The scope of this package consisted of: 1) deleting the automatic loading of the Control Rod Driver Mechanism (CRDM) cooler fans from -

the Emergency Diesel Generators (EDG's) and 2) revising the CRDM '

cooler- fan damper power feed to ensure damper closure after a thermal overload trip occurs on the CRDM cooler fan.

Safety Evaluation:

The CRDM cooler. fans-are required for normal plant operation but not.for safe shutdown of the plant. Failure of the fans will not prevent the CRDM from performing its safety related function nor will it prevent natural circulation cooldown. This modification prevents automatic EDG loading of the CRDM cooler fans which could adversely af fect EDG ' operation. However, the fans can still be loaded manually ' on the EDGs if required and diesel capacity 'is available. The safety related sequencer circuits and components are isolated from the non-safety related' fan' circuit by the use of safety related circuit breakers.

The CRDM cooler fans and associated dampers share'a common control circuit. This modification only relocates the connection of the damper's power feed and does not change the damper's power feed and does not change the damper's control logic. The damper's new power feed is connected directly to the fan's control power transformer through a separate protective fuse. CRDM cooler fan dampers, which work in conjunction with their associated fans, are required for normal operation but not for safe shutdown of the plant. Failure of the damper will not prevent the CRDM from performing its safety related functien nor will it prevent natural circulation cooldown.

The vital Motor control Center (MCC) buses are also protected from a fault on the non-safety related' damper circuit by safety related

. circuit breakers.

54  ;

I L l_ _E___ ___ _.. _ _ _ _ . . . _ _ l

p g

' ,q PLANT CHANGE / MODIFICATION 87-100 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 5/15/89 REACTOR CAVITY SEAL REPLACEMENT SummerTi This package provided for the replacement of the Unit 4 reactor .

cavity seal system and the installation of a restraint system for -

the reactor cavity seal ring plate in its storage location. The seal system was modified to provide redundant passive seals to prevent leakage in excess of 50 GPM. An inflatable seal was also provided to reduce seal leakage to as low as practical in line with

.ALARA and housekeeping requirements.

This . package also addressed the' replacement of twelve reactor cavity seal ring anchor bolts which were not in accordance with the original design drawings. These anchor bolts were identified as "Wej-it" type expansion' anchors. This package provided for the replacement of these expansion anchors with grout-in-place anchor

. bolts in order to meet the original design intent of the sealing system.

Safety Evaluation:

.The proposed design retains the original design concept of'two passive seals, and also includes the inflatable seal concept added in 1974 and used with success thereafter. The compression seals have been proven through vendor testing to have a margin of safety of at least 2.77 without leakage. A failure analysis of the non-safety related inflatable seal indicates it will not adversely affect the safety related compression seals.

The proposed reactor cavity seal ring anchorage modification (from

.use of expansion anchors to grout-in-place anchors) wil1 increase the anchorage system capacity safety factor to conform to the original design intent of the sealing system.

The reactor cavity seal ring storage restraint system does not perform a nuclear safety related function. The safety related function of - the ' missile shields is to prevent a breach of the containment dome by a postulated control rod drive missile 55 i

- ____--____--_____-__-_-____-___-__-____-___a

N < -

p ,

i

^

h' ;1 M

generation event.. .The.installationLof;the restraint: system does

, not'cimpair the missile shields from performing this' function. The-l storing of the seal: ring . on Atop of := the. shields during : plant .

operation:provides additional mass lto. improve the ability of the-missile' shields.:

11 0:.

l-4 l8-56

j PLANT CHANGE / MODIFICATION 87-212 PC/M CLASSIFICATION: BR UNIT: 3 E4 TURNED OVER DATE: . 2/10/89 EDG ENHANCEMENT - SITE PREPARATION Summary:

This package details the requirements for the preliminary site work to be performed prior to the construction of a new diesel generator ,

building located between unit 2 and 3, just west of the water -

treatment area. The new building will contain two new diesel generators which will enhance the emergency power capability.

Safety Evaluation:

The only safety related components associated with these-modifications are the existing Intake Cooling Water (ICW) piping and a electrical ductbank. There are no accidents previously postulated in the FSAR involving these safety related components.

Special precautions have been specified to restrict loading on the ICW piping and excavation in its vicinity. The existing electrical ductbank located in the vicinity of these modifications has been

'found to be adequate to withstand the revised loading due to the modification of the ramp.

Most of the modifications do not affect the function of the systems involved; the maintenance building will be removed entirely; and the portion of the diesel oil transfer piping being installed does not perform any safety related function. Furthermore, the temporary modifications to the Radiation Control Area fence and site lighting are designed to maintain the original design function of the components. None of these modifications will reduce the margin of safety as defined in the basis of any Technical Specification.

57

h

. PLANT CHANGE / MODIFICATION 87-310 PC/M CLASSIFICATION: QR UNIT: 3 &4 TURNED OVER DATE: 6/23/89 EMERGENCY RESPONSE DATA ACQUISITION & DISPLAY SYSTEM UPGRADE Summary:

The enhancement of the Emergency Response Data Acquisition and .

Display -System _(ERDADS) is a result of startup testing and -

debugging of the computer software, and to satisfy items identified in FPL letter L-86-264, " Update on the Safety Parameter Display System." This package provides for the installation of the updated computer and peripheral equipment to support software revisions provided by Energy Incorporated.

Safety Evaluation:

The inherent function of the' system to provide selected plant data in the form of. video displays and printed reports to the control room, Technical Support Center, and Emergency Operations Facility staff ' as described in the FSAR is unaffected by the equipment installation.

Seismic design criteria -for the installation of equipment and raceway have been applied to preclude adverse interactions with safety related equipment. The new equipment installations will not adversely affect any block walls. Also, this modification does not reduce the dispersion of the plant paging system, affect any fire protection features or the Fire Hazards Analysis, or impact the existing flooding analysis.

Cables for power and computer communications circuits were reviewed for compliance with project ampacity and voltage drop calculations, as well as vendor requirements, and are acceptable.

This package does not interface with the signal isolator portion of the ERDADS; thereby, maintaining electrical separation for new I and reused circuits.

58 l

10 V

I -PLANT CHANGE / MODIFICATION 87-317 F , PC/M CLASSIFICATION: QR UNIT: 3 & 4 -.

TURNED OVER DATE: 11/08/88 PERMANENT INSTALLATION OF WHOLE BODY COUNTER Summary:

The whole body counter equipment was temporarily located in a trailer adjacent to the Health Physics Building. This package relocated the whole body counter equipment in a permanent facility; . -

removed the trailer. from the area; and added office space as ~

required for Health Physics personnel. The west end of the Health Physics (HP) building was remodeled to accommodate additional offices and the whole body counter equipment.

Safety Evaluatigni, The modifications to the HP building will not add to, modify, or affect any . plant nuclear safety related systems nor will they perform a safety related function.

The modifications to the HP building do not change any assumptions made or conclusions drawn in the Turkey Point FSAR.

. .The probability of occurrence or the consequences of an-accident or malfunction of equipment important to the FSAR has not been increased since the modifications provided in this package do not affect any safety related features or systems.

- The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR has not been created since no new safety related functions or interfaces with safety related systems are created by this engineering package.

• The margin of safety as defined in the basis of any Technical Specification has not been reduced since the modifications to -

the west end of the H.P. building perform no safety related function and are not included in the bases of any Technical e Specification.

59

p I

p

PLANT CHANGE / MODIFICATION 87-320 p

PC/M CLASSIFICATION:- SR

-UNIT: 4

!: . TURNED OVER DATE: 7/21/88 CCW HEAT EXCHANGER CHANNEL HEAD REPLACEMENT i-Summary

- This : package replacedL the' Component Cooling Water. (CCW) channel heads with'new -channel heads of a more corrosion resistant *.

material ~.

4 In addition, this package changed the.3/4" drains and vents to 1-1/2" or 1" drains and vents respectively. The change in valve size-affected only ' the . time required to drain the channel' head for maintenance.

Safety Evaluation:

The original channel heads and the-new were both. built to ASME Section VIII. The new material has been shown to be more corrosion resistant and to be as good or better than the original in other areas. Also,.the vendor has stated that the new-channel heads conform to the seismic analysis performed on the previous channel

. heads. The new valves being installed as part of this modification meet the system design req'irements u and.are identical in quality to the original valves.

60

L' PLANT CHANGE / MODIFICATION 88-069 i

PC/M CLASSIFICATION: SR UNIT: 3 &4 TURNED OVER DATE: 5/4/89 AFW AND MAIN STEAM SYSTEM TRAP AND DRAIN DRAWING UFDATE Summary:

.This package ' corrects drawings to resolve differences betwe'en drawings and actual field conditions. ,

Safety Evaluation:

None of the. items identified are considered operability concerns based on design equivalence. Therefore, they do not impact system operation nor do they create the potential for mis-operation of any

( safety related system. Drawing corrections are necessary only to reflect the as-built configuration of the plant, no new equipment or components are being installed. The Auxiliary Feedwater (AFW) steam trap and drain system, where it was not previously shown on the T-E drawings,'has been reviewed for seismic interaction and found ' to be acceptable. The additional detail provided by the drawing changes instituted by this package are drawing enhancements which will facilitate the operation of the . existing system and serve only to reflect the as-built configuration of the plant. As such, this package does not impact plant safety, system functional requirements, plant Technical Specifications, system or component operation.

61

L PLANT CHANGE / MODIFICATION 88-078 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 2/11/89 RESIDUAL HEAT REMOVAL PUMPS MECHANICAL SEALS AND SEALS COOLER REPLACEMENT

1. _umm4rYi This package provided details for the replacement of the mechanical seals and the seal coolers on the Residual Heat Removal (RHR) ;

pumps. The old seals have proven to have an unsatisfactory seal life. The new seals are an upgraded design, and the seal coolers have a larger heat removal capability,which should provide a longer

. operational life.

-Safety Evaluation:

The specification for these new seals was specifically prepared to include the design requirements for seal life by taking into account operating parameters such as process fluid chemistry, pressure, temperature, uninterrupted running time, cooling water telrr trature and availability and radiation doses. These operating parameters were specified for several modes of operation including accident conditions when the RHR system may be required to operate for an extended period. The new seals meet the requirements of the

.specificellon. No change is ' being made in the operation or funct'ais of the RHR system.

62

I l

PLANT-CHANGE / MODIFICATION 88-131

,PC/M CLASSIFICATION: SR  !

UNIT: 3 &4 TURNED OVER DATE: 8/11/88 CONTAINMENT SPRAY PUMP PULL OUT ASSEMBLY REPLACEMENT pummary:

This package only provided the document updates necessary to allow interchangeability of the originally supplied containment spray *,

pump pullout assemblies and the recently obtained " spare" pullout assemblies. (A pump pullout- assembly is the entire pump less coupling, motor, and casing.) Part by part interchangeability is not-provided, but the entire assembly may be changed as a whole.

A s .' s h o w n on the affected drawing, 5610-M-470-029A/88-131, the appropriate testing must be performed when an assembly is completely changed prior to returning the affected pump to service.

Safety Evaluation:

The new materials used in the pullout assemblies are equivalent or better than the original materials since they reflect the more current standards and ' specifications for this pump. The installation of the new pullout assemblies decreases the weight of the pump. Therefore, no additional loads are introduced by this change. The proposed modification has been evaluated and determined not to adversely affect the seismic qualification of the pump-or its interfacing parts. This modification does not alter the plant's design basis and is bounded by existing design analysis.

63

1: 2

i l

g PLANT CHANGE / MODIFICATION '88-160 q

PC/M CLASSIFICATION: SR UNIT: 3&4 TURNED OVER DATE: 7/2/88 RER/8AMPLING VALVE UPGRADE Summary:

This package' addresses revision of drawings / documents.to reflect the existing.. configuration, replacement'of drain valves (*-958A, -

• -958B, 3-999, 3-1000) with valves to meet the Sampling System.'

design requirements, revision of the ASME Code Boundary Drawings, and revision of the associated items in the FSAR, TEDB and Q-List.

Safety Evaluation:

Modifications made.by this package do not alter the. plant's design-basis, i.e., functional, performance, operating and regulatory requirements, and are bounded by the existing-design criteria.

64

T' i I+

l PLANT CHANGE / MODIFICATION.88-172 PC/M CLASSIFICATION: SR

. UNIT: 3 TURNED OVER DATE: 10/15/88 CONTAINMENT PURGE VALVE ACTUATOR VENTING MODIFICATION Summary:

This package installed a check valve bypass around the' existing ~

solenoid valve SV-3-2602 outside containment and solenoid valve SV .

2-2603 inside containment.It also installed larger' tubing between ~

the first. solenoid valve and the purge valve (actuator) to provide an additional venting capability. Included in this modification was the reconfiguration of the actuator supply Instrument Air (IA)..

piping, tubing and associated valves.

PMfety Evaluation:

The modification of the Actuator Venting System for the Containment Purge Valves does not increase:the probability of a malfunction of equipment as previously evaluated. This modification enhances th9 capability of the Containment Purge Valves to meet'their safety function to close within five seconds upon receipt of safety '

actuation signal.

Also, the failure of the Containment Purge Valves is not one of the postulated initiating events of any design bases accident or event described in the FSAR. This modification does not adversely impact the Containment Purge Valve Single Failure Analysis provided in Table 6.6-2 of the FSAR. as documented in the Failure Modes and Effects Analysis.

65

' PLANT CHANGE / MODIFICATION 88-193 PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: 4/21/89 DETECTION SYSTEM FOR MONITORING RCS LEAKS IN THE REACTOR HEAD AREA Summary:

This modification installed a. non-safety related. Reactor Vessel Head Area Leakage Detection System in Unit 4. The leak detection system . draws samples from the reactor head area or from the -

containment atmosphere, into a skid-mounted particulate sampling ~

system located inside containment. The sampling system skid consists of a particulate detector, sample pump, motor operated valves, and associated apparatus. A provided remote control and display rack combines the Unit 3 & 4 monitoring instrumentation into a common rack. This remote control and display rack is located in the computer room and will include a programmable strip chart trend recorder, controller, digital rate meter, and associated apparatus. .The sampling system skid is provided with 120 VAC power and 480 VAC power. The remote rack is provided with 120 VAC.' The power is from non-vital distribution panels.

Safety Evaluation:

The Reactor Vessel Head Leakage Detection System is not required for safe shutdown of the plant. The system including its associated tubing and raceway has been designed and will be installed to preclude any adverse interaction with safety related components.

Although some components to be installed inside containment may be coated with unqualified coating systems, it has been demonstrated that they do not impair the ability of the Emergency Core Cooling

. Systems (ECCS) to perform as intended.

The sampling system skid is installed inside containment. The effect on containment heat sink, containment free volume, and hydrogen generation due to addition of this system has been adequately addressed and is concluded to be negligible. The design bases established in the FSAR have not been changed.

66 i

s-

-4 This modification does not ~ involve any . safety related ' snubbers, does not impact the spent fuel cooling operations of the plant, and does not affect'the. Radioactive Waste Treatment System.

The components associated with this modification will 'not be attached to any block walls unless the structural adequacy of the

' wall has been demonstrated.

No devices installed by this modification penetrate an existing nuclear safety related pressure - boundary or affect any existing piping analysis.

l l

l l

l 1

i 67 1

a i

PLANT CHANGE / MODIFICATION 88-197 l

l PC/M CLASSIFICATION: BR UNIT: 4-TURNED OVER DATE: 4/21/89 INADVERTENT ACTUATION OF PRMS RELAYS R-11 AND R-12 Summaggi This package modified the Process Radiation Monitoring System (PRMS) cabinet 4QR66 by providing separately fused power supplies, ;

improved connections to system ground, metal-oxide resistors and bypass switches for PRMS channels R-4-11 and R-4-12. These modifications were necessary to prevent inadvertent alarms and system actuations during maintenance, testing, and surveillance activities.

Safety Evaluation:

This modification increases the reliability of the PRMS power supply by adding a system ground and equipment that will be seismically qualified. All equipment has met the necessary environmental qualification criteria for the area in which it will be used. The bypass switches also meet the necessary seismic and environmental requirements, and are installed as a maintenance aid only. The load on the Vital 120V AC system has been reviewed and determined to be acceptable. The system / equipment protection features have not been deleted or modified, nor has the support system performance been downgraded.

The PRMS is used to monitor radiation levels in fluid processes, and provide control to stop or divert these processes. Failure of the monitoring system cannot cause an accident, and the modification provided by this package does not impact the ability of the PRMS to perform its design function. The reliability of the PRMS is increased by adding separately fused power supplies for the individual drawers and administratively controlled bypass switches used only when the associated drawer is removed from service. The PRMS is designed to fail in the safe condition; therefore a power failure to this system will automatically actuate the required functions. If either R-4-11 or R-4-12 is in bypass mode, indication of BYPASS MODE is provided on the annunciator panel, while the other channel's operation is not affected. Also, the 68

-- --- - - - - -- - - - - - - - =

y h.. ,

1. i-cf.-t f).(: ,

l . r l_- .

s.

4 p gy L bypassed channel: is. not restricted from indicating ~ a HIGH AIARM on .

J the'r same' ( annunciator L panel :. and : then': the : operator ' may initiate

' actuation: manually.

., , .e -

.)

4 a g; . , < .

Y

i. -

6

., h - :

}- ,

i s

G t

4

% i{

f 69 A

PLANT CHANGE / MODIFICATION 88-212 PC/M CLASSIFICATION:- SR-UNIT: 3 TURNED OVER DATE: 4/26/89 l

PRESSURIZER LEVEL INDICATOR DRAWING DISCREPANCY, LI-459B Summery 3 Drawing 5610-7-D-15 incorrectly showed the local (Charging Pump -

Room) Pressurizer Level Indicator ' to be LI-459D, whereas other documentation correctly showed this instrument to be' LI-459B. This -

packags corrected this error and enhanced other drawings related '

to this loop.

Safety Evaluation:

The evaluation proved.that the existing field condition was correct and therefore bounded by the existing design analysis.

I 1

l-L 70 l

l-

PLANT CHANGE / MODIFICATION 88-245 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 4/21/89 MSIV AIR ACCUMULATOR SYSTEM Summary:

This modification consisted of improving independent pneumatic controls, redundant electrical control circuits and dedicated air reserves which ensure that each Main Steam Isolation Valve (MSIV) will close in five seconds or less upon receipt of a closure signal coincident with the most limiting process conditions for valve -

closure. In addition, the modification ensures MSIV closure can be maintained for a minimum of one hour without the need for operator action independent of the availability of instrument air.

The air accumulator and modified controls ensure that the MSIVs have the capability to meet the requirements of the Technical Specifications.

Safety Evaluation:

The. modification maintains the original design bases of the MSIVs by providing an independent accumulator for each MSIV air operator.

Protection against missiles generated by environmental conditions or plant equipment failures is provided by spatial separation, shielding and intervening structures. This approach precludes a common mode mechanical failure and thus does not reduce the ,

reliability of the MSIVs. I I

This modification maintains the electrical redundancy of the MSIV l actuation system. In order to prevent a common mode electrical ,

failure from adversely affecting the closing performance of the I MSIVs, redundant 125V channels power redundant solenoid valves for l each Actuation Assembly Rack. Therefore, the ability of the MSIVs to perform required safety functions is not affected by a common -

mode electrical single failure of the solenoid valves. This modification is an enhancement to ensure that the MSIVs perform the  ;

intended safety functions as required by the FSAR and does not adversely affect any safety related eq.Hpment or other safety related systems or components.

71 l

__.__._______._________.m

PLANT CHANGE / MODIFICATION 88-263 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 1/25/89 COMPONENT COOLING WATER HEAT EXCHANGER REPLACEMENT ggstnary:

This package provided for the replacement of the Component Cooling Water (CCW). Heat Exchangers. The replacement heat exchangers are -

siuilar to the existing _ units except the new units will y rated '

to withstand a maximum-shell side flow rate of 4.00 x 10 lbm/hr without damaging vibration. In addition, the new heat exchanger shell side vent and drain connections will be increased from 'J/4" to.2" to facilitate maintenance activities. This package also provides modifications.to the Intake Cooling Water (ICW) piping system by replacing the existing, cement lined, ductile iron 90 elbows, with new elbows, on the CCW Heat Exchanger channel head inlets and outlets.

