Text

Markup of Final Draft Tech Specs
	ML17347B114
Person / Time
Site:	Turkey Point
Issue date:	06/05/1989
From:	; FLORIDA POWER & LIGHT CO.
To:	;
Shared Package
ML17347B113	List: ML17347B114;
References
	NUDOCS 8906070247
	Download: ML17347B114 (532)
	v • d • e

ATTACHMENT I Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendment Revised Technical S ecifications MARKUP OF