Safety Evaluation:

During implementation of this modification, the remaining portions of the CCW/ICW systems were stress analyzed and provided with temporary supports to maintain seismic qualification in accordance with FSAR, Appendix 5A. Also, administrative controls were in place, in the form of existing plant procedures, to preclude handling of hesvy loads over redundant safe shutdown equipment and for approving deviations from safe load handling paths.

The replacement components were designed in accordance with later editions of the eriginal Codes of Construction, which were reviewed and found suitable for use in this application. It was demonstrated that all requirements of the ASME Section XI Code, for use of lator editions of the Construction Codes, were met. The replacement components were designed to meet the same functional and perform 0nce requirements as the original equipment. The replacement heat exchangers and the final system configuration have been stress analyzed and seismically qualified in accordance with the requirements of the FSAR Appendix 5A.

72

7 r -J ~New: .ICW system performance. curves were generated- for .the replacement heat. exchangers .to ensure adequate heat removal capabilities ~ exist during normal and accident conditions. . This

< modification;presentscno new operational requirements for the ICW

!' ' and CCW systems and the replacement components do-not present any

~

new failure modes or. increase the probability of occurrence of any existing . failure ' modes. . This modification does not degrade, prevent -.: any required - actions, or - alter - any accident. analyses

-assumptions previously evaluated in the FSAR.

l l

~_ .

k 73 1

L 1

fi r i r j' f PLANT CHANGE / MODIFICATION 88-282 PC/M CII. OSSIFICATION:.. QR UNIT: 4 TURNED OVER DATE: 4/30/89 REPLACEMENT OF SGBD FLOW INSTRUMENTATION AND CHANGE OF SAFETY CLASSIFICATION 0-BASIS Summary:

This package provided for the replacement of Steam Generator ~

Blowdown Flow Indication pneumatic transmitters, square root extractor, P/I, and Flow Recorder with electronic instruments.

Additionally, the reason for the safety classification of the FlLw Transmitters is changed.

Safety Evaluation:

The accuracy, drift, and. stability of the signal will be improved since the electronic instruments are less affected by vibration.

The accuracy will further be improved since the transmitter output signal will go directly to the control room instead of through two instruments which adds additional error.

The transmitter output will be a 4-20mA signal. This matches the 4-20mA signal of the existing P/I. Therefore, the output of the flow instrumentation remains uncnanged. It was determined that sufficient space is available to fit the new instruments and associated raceways without interference with existing plant systems, structures and components.

The combined weight of the new instruments, their associated-raceways and mounting brackets being added to the instrument panel exceed those they replace. The resulting applied load to the instrument panel as a result of this modification has been evaluated and determined not to cause an adverse effect on the instrument panel or the existing supporting structures.

The mounting details for the new instruments comply with the vendors' mounting requirements.

74

f. '

PLANT CHANGE / MODIFICATION 88-287 PC/M CLASSIFICATION: BR UNIT: 4 TURNED OVER DATE: 5/2P/89 AFW FCV STEM TRAVEL LIMIT STOPS Summary:

The Auxiliary Feedwater (AFW) flow control valves (FCV's)-function to automatically or manually throttle flow to the steam generators upon initiation of the AFW system. The installed valves (CV-* '.

2816, 2817, 2818, 2831, 2832, and 2833) are 4 inch 900 lb. rated air. globe valves supplied by Valtek. Normal valve control is through an air actuator mounted on the valve with automatic and remote manual control. A handwheel is also provided to take local manual control if necessary.

The AFW FCV's (CV-4-2816, 2817, 2818, 2831, 2832, 2833) handwheels were locked to restrict valve stem travel to a maximum of 85% open.

This package installed a stroke reducing ring to limit the stem travel.

Safety Evaluation:

The valve stroke reducing rings to be installed on the FCV's were designed by the valve manufacturer, Valtek. The ring is 2.5 inches in diameter and weighs less than 1.0 lbs. This ring will be welded to the bonnet to limit the valve stoke by causing the top of the plug to seat against it. The additional weight is small when compared to the weight of the valve (approximately 860 lbs.) and will not adversely affect the seismic qualification of the valve or piping system. The FCV's are currently locked to limit the flow through the valve to 85%. This ring will also limit the flow to 85% but will eliminate the need to lock the handwheel at 85%.

Therefore, since the maximum throttling capability of the valve remains the same, this change is acceptable.

75

L L 88-290 PLANT CHANGE / MODIFICATION PC/M CLASSIFICATION: QR UNIT: 3 &4 TURNED OVER DATE: 9/27/88 l

l l

l CHANGING OF ' METEOROLOGICAL TOWER DATA FROM INSTANTANEOUS TO 15 MINUTE AVERAGED DATA Summary

• This package provided the installation of a microprocessor at the 10 meter tower and 60 meter tower. This microprocessor samples -

instantaneous meteorological data at 5 second intervals and '

mathematically averages the data over 15 minute intervals. The microprocessor updates its output at 1 minute intervals.

The meteorological data is required to be averaged over a 15 minute interval for use. in the Emergency Procedure EP-20126, "Off Site Dose Calculations," and for- use in the radiological dose calculations by the Emergency Response Data Acquisition and Display System (ERDADS). This package installed the equipment required to average the instantaneous meteorological data.

Safety Evaluation:

The modification does not impact any safety related or quality related systems. Also, none of the accidents analyzed in this FSAR requires meteorological data. This modification does not constitute an unreviewed safety question or require change to the Plant Technical Specifications.

76

b-1 PLANT CHANGE / MODIFICATION 88-310 PC/M. CLASSIFICATION: BR UNIT: 3 TURNED.OVER DATE: 8/27/89-REMOVAL OF PAHM-3-008B AND P.Of-3-009' FROM THE IXS7JJi t'

Summary:

The valve lineup to the hydrogen nonitor ~ (AE-3-6307B) .. in the Post. .

' Accident Hydrogen Monitoring System (PAHMS) was not consistent- for: -

valve PAHM-3-008B as shown on.5610-T-E-4534, sheet 1, 5610-M-11 ~

and 3/4-OP-094. ' Valve PAHM-3-002Band a valve located in . the hydrogen.' monitor are provided for -isolation of Lthe hydrogen

. monitor.-

' Safety Evaluation:

This valve is not required'for-isolation of the hydrogen monitor or any' maintenance activity. The removal of this valve will' return,

.the' system-to its original seismically analyzed design.

77 t..

5

-- . - - .____..___._..-...m.._.____._m._-______--___.m_____._._ mmm_

(

k -

k; i

l- . PLANT-CHANGE / MODIFICATION'88-311 l.:

'PC/M CLASSIFICATION: 8R

. UNIT:- 4 TURNED OVER DATE: 8/27/89 REMOVAL OF PAHM-4-008B FROM THE SYSTEM Summarvt The~ valve ~ lineup to' the hydrogen monitor (AE 4-6307B)~; in the Post-Accident Hydrogen Monitoring System (PAHMS)-was not consistent for, valve PAHM-4-008B as shown on~5610-T-E-4534, sheet.1, 5610-M-11 ',

and 3/4-OP-094.- Valve ~ PAHM-4-002B and a valve located in the

- hydrogen monitor' are provided.-for isolation .of the hydrogen

. monitor.'

~ 8afety Evaluation

. This' valve is'not required for isolation'of the Sydrogen monitor or any maintenance activity. The removal. of this va: 1.ve will-return the system to its original seismically analyzed des (qn.

78

! PLANT CHANGE / MODIFICATION 88-336

, PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: 3/1/89 REPLACEMENT OF PCV-4-118 AND PCV-4-119 ON THE HYDROGEN AND NITROGEN LINES TO THE CVCS VOLUME CONTROL TANK (VCT)

Summary:

The existing regulators on the Hydrogen and Nitrogen lines to the -

Volume control . Tank, PCV-4-118 and PCV-4-119, are obsolete and '

require replacement. This package replaces the existing regulators with MECO Model P-1-DD Regulators. In addition, the piping and' support analyses have~been upgraded to consider seismic category II over I considerations.

Safety Evaluation:

The characteristics of the new regulator and gauges are identical to the existing components in quality, form, fit and function with the exception of areas which have been evaluated and found to be acceptable. These changes have been evaluated.and found not to alter the plant's design basis and to be bounded by the existing design bases, The support modifications will restore or increase the load capacity of the existing supports by replacing defective / missing material, and by installing new support components increasing the load carrying capacity of the support. The changes made have been evaluated and it has been determined that no new interference with existing systems, components, or the operation of any such systems or components are introduced.

The re-analysis /re-design of the piping / supports eliminates potential interaction of this non-safety related system with other safety related systems. i 1

79

g PLANT CHANGE / MODIFICATION 88-359 3 PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: '4/24/89 l' RJ RCCUMT"ATOR NITROGEN PRESSURE REGULATOR PCV-4-846 REPLACEMENT Ensparv:

The pressure regulator valve PCV-4-846 required repair, but was ,

' obsolete. This package replaced the existing Union carbide model R-1745 with a Victor.. model SR4G-580. As a result of the configuration-of the replacement regulator, a minor pipe routing change was required. This package also updates the' FSAR's information on this information.

Safety Evaluation:

PCV-4-486 serves no safety function. The change in pipe routing and-- the new _ regulator have been . valuated and- it has been determined that they have no adverse '.mpact on the safety related piping for containment isolation, will not result in a significant loading change to the - pipe supports and that the structural integrity of system is maintained. The replacement of the regulator and the update of plant documentation do not alter the plant's design basis and is bounded by. the existing -design analysis.

p 80

PLANT CHANGE / MODIFICATION 88-404 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 2/27/89 DEMINERAL73ER CONNECTIONS FOR THE COMPONENT COOLING WATER SYSTEM p_ummary:

The Unit 4 Component Cooling Water (CCW) system had a high concentration of suspended solids which were analyzed as iron and copper corrosion products. In order to remove this material ~from .

the_ system, demineralizers were temporarily _ connected to a side '

stream from the CCW system and functioned as a closed cleanup loop.

The CCW train to which the demineralized ~was connected remained operable. To allow future installation of the temporary demineralized system, the 2" supply and return line connections were analyzed and installed as permanent plant equipment.

The temporary cleanup loop was composed of supply and return lines from the CCW system, separate pumps and a Duratek anion / cation demineralized system. The supply line for the subsystem pump suction was an isolated CCW surge tank line. The return line from the demineralized was via a tap in the suction to the 4B Component Cooling - Water Pump (4P211B). The demineralized includes independent flow and pressure measurements and control to limit pressure to the design limit of 150 psig. The power source for the pump utilized non-vital 480V welding outlets available in the CCW Heat Exchanger area.

Safety Evaluation:

The supply and return configurations have been analyzed to meet the allowable stresses stated in the FSAR. The supply configuration connects to an out of service section of the CCW System which will not affect the remainder of the CCW system'in service. The return assembly is isolated from the CCW System by use of a check valve.

The unit shall be in either Mode 5 or 6 for operation of the temporary demineralized system.

81

i i

i-PLANT _ CHANGE / MODIFICATION 88-427 PC/M CLASSIFICATION: SR UNIT: 3 TURNED OVER DATE: 12/10/88 PRESSURIZER PORY AIR SUPPLY TUBING ENHANCEMENT Summary:

This package provided a modification to the Unit 3 Power operated Relief Valves -(PORV's) to enhance the ability of these valves to *.

meet-the. opening time requirements and to ensure reliability-by avoiding a single failure from preventing valve closure.

. Safety Evaluation:

.The overpressure mitigation system is described in the FSAR Sections 4.2 and 4.3. The modification to the air lines and valves has no adverse effect on the overpressure Mitigation-System (OMS).

The failure of.the PORV's is not one of the postulated initiating events of any design bases accident or event described in the FSAR.

.The modification of the Actuator Air and Nitrogen Supply System for the PORV's does not increase the probability of a malfunction of equipment as previously evaluated. This modification enhances the capability of the PORV's to meet their safety function to open upon receipt of an open actuation signal as analyzed. This modification

-does not adversely impact the reliability of the PORV's or add any new failure modes to the PORV actuation instrument air or nitrogen supply system.

The addition of the check valves and the utilization of the larger diameter tubing. for the air and nitrogen supply increase the reliability of the PORV Air Supply System, while maintaining the original design intent. The pressure test points ~ downstream of the nitrogen system pressure control valves provide points for attachment of equipment for the testing and calibration of the valves.- The isolation and bleed valves are normally closed, and have been designed to meet the system design conditions.

82

I PLANT CHANGE / MODIFICATION 88-442 I'

PC/M CLASSIFICATION: BR UNIT: 3 TURNED OVER DATE: 2/2/89 l

DEMINERALIZED CONNECTIONS FOR THE COMPONENT COOLING WATER SYSTEM i

Summary:

The Unit 3 Compohcat Cooling Water (CCW) system had a high concentration of suspended solids which were analyzed as iron and copper corrosion products. In order to remove this material from ,

the system, demineralizers were temporarily connected to a side-stream from the CCW system and functioned as a closed cleanup' loop.

The CCW train to which the demineralized was connected remained operable. To allow future installation of the temporary-demineralized system, the 2" supply and return line connections were analyzed and installed as permanent plant equipment.

The temporary cleanup loop was composed of supply and reeturn lines from the CCW system, separate pumps and a Duratek anion / cation demineralized system. The supply line for the subsystem pump suction were in isolated CCW surge tank line. Water supply will be via spilloser in the CCW surge tank. The return line from the demineralized will be via a tap in the suction to the 3B Component Cooling Water D.'mp (39211B) .The demineralized includes independent flow and pressure measurement and control to limit pressure to the design limit of 150 psig. The power source for the pump will utilize non v.ltal 480V welding outlets presently available in the CCW Heat Exchanger area.

Safety Evaluation:

The supply, return and alternate temporary return configurations have been analyzed to meet the allowable stresses stated in the Turkey Point FSAR. The supply configuration connects to an out-of-service section of the CCW system which will not affect the remainder of the CCW System in service. The supply line is isolated by the surge tank baffle and 3-701B. The return assembly and alternate temporary return assembly are isolated from the CCW System by using a check valve.

83

PLANT CHANGE / MODIFICATION 88-448 PC/M CLASSIFICATION: QR.

UNIT: 4 TURNED OVER DATE: 1/25/89 REPLACEMENT OF TUBING, FITTINGS AND ISOLATION VALVES ASSOCIATED WITH ICW PRESSURE INDICATORS Summary:

This package provided for replacement of the existing tubing, fittings and instrument valves associated with pressure indicators -

on the Intake Cooling Water (ICW) piping to the Component Coolir,g ~

Water (CCW) heat exchangers. The brass piping, fittings &

isolation valves are being replaced with stainless steel.

Safety Evaluation:

The replacement tubing and valves are installed on a dead leg to the pressure indicator, therefore, flow capacity and pressure drop have-no effect on the design function of the pressure indicators.

-These valves do not provide any throttling capability and are fully open during operation. Therefore, the change in size of these valves has no effect on the design function of the pressure indicators.

The piping class for this service is Class L per Specification 5177-PS-10. Specification 5177-PS-11 for Class L specifies stainless steel for small pipe 2" and under, therefore, stainless steel tubing is acceptable for this service. The replacement of existing brass piping, fittings and brass isolation valve with stainless steel components has negligible effect upon the equipment's weight.

The replacement instrument valve is rated for 3000 PSI at 100*F and 450*F maximum temperature. Since the design pressure and l temperature of the valve exceeds that of the ICW system to the CCW Heat Exchanger, the replacement valve is considered to be acceptable for this service.

84

(

\

I

t. -

[.

PLANT CHANGE / MODIFICATION 88-453 p

PC/M CLASSIFICATION: SR UNIT: 3 &4 TURNED OVER DATE: 12/7/88 t

UNIT 3 DRAWING DISCREPANCIES Summary:

. Several drawing discrepancies were noted during a walkdown of .

various' systems. This package corrected the-drawing discrepancies '

and added pipe. supports resulting~from the walkdown evaluation.

Bafety Evaluation:

The drawing- revisions provided in this package . correct the discrepancies - identified. .The discrepancies were evaluated and determined not to adversely impact. system . design bases.

Engineering analyses ' determined that pipe support modifications were required on two vent lines upon completion ~ of -the modifications provided within this engineering package, the affected piping ~will meet the requirements of Appendix 5A of the Turkey Point FSAR for Class I structures, systems and equipment.

o 85 c_=___

PLANT CHANGE / MODIFICATION 88-521 PC/M CLASSIFICATION: SR UNIT: 3 &4 TURNED OVER DATE: 4/30/89 UNIT 3 AND 4 DRAWING UPDATE FOR SYSTEM 023 EMERGENCY DIESEL GENERATOR Summary:

During a walkdown, various discrepancies were found on the Diesel :

Generator System operating diagrams. There were items common to several drawings. Some di;crepancies were drawing specific and others were documentation discrepancies only. The package evaluates and corrects identified discrepancies.

Safety Evaluation:

Items evaluated were found to be acceptable. Based on this evaluation, these discrepancies were determined not to adversely impact plant systems, structures or components. Further, the I drawing changes will not alter the Plant's Design Basis and are bounded by existing design analyses.

l l

86

[y'g s;'(, ,

L PLANT CHANGE / MODIFICATION 88-527 PC/M CLASSIFICATION: BR UNIT: 3&4 i TURNED OVER DATE: 4/27/89 RESOLUTION OF DRAWING CHANGES ASSOCIATED WITH 5610-T-E-4503 33mmary:

This package resolves discrepancies in drawings and procedures and - ~

the as-built condition of the plant.

Safety Evaluation:

All items evaluated were found to be acceptable. It can be concluded from this evaluation that the drawing changes provided by this package-will not alter the plant's design basis and are bounded by existing design analysis. Further, there will be no adverse effects.on the plant's systems, structures or components.

87

5 l

i PLANT. CHANGE / MODIFICATION 88-528 i PC/M CLASSIFICATION:- SR I

UNIT: 3 &4 TURNED OVER DATE: 5/1/89 -

DRAWING UPDATE 5610-T-E-4065, SHEET 1 AND SHEET 2, LUBE WATER AND' CIRCULATING WATER SYSTEMS  !

Summary:

This package resolves discrepancies between drawings and the as ,

' built condition of the plant.

33fety ' Evaluation:

Items evaluated were found to be acceptable. Based on this evaluation, these discrepancies were determined not to adversely impact plant systems, structures or components. Further, these

. drawing changes will not alter the plant's design basis and are bounded by existing design analyses.

l-l 88

I' ,

L PLANT CHANGE / MODIFICATION 88-530 l PC/M CLASSIFICATION: SR UNIT: 3 &4 TURNED OVER DATE: 4/27/89 4 l l 23EAKER LIST UPDATING FOR NON-CONFORMANCE REPORTS (NCRs) AND '

REOUEST FOR ENGINEERING ASSISTANCE (REAs)

Summary 8 This package provides documentation to update the Breaker List drawings to correctly depict the.as-installed plant conditions.

Numerous Requests for Engineering Assistance (REAs). and Non-Conformance Reports (NCRs) have been_ generated in the past several years identifying drawing discrepancies between the Breaker List (5610-E-855) and other design documentation or as-built plant conditions. The majority of the NCRs were generated as a result of a comprehensive walkdown of all plant electrical distribution panels in order to provide the as-built verification of the Breaker List.

Safety Evaluation:

Each of the specific problems and/or non-conforming conditions which necessitates a Breaker List revision for vital distribution panels- were evaluated with respect to technical adequacy and r,tential operability concerns. Only one item represents a technical or functional non-conformance between design documentation and as-built plant conditions. This problem was evaluated and determined to be acceptable. The remaining items are minor in nature involving only design documentation changes and not functional non-conformances between plant design and as-built conditions. All of the problem items and corresponding Breaker List corrections are technically acceptable and do not involve a potential operability concern. None of the drawing discrepancies and/or plant conditions have an effect on a component or system that could impact plant safety. -The Breaker List revision is bounded by existing design analysis and does not alter the design basis of the plant.

89

PLANT CHANGE / MODIFICATION 88-533 PC/M CLASSIFICATION: SR UNIT: 3 &4 TURNED OVER DATE: 4/28/89-UNIT 3 FED 4 DRAWING ~ DISCREPANCIES ON DRAWING 5610-T-E-4510,~ SHEETS 1&2 Summary:

~

This package resolves discrepancies between drawings and the as built condition of'the plant.

Safety Evaluation:

Items evaluated were found to be acceptable. Based on this evaluation, these discrepancies were determined not to adversely impact plant system, structures or components. Further, drawing changes will not alter the plant's design basis and are bounded by.

existing design analyses.

1 l

l 90 l.

L -

f PLANT CHANGE / MODIFICATION 88-534 PC/M CLASSIFICATION: SR UNIT: 3 &4 TURNED OVER DATE: 4/28/89 DRAWING DISCREPANCIES ON 5610-T-E-4534 SHEETS 1 AND 2 - CONTAINMENT VENTILATION SYSTEM Summary:

This package resolves' drawing discrepancies identified - via a'

. walkdown of the Containment Ve tt116 tion System and a review of Operations Diagram 5610-T-E-4536, Sheets 1 and 2.

Safety Evaluation:

Based on the evaluations performed, the discrepancies were determined not-to adversely . impact plant system, structures or components. Further, the drawing changes were determined not to-alter the plant's design basis and are bounded by existing design analyses.

i j

I i

j 91

PLANT CHANGE / MODIFICATION 88-535 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 3/2/89 PRESSURIZER PORY AIR AND NITROGEN SUPPLY TUBING ENHANCEMENT Summary:

This package provides for modification of the existing Power Operated Relief Valves (PORV) actuator covers and replacement of ,

the pressure regulators and indicators, relief valves with rupture ,

discs, and increases the size of the solenoid valves, tubing, in-line valves, and fittings for the air and nitrogen supply lines to enhance the opening times for the PORVs. In addition, an accumulator will be installed in the PORV actuation supply system.

A bypass with an in-line check valve will also be installed around the solenoid valve closest to the PORV, to assure that a single failed solenoid valve will not prevent a PORV from closing. In addition, all manual valves will be relocated from inside the pressurizer cubicle.to outside the pressurizer cubicle to enhance access and system maintenance.

Safety Evaluation:

The modifications of the PORV actuator supply enhance the ability of the PORVs to perform their required safety function and do not change the function or the operation of the valves. This modification does not adversely impact the reliability' of the PORV's or add any new failure modes to the PORV actuator instrument air or nitrogen supply system. The replacement of the nitrogen pressure regulators, the utilization of the larger diameter tubing for the Instrument Air (IA) and nitrogen supply, and the addition of the check valves increase the reliability of the PORV actuation supply system, while maintaining the original design intent. The new or replacement components have been procured to the appropriate L quality level, seismically qualified as required, and installed in accordance with the requirements of the FSAR, Appendix 5A.

I 92

PLANT CHANGE / MODIFICATION 88-536 PC/M CLASSIFICATION: BR UNIT: 4 TURNED OVER DATE: 4/1/89 SAFETY INJECTION ACCUMULATOR LEVEL TRANSMITTER REPLACEMENT Bummary:

This package provides for replacement and relocation of existing Safety Injection (SI) Accumulator Level transmitters LT-4-920, LT-4-922, LT4-4-924, LT-4-926, LT-4-930. This package also provides ;

for the rerouting and resupporting of the tubing associated with PT-4-925 to preclude interference with the support for LT-4-924.

The replacement transmitters will have a new range with a change in span which involves a change in the Safety Parameter Display System (SPDS) measured range and indicated range on the control room indicat' ors.

Safety Evaluation:

Based on the field walkdown which considered these changes and the  !

required spatial envelope, the changes in support length, cross section, and mounting location of the level transmitters are i considered acceptable. The spatial envelope has been reviewed and I these changes will not affect other structures, systems and I components.

As discussed in the engineering evaluation, the function and fit; the weight of the new supports, including loads due to the {

attachment of the new level transmitters, tubing, and conduit; were 1 considered in the design for the support. Since the support design {

meets the seismic requirements for Class I structures of the FSAR, I Appendix 5A, the changes in support loads are considered acceptable.

No new interferences with, or adverse effects on, existing plant I systems, structures, or components are created by this modification. The changes do not alter the plant's design basis I and are bounded by the existing design analysis.

l

{ 93

j l 1 88-584 i L. PLANT CHANGE / MODIFICATION PC/M CLASSIFICATION: SR UNIT: 3 E. 4 .,

TURNED OVER DATE: ' 4/29/89-DRAWING REVISION FOR LIOUID WASTE DISPOSAL SYSTEMS OPERATING DIAGRAM Summarvt This package revised drawing 5610-T-E-4518 Sheet 2 to indicate

1) Deletion of Flange Leak Detection System from the drawing *,

and addition of flags notifying this system is on Drawing 5610-T-E-4501 Sheet 1

2) A capped connection just downstream of the- branch 'for valve 3-4670 and correct type for valve 4668D, and

3) The correct- locations of - the -level controllers and indicators on the Laundry Tank Level Loop.

This package also adjusted the Laundry Tank's level indicators setpoints.

Safety Evaluation:

The changes made are either administrative in nature or do not adversely change the design of the components affected. The

. evaluation determined that no unreviewed safety question results from these changes.

94

1 g

PLANT CHANGE / MODIFICATION.88-586 PC/M CLASSIFICATION: BR.

UNIT: 3 &4 TURNED OVER DATE: 5/3/89 CONTROL ROOM HVAC T-E DRAWING UPDATE Summary:

This package resolves drawing discrepancies identified in a Non-Conformance Report. It revises the Control Room HVAC T-E Drawing-.

(5610-T-E-4635, Sheet 1) to reflect the as-built condition.

Safety Evaluation:

These discrepancies were deaermined not to adversely impact plant systems, structures or components. Further, these drawing changes will not alter the plant's design basis and are bounded by existing design analysss.

95

PLANT CNANGE/ MODIFICATION 88-597 PC/M CLASSIFICATION: NN8R UNIT:- 3 &4 TURNED OVER.DATE:. 4/27/89 L

DRAWING DISCREPANCIES ON DW6 5610-T-E-4063, SHEET 1 SunuB8ry I

This package corrects, drawing discrepancies on drawing 5610-T-E-4063 Sheet 1. . .This ~ package deleted a Y strainer shown . located in -

a. drain line from the condenser, deleted a vent. valve shown on the ^'

condenser side of the Steam Jet Air Ejector (SJAE) , .and changed the location shown"for a. local temperature Indicator on:the SJAE for both units.

Safety Evaluation The changes were. evaluated and it was determined the package did-not affect-any nuclear safety related functions of the plant or adversely impact.the system design bases.

l'

.~

l, i

)

96 i

_:-_-_--_--_______---__---______ . _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ - _ _ - _ _ - - _ _ - _ _ _ _ _ _ _ --_______-_-__-____-____-_-____-_____-_____-____J

PLANT CHANGE / MODIFICATION 88-606 1

1 PC/M CLASSIFICATION: NN8R-UNIT: 3&4 TURNED OVER DATE: 5/1/89' -1 1

n ' DRAWING REVISION FOR CONDENSATE SYSTEM (5610-T-E-4062 SH 1)

I summary- i This package resolves discrepancies between the drawing and the as- .i built condition. .

L.

Safety Evaluation:

These_ discrepancies corrected by this drawing change were determined not to adversely impact plant systems, structures or components.- Further, these drawing changes will not alter the plant's design basis.

97 L _ _ . - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

PLANT CHANGE / MODIFICATION 89-029 PC/M CLASSIFICATION: QR UNIT: 3 &4 TURNED OVER DATE: 5/1/89 SPENT FUEL PIT (SFP) LEAK RECOVERY SYSTEM AS-BUILT VALVE TAGS Summary:

This package provides the documentation to incorporate drawing discrepancies identified and install tags on valves-in the Spent Fuel Pit (SFP) Iaak Recovery System. '.

Safety Evaluation:

The discrepancies identified and corrected were for components located outside of the Quality and Safety boundaries of the SFP Leak Recovery Syston. These discrepancies were evaluated and determined not to' adversely impact the system design bases or the operation of the plant.

l l

l .

98 f.

I

b; I

PLANT CHANGE / MODIFICATION 89-091 j PC/M CLASSIFICATION: QR UNIT: 3&4 TURNED OVER DATE: 4/27/89 PRIMARY WATER AND DEMINERALIBER WATER SYSTEM DRAWING UPDATE. 5610-T-E-4531 Summary:

The package evaluates and corrects drawing discrepancies fr.c Units 3 and 4 identified in a Nonconformance Report (NCR) and by plant -

walkdowns. . Where appropriate, Drawing 5610-T-E-4531, Sheet 1, has been updated to reflect actual plant conditions.

Safety Evaluation:

None of the items identified in the NCR affects plant operability

, or safety. They correct discrepancies between the drawing -and field conditions and clarify the information on Drawing 5610-T-E-4531, Sheet 1.- Other drawing changes identified in this package are still being. evaluated by Engineering.

L 99

PLANT CHANGE / MODIFICATION 89-100

~PC/M CLASSIFICATION: QR UNIT: 4 TURNED OVER DATE: 5/1/89 MAIN STEAM DRAWING UPDATE OF DRAWING 5610-T-E-4061, SHEET 1 OF 4 Suamary1 This package identifies and incorporates the corrective actions-of a Non-Conformance Report (NCR) to 5610-T-E-4061, Sheet 1 of 4, Rev. .

.56, based on "as built" verification to P&ID 5610-M-1, Rev. 45 for -

Unit #4 (outside containment), Main Steam System. Correspondingly,-

other applicable plant documents were reviewed and respective update forms initiated.

~

Safety Evaluation The items evaluated by che package were determined to be acceptable. Based on the evaluation, these discrepancies were determined not to adversely impact plant systems, structures or components. Further, these drawing changes will not alter the plant's design basis and are bounded by existing design analyv ,

~

4 100

f '

t 89-105 PLANT CHANG 3/ MODIFICATION PC/M CLASSIFICATION: SR UNIT: 44 TURNED OVER DATE: 5/17/89 DRAWING UPDATE FOR THE CHEMICAL'AND VOLUME CONTROL SYSTEM Summary:

This package addresses discrepancies in the drawings for . the

-Chemical and Volume Control Syntem (CVCS) identified . -in a.Non ,

Conformance Report (NCR) and two Requests for Technical Assistance -

(RTAs). Drawings 5610-T-D-19, theet 1, and 5610-T-E-4505, Sheet 3, were . updated to correct valve tagging and valve failure position.

Safety Evaluation:

The drawing and document revisions provided in this package correct the drawing discrepancies identified in the RTAs and-the NCR. The discrepancies have been evaluated and determined not to adversely impact the FSAR design basis.

101

L l

PLANT CHANGE / MODIFICATION 89-127 PC/M CLASSIFICATION: BR UNIT: 4 TURNED OVER DATE: 4/25/89 l  !

UNIT 4 RESIDUAL HEAT REMOVAL fRHR) SYSTEM NO. 50 AND SAFETY INJECTION (SI) SYSTEM NO. 62 DRAWING UPDATE FOR DRAWING 5610-T-E-4510, SHEET 1 AND 2 Sumsc ri -

This package is issued to resolve drawing discrepancies identified, 1

- and to update the included drawings to depict the plant as-built condition.

Safety Evaluation With the exception of those items .needing further vendor information and therefore requiring further engineering review, all other items were evaluated and found to be acceptable. Based on this evaluation, these discrepancies were determined not' to adversely impact plant systems, structures or components. Further,

. drawing changes will not alter the plant's design basis and are bounded by existing design analyses.

I i

102

PLANT CHANGE / MODIFICATION 89-139 PC/M CLASSIFICATION: BR UNIT: 4 TURNED OVER DATE: 4/30/89 DRAWING DISCREPANCIES ON DRAWING 5610-T-E-4512, SHEETS 1 AND 2,

" COMPONENT COOLING WATER SYSTEM" Summary:

This package resolves discrepancies between the drawings and the .-

as-built condition of the plant.

Safety Evaluation:

The discrepancies evaluated dere found acceptable to the requirements of the system design bases. The evaluation demonstrates that the (*.rawing revisions and tagging of equipment do. not affect the fur.ctional requirements of the system as described in the FSAR. These changes do not alter the Plant's Design Basis and are bounded by the existing design analysis.

Further, there will be no adverse effects on plant systems,

- structures or components.

a W

e 103 4w--_-- --~_ -- . _ _ . .

PLANT CHANGE / MODIFICATION 89-264 PC/M CLASSIFICATION: SR

-UNIT: 4 TURNED OVER DATE: 4/12/89 REPLACEMENT OF YARWAY CONTAINMENT ISOLATION STOP VALVF8 Summary:

This package evaluates replacement of a 1 inch Yarway Figure 5551B-F315 valve with a Rockwell Model B3627KT4-F316 ' valve in Unit 4 -

valve location 945E, Nitrogen Supply te Accumulators Check Valve. '

Safety Evaluation:

The existing . valve is a stop check type design valve and the replacement is a check valve. The function of a check valve is all that is required in order for valve 4-945E to meet its safety related function of containment isolation. .The stop valve type

- function offered by.the design of the existing valve was used for maintenance or special testing and is redundant to that provided by other valves in the line and therefore is not required. It was determined that the replacement model valve is an equivalent replacement for the original valve, meeting or exceeding the

- requirements of the original valve.

l l

L 104 9

PLANT. CHANGE / MODIFICATION 89-373 PC/M CLASSIFICATION: SR UNIT: 4 TURNED OVER DATE: 6/5/89-RQMM5T EOUALIIING LINES FOR MOV-4-750 AND MOV-4-751 Summarvt.

This package removed the- lower set of valve packing and installed bonnet equalizing lines from the valve packing gland leak-off -

connection of both valves MOV-4-750 and MOV-4-751 to-the Reactor "

Coolant System (RCS) side of the valve. For MOV-4-750, this source is the RCS bypass loop manifold instrument header at drain valve 4-562A. For MOV-4-751, this is the instrument line at valve'4-750C. Therefore, the bonnet will be at the same or-lower pressure than-that associated with the high pressure side.

Safety Evaluation:

The ~ . changes meet the design, material and construction codes,.

identified in the evaluation, for the RCS presuure and temperature design limits-of 2485 psig and 650*F. The equalizing line shall be constructed of stainless steel which is compatible with the -

Residual Heat Removal (RHR) and RCS systems. As'a result, the line does' not interfere with the capability of maintaining the RCS system integrity. Furn ermore,. the addition of the equalizing line does not alter the func.ionality of the MOVs.

The changes do not impact the safety related, quality related, nor non-safety related functions of MOV-4-750 and MOV-4-751. The isolation interface between the RCS and RHR is maintained. The tubing is seismically supported as e Class I component to ensure

- safe operation of the valves and to prevent interaction with other safety related components.

-)

105

- j)

Im __.a - __ .__ __ _ . _ _ _ _ . _ _ ~_

l w .j

r. >

. PLANT CHANGE / MODIFICATION'N '89-396 l i

'PC/M CLASSIFICATION: SR UNIT:- 4 TURNED OVER DATE: 6/16/89-RCS RTD MANIFOLD FLOWRATE CHANGE SummarYi This change-reduced the Reactor Coolant System (RCS) RTD manifold flowrate1to allow operation with the.flowrate achievable in the field. . This modification-is classified as safety related as the .

change affects the Reactor Protection System. This only modification involved drawing and procedure changes.

Safety Evaluation:

Each change-identified was appropriately evaluated with respect to the system design basis above and found to be acceptable. Included in'this~ evaluation was a review of the effects of the increase in the. overall response time for the protection' functions that use temperature measurement.by the narrow range RTDs. The conclusions

. of the FSAR were found to remain valid. The evaluation established

-design equivalence and conformance-to the original design basis.

106

2;. e <

) .I. I

\

,3 l- -

f

.)

f

'ii-SECT.',0N . 2 changes to the . facility or procedures as described' in 'the Safety .'

An< lysis Report not performed by a PC/M, and tests'or experiments not.

described in the Safety Analysis Report.

'I a$ y2

\ < +

e 107

_..__..._.__________m._._ _______. _ _ _ _ . _ _ _ _ _ __

L r

r l SAFETY EVALUATION: JPE-LR-87-038 Revision 0 INTERCONNECTING 125 VDC. BUSSES 3B AND 4A WHILE BATTERY 3B(4A) IS RENOVED FOR TESTING AND THF 'THITS ARE AT FULL POWER.

- The existing and proposed echnical Specification Surveillance Requirements-for the 125V DC System for Turkey Point Units 3 and 4 require each Station Battery to be load tested annually. Battery load test re pires removal of a battery from service which invokes

. a 24 Limiting : Condition for Operation (LCO) . Although both units are allowed by Technical Specifications to continue operating with one battery out of service, the preferred method is to keep the bus associated -with the battery under. test energized by cross connecting with'an energized bus of the other unit. The purpose of this evaluation is to demonstrate the l acceptability of interconnecting DC Busses 3B and 4A during the 24-hour LCO period

when Battery 3B(4A) is removed for load test or maintenance..

Safety Evaluation Summary:

This interconnection does not constitute an . unreviewed safety '-

question for the following reasons:

The removal of a battery from service invokes a 24-hour DC System LCO. While Busses 3B and 4A are interconnected during the LCO period, a single failure is not postulated.

No new equipment or component has been adaed to the system..

Existing installed equipment will be reconfigur~1 for interconnection of DC busses during the annual load .f.-t for a period not exceeding the DC System LCO of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The capability of Battery 3B(4A) and associated system hardware / equipment to supply the loads of both DC busses during a design basis event (SI concurrent .rith LOOP) has been demonstrated to be acceptable.

Although Battery 3B(4A) will be removed frem service, the equipment on its associated bus will be kept energized by interconnecting Busses 3B and 4A and allowing Battery 4A(3B) to supply the combined loads. Therefore, the DC system will function as normal while the busses are interconnected.

, Issued: September 24, 1987

~

The following Temporary Procedure was issued during this reporting period as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

TP-496 Removal and Return to Service 3B 125V DC Battery for Load

- Test 108 -

_m __-____ms.-_. _____m. -

o f

SAFETY EVALUATION: JPES-E-87-2484 Revision 0 THE INDIVIDUAL CELL EQUALIZING CHARGE AND ELECTROLYTE LEVEL CORRECTION-3A, 3B, 4A, AND 4B A review of the monthly maintenance records on station battery '4B' indicated' that over the period of several months, the float voltages of a few cells have drifted downward to the minimum acceptable value of 2.13 volts. Also, since the cells were never completely filled with electrolyte to the high mark when new, specific gravity readings are at the lower end of the specification when calculated for level correction.

During the monthly surveillance an equalizing charge to the entire battery is given, however the attempts to elevate the voltages of the weak cells are not succeeding due to the limitation of 138 volt maximum operating voltage permitted at the battery terminal.c' (without isolating the battery from the respective DC bus).

The purpose of this evaluation is as follows:

Provide guidelines to evaluate the low voltage readings of the individual battery cell by individual. cell equalizing method as recommended by the battery manufacturer, GNB, Incorporated.

Provide guidelines to correct the electrolyte level of the low specific gravity cells on a one time basis after performing the individual cell equalizing or on completion of the scheduled monthly equalizing of the entire battery.

Egf.ety Evaluation Summary:

Performing this activity would not result in an unreviewed safety question for the following reasons:

The portable battery charger will be temporarily connected in parallel to the individual cell of the battery. During this process should a design basis event occur, the 125 VDC battery system has adequate capability to supply the vital DC loads without any adverse effect.

The potential failure modes of the portable battery charger have been reviewed and the effects of such failures will not adversely affect the performance of the safety related battery system. i The 125 DC System, with Class 1E batteries operational along with the associated battery chargers and DC buses, is unaffected during the equalizing process and therefore, the margin of safety as defined in the bases for any technical specification has not been reduced.

109

l L No new equipment ' or component (except the temporary use of a

.]

portable battery charger) has been added to the system. Existing 1 installed equipment is not required to be reconfigure 'for the .

. individual cell equalizing. The operation or function of the plant  ;

~

safety. system is not altered. .

The . battery 'and associated Class 1E battery charger and DC bus-remain energized to supply the vital DC loads. The portable

' charger is to be located and seismically supported outside tl.a respective. battery rooms. Therefore, the integrity, operation or function of the 125 VDC Battery System and other safety related systems addressed in the Technical Specifications, is not affected.

Issued' November 17, 1987 The following procedure was issued during this reporting period ac allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

0-PME-003.16 Individual Cell Equalizing Charge for Station '-

Batteries 3A, 3B, 4A, and 4B 110 l

I

'8AFETY-EVALUATION; JPE-PTN-SELJ-38-030' Revision 1 BATTERIES 3A, 3B, 4A, AND 4B MINIMUM TERMINAL VCLTAGE HIGHER THAN 105VDC

'This evaluation provides the basis for continued operation ' for circuits requiring minimum battery terminal voltage higher than the FSAR minimum of 105vdc from batteries 3A, 3B, 4A, z and 4B.

Referenced documents provided acceptability for continued operation 1 for certain 4.16KV breaker closing circuits and for motors of AFW motor operated valves. With a minimum battery terminal voltage of 105vdc as specified in' the FSAR, the voltage drop in these circuits resulted in lower than the vendor minimum specified voltage required to operate the equipment.

In addition to higher than minimum battery terminal voltage requirement, other items such as battery load profile and battery loading were addressed in the referenced documents. Between the period of issuance . of the referenced documents and the present, .

modifications were implemented for some circuits to meet the FSAR '

minimum battery terminal voltage requirement and the FSAR was updated to correct the battery load profile and the loading data.

This Safety Evaluation consolidates into a comprehensive document all the remaining circuits addressed in the existing documents that have not yet been corrected thorough plant modifications. This comprehensive Safety Evaluation will therefore provide-acceptability for continued operation until plant modifications have been implemented for the remaining circuits (PCM 88-094 for Unit'3 and PC/M 88-095 for Unit 4).

Safety Evaluation Summary:

Based on test data and vendor data, it has been shown in that 4.16KV breaker closing coils rated at 125vdc are capable of operating at 80Vdc minimum pickup voltage. Calculations and Sefety-Evaluations have shown that with the battery load profile daring accident scenarios and battery load margin available, that the battery minimum terminal voltages are above that which is required to operate the 4.16KV breaker closing coilc. Therefore, safety related batteries 3A, 3B, 4A, and 4B are capable of providing sufficient terminal voltages for the operation of. equipment necessary for the different accident scenarios evaluated in the FSAR. Therefore, the probability of an occurrence of an accident is not increased, the consequence of an accident is not increased, or the malfunction of equipment important to safety previously avaluated in the FSAR is not increased.

The safety related batteries minimum battery terminal voltages are above the minimum required to operate the safety related equipment as shown above. Therefore, the possibility of an accident or a 111 i

i

g 1

E-L p

malfunction of a different type than any. ' evaluated previously in:

i- 'the FSAR is not created.

Technical Specification bases requires that the batteries

. surveillance,-testing, maintenance r and operation ensures that-the batteries are at'ful1~ charge, to ensure-that adequate DC power'is available 1 for ' emergency use in anticipation of less-of-AC-power "q  : incident. Calculations show that.with the battery load profile,

- worse case cell electrolyte temperature for a battery less than'14 years old,- the battery terminal voltages will be above that -l required to operate the safety-related equipment during accident' scenarios. Therefore, the margin of safety as defined in the basis

for the technical specifications is not reduced.

Issued: September 13, 1989 i-s

'A i s

112 '

SAFETY EVALUATION: JPE-PTN-SELS-88-031 Revision 0 REMOVAL OF AIRBORNE CONTAMINANTS USIVG THE EMERGENCY CONTAINMENT FILTERS The only mcans currently available to reduce the airborne radioiodine concentrations in containment, specifically prior to personnel entry, is to purge containment atmosphere to the environment. This iodine removal function can be performed by the Emergency Containment Filtering System (ECFS). By using the ECFS in lieu of or in combination with the containment purge, both the quantity of radioactivity released to the environment and the cleanup time can be reduced.

Safety Evaluation Summary:

System components are not subject to rapid deterioration, having lifetimes of many years, even under continuous flow conditions.

Visual inspection and operating tests provide assurance of the system reliability and will insure early detection of conditions ~

which could cause the system to fail to operate properly. The performance tests prove conclusively that filters have been properly installed, that no damage or deterioration has occurred, and that all components and subsystems operate properly. The tests are performed in accordance with the methodology and intent of ANSI N510 (1975) and provide assurance that filter performance has not deteriorated below required specification values due to aging, contamination or other effects.

Operation of the Emergency Containment Filtering System for containment atmosphere cleanup during normal operation, provided the applicable Technical Specification surveillance are satisfied, will not affect the ECFS credit assumed in'the accident analyses presented in Chapter 14 of the FSAR. Further, the present Technical Specifications remain valid and the ECFS design basis l

presented in Section 6.3.1 of the FSAR would not be altered and would still be bounding.

Issued: August 17, 1988 113

r. 1 l

I L -SAFETY EVALUATION: JPE-PTN-SEMS-68-039 Revision 1 TEMPORARY SYSTEM ALTERATION TO FILTER FUEL OIL ~IN DIESEL FUEL OIL l STORAGE TANK 1

)

The current and Interim Technical Specifications require that the diesel fuel oil storage tank be periodically sampled to verify that fuel viscosity, water e.nd sediment levels are within acceptable limits.- Recent sample results- indicate that sediment concentrations are acceptable but close to the limit. A temporcry filtering system will be used to reduce the sediment concentration to as low as practical to ensure that the acceptable limit is not exceeded. The temporary filtering system will consist of a pump, filters,-valves and hoses to connect the system to existing drain fill connections on the diesel fuel oil storage tank. This temporary filtering system will be installed under (Temporary System Alteration 88-23-13) and operated using a temporary operating procedure.

Safety Evaluation Summary:

~

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR have not been increased. The safety function of the diesel fuel oil storage tank has not been affected since there is sufficient time for an operator to isolate the temporary filter system from the tank. In addition, the connection of the temporary filter system to the tank will not adversely impact the tank itself.

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

Measures will be taken to prevent draining of the tank in the event of equipment failure or damage caused by movement of the trailer assemble.

The margin of safety as defined in the basis for any Technical Specification has not been reduced since the temporary filter system will not adversely impact the availability of the required 7 day supply of emergency fuel or the cleanliness of that fuel.

In addition, the temporary system will not adversely affect any equipment addressed in the basis for the Technical Specifications.

Issued: August 17, 1988 114

I SAFETY EVALUATION: JPE-PTN-SENJ-88-040 Revision 0 l

.LPT ROTOR PRE-INSTALLATION PERFORMANCE EVALUATION

Test Procedure TOP-464 involves the performance testing of the secondary . plant to determine the baseline heat balance for comparison to the heat balance to be~ . determined after' the installation of Westinghouse fully integral low pressure. turbine rotors. This testing is required to verify the Westinghouse contract guaranteed performance increase of 8.1 MWe due to the modifications.

~

~The valve alignments for this testing are required to provide a.

controlled steam and feedwater flow path that can be reproduced for the post-modification testing. Steam generator blowdown and Unit 4 auxiliary steam will be isolated. Auxiliary steam from Unit 3 will be supplied, if available. The secondary plant will be monitored closely during the procedure to insure system stability and proper operation.

Safety Evaluation Summary: -

This procedure will affect the operation of limited secondary systems. (The steam generator blowdown and Unit 4 auxiliary steam supply will . be isolated) . However, safe operation of the plant systems, as described in chapter 14 of the FSAR, will not be

impaired. The system alignments and test connections do not adversely. affect the operation of the secondary plant systems and-equipment used for safeguard functions will remain operable at all

-times during the test. The temporary system alterations due to this procedure-. do not affect the Turkey Point Technical Specifications.

Issued: August 26, 1988 115 l

l

. _ _ _ _ _ _ _ _ _ _ _ _ .__-_-___-____ _ __ D

' SAFETY EVALUATION: JPE-PTN-SEMS-88-041 Revision 1 CHEMICAL CLEANING OF THE TUBE SIDE OF THE COMPONENT COOLING WATER HEAT. EXCHANGER To clean the tube side of the Component Cooling Water (CCW) heat exchanger, the tube side of the CCW heat exchanger will be isolated from the. Intake Cooling Water (ICW) system concrete lined piping by installation of a temporary plug in the channel head nozzles as well as by the normal isolation valves. The channel head vent and drain valves will be used for supply and return points for the chemicals. The pressure in the CCW heat exchanger tube side will be monitored and maintained less than the design pressure of the pipe plug to prevent the cleaning fluids from entering the concrete lined pipe of the ICW sy6 tem.

Safety Evaluation Summary:

The chemical cleaning process will be used to maintain the CCW heat .

exchangers so that they will be capable of removing their design ~

basis heat load. The safety evaluation demonstrated that the cleaning process will not degrade the pressure boundary integrity, structural integrity, or performance of the CCW heat exchanger or any other safety related components.

Issued: August 28, 1989 116

l SAFETY EVALUATION: JPE-PTN-SEEJ-88-042 Revision 0 DE-ENERGIZATION OF UNIT 4 4160 VOLT SAFETY RELATED BUSES The de-energization of'a 4160 volt safety related bus at Turkey Point Plant during a single unit outage is sometimes Lecessary to allow the periodic maintenance, testing, or design modifications of the 4160 volt switchgear. 'De-energization of a'4160 volt bus would cause de-energization of the 480 volt load centers and motor control centers fed from that bus and a loss of power to equipment which may be required to maintain cold / refueling shutdown, perform outage related activities, or sOpport operation of the opposite unit. This condition can be. alleviated by closing the tie-breakers bet'9en opposite train load centers while a 4160 volt bus is de-energized.

Both 4A and 4B 4160 volt safety related buses may be de-energized (non-concurrently) during a Unit 4 outage while Unit 3 is in operation. In addition, the corresponding train Emergency Diesel Generator (EDG A-or EDG B) may also be taken out of service .

concurrent with the 4160 volt bus de-energization. During the 4160 '

volt bus de-energization, the associated 480 volt load centers will be kept energized by closing the tie-breakers to the opposite train load centers.

Safety Evaluation Summary:

The de-energization of a Unit 4 4160 volt safety related bus with or without removal from service of an Emergency Diesel Generator (EDG) while Unit 4 is in cold or refueling shutdown and Unit 3 is in power operation is allowed by Interim Technical Specifications for a period of seven days provided that Technical Specification 3.8.1.1 action item b is performed. The availability of Unit 4 480 volt equipment will be increased during 4160 volt bus de-energization by cross-connecting opposite train Load Centers (LCs) .

The breaker positions specified and plant restrictions imposed will ensure the LC cross-connection does not adversely affect the normally aligned electrical distribution system. Prior to de-energizing a 4160 volt bus, Unit 4 will be placed in cold or refueling shutdown which limits the accidents analyzed in the FSAR to fuel handling, loss of spent fuel cooling, and loss of AC power accidents. These accident analyses are not affected because the Unit 4 Spent Fuel Pit Pump (SFPP) will remain energized during bus de-energization and there- is no impact on any fuel handling equipment. In addition, the probability of occurrence of a total loss of AC power is not increased since the proposed configuration is allowed by technical specifications. Although Unit 3 will be in power operation during the proposed Unit 4 electrical configuration, the failure or de-energization of any component associated with Unit 4 cannot increase the probability of an accident on Unit 3. The loading of the operable EDG with the cross-connection of the Unit 4 LCs while a 4160 volt bus and the 117 9

_m. - - . _ _ _ _ - - . _ _ _ _ . _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - -

a

1. 1 -: .

,Y. -

[ ,

i associated EDG~was ou't of service ~was. evaluated and: determined to (be. acceptable.

" The: cross-connection of Unit 4'LCs while a 4'160 volt bus.and EDG of-l.the same train is - out'. of service; . ensures that , systems, structures, or. components: required to mitigate.the. consequences of an . accident on' either Unit 3 or.147 will . perform their ~ intended -

safety functions', 'provided the' breaker positions specified, as applicable, and plant restrictions outlined are observedLduring

the proposed Unit 4 configuration.

Issued: October,14, 1988' s '.

118-

1 I

l SAFETY EVALUATION: JPE-PTN-SEEJ-88-047 Revision 0 BACKFEED OF POWER THROUGH MAIN AND AUXILIARY TRANSFORMERS During an upcoming unit outage it will be necessary to perform extensive maintenance on the startup transformer associated with that unit. In order to provide an equivalent path between the switchyard and the 4160 V distribution system it has been proposed to backfeed power through the main and auxiliary transformers. To successfully operate in this configuration it is necessary to

. hysically disconnect the main generator and disable the generator protective relaying. The disconnection of the generator removes the grounding reference requiring the establishment of a separate ground detection scheme.

Safety Evaluation Summary:

The backfeed of power through the main and auxiliary transformers makes use of equipment whose reliability is demonstrated during -

normal operation. The reliability of the main and auxiliary "

transformers is comparable to that of-the startup transformer.

Since the temporary modifications associated with the backfeed operation will be performed on the unit after it has reached a cold shutdown condition and is independent of the operating unit, there will be no effect on the accident initiating events for the operating unit. As demonstrated, the main and auxiliary transformer combination, is capable of performing the startup transformer function during shutdown; therefore, the normal and emergency power requirements for equipment required during shutdown and refueling conditions is ensured. All equipment required for .

mitigation of an accident on e3thr the shutdown or operating unit will-be available.

Issued: September 14, 1989 The following Temporary Procedure was issued as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

TP-475 Temporary Operational Procedure for the Backfeed of 4A and/or 4B 4160V Buses via Main Transformer No. 4, Aux. Tx. No. 4, and 22kV Isophase Bus 119

BAFETY EVALUATION: JPN-PTN-SENJ-88-052 Revision 3 CON'lAINMENT BULK AMBIENT TEMPERATURES ,

i Containment bulk ambient temperatures at Turkey Point Units 3 and l 4 have recently been trending upward toward the design basis  ;

temperature limit of 120*F. Based on these observations, there is a potential for exceeding the design basis limit during the hotter summer months for short periods of time. In anticipation of such l occurrences, the effects of operating with a higher containment l temperature limit were reviewed, evaluated and in some specific i instances analyzed.

Safety Evaluation Summary: ,

j' No physical plant modifications are involved. The only change involved is to the temperature conditions used as a basis for plant operation. Considerations included, but were not limited to, evaluating the effects of elevated ambient temperature conditions on structural integrity, cable ampacities, effects on accident ~

analyses, equipment qualification, and instrumentation accuracy.

This Safety Evaluation concludes that raising the containment bulk ambient temperature limit from 120*F to 125'F for a cumulative period of two weeks per year is considered acceptable with no adverse impact on plant safety or operation.

Issued: April 14, 1989 The following were procedures revised as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

3-OSP-201.1 RCO Daily Logs 4-OSP-201.1 RCO Daily Logs 0-ADM-021 Technical Specification Implementation Procedure 120

SAFETY EVALUATION: JPN-PTN-SELS-88-053 Revision 0 SPENT FUEL POOL LEVEL REDUCTION The dummy Fuel Assembly used to test refueling equipment must be transferred to Unit 4 to support preparations for the upcoming refueling outage. The dummy Fuel Assembly is presently in the Unit 3 SFP. The Bridge Crane, using the long fuel handling tool, will be used to move the assembly to the Transfer Canal where it will be placed in the fuel elevator. The short fuel handling tool will then be attached to the Bridge Crane to allow the dummy Fuel Assembly to be lifted to a height that will allow transfer to the Cask Handling Crane. The Bridge Crane will then move the dummy assembly back to the Unit 3 SFP, where it will be transferred to the cask Handling Crane. The dummy assembly will then be transferred to the Unit 4 SFP by the Cask Handling Crane. The keyway gate separating the Transfer Canal and the SFP must be removed to provide a transfer path. The Transfer Canal will be filled, in accordance with existing plant procedures, from the SFP. -

This removes the static head against the gaue, allowing the gate '

to be withdrawn and moved to its normal storage location, in accordance with existing plant procedures. The normal source of water for filling the transfer Canal is the Refueling Water Storage Tank (RWST), however, since Unit 3 will be in operation, RWST water will not be used due to Technical Specification restrictions on RWST level. Instead, the transfer Canal will be filled using SFP water, until levels equalize.

Safety Evaluation Summary There is no effect on the design bases of any component, equipment, or system. The FSAR allows for the SFP cooling loop to be temporarily shut down for maintenance, and the requirements of this evaluation will ensure that the SFP temperature will remain below the design margin. No fuel assembly movement will occur when the SFP level is below 56'10". All credible failure modes are bounded by existing FSAR analysis. ,

Issued: September 14, 1988 The following Temporary Procedure was issued as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

TP-474 Unit 3 SFP Level Reduction for Maintenance i

1 l 121 l

l l

l

.___-_______-___-____--________D

- +

n;

.i' g SAFETY EVALUATION: JPN-PTN-SECS-88-060 Revision 0 " . ' - .. :

, - A, TEMPORARY REMOVAL OF STEAM GENERATOR THRUST BEAN AND FLOOR GRATING = '.

During the Fall 1988 Unit 4 refueling outage, the Reactor Coolant ,[ [ * '

Pump "B" motor was replaced. To provide adequate clearance for .

rigging the in-place pump motor out through the hatch, the Steum .N

~

Generator "C" thrust beam support, floor steel, handrail, grating ~

-and pipe supports for the 2" containment primary water service T;l' connections and 2" containment service air piping above the ~j W

..guipment hatch had to be temporarily removed. . d, Safety Evaluation Summarv .f The work involved the temporary removal and reinstallation of . ,

structural items in accordance with original plant configuration and design. Plant mode restrictions and restrictions on heavy load f movements were imposed to assure that the plant was maintained in l >

a safe condition and to preclude any adverse interactions with - ' ' '

safety related systems, structures, and components.

GNR Issued: September 22, 1989 g.

==

?

,[ y ~

- 4 Yt W 'f n

s - ,

l

'. ', ,L -}

'n

.[

i -r

..*r l

l I .

122 7

?

sp L

SAFETY EVALUATION: JPN-PTN-3ENJ-88-067 Revision 0 EMERGENCY DIE 8EL GENERATOR FIVE STARTING ATTEMPT TEST The special test described by TP 488 involves five starting attempts of a diesel generator. The purpose of this test is to

. demonstrate the capability of the air start system to meet the five starting attempts criteria established by the FSAR and NUREG-0800.

The-current system of two-air receivers valved in with two air

. receivers in reserve meets'this' criteria. PC/M 86-155, Addition of EDG Air Start Motors, adds a redundant set of air start motors in~ operation. The new configuration will have both sets of air start motors on one pair of air receivers with the other pair of

- air receivers as back-up, if the test are successful.

Safety Evaluation Summarv This procedure requires the affected EDG to be placed Out of service-(OOS). The other EDG shall remain operational and in the ready to start mode as per Operational Procedure O-OP-023, and the daily start test will be performed in accordance with Technical

' Specifications (TS) 3.7.2.b. TS 3.7.2.b further requires that the plant be shut down if an EDG is OOS greater than seven days. Both units will already be in cold shutdown for the duration of the test, which is expected to take less than twelve hours. Power requirements are minimal in cold shutdown and one EDG is sufficient to provide power to both units for any normal or accident condition. The EDG out of service will be equivalent to having an EDG out of service for' maintenance. The EDG out of service does not. affect the EDG in service to provide power for emergency loads.

The EDGs are completely independent except the Diesel Oil Storage

- Tank. The EDG to be tested will use no fuel.

Individual Technical Specification requirements, for systems required in mode 5 and/or mode 6, must also be adhered to prior to placing the EDG OOS. If required, the EDG out of service could be used to power emergency loads within an hour.

Issued: September 17, 1988 l

l 123 l

h 1

SAFETY EVALUATION: JPN-PTN-SELS-88-070 Revision 0 REACTOR CAY 1TY FILTRATION SYSTEM OPERATION

~

Turkey Point Procedures 3/4-OP-38.9, " Refueling Activities Checkoff List", require that the Reactor Cavity Filtration System be in operation prior to the commencement of refueling operations, and the FSAR, Section 9.5.3, states that the systems will operate

...throughout the duration of refueling operations...". The question was raised if it was required by the FSAR for refueling operations to be suspended in the event the Reactor Cavity Filtration System became inoperable.

The proposed changes are to remove the requirement for the Reactor Cavity Filtration System to be in operation prior to the commencement of refueling operations and to change the FSAR to reflect this operational situation.

Safety Evaluation Summary:  ;

The probability of occurrence or the consequences of an accident or malfunction of equipment important. to safety previously evaluated in the FSAR have not been increased since the design basis of any component, equipment, or system is not affected. The design basis of the Reactor Cavity Filtration System remains to provide water clarity in the reactor cavity during refueling operations.

The possibility of an accident or malfunction of a different type than evaluated previously in the FSAR has not been created because no new failure modes have been introduced by this change. All credibic failure modes are bounded by existing FSAR analyses.

The margin of safety as defined in the basis for any Technical Specifications has not been reduced since any components, equipment or structures addressed by the Technical Specification have not been adversely affected.

Issued: October 14, 1988 The following were procedures revised as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

3-OP-038.2 Reactor Cavity Filtration System Operation 4-OP-038.2 Reactor Cavity Filtration System Operation 124

SAFETY EVALUAT50N: JPN-PTN-SENS-88-081 Revision 0 i

FREEEE SEAL SAFETY EVALUATION FOR REPAIR 0 P VALVE 3-312A AND 3-312B j This evaluation ' addresses the use of a freeze' seal in order to j repair valves .3-312A and 3-312B. These valves are 3". check valves )

.directly off the Reactor Coolant-System (RCS) and require a freeze 1 seal to isolate. i Safety Evaluation Summary j The probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety previously evalue'ed in the FSAR have not been increased since it has been shown that a failure of the freeze seal plug is unlikely and is bounded by the existing design analysis. The operability of the Residual Heat Removal (RHR) System will not be affected.

~

The possibility of an accident or malfunction of a different typo .

than evaluated previously in the FSAR has not been created because -

no new failure modes are introduced. A freeze seal plug failure is similar to mid nozzle shutdown operation.. The RHR system would remain operable and be capable of providing core cooling.

The margin of safety . as defined in the basis for any Technical Specification has not been reduced. The requirements of the evaluation ensure that the freeze seal is established in accordance with approved procedures and that the maintenance is performed in ac4:ordance with PC/M 88-164. The operability of the RHR system is not affected and no other equipment required for modes 5 or 6 is adversely affected.

Issued: November 15, 1988 125 l

~ SAFETY EVALUATION: JPN-PTN-SENJ-88-087 Revision 0 THE DELETION OF' BACKFLOW DAMPERS IN THE CONTROL ROOM VENTILATION

-8YSTEM A recent program evaluating the incorporation of Nonconformance Reports onto selected plant drawings noted a discrepancy regarding

.the control Room. Ventilation System.- It was noted that several drawings did not match the existing system configuration. These drawings showed'backdraft dampers installed on the discharges of the three air handling units supplying the control room. System walkdowns have- confirmed that the backdraft dampers are not installed. This evaluation of the proposed changes to the drawings to match the existing system configuration determines the ability of' the ' existing system configuration to meet its design basis requirements as described in the FSAR.

Safety Evaluation Summary As demonstrated by testing of the installed system, the lack of the backdraft dampers shown on the drawings will not prevent the Control Room Ventilation System from performing its safoty Function in the event of an accident, nor does the lack of backdraft dampers result in new or more severe failures of the Control Room Ventilation System than those previously evaluated. Also, no new initiating . events have been created by the lack of backflow dampers. Therefore, ' the ability of the Control Room Ventilation System to perform its safety-function has not been changed by the lack of backflow dampers.

Issued: December 6, 1988 l

L 126

1 l

i SAFETY EVALUATION: JPN-PTN-SEEJ-88-088 Revision 0 125 V DC CONTROL POWER FOR LOAL CENTER 3B41 FROM DC PANEL 3D01 PC/M 82-100 provided the design to transfer the source of DC control' power for Condensate Polishing Load Center 3B41. The original control power for this non-class 1E load center was from Vital Bus 3D01, Breaker 9. The new source . of control power assigned for this load center was non-class 1E Bus 3D31, Breaker

4. The transfer of the control power source for this load center  !

was one of several loads being transferred as part of the Auxiliary Power Upgrade and was intended to relieve the heavily loaded class 1E DC buses.

The cable for the control power source to Load Center 3B41 was ins'.alled and terminated at Bus 3D31, Breaker'4. However the '

ex, ting cable from Vital DC Bus 3D01 was never disconnected in the field and continues to provide DC control power to Load Center 3B41. All of the drawings affected by the power source transfer .

have be.en updated to reflect the new non-class 1E power source. ~

This evaluation is to determine the impact on plant safety of not having transferred the power source to non-class 1E Bus 3D31.

Safety Evaluation Summary This existing condition does not increase the probability of occurrence or the consequences of an accident previously evaluated in the FSAR since this is the original plant design condition.

This condition results in a negligible increase to the vital battery . loading profile. This small increase is within the battery's capacity and will not measurably degrade the vital DC system voltage. This exiting condition does not increase the probability of occurrence or the consequences of a. malfunction of equipment important to safety previously evaluated in the FSAR since Load Center 3B41 is non-class 1E and adequate electrical protection is provided in the original design to protect Class 1E Bus 3D01 against fault currents.

The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR is not created by this existing condition.

This condition is the original design for this circuit. The failure of this circuit will not have an adverse effect on any safety related equipment, components or functions.

This existing condition does not reduce the margin of safety as defined in the basis for any Technical Specification. Adequate DC power will remain available for emergency use. This condition does not change or modify the operation of any plant equipment.

Issued: December 6, 1988 127

_ - _ _ _ = - _ - - __---___-__ _ _

i SAFETY EVALUATIONS' JPN-PTN-8EES-89-002 Revision 0 UNITS 3.&.4 TEMPORARY SYSTEN ALTERATION POR DISABLING . OF LOCAL-INDICATION FOR PRN8 R-18

' A Temporary System Alteration to lift both leads at the local. i radiation level indicationLof R-18, Waste. Disposal System Liquid

. Effluent Monitor, .has been-requested. The FSAR identifies that-tne R-18 ' channel provides 1) . remote indication on the Waste Disposal' System control board,'2) control room indication,'and 3) i an interlock to operate RCV-018. Lifting.these leads will disable the local indicator and will have no effect on the operation of the -

remainder of the system.

Safety Evaluation Summary The probability of occurrence or the consequences.of an accident-or malfunction of equipment important- to: safety- previously evaluated in the FSAR have not been increased since the parameter ^

displayed by this device.is still available in the Control Room .

and the automatic action of R-18 has not been impacted.

The possibility of an' accident'or' malfunction of a different type than evaluated previously in the FSAR has not been created because-the local indicator does not serve any safety function and this modification does not affect any safety related components or systems.

The margin of safety as defined in ' the basis for any Technical Specification has not been reduced. The requirements of the e';aluation ensure that indication of the Waste Disposal System Radioactive Effluent is still provided in the control room and the operability requirement- (automatic closure' of RCV-018 upon high radiation level) for this component imposed by the Technical-

. Specifications is still met.

Issued: January 12, 1989 l

128 i

-l SAFETY EVALUATION: JPN-PTN-SEMS-89-008 Revision 0 TENPORARY SYSTEM ALTERATION FOR CLEANING OF THE TURBINE LUBE OIL I The oil in the Unit 3 Turbine Lube 011 system has been sampled and determined to be outside ' the guidelines for particulate count provided by Power Resources. In order to remove the particulate contamination from the system, an oil filter system will be

. temporarily connected to function as a cleanup loop. 1 This temporary clean-up loop will be composed of supply and return lines from the Turbine Lube 011 reservoir, a vacuum dehydrator, a booster pump,.a pump skid and two parallel filters. ~

One supply line for the subsystem pump suction will be located downstream of valve 3-040-077' (a spool piece has been removed to provide a connection). The other supply line and the return lines from the filters'will be routed into the manway on the top of the reservoir.

Safety Evaluation Summarv:

The probability of occurrence or the consequences of an accident '

or malfunction of equipment important to safety- previously evaluated in the FSAR has not been increased. The Turbine Lube Oil system is classified as Not Safety Related. Measures will be taken to minimize the likelihood of draining the reservoir. If the reservoir were ellowed to drain, the turbine and reactor would automatically trip.

The possibility of an accident or malfunction of a different type than evaluated previously in the FSAR has not been created. As noted 'above, the turbine lube oil system is classified . as Not Safety Related. Measures will be taken to preclude damage to adjacent safety related equipment.

The margin of safety - as defined in the basis for any Technical Specifications has not been reduced since the basis of any Technical Specification is not affected. The requirements of the evaluation ensure that this temporary system will not adversely affect equipment addressed in the Technical Specifications.

Issued: February 2, 1989 129

SAFETY EVALUATION: JPN-PTN-SEMS-89-010 Revision 0 DELETION: OF BACKSEATING OF SAFETY INJECTION / CONTAINMENT SPRAY VALVES INSIDE CONTAINMENT The FSAR describes the design features of Safety Injection valves as having backseats that limit leakage. This evaluation reviews whether the Safety Injection and containment Spray valves inside containment could be taken off and left off their backseats.

Safety Evaluation Summary The probability of occurrence or the consequences of an accident or malfunction of equipment .important to safety previously evaluated in the ' 3AR have not been increased since the change does not affect offsite doses post-LOCA and the design of the affected valves is such as to limit leakage. Leakage that may occur would not be significant enough to adversely affect system performance.

The possibility of an accident or malfunction of a different type than evaluated previously in the FSAR has not been created because no new failure modes are created. The original Design codes for the valves are upheld.

The margin of safety as defined'in the basis for any Technical Specification has not been reduced. The maximum allowed Reactor Coolant System leakage rates are not affected. System performance criteria is not affected.

Issued: January 28, 1989 130 i

J

, SAFETY EVALUATION: JPN-PTN-SENJ-89-016 Revision 0 L.

ENERGENCY DIESEL GENERATOR LOADING EVALUATION UPDATE During preparation of the revised Emergency Diesel Generator (EDG) load evaluation, ' an assessment was made of the various plant

' modifications and plant information which had . been developed subsequent to its last issuance and.affected EDG loadings. . Among the revisions were,the following:

Residual Heat Removal (RHR) and Containment Spray (CS) pump motor loads were reduced based upon the latest safeguards pump flow analyses. These analyses modeled the various post accident injection and recirculation phase alignments including the Emergency Core' Cooling Systems (ECCS) " piggy-'

' back" configuration of RHR, CS, and High Head Safety Injection (HHSI) pumps. The flow rates for specific configurations were then used to determine the pump motor loads consistent with the methodology previously utilized in the EDG load evaluations. ,

EDG Starting Air System compressor loads were increased as a result of recent in plant testing which indicated that the t.ctual motor loads are higher than the name plate values

>reviously assigned.

Ioads sr.,sociated with the Normal Containment Coolers (NCCs) were increased due to the modification of the NCC tube bundles and fan motor sheaves.

Battery Room HVAC loads were increased to reflect latest plant information.

Safety Evaluation Summary:

In their EDG Loading Safety Evaluation Report, the NRC approved EDG loads which' complied with the following limits:

Auto-connect limit 2750 kW Short-term continuous limit (2000 hr. rating)2850 kW Transient load limit 2950 kW As stated in JPE-L-86-74 Rev. 3, the peak EDG loads are as follows:

Auto-connect 2714 kW 1 - 30 minutes 2840 kW*

30 minutes - 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2619 kW

(* One unit at power with one unit shutdown for less than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.)

These loads are within the limits established by the NRC.

Issued: April 13, 1989 131

SAFETY EVALUATION: JPN-PTN-SEMJ-89-023 Revision 0 r ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C In the Appendix R Safe Shutdown Analysis, credit is taken for one charging. pump per unit (Charging Pumps 3B/4B) as being available for alternate shutdown. This charging pump is required to maintain hot standby and achieve cold shutdown in the event of a fire in an alternate shutdown fire zone. The charging pump takes suction from either the Volume Control Tank (VCT) or the Refueling Wai.w Storage Tank (RWST) through interlocked valves. The VCT valve (LCV-115C) is normally'open and the-RWST valve (LCV-115B) is normally-closed.

An interlock is provided such that at least one valve is open to ensure continuous suction flow to the charging pump. A fire postulated.in certain alternate shutdown fire zones could cause spurious closure of'the normally open LCV-115C and failure of LCV-115B to open; the pump (s) would be starved of suction flow. This could result in the inability of the pump (s) to perform the required alternate shutdown function due to the possible damage -

resulting from flow starvation. Credit is taken in the Appendix '

R Safe Shutdown Analysis for an operator action to mitigate the adverse effects of spurious closure of the normally open valve.

Without administrative controls or a permanent design change, this operator action may not be taken.in time to preclude pump damage.

The temporary administrative controls prescribed in this safety evaluation, for fire zones 98 and 106', are necessary to meet FPL's commitment for Turkey Point Units 3 and 4 alternate shutdown capability.

Safety Evaluation Summary The proposed administrative controls ensure that equipment required to be available to provide alternate shutdown capability, in the event of a fire, will be operable. In addition, the presence of fire watches is designed to provide early detection and warning of a fire in its incipient stages, thereby initiating Control Room operator action to preclude possible pump damage resulting from a spurious closure of LCV-115C and failure of LCV-115B to open due ,

to a fire. Shutting off Charging Pump (s) 3B/4B, upon identification of a fire, will prevent damage to the pump (s) due to flow starvation. In the case where Charging Pump (s) 3B/4B is(cre) shut off, one other charging pump will always be available l to perform its required function, since the Technical I specifications require two charging pumps to be operable in Modes 1 thru 4. Also, because the Reactor Coolant Pumps (RCP) have a redundant.means of cooling, shutting off Charging Pump (s) 3B/4B, if operating, will have no adverse effect on the RCP seals.

Issued: April 19, 1989 l

l l

132

g3, ;

, 3- ,

t r r, x tr

,. 'i .

'The following was a.' procedure revised as n'11 owed by_'10CFR50.59.as

1. ', :-)ustified by the preceding Safety- Evaluation:

[, _ ' 14-ONOP-016.9 Response to a' Reported Fire .in the Charging' Pump Room, .

b .MCC B Room,;. Cable.SpreadingERoom,.or Control ~ Room I

1'

l. '

4 4 .

e i

i 133

'~

=_ _ _

r l.

SAFETY EVALUATION: JPN-PTN-SEMJ-89-031 Revision 0 TOP-514 VARIABLE BACK PRESSURE PERFORMANCE TEST Test Procedure TOP-514 involves testing of the Unit 4 ' secondary plant to determine the Low Pressure Turbine post-modification heat balance for comparison to the pre-modification heat balance. This comparison _is necessary to validate the predicted performance increase of at least 8.1 MWe as guaranteed for this modification.

l l- This~ test will acquire performance data ~ from selected instrumentation at various back pressure values between 2.50 inches Hg and 3.50 -inches Hg, inclusive. The condenser back pressure will be modulated by use of a fabricated blank flange / throttle - plate temporarily mounted over the end of condenser vacuum breaker valve e 4-30-001. No instrumentation will be installed on safety related systems for this test. However, data from safety related l

instrumentation will be used for this test.

Safety Evaluation Summarv -

This procedure affects the operation of limited secondary systems.

l- (The Steam Generator Blowdown and the Unit 4 Auxiliary Steam Supply will be isolated during the test.) Also, this test does not adversely affect any safety related equipment or equipment used to mitigate the consequences of an accident. If, during performance of the test, the condenser vacuum should be reduced to below 20.0 inches Hg, the turbine / generator will automatically .be tripped.

An automatic trip of the turbine / generator results in-an automatic transfer to connect the A and B 4.16 kV busses to the unit's start-up transformer (the start-up transformer and C-bus transformers serve' the unit after- a reactor trip). Thus, automatic turbine / generator trip would be the most serious event which could be a consequence of 'this temporary operating procedure and this '

event has been considered in the FSAR.

^

Issued: May 22, 1989 134 1

SAFETY EVALUATION: JPN-PTN-SEMJ-89-035 Revision 0 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C In the Appendix R Safe Shutdown Analysis, credit is taken for one charging pump per unit (Charging Pumps 3B/4B) as being available for alternate shutdown. This charging pump is required to maintain hot standby and achieve cold shutdown in the event of..a fire in an alternate shutdown fire zone. The charging pump takes suction from either the Volume Control Tank (VCT) or the Refueling Water Storage Tank (RWST) through interlocked valves. The VCT valve (LCV-115C) is normally open and the RWST valve.(LCV-115B) is normally closed.

An interlock is provided such that at least one valve is open to ensure continuous suction flow to the charging pump. A fire postulated in certain alternate shutdown fire zones could cause spurious closure of the normally open LCV-115C and failure of LCV-115B to open; the pump (s) would be starved of suction flow. This could result in the inability of the pump (s) to perform the required alternate shutdown function due to the possible damage ;

resulting from flow starvation. Credit is taken in the Appendix R Safe Shutdown Analysis for'an operator action to mitigate the adverse effects of spurious closure of the normally open valve.

Without administrative controls or a permanent design change, this operator action may not be taken in time to preclude pump damage.

The ' temporary administrative controls prescribed in this safety evaluation, for fire areas O and T, are necessary to meet FPL's commitment for Turkey Point Units 3 and 4 alternate shutdown capability.

l Safety Evaluation Summary The proposed administrative controls ensure that equipment required to be available to provide alternate shutdown capability, in the l

event of a fire, will be operable. In addition, the presence of fire watches is designed to provide early detection and warning of a fire in its incipient stages, thereby initiating Control Room operator action to preclude possible pump damage resulting from a spurious closure of LCV-115C and failure of LCV-115B to open due l to a fire. Shutting off Charging Pump (s) 3B/4B, upon identification of a fire, will prevent damage to the pump (s) due to flow starvation. In the case where Charging Pump (s) 3B/4B )

is(are) shut off, one other charging pump will always be available l to- perform its required function, since the Technical l Specifications require two charging pumps to be operable in Modes 1 thru 4. Also, because the Reactor Coolant Pumps (RCP) have a redundant means of cooling, shutting off Charging Pump (s) 3B/4B, if operating, will have no adverse effect on the RCP seals. ]

Issued: April 28, 1989 135 l

l

g , ,

--, m;:; +,,-

---

,----- -----------y . -

g*-, ' '

, n. .:

QR ,

ll?p: ', <  :

%<T, ,

~.

o .,

I' -;

.. , + - , , . . , . ,,

w The:following. procedure was developed as allowed;by"100N50.59'as-f !U ,

(partially. justified byfthe/ preceding; Safety; Evaluation:

L -

3-ONOP-O'1659 Respons's to a Reported Firelin the Charging Pump Roomb MCC'B: Room,' Cable;. Spreading Room-orfControl. Room- '

o e

j ',

si Eii 3, . e '

f .f i

Y f., _ 2 i.-? ,

136 hm-mumm _ _ _ . ._..m._. ____.m_._______-- _m

SAFETY EVALUATION: JPN-PTN-SENJ-89-036 Revision 0 ALTERNATE SEUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C In the. Appendix R Safe Shutdown Analysis, credit is taken for one charging pump per unit (Charging Pumps 3B/4B) as being available for alternate shutdown. This charging pump is required to maintain hot standby and achieve cold shutdown in the event of a fire in an alternate shutdown fire zone. The charging pump takes suction from either the Volume Centrol Tank (VCT) or the Refueling Water Storage Tank (RWST) through interlocked valves. The VCT valve (LCV-115C) is normally open and the RWST valve (LCV-115B) is normally closed.

An interlock is provided such that at least one valve is open to ensure continuous suction flow to the charging pump. A fire postulated in certain alternate shutdown fire zones could cause spurious closure of the normally open LCV-115C and failure of LCV-115B to open; the pump (s) would be starved of suction flow. This could result in the inability of the pump (s) to perform the required alternate _ shutdown function due to the possible damage resulting from flow starvation. Credit is taken in the Appendix -

R Safe Shutdown Analysis for an operator action to mitigate the '

adverse effects of spurious closure of the normally open valve.

Without administrative controls or a permanent design change, this operator action may not be taken in time to preclude pump damage.

The temporary administrative controls prescribed in this safety evaluation, for fire areas N and R, are necessary to meet FPL's commitment for Turkey Point Units 3 and 4 alternate shutdown capability.

Safety Evaluation Summary The proposed administrative controls ensure that equipment required to be available to provide alternate shutdown capability, in the event of a fire, will be operable. In addition, the presence of fire watches is designed to provide early detection and warning of a fire in its incipient stages, thereby initiating Control Room operator action to preclude possible pump damage resulting from a spurious closure of LCV-11SC and failure of LCV-115B to open due to a fire. Shutting off Charging Pump (s) 3B/4B, upon

, identification of a fire, will prevent damage to the pump (s) due l to flow starvation. In the case where Charging Pump (s) 3B/4B is(are) shut off, one other charging pump will always be available to perform its required function, since the Technical Specifications require two charging pumps to be operable in Modes 1 thru 4. Also, because the Reactor Coolant Pumps (RCP) have a redundant means of cooling, shutting off Charging Pump (s) 3B/4B, if operating, will have no adverse effect on the RCP seals.

Issued: April 19, 1989 137

~

=

7- 7. ,

, , x-

.

• i f.

r.

L'-

v. , !0 >.

!4

!{ i

The'following was'a procedure revised as'allcwed byf10CFR50.59.'as J

b 4 ~ justified by.the: preceding; Safety Evaluation:

4-ONOP-016.9 Response to a Reported Firein the Charging Pump Room, y ,

cMCC B Room,fCablel Spreading.. Room, or Control' Room.

j, ,

I' I

l O

e

.},

I 138 i

.! [

- _-- J L. _ _ _ _ . _ _ _ _ _ -

e SAFETY EVALUATION: JPN-PTN-SEMJ-89-038 Revision 0 j i

RCP OIL COLLECTION SYSTEM ]

i During ' the effort to prepare for an Appendix R audit, a non-compliance pertaining to the capacity of the Reactor Coolant Pump (RCP) motor oil collection tank for Turkey Point Unit 4 was identified. The FSAR Section 3.10.3 of Appendix 9.6A currently states that the 1.5 inch drain lines from each RCP motor converge to a common 1.5 inch line that drains (gravity flow) to a 265 gallon vented collection tank located outside the biological shield '

wall in the containment building. The calculated capacity of the RCP motor oil collection tank based on the tank dimensions is 251 gallons. The proposed change is to revise the capacity of the RCP motor oil collection tank in Section 3.10.3 of Appendix 9.6A from 265 to 251 gallons to reflect the as-built capacity of the tank.

Corresponding changes are proposed to Sections 2.5 (Paragraph III. 0) , and 4. P. 2 (Exemption P.4, Justification for Exemption), of Appendix 9.6A of the FSAR.  ;

Safety Evaluation Summary:

The proposed change revises the FSAR to reflect the actual capacity of the RCP Motor Oil Collection System. No new equipment is added or deleted. The RCP Oil Collection System maintains the ability to:

a. collect lubricating oil from pressurized and unpressurized leakage sites,

b. stay in place following a Safe Shutdown Earthquake (SSE) in accordance with Regulatory Guide 1.29, and

c. collect and contain the entire inventory of one RCP Motor Lube Oil System and normal leakage from the two remaining RCP motors during one fuel cycle.

Lube oil will continue to be collected in the RCP oil collection tank, to minimize the possibility of fire. There is no change to the system or associated systems used in mitigating the consequences of an accident described in the FSAR. The proposed change does not make any changes to the existing design criteria and regulatory commitments.

Issued: April 28, 1989 139

SAFETY EVALUATION: JPN-PTN-SEMJ-89-041 Revision 0 RCP OIL COLLECTION SYSTEM During the effort to prepare for an Appendix R audit, a non-compliance pertaining to the capacity of the Reactor Coolant Pump (RCP) motor oil collection tank for Turkey Point Unit 3 was identified. The FSAR Section 3.10.3 of Appendix 9.6A currently states that the 1.5 inch drain lines from each RCP motor converge to a common 1.5 inch line that drains (gravity flow) to a 265 gallon vented collection tank located outside the biological shield wall in the containment building. The calculated capacity of the RCP motor oil collection tank based on the tank dimensions is 251 gallons. The proposed change is to revise the capacity of the RCP motor oil collection tank in Section 3.10.3 of Appendix 9.6A from 265 to 251 gallons to reflect the as-built capacity of the tank.

Corresponding changes are proposed to Sections 2.5 (Paragraph III . 0) , and 4. P. 2 (Exemption P.4, Justification for Exemption), of Appendix 9.6A of the FSAR. In addition, the spare RCP motor, which -

has a total lubricating oil capacity of 275 gallons, has been' installed on RCP-3A. Changes to the above section are also proposed to reflect the increase in the lubricating oil capacity of RCP-3A which is in excess of the holding capacity of the oil collection tank.

Safety Evaluation Summarv:

The proposed change revises the FSAR to reflect the actual capacity of the RCP Motor Oil Collection System. No new equipment is added or deleted. The RCP Oil Collection System maintains the ability to:

a. collect lubricating oil from . pressurized and unpressurized leakage sites,

b. stay in place following a Safe Shutdown Earthquake (SSE) in accordance with Regulatory Guide 1.29, and

c. collect and contain the entire inventory of RCP-3B or RCP-3C Motor Lube Oil System and normal leakage from the two remaining RCP motors during one fuel cycle.

d. collect and contain the majority of the contents of RCP-3A Motor lube oil system and the normal leakage from the two remaining RCP motors during one fuel cycle while transferring the overflow to a location safe from ignition. The consequences of a 54 gallon overflow of the RCP oil collection tank has been evaluated and determined to be acceptable. In addition, the NRC has evaluated in an earlier exemption approval letter a potential RCP oil collection tank overflow and has determined this overflow to be acceptable.

140

y , - -- -

.y - ,

f(

1 4 , H TV , .

[ .

e l

. Lube - oil" will . ' continue : to" b' collected in the.RCP-oil collection  !

tank, to minimize the possibility of. fire. There is<no. change.to .

the.' system or . associated.. systems 1 used . in ' mitigating. the 11

. consequences of an accident ~ described-in the FSAR..

. The~ proposed

' change.does.'not make'any changes to-the existing design criteria-

and regulatory' commitments.

. Issued:. June ' .15 , .~ 19 8 9 '

N

--),'

m i

141

___x_-____-__ _ _ 1

SAFETY EVALUATION: JPN-PTN-SEMJ-89-043 Revision 0 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C In the Appendix R Safe Shutdown Analysis, credit is taken for one charging pump per unit (Charging Pumps 3B/4B) as being available for alternate shutdown. This charging pump is required to maintain hot standby and achieve cold shutdown in the event of a fire in an alternate shutdown fire zone. The charging pump takes suction from either the Vcl.ime Control Tank (VCT) or the Refueling Water Storage Tank (RWST) through interlocked valves. The VCT valve (LCV-115C) is normally open and the RWST valve (LCV-115B) is normally closed.

An interlock is provided such that at least one valve is open to ensure continuous suction flow to the charging pump. A fire postulated in certain alternate shutdown fire zones could cause spurious closure of the normally open LCV-115C and failure of LCV-115B to open; the pump (s) would be starved of suction flow. This could result in the inability of the pump (s) to perform the required alternate shutdown function due to the possible damage -

resulting from flow starvation. Credit is taken in the Appendix ~

R Safe Shutdown Analysis for an operator action to mitigate the adverse effects of spurious closure of the normally open valve.

Without administrative controls or a permanent design change, this operator action may not be taken in time to preclude pump damage.

The temporary administrative controls prescribed in this safety evaluation, for fire areas N and R, are necessary to meet FPL's commitment for Turkey Point Units 3 and 4 alternate shutdown capability.

Safety Evaluation Summary The proposed administrative controls ensure that equipment required to be available to provide alternate shutdown capability, in the event of a fire, will be operable. In addition, the presence of fire watches is designed to provide early detection and warning of a fire in its incipient stages, thereby initiating Control Room operator action to preclude possible pump damage resulting from a spurious closure of LCV-115C and failure of LCV-115B to open due to a fire. Shutting off Charging Pump (s) 3B/4B, upon identification of a fire, will prevent damage to the pump (s) due to flow starvation. In the case where Charging Pump (s) 3B/4B is(are) shut off, one other charging pump will always be available to perform its required function, since the Technical Specifications require two charging pumps to be operable in Modes 1 thru 4. Also, because the Reactor Coolant Pumps (RCP) have a redundant means of cooling, shutting off Charging Pump (s) 3B/4B, if operating, will have no adverse effect on the RCP seals.

Issued: April 28, 1989 142

BAFETY EVALUATION: JPN-PTN-SEMJ-89-044 Revision 0 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C In the Appendix R Safe Shutdown Analysis, credit is taken for one charging pump per unit (Charging Pumps 3B/4B) as being available for alternate shutdown. This charging pump is required to maintain hot standby and achieve cold shutdown in the event of a fire in an alternate shutdown fire zone. The charging pump takes suction from either the Volume Control Tank (VCT) or the Refueling Water Storage Tank (RWST) through interlocked valves. The VCT valve (LCV-115C) is normally open and the RWST valve (LCV-115B) is normally closed.

An interlock is provided such that at least one valve is open to ensure continuous suction flow to the charging pump. A fire postulated in certain alternate shutdown fire zones could cause spurious closure of the normally open LCV-115C and failure of LCV-115B to open; the pump (s) would be starved of suction flow. This could result in the inability of the pump (s) to perform the required alternate shutdown function due to the possible damage resulting from flow starvation. Credit is taken in the Appendix -

R Safe Shutdown Analysis for an operator action to mitigate the '

adverse effects of spurious closure of the normally open valve.

Without administrative controls or a permanent design change, this operator action may raot be taken in time to preclude pump damage.

The temporary administrative controls prescribed in this safety evaluation, for fire areas O and T, are necessary to meet FPL's commitment for Turkey Point Units 3 and 4 alternate shutdown capability.

Safety Evaluation Summary The proposed administrative controls ensure that equipment required to be available to provide alternate shutdown capability, in the event of a fire, will be operable. In addition, the presence of fire watches is designed to provide early detection and warning of a fire in its incipient stages, thereby initiating Control Room operator action to preclude possible pump damage resulting from a spurious closure of LCV-115C and failure of LCV-115B to open due to a fire. Shutting off Charging Pump (s) 3B/4B, upon identification of a fire, will prevent damage to the pump (s) due to flow starvation. In the case where Charging Pump (s) 3B/4B is(are) shut off, one other charging pump will always be available to perform its required function, since the Technical Specifications require two charging pumps to be operable in Modes 1 thru 4. Also, because the Reactor Coolant Pumps (RCP) have a redundant means of cooling, shutting off Charging Pump (s) 3B/4B, if operating, will have no adverse effect on the RCP seals.

Issued: May 19, 1989 l

l 143

~

g -y ..

l , ytl '

-The;following procedure-was developed'as allowed by110CFR50.59.as cpartially; justified by the preceding Safety Evaluation:

.c

'fr . 3-ONOP-016.9 Response to a Reported Fire in the Charging Pump Room, .

=MCC.B Room,' Cable. Spreading: Room or Control.-Room.-

S t

j i I. (.

.i e

s 144

- 2-_ -_- _-_--___ -__

SAFETY EVALUATION: JPN-PTN-SEMJ-89-045 Revision 0 ALTERNATE SHUTDOWN CAPABILITY WITH SPURIOUS CLOSURE OF LCV-115C In the Appendix R Safe Shutdown Analysis, credit is taken for one charging pump per unit (Charging Pumps 3B/4B) as being available for alternate shutdown. This charging pump is required to maintain hot standby and achieve cold shutdown in the event of a fire in an alternate shutdown fire zone. The charging pump takes ' suction from either the Volume Control Tank (VCT) or the Refueling Water Storage Tank (RWST) through interlocked valves. The VCT valve (LCV-115C) is normally open and the RWST valve (LCV-115B) is normally closed.

An interlock is provided such that at least one valve is open to ensure continuous suction flow to the charging pump. Afire postulated in certain alternate shutdown fire zones could cause spurious closure of the normally open LCV-115C and failure of 1CV-115B to open; the pump (s) would be starved of suction flow. This could result in the inability of the pump (s) to perform the required alternate shutdown function due to the possible damage -

resulting from flow starvation. Credit is taken in the Appendix ~

R Safe Shutdown Analysis for an operator action to mitigate the adverse effects of spurious closure of the normally open valve.

Without administrative controls or a permanent design change, this operator action may not be taken in time to preclude pump damage.

The temporary administrative controls prescribed in this safety evaluation, for fire areas HH and MM, are necessary to meet FPL's commitment for Turkey Point Units 3 and 4 alternate shutdown capability.

Safety Evaluation Summary The proposed administrative controls ensure that equipment required to be available to provide alternate shutdown capability, in the event of a fire, will be operable. In addition, the presence of i' ire watches is designed to provide early detection and warning of a fire in its incipient stages, thereby initiating Control Room operator action to preclude possible pump damage resulting from a spurious closure of LCV-115C and failure of LCV-115B to open due to a fire. Shutting off Charging Pump (s) 3B/48, upon identification of a fire, will prevent damage to the pump (s) due to flow starvation. In the case where Charging Pump (s) 3B/4B is(are) shut off, one other charging pump will always be available to perform its required function, since the Technical Specifications require two charging pumps to be operable in Modes 1 thru 4. Also, because the Reactor Coolant Pumps (RCP) have a redundant means of cooling, shutting off Charging Pump (s) 3B/4B, if operating, will have no adverse effect on the RCP seals.

Issued: May 19, 1989 145

- , y' F ._. _ . _ _ . _ . _ .

.t.-l:c:4'; ' ( 'j ,-

": , ., t ^ 36 4

.d ? -

II:(['r 2f ?-; ,

F .->;;q. &

I 4 --

g y

...--- .. e Ttie J following: procedure was : developed 'as allowed by.10CFR50.59. - as 2

ef:M:l ' < partially justified by the preceding. Safety, Evaluation: , ,

,n '

i;j'~ .

, , L3-ONOP-Ol6.9 Response to a' Reported Fire' in the Charging Pump Room, MCC,B Room,. Cable Spreading' Room or. Control'Acom

+

, r" t

Y I ;4 k.

I J i e

1 6

H L

l l'

146 p,

l. SAFETY E*#ALUATIONT . JPN-PTN-8ENJ-89-047 Revision 0 l

l' COORDINATED LITRIUN/ BORON CORRELATION . POR BORON CONCENTRATIONS-BETWEEN 1200 AND 2000 PPN Turkey Point Units 3 & 4 currently use a coordinated lithium / boron treatment program for reactor coolant pH. This program was specified by Westinghouse in Standard Information Package, Document 5-1, Revision 4. The coordinated lithium / boron curve supplied by the above document maintains a constant hot reactor coolant pH of l 6.9 but does not include boron concentrations above 1200 ppm. With i

extended fuel cycles, the boron concentration after refueling exceeds 1200 ppm for.the first three to five effective full power days. Westinghouse has extended the existing coordinated lithium / boron program to maintain a constant hot rea<: tor coolant l pH of 6.9 for boron concentrations up to 2000 ppm which would

result in a maximum lithium concentration of 3.75 ppm.

Safety Evaluation Summary:

The proposed change increases the reactor coolant system lithium i

concentration and pH. durir.g operation with boron concentrations l above 1200 ppm. This change does not degrade any systems, components or materials that come in contact with reactor c,olant i

either during normal or LOCA conditions. Therefore, the.

l probability or consequences of an accident or malfunction of equipment important to safety are not increased.

This change will have no effect on core performance and will not cause any material degradation; therefore, there is no possibility that making this change could create an accident or malfunction of a different type than those previously evaluated.

The reactor coolant system Technical Specification basis is to minimize corrosion of systems and components that come in contact with reactor coolant. As presented in the evaluation above this change does not have any adverse effect on corrosion rates.

Therefore, this change does not reduce the margin of safety as defined in the basis for any technical Specification.

Issued: May 1, 1989 147

SAFETY EVALUATION: JPN-PTN-8ENJ-89-055 Revision 0 CONTROLLED GA8 DECAY TANK RELEASE 8 USING ALTERNATE RADIATION NONITORS In the FSAR section regarding monitoring of gaseous releases from the gas decay tanks through the plant vent, there is a discussion of automatic closure of the isolation valve RCV-014 on high radiation level detected by. radiation monitor R-14. Plant procedures, however, permit the performance of gas decay tank releases using the SPING-4 radiation monitor when R-14 is out of service. In this case there would be no automatic function.

Safety Evaluatit.1 Summary:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR are not increased by this procedural modification. The waste gas system is a non-nuclear safety related -

system which is not part of the RCS pressure boundary, and is not '

required to mitigate any accident.

Release of waste gas occurs only under a controlled procedural process, which includes monitoring at a predetermined release rate, therefore there is no possibility of an accident or malfunction of a different type than any previously analyzed in the FSAR.

The margin of safety as defined in the basis of Technical Specifications is not reduced by this change. Independent sampling of the gas decay tank and independent verification of the valve line-up for release are the primary methods of ensuring that the requirements of the Tech. Specs. regarding effluent releases are met.

Issued: May 17, 1989 l

l I

148 1

l SAFETY EVALUATION: JPN-PTN-SFMS-89-059 Revision 0 EVALUATION FOR MOV-4-859 MATERIAL CHANGE Motor Operated Valve 4-869 is a 1500 pound rated stainless steel motor operated valve located in the safety injection system to block the hot leg recirculation flow path. FPL discovered that l the existing valve discs installed in MOV 4-869 are made from carbon steel instead of stainless steel. This configuration is not in accordance with the original valve design. This evaluation addresses the safety significance of operating the MOV 4-869 with the carbon steel discs for one fuel cycle at which time stainless steel discs will be installed.

Safety Evaluation Summarv l

l Anchor Darling manufactures this particular valve design in both stainless steel and carbon steel versions. Anchor Darling has evaluated that the carbon steel discs are identical in form, fit and function to the stainleso steel discs. _

The structural integrity of the discs is not expected to be affected based on the calculated corrosion rate with the associated chemistry conditions remaining the same as referenced in this evaluation. This evaluation was performed for the period of one fuel cycle. The leak tightness is unaffected with these hardfaced carbon steel discs.

The carbon steel corrosion products may be transported into the reactor coolant system and may result in an increased concentration of suspended solids, potential increased dose rates and a greater potential of contamination. These conditions will have no effect i on plant safety.

1 Issued: May 26, 1989 l

149

l SAFETY EVALUATION: JPN-PTN-SEMS-89-060 Revision 0 EVALUATION FOR MOV-3-869 MATERIAL CHANGE Motor Operated Valve 3-869 is a 1500 pound rated stainless steel motor l operated valve located in ths . safety injection system to block the hot leg recirculation flow path. FPL discovered that-the existing valve discs installed in MOV 3-869 are made from l carbon steel instead of stainless steel. This configuration is not in accordance with the original valve design. This evaluation addresses the safety significance of operating the MOV 3-869 with the carbon steel discs for one fuel cycle (up to a total of 30 months from initial installation) at which time stainless steel df.scs will be installed.

Safety Evaluation Summary hnchor Darling-manufactures this particular valve design in both stainless steel and carbon steel versions. Anchor Darling has evaluated that the carbon steel discs are identical in form, fit -'

and function to the stainless steel discs.

The structural integrity of the discs is not expected to be affected based on the calculated corrosion rate with the associated chemistry conditions remaining the same as referenced in this evaluation. This evaluation was performed for the period of one fuel cycle. The leak tightness is unaffected with these hardfaced carbon steel discs.

The carbon steel corrosion products may be transported into the reactor coolant system and may result in an increased concentration of suspended solids, potential increased dose rates and a greater potential of contamination. These conditions will have no effect on plant safety.

Issued: May 31, 1989 i

150 1

l SAFETY EVALUATION: JPN-PTN-SEIJ-89-061 Revision 0 TSA INSTRUMENTATION LOOP MODIFICATIONS i

To alleviate the effect of common mode voltage concern on the primary loops, the affected Safety Parameter Display System (SPDS) signals for Volume Control Tank (VCT) Level, Reactor Coolant System (RCS) Bypass Loop Temperature, Steam Generator Blowdown and Heat Exchanger Flow will be disconnected from the primary loop at the signal conditioner. These instrument loops provide control functions for the plant and are not associated with the protection channels. The modification consists of lifting the leads at Panel 3 (4) C245 located in the Computer Room and will be done under a temporary system alteration (TSA). The affected loops are listed below:

Loon No. Description LT-*-112 Volume Control Tank -

TE-*-411B & C RCS Bypass Loop Temperature '

TE-*-421B & C RCS Bypass Loop Temperature TE-*-431B & C RCS Bypass Loop Temperature FT-*-6277A, B, &C Stcam Generator A/B/C Blowdown Flow FT-*-6274 Blowdown Heat Exchanger to Condenser Flow As described in FSAR Section 7.2 and Appendix 7A, the Safety Parameter Display System (SPDS) aids operating personnel in the control room in making rapid assessments of plant status. It is a non-safety related system but designed to meet the reliability requirements of NUREG-0696, " Functional Criteria for Emergency Response Facilities."

Of the specific instrumentation described in FSAR Section 1.0, only the Volume Contrcl Tank (VCT) level instrument (LT-*-112) shown in figure 9.2-la of the FSAR, is used for Post Accident Monitoring as required by Regulatory Guide 1.97, Revision 3, " Instrumentation for 1-Light Water Cooled Nuclear Power Plants to Assess Plant and s Environmental Conditions during and Following an Accident". The Regulatory Guide 1.97, Revision 3 requirements described in the FSAR, Section 7.5.4, covers the requirements of 10 CFR 50.49, NUREG 0737 and Generic Letter 82-33.

FSAR Tables 7.5-1 and 7.5-2, which include LT-*-112, present the parameter listing summary sheets for Turkey Point Units 3 and 4, l developed in response to Regulatory Guide 1.97, Revision 3, and transmitted to the NRC via FPL Letter L-85-176A, J. W. Williams, Jr. (FPL) to S. A. Varga (NRC), dated May 10, 1985.

i 151 1  :

e,o p

I.,

' Safety Evaluation Summary:

The proposed change removes the SPDS' input signals from the signal conditioner in the primary . control loop.

Primary control loop instruments continue' to provide their related control- and

indication functions. -The change-does not affect; (1).the Steam Generator Blowdown Flow- Control Loop's ability - to isolate . in

-response to a high~ radiation signal; (2) RCS Delta-Tavg control of f the RCS Bypass Loop; (3) Steam Generator Blowdown Heat-Exchanger flow to the. condenser; or (4) VCT Level _ due . to the redundant instrumentation. The proposed: change will enhance the capability of the_ primary instrumentation loops by eliminating the l-- cause of instrument ~ deviations'due to excessive loading. 'This configuration was previously evaluated as acceptable prior to the addition of the SPDS. modification.

Issued: . June 13, 1989 1

152

SAFETY EVALUATION: JPN-PTN-SENJ-89-064 Revision 0 INCREASED PRESSURIZER PRESSURE UNCERTAINTY For the Turkey Point Units 3 &4 accident analyses, initial plant condition assumptions are obtained by applying maximum steady-state uncertainties to the nominal rated values. For pressurizer pressure, an uncertainty of +/- 30 psi to account for steady state fluctuations and measurement error has been applied. This is an historical generic value that is not derived from detailed control channel uncertainty calculations. Recently, detailed calculations have been performed for Turkey Point that indicate that the pressurizer pressure control channel uncertainty is greater than

+/- 30 psi. The following evaluation documents the acceptability of a pressure uncertainty of +/- 50 psi with the respect to the FSAR Chapter 14 accident analyses.

Safety Evaluation Summary The effects of an increased steady pressurizer uncertainty from -

+/- 30 psi to +/- 50 psi have been evaluated for each of the '

licensing basis safety analyses for Turkey Point Units 3 & 4. The Pressurizer Pressure uncertainty for Loss of Coolant Accidents increase will not result in exceeding any design or regulatory limits of 10 CFR 50.46. The conclusions.of the Turkey Point Units 3 & 4 FSAR for all non-LOCA events remain valid for an increase in the pressurizer pressure uncertainty from +/- 30 psi to +/- 50 psi.

l Based on the evaluation, it is concluded that a +/- 50 psi uncertainty for the steady state pressurizer pressure does not adversely affect the results of the licensing basis FSAR accident analyses or the conclusions of the FSAR and is acceptable with the respect to those accidents addressed in the evaluation. This increased uncertainty will be considered for all future safety analyses and evaluations.

l Issued: June 2, 1989 l

l l

153 l

SAFETY EVALUATION: JPN-PTN-SEMJ-89-067 Revision 0 CHANGE TO THE ADMINISTRATIVE TEMPERATUF.E LIMITS ON HEATUP AND

=COOLDOWN RATES TO CORRECT AN FSAR ERROR I During a review of the FSAR a discrepancy was noted in Section

~

4.2.6. -The FSAR states that the administrative limit for the heatup and cooldown rates for the reactor coolant system (RCS) is maintained at 50 F/hr. This limit is incorrect. The correct limit for administrative purposes is 90 F/hr.

Safety Evaluation Summary This proposed change does not increase the probability:

• of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR inasmuch as the design limit for the RCS heatup and cooldown rate is 100 F/hr. and this -

administrative limit is set at 90 F/hr.

~

• for an accident or malfunction of a differant type than any previously evaluated in the FSAR because the design limit for heatup and cooldown rate is not changed or exceeded. In addition, the limits of the Technical Specifications are not exceeded.

The margin for safety as defined in the basis for any Technical Specification is not reduced since the RCS heatup and cooldown I

rates are governed by the curves in the Technical Specifications.

Issued: June 16, 1989 l

154

l SAFETY EVALUATION: SECL 88-530 ULTRASONIC INSPECTION OF NUCLEAR FUEL In preparation of the repair of spent fuel, ultrasonic inspection of the spent fuel will be conducted to identify failed fuel rods.

This inspection will be performed using the Asea Brown Boveri I Failed Fuel Rod Detection System (FFRDS). During the inspection ,

process the FFRDS is supported on the_ top of the spent fuel racks. 1 Safety Evaluation Summary There are adequate safety features built into the FFRDS to preclude fuel damage. The loads _ imposed on the spent fuel storage racks by the FFRDS under normal and seismic conditions are well within'the i

- acceptable limits set forth in the FSAR. Assurance is provided l that a drop accideret involving the FFRDS is enveloped by the fuel drop accident discussed in the FSAR. Spent fuel cooling ' is ,

provided for while the equipment is in place and while UT -' )

inspections are being conducted. l Issued: October 7, 1988 l The following procedure was issued as allowed by 10CFR50.59 as  !

justified by the preceding Safety Evaluation:

GBRA 010618 Fuel Assembly Inspection Procedure l

i 1

1 155 E ____ _ ________ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

j

l SAFETY EVALUATION: SECL-88-647 Revision 1 ALTERNATE BORATION PATH DURING THE REPAIR OF CVCS VALVE 3-268 1

. Valve 3-268 is located in the Chemical Volume and Control System (CVCS):in.the common charging pump suction header from the volume

~ Control Tank (VCT), Refueling Water Storage Tank (RWST) , and the l

Boric . Acid Tanks (BAT's)'. Therefore, to work on the: valve the flowpath must be isolated, thus making the. charging _ pumps l inoperable. With the charging pumps inoperable, the.boration path from the BAT's to the Reactor Coolant System (RCS) via the charging pumps and the boration path from the RWST to.the RCS : via the charging pumps is not available. Therefore, to meet'the intent.of Tech.. Spec. 3.6.a, which requires at least one flow path to the core for boron injection, an alternate acceptable boration path.

must be justified. This evaluation reviews the use of the flowpath from the RWST to the RCS via the High Head Safety Injection (HHSI) l-System.as an alternate boration flowpath -to meet the intent of Tech. Spec. 3.6.a, during Mode 5.

Safety Evaluation Summary:

A flowpath from the RWST to the RCS via a HMSI flow path is viable boration flowpath. The boron concentration in the RWST is sufficient to borate the RCS to cold shutdown conditions. There are two options for HHSI flow paths, (1) a RCS cold leg injection flow path, or (2) a RCS hot leg flow path. The possibility of a l boron dilution event from the Boric Acid Tank (BIT) is precluded I by maintaining an adequate BIT boron concentration when the BIT is in the HHSI injection path. Other boron dilution events are precluded anri adequate RCS low temperature overpressure protection 1 is provided by the requirement of this evaluation.

1 L

The following procedures were temporarily revised as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

AP 0103.3 Reactor Cold Shutdown Conditions 0-ONOP-046.3 Loss of Boration Flowpaths l

l=

156

SAFETY EVALUATION: 4-EOP-ES-1.3 TRANSFER TO COLD LEG RECIRCULATION The sequence of time-critical steps was modified to ensure that all interruptions of injection flow are kept below 2 minutes.

Safety Evaluation Summary This procedure is only used to mitigate the consequences of an accident. This procedure revision ensures that interruptions of injection flow are prevented from exceeding 2 minutes. Following this revision the same equipment is used to mitigate an accident as is used in the current revision of the procedure.

Issued: September 7, 1988 The following procedure was revised as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation: .

4-EOP-ES-1.3 Transfer to Cold Leg Recirculation l

157

A SAFETY EVALUATION: TP-523 ENERGENCY DIE 8EL GENERATOR FUEL OIL DUPLEX FILTER FOULING AND-CONFIGURATION TEST The Emergency Diesel Generator (EDG) Fuel Oil Filters are normally.

lined up for parallel operation. The present preferred method for changing out' the filter ' elements is to first shut down the engine and then take the engine out.'of service. Following changeout of the filter : elements, the engine must be. retested 'to prove operability in accordance with 0-OSP-023.1.

It is tiot desirable to shut down the engine in' order to change out' the filter.because.the EDG is then unavailable to supply emergency loads. In order to keep the engine operable and allow for the normal' impurities associated with diesel fuel oil', a method-for changing out the filters-with the engine operational is needed.

The purpose of.this test procedure is to provide.the operational data; involving filter element changeout procedure.' The data' .

results obtained in the test will be used in-the establishment of4

a. permanent method of changing out fuel oil filter .elementre with the emergency diesel generator fully operational.

Safety Evaluation Summary:

This test does not involve an unreviewed safety question due to the following requirements of this test:

The EDG will be. removed from service during the performance of the test. Following the performance of the test, it will be tested in accordance with 0-OSP-023.1 prior to placing the EDG back into service.

Shim stock material and filter elements will be inspected prior to use to insure - no foreign material or residue is introduced into the fuel oil system.

The EDG duplex oil filter assembly will be inspected to insure that no foreign material is left inside the assembly following the test.

Any leakage of residual oil from the filter element assembly will be wiped away prior to running the EDG for the 0-OSP-023.1 test.

l Issued: April 7, 1989 The -following . Temporary Procedure was issued as allowed by 10CFR50.59 as justified by the preceding Safety Evaluation:

TP-523 Emergency Diesel Generator Fuel Oil Duplex Filter Fouling and Configuration Test 158

l t

l l

l b

1 1'

SECTION 3 .

Unit 4 cycle 12 Core Load 159 l

l l

l i

I

_ _ _ . 1

UNIT 4 CYCLE 12 CORE LOAD The Turkey Point Unit 4 Cycle 12 core loading consists of 52 fresh Region 14 Optimized Fuel Assemblies (OFA), 76 previously burned OFA, and 29 previously burned Low Parasitic fuel assemblies. This cycle design also incorporates 2864 Integral Fuel Burnable Absorbers (IFBA) and 144 Wet Annular Burnable Absorbers (WABA).

In addition, 12 assemblies contain the reduced length hafnium pressurized thermal shock absorber rods. Forty of the Region 14 assemblies contain 60 IFBA pins each, while four assemblies contain 116 IFBA pins. The IFBA coating covers 108 inches of active fuel centered on the core mid-plane. The WABA are arranged in 16 clusters containing 8 pins each and 4 clusters containing 4 pins.

The WABA pins are 134 inches in length, centered on the core mid-plane. This core loading will provide a cycle length of 14,300 MWD /MTU including a 500 MWD /MTU power coastdown.

The Region 14 fuel assemblies are similar to the Region 13D and 13E assemblies used in cycle 11 except for the following: IFBA -

rods, extended burnup modifications, reconstitutable top nozzles,'

standardized fuel pellets, reduced fuel rod backfill pressures, 4g fuel rod plenum springs, and 304L stainless grid sleeve material.

These modifications provide either improved fuel performance, fuel utilization or manufacturing reliability. These features have been reviewed and approved by the NRC and have been shown to be compatible with the existing fuel assemblies in Turkey Point Unit 4.

Safety Evaluation Summary:

Based on the technical evaluation / analyses performed by Westinghouse, it can be concluded that the Turkey Point Unit 4 Cycle 12 reload design meets all design criteria, is bounded be the results of the referenced analyses, and can be implemented with no changes required to the existing Turkey Point Unit 4 Technical Specifications. Therefore, it can be stated that:

The Turkey Point Unit 4 Cycle 12 reload design does not change the overall configuration of the plant. The mode of operation of the plant remains unchanged. Changes to the fuel assembly design features, i.e., IFBA rods, extended burnup l modifications, reconstitutable top nozzles, standardized fuel pellets, reduced fuel rod backfill pressures, 4g fuel rod plenum springs, and 304 stainless grid sleeve material have been reviewed and approved by the NRC and have been shown to be compatible with the existing fuel assemblies in the Turkey Point Unit 4. The reload safety evaluation demonstrates that the consequences of an accident or malfunction have not been increased beyond those evaluated in the previous analyses since all transients meet current criteria. Therefore, the )

probability of occurrence or consequences of an accident or i 160 l

I

! 4

i malfunction of equipment important to safety previously evaluated in the safety analysis report, is not increased.

The Turkey Point Unit 4 Cycle 12 reload design does not change the overall configuration of the plant, or the mode of operation of the plant. The changes to the design features have been reviewed and approved by the NRC and have been shown to be compatible with the existing fuel assemblies in Turkey Point Unit 4. Therefore, a possibility for a new accident or equipment malfunction has not been created.

The Turkey Point Unit 4 Cycle 12 reload design neutronics input and the resulting safety analysis has been reviewed, and in all cases the results are well within the acceptance criteria of the design basis. Based on FPL's independent review of the RSE report it can be concluded that the Turkey Point Unit 4 Cycle 12 reload design does not result in a reduction to the margin of safety relative to the Technical Specification basis for Turkey Point Unit 4.  ;

161 l

1

SECTION 4 Annual Report of Power Operated Relief Valve (PORV) actuations r

162 l

?

l

ANNUAL REPORT OF SAFETY AND RELIEF VALVE CHALLENGES By letter dated June 18, 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 power operated relief valve (PORV) actuations for Turkey Point Units 3 and 4 from July 1,1988 to June i

30, 1989.

Procedure Title Kev 3-OP-041.4 and'4-OP-041.4 Overpressure Mitigating System 3-OSP-041.4 and 4-OSP-041.4 Overpressure Mitigating' System Nitrogen Backup leak and .

Functional Test OP O209.1 Valve Exercising' Procedure Unit 3 October 2, 1989 PORV 455C would not open when cycled per 3-OP-041.4.

October 2, 1989 PORV 456 was cycled per OP 0209.1 and 3-OP-041.4 October 3, 1988 PORV 455 was cycled twice per OP 0209.1.

October 4, 1988 The Over Pressure Mitigation System cycled a PORV. Which PORV operated is unknown.

October 4, 1988 PORV 455C was cycled per 3-OP-041.4  ;

October 15, 1988 PORV 455C was cycled per OP 0209.1.

October 18, 1988 PORV 455C and 456 were cycled per 3-OSP-041.4. I 163 i

____ .__ _ _ --__ _ i

October 19, 1988 PORV 455C and 456 were cycled per OP 0209.1 and 3-OP-041.4.

October 27, 1988. PORV 455C and 456 were cycled per 3-OSP-041.4.

October 28, 1988 PORV 456'was' cycled.12 times per 3-OSP-041.G.

October 30, 1988 PORV 455C was cycled 4 times per-3-OSP-041.4.

November 1, 1988 PORV 455C.was cycled per_OP 0209.1 and 3-OP-041.4.

November 2, 1988 PORV 456 was cycled per OP 0209.1, 3-OSP-041.4, and 3-OP-041.4.

November 17, 1988 PORV 455C and 456 were cycled per - -

3-OSP-041.4 and OP 0209.1.

November 18, 1988 PORV 455C and 456 were cycled per 3-OP-041.4.

November 19, 1988 PORV 456 was opened to provide a vent path.

November 25, 1988 PORV 455C and 456 were cycled per.

3-OP-041.4.

December'2, 1988 PORV 456 was cycled per OP 0209.1.

December 8, 1988 PORV 455C automatically opened on sensed high pressure.

December 20,'1988 PORV 455C and 456 were cycled per OP 0209.1 and 3-OP-041.4.

December 30, 1988 PORV 455C and 456 were cycled per 3-OSP-041.4.

l January 3, 1989 PORV 455C and 456 were cycled per 1

3-OSP-041.4.

January 17, 1989 PORV 455C and 456 were cycled per 3-OP-041.4. l January 25, 1989 PORV 455C and 456 were cycled per L 3-OP-041.4. 4 April 2, 1989 PORV 455C and 456 were cycled per 3-OP-041.4.

164 {

1

3 c

~

3 c

,m LApril 4,.1989 PORV 455C and 456'were' cycled'per OP 0209.1. ,

'Aprill5,1989 PORV 456 was cycled and 455C' .

)

failed to fully open when cycled-per 3-OSP-041.4

' April 6, 1989 PORV~455C was; cycled per OP'0209.1 and 3-OSP-041.4.

May.3, 1989 PORV 456'was cycled per

.3-OSP-041.4. ..

June 7, 1989. PORVI 455C and 456'was cycled per 3-OSP-041.4'.

. June 17,.1989 PORV 455C inadvertently opened while' drawing a steam bubble in ,

the Pressurizer.- -

l 165

~

a. .

l l

-Unit 4, PORY Cyclings I

September 20,.1988 PORV 455C and 456'were cycled per OP 0209.1 and 4-OP-041.4.

September 21, 1989 PORV 455C inadvertently opened due to incorrect sensed high' pressure due to the running of the B RCP..

March 3, 1989 PORV>455C and 456 were twice

opened per OP O209.1.

March 4, 1989 PORV 455C was inadvertently opened during-testing.

March 6, 1989 PORV 455C was manually opened and-closed.

March 13, 1989 PORV 456 failed to fully open.when cycled per 4-OSP-041.4.

April 4, 1989- PORV 455C would not cycle from the test rack but cycled from control board per 4-OSP-041.4. PORV 456 stroke times were too fast when cycled-per 4-OSP-041.4.

~ April 8, 1989 PORV 456 was cycled per 4-OSP-041.4.

April 9, 1989 PORV 456 was cycled per 4-OSP-041.4.

May 29, 1989 PORV 455C and 456 were cycled per OP 0209.1.

June 5, 1989 PORV 455C was cycled and 456 failed to operate correctly due to an excessive opening time.when cycled per 4-OSP-041.4.

June 6, 1989 PORV 456 failed to operate correctly-twice due to excessive opening time when cycled per 4-OSP-041.4. PORV 455C cycled correctly per 4-OSP-041.4. PORV 455C and 456 were cycled per OP 0209.1.

June 6, 1989 PORV 455C and 456 were cycled per 4-OSP-041.4 and OP 0209.1.

166

_ _ _ _ _ _ _ _ . _ _ _ _ _ . . . _ _ _ .t

June 7, 1989 PORV 456 was cycled per 4-OP-041.4 and 4-OSP-041.4. PORV 455C was cycled per 4-OP-041.4.

167 i

SECTION 5 Summary of Unit 4 Steam Generator Tube Inspection Results 168 i

FORM NIS-BB OWNERS' DATA REPORT FOR EDDY CURRENT EIANINATION RESULTS A3 rCquirOd by th3 prCvioiOC3 Of th3 ASME Cado Ru1GD Page 1 Of 5 8UMMARY OF EDDY CURRENT EXAMINATION RESULTS

=====____=================_========================= ============== _

PLANT TURKEY POINT NUCLEAR POWER PLANT UNIT NO. 4 EXAMINATION DATES: October 27, 1988 TERU November 15, 1988 STEAM TOTAL TOTAL TOTAL TOTAL TUBE 8 TOTAL GEN TUBE 8 IND. IND. > PLUGGED A8 TUBES NUMBER INSP FROM 20% TEAN PREVENTIVE PLUGGED

+0 39% OR = 40% -MAINT 4E210A 3,199 7 0 1 1 4E210B 3,207 5 0 0 0 4E2100 3,205 5 0 0 0 LOCATION OF INDICATIONS STEAM AVB SUPPORT 8 TOP OF TUBE SEEET GENERATOR BARS 1 THROUGE 6 TO FIRST SUPPORT BOT LEG COLD LEG BOT LEG COLD LEG 4E210A 0 4 3 0 0 4E210B 0 4 1 0 0 4E210C 1 1 3 0 0 CERTIFICATION OF RECORD We certify that.the statements in this record are correct and the tubes inspected were tested full length in accordance with the requirements of Section EI of the ASME Code.

FLORIDA POWER and LIGHT COMPANY (Organization)

Yl7 94 '

D. _ ? _'

wuE SUPERVISOR

[

169

STEAM GENERATOR TUBES PLUGGED page 2 of 5 STEAM GENERATOR STEAM GENERATOR STEAM GENERATOR 4E210A 4E210B 4E210C ROW COLUMN REMARK 8 ROW COLUMN REMARK 8 ROW COLUMN REMARK 8

~

'2 5 EOT LEG 405B4 O

m 170

__ _ _ _ _ _ _ _ _ _ __ _ i

FORM NIS-CD CWNERS' DATA REPO.T FOR EDDY CURRENT EIAMINATICN RESULTS Ao requircd by tho pr;vicicn3 Of th3 ASME C;d3 RulC3 page 3 of 5 EDDY CURRENT EXAMINATION RESULTS PLANT: TURKEY POINT NUCLEAR POWER PLANT UNIT NO. 4 STEAM GENERATOR: 4E210A EXAMINATION DATES: 11-07-88 THRU 11-15-88

% TUBE WALL ROW COLUMN PENETRATION ORIGIN LOCATION 28 14 37 O1B 42.1 ROT LEG 34 02C 2.6 COLD LEG 33 19 24 05H 42.8 HOT LEG 26 24 25 03H .0 HOT LEG .

?

29 25 35 04H 5.8 HOT LEG 36 01C .0 COLD LEG

14. 82 27 04C 9.2 COLD LEG l__

171

FORN NIS-CB CWNER8' DATA REPORT 10R EDDY CURRENT EIANINATICN RESULTE As required by th3 provicies.A of thO ASNE CIC3 RulCO page 4 of 5 EDDY CURRENT EXAMINATION RESULTS

===============_____=_==================___===____=========___

PLANT: TURKEY POINT NUCLEAR POWER PLANT UNIT NO. 4 l STEAN GENERATOR: 4E2108 EXAMINATION DATES: 10-27-88 TNRU 11-05-88

% TUBE WALL ROW COLUNN PENETRATION ORIGIN LOCATION 3 24 21 03C 25.5 COLD LEG 37 59 37 02E 19.3 EOT LEG 13 75 34 01E 48.8 EOT LEG .

22 02E 15.0 EOT LEG '-

14 82 29 02E 15.7 EOT LEG 172 I4

FORM NIC-BB CNNERS' DATA REPO.T FOR EDDY CURRENT EXAMINATION RESULTS As requirCd.by th3 proviCicn3 Cf tho A8ME C;C3 RulC3 page 5 of 5 EDDY CURRENT EXAMINATION RESULTS

======_===____ ===========================-=========

PLANT: TURKEY POINT NUCLEAR POWER PLANT UNIT NO. 4 STEAM GENERATOR: 4E210C EIANINATION DATES: 10-29-88 THRU 11-05-88

% TUBE WALL ROW COLUMN PENETRATION ORIGIN LOCATION 17 16 35 05E 47.3 BOT LEG 38 22 27 05C 38.5 COLD LEG 26 37 24 05C 31.2 COLD LEG ,

43 48 30 06C .6 COLD LEG 17 73 24 AV2 2.1 ROT LEG 173

j' ll CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT 4 10/88 COMPONENT : S/G A Page : 1 of 1 Date : 04/17/89 Time : 1:58 PM l

l l' Examination Dates : 10/27/88 thru 11/15/88 l

Total Number'of Tubes Inspected .....: 3199 Total Indications Between 20% and 39% ............: 7 Greater than or equal to 40% ...: 0 Total Tubes Plugged as Preventive Maint : 1 Total Tubes Plugged ....................: 1 s

Location Of Indications 20% to 100% -

Hot Leg Cold Leg TSH .5 to O1H -2.1 : 0 TSC .5 to 01C -?. 1 : 0 O1H -2.0 to 06H +2.0 : 4 Olc -2.0 to 06C +2.0 :- 3 06H +2.1 to AV1 -3.1 : 0 06C +2.1 to AV4 -3.1 : 0 AV1 -3.0 to AV4 -3 0 : 0 1

174

t'.'

h PLUGGABLE TVSES LIST TURKEY POINT 4:

10/08 C3NPONENT :$/G A Page : 1 of.1 Date :- 06/18/89 Time : 8:05 AN ~

I 1. I: lRE0l l TESTED l l. l l l l l j. j

.lROWlLINEl ZONE l EXT lLE3l EXTENT l REEL jCHl PROSE l ' LOCATION l%l Volts l Dog lDateset l^

2 l

l : 2l. 5l: 3l06ClNl 1045AN l l680-SF/RN l 0.0 - lTR$l 0.0l 0l BALANCE -l

'Nunter of Pluggebte Tubes : 1 Numer of Indications  : 1 Selection Criterie :

Percent T.W. ..: 40 Includes ADI,WOI,051,DTI,DRI,LPI,TBP,TRS,PTP Codes r

.t' e

s . ..

175

= ______ _ _-_

l'

'CUMMULATIVE EKAK!NAT10N REPORT L

TURKEY PO!".T 4 OUTAGE : 10/88 CDMPONENT : S/G A Page : 1 of 1 DESCRIPTION : 20% 70 100% Date : 4/17/89

! Time 2 3:05 Pu l' <

e......................................................................................................................

l , 'l l Extent l l 10/88 l- j N/A l' l Row l Col l Leg l Reg lTst/ Note l Reet l Probe l Location l Volts lDeglChl%lDiffl Location l Volts lDeglCh l % l l g...l...l...l...l........l........l............l..............l.....l...l...l...l....l..............l.....l...l...j...l l 2&l 14l N lTEClTEC ,PSl004AH l720SF/RN l01N42.1 l 1.1l145l 1l37l l l' l- l l l

l. .l l N lTEClTEC SSl004AN- l720$F/RM l02C 2.6 l .5l149l-1l34l l .l l '. l l l l33l19lNlTEClTEC, PSl006AN l720$F/RM l05N42.8 l 1.7l159l 1l24l l. l l l j. l l26l24lNlTEClTEC PSl008AM - l720SF/RN l03M .0 l .5l142lm1l25l l l j. l' l l

. l 29l 25l N jiEClTEC PSl009AM l720SF/RM l 04N 5.8 l 1.2l147l 1l35l_ l l l l l l l

, 'l l'-lNlTEClTEC PSl009AN l720-SF/Rn l01C .0 l .6l132lu1l36l- l l l j j' l

-l .81155l 1l27l l14l82lNlTEClTEC' PSl039AN .l720SF/RM l04C 9.2 l l l l l }

j...l...l...l...l........l........l............l..............l.....l...l...l...l....l..............l.....l...l...l...l N m r of INDICATIONS Selected from Current Outage : 7  ?,

stunner of TUBES Selected from current Dutage : 5 176

i

\

6 1 CUNMULti!VE EEAM!hATIDW REPORT TURKEY Po!NT 4

.0UTAGE : 10/8B C3ePONENT 2 $/G A Page 2 1 Of 1 DE$CRIPTION : 2DE TD 391 Date 3 4/17/99 I Time : 3:05 PM e.......................................................................................-..............................

l lEntent -l l 10/88 l l N/A l.

,, { Row l Col l Leg l Req lT&t/ Note l Reet l Probe j Location l Volts lDeglChl1lDiffj Location l Volts lDegJCh l % l g...-l...l...l...j........l........l............l..............l.....l...l...l...l....l..............;.....j...l...l...l l 2Bl 14) : 'lfEClTEC P5l004AH l720-$F/RM l01N42.1 l 1.1l145l 1l3Tl l l l .l l l l l l N jfEClTEC $$l004AM lT20-sF/RN l02C 2.6 l .5l149l 1l3&l l l l l l l j33l19lNlTEClTEC- P$l006AN- l7205F/an l05N42.8 l 1.7l159l 1l24l. l- l l l l j' l26l26l.NlTEClTEC PSlDOBAN l72G IF/RM l03N 0 l . .5l142lm1l25l l l l l l l

. l 29l 25l N lTEClTEC .Psl009AN l72Li-$F/am lMN 5.8 l 1.2l147l 1l 35l l l l_ l l l 'l l -l l N lTEClTEC P5l009AN l720SF/AM l01C .0 l .6l132ln1l36l l l l l l l )

l 14] 82l N lTEClTEC P$l039AN l720sr/nm l%C 9.2 l .8l155l 1l27l l l 'l l- l l j...l...l...l...l........l........l............l..............l.....l...l...l...l....l..............l.....l...l...l...l

.Cueer of 1N0! CATIONS Selected from Current Outage : 7 Tc Nu eer of TUBES Selected from current outage : 5 1

l 177  !

u_ _ _ _ _ _ _ _ _ __.._ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ __ _____. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _

___j

x. '

CUMMUIATIVE DISTRIBUTION

SUMMARY

TURKEY POINT 4 10/88 COMPONENT :-S/G B Page : 1 of 1 Date : 04/17/89 Time : 1:59 PM Examination Dates : 10/27/88 thru '11/15/88 Total' Number of Tubes Inspected .....: 3207 Total Indications Between 2 0 % ' and 3 9 % . . . . . . . . . . . . : 5 Greater than or equal to 40%-...: 0 Total Tubes Plugged as' Preventive Maint : 0 Total Tubes P1ugged'....................: O Location Of Indications 20% to 100%

Hot Leg Cold Leg TSH .5 to O1H -2,1 : 0 TSC .5 to 01C -2.1 : 0 O1H -2.0 to 06H'+2.0 : 4 01C -2.0 to 06C +2.0 : 1, 06H-+2.1 to AV1 -3.1 : 0 06C +2.1 to AV4 -3.1 : 0 AV1 -3.0 to AV4 -3.0 : O'

[

1 l-178

n__ __ ._

i I

CUMMULATIVF EXAMINATION REPORT TURKEY POINT 4' DUTAGE .: 10/88

. COMPONENT : $/C 8 Page : 1 of 1 DESCRIPTION : 20X TO 1001 Date : . 4/17/89 Time : 3:06 PM l: l Extent l l. 10/88 .l l N/A -l l Row l col l Leg l Reg lTst/ Note l Reet l Prebe l' Location l Volts lDeglChjXlDiffl LOcetion l Volts lDeglChlXl j...l...l...l...l........l........l............l..............l.....l...l...l...l....l..............l.....l...l...l...l j 3l 24l N lTEClTEC PCl0098N l720.$F/RM l03C 25 5- l .6l155l 1l21l l. l l l l l l

l37l69)NlTEClTEC PSl0348W l720SF/RM l02N19.3 l .6l142l 1l 37l l l l l- l j l 13l 75l~ H ltEClTEC PClC368N l720$F/RM 101H48.8 l .6l141l 1l34l l -l l l' l l l_ l l N lTEClTEC P5l0368N l720-5F/RM l02H15.0 l 1.6l154l 1l22l l l l l l l

- ltij82lNlTEClTEC P5l0398N l720.SF/RM l02H15.7 l 1.1l149l 1l29l l l. l .l l 1 l...l...l...l...l........j........l............l..............l.....l...l...l...l....l..............l.....l...l...l...l Nunterof INDICATIONS Selected from Current Outage : 5-Nupter of TUSES Selected from Current Dutage : 4 4D l

~

f.

1 I

l 179

__-___.________________.-_m______-__-._..m _______.m .m______.______________.__________.m__.__.m ______ _ _ _ _ _ _ _ - _ _ - _ - _ - . _ _ _ . _ . , _ _ _ ,

v x r.;

"if 9

CUMMULATIVE EXAMINATION REPORT.

TURKEY POINT 4 OUTAGE : 10/88-WNT. : S/G B Pe9e : 1 of 1-

~2 DESCRIPfl0N 20X TO 391 Date t 4/17/89 Time : 3:06 PM

, o.....................................................................................................................+

.l- 'l Extent l l 10/88 l- l' N/A l.

. .l Row l COL l Leg l Req lTst/Notej . Reet l Probe l Location l Volts l Dog lChl%lDiffl Location l Volts l Dog lChl%l

J...j...l...l...l........l........l...........,l..............l.....l...l...l...l....l..............l.....l...l...J...l

-l 3l24lNlTEClTEC. PClG098H l720-$r/nN l03C25.5 l .6l155l 1l21l l l l l l- l.

l37l69lHlTEClTEC PSl0348H l720SF/RM l02H19.3l l .6l142l'1l37l l l l l l l l 13l 75l H,lTEClTEC PC]D36BN Ll720SF/RM j01H48.0 l .6l141l 1l34l- l } l l l l

_l l- lN.lTEClTEC -PSl0368N l720SF/RM l02H15.0 l 1.6l158] 1l22l l l l. l' l l l14l82l.HlTEClTEC- PSj0398H l720.SF/RM l02H15.7 l 1.1l149l 1l29l l l l l j. l

[...l...j...l...l........l........l............l..............l.....l...l...j...l....l..............l.....g...l...l...l-

'NWher of INQlCATIONS Selected fM W Current Dutage : 5

.Naber of TU8ES Selected from Current Dutage 4 s

1 180 t

CUMMUIATIVE DISTRIBUTION

SUMMARY

TURKEY POINT 4 10/88

-COMPONENT.: S/G.C' Page : 1 of 1 Date : 04/17/89' Time : 1:58 FM Examination Dates :' 10/27/88 thru 11/15/88 1

l: Total Number of Tubes Inspected .....: ~3205 Total Indications Between 20% and 39% ............': 5 Greater than or equal to 40% ...: 0 Total Tubes Plugged as Preventive Maint :- 0 Total Tubes Plugged ....................: 0 Location of Indications 20% to'100%

Hot Leg Cold Leg TSH .5 to O1H -2.1  : 0 TSC .5 to Olc -2.1 : 0 O1H -2.0 to 06H +2.0  : 1 01C -2.0 to 06C +2.0 : 3

! 06H +2.1 to AV1 -3.1  : 0 06C +2.1 to AV4 -3.1 : 0 AV1 -3.0 to AV4'-3.0  : 1 l

l 181 1

r CUMMULATIVE EXAMINATION REPORT TURKEY POINT 4 OUTAGE : 10/88 COMPONENT : S/G C 889e : 1 of 1 DESCRIPTION : 20% TO 100% Date : 4/17/89 Time 3:06 PM o.'...................................................e................................................................+

.l l Extent l- l 10/88 l l N/A l l Row lColl Leg l Req lTst/ Note l Reet l Probe l~ Location [ Volts lDeglC4l1lDiffl Location l Volts lDeglChl%l g...l...l...l...l........l........l............l..............l.....l...l...l...l....j..............l.....l...l...l...l.

- l 17] 16l N lTEClTEC PSl005CH l720SF/RM l05N47.3 l 7l147l 1l35l 'l l l l l l l38l22lNlTEClTEC SSl008CH . l720-$F/RM l05C38.5 l .6l15Tl 1l27l l l l ll l l 26l 3Tl N lTEClTEC . PSl016CH l720-SF/RM l05C31.2 l .5l159l 1l24l l l l 'l l l l 63l 48l N [TEClTEC PSl023CH l720SF/RM 106C .6 l .8l1kdlM1l30l l l l l l j.

l l17l73lNlTEClTEC PSj036CN l720SF/RM lAV2 2.1 l .9l125lM1l24l j. l l l l -l l...j...l...l...l........l........l............l..............l.....l...j...l...l....l..............l.....l...l...l...:

N N r of INDICAT!DNS Selected from Current outage : 5 Cueer of TUBES selected from current outage : 5 l

l l

l l

l

' 182 l

m.___ _ _ _ . _ ._ _ _ . _

CUIgULATIVE EXAMINATION REPORT TURKEY POINT 4 OUTAGE : 10/88 COMPONENT $/G C '# age : 1 of 1 DESCRIPT10N : 20% TO 391 Date 4/17/89 Time : 3:07 PM o............................. ........................................................................................

-l l Extent _j~ l 10/88 l l N/A l l Row l Col l Leg l Req lTst/ Note l Reel-l Probe l Location l Volts l Dog lChl1j0iffl Locetion ' l Volts lDeglChl1l, l...l...l...l...l........l........l............l..............l.....l...l...l...l....l..............l.....l...l...l...l lj17l16lNlTEClTEC. PSl005CN l720SF/RM l05N 47.3 l .7l147l 1l35l l l l l. l l l38l22lNlTEClTEC SSl008CN l720SF/RM l05C38.5 l .6l157l 1l27l l .l l- l l l 126l37lNlTEClTEC PSl016CH l720.SF/RM l05C31.2 l .5l159l 1l24l l l l l l l l 43l 48l N lTEClTEC 'PSl023CN l720-SF/RM l06C .6 l .8l130lM1l30l l l l l' l l l17l73lNlTEClTEC PSl036CN l720SF/RM lAv2 2.1 l .9l125lM1l24l l l l l l l.

.j...j...l...l...l........l........l............l..............l.....l...l...l...[....l..............l.....l...l...l...l.

Number of INDICAT!DNS $ elected from Current Outage : 5 Nweer of TUBES Selec'ted from current outage : 5-183

- - _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ __- _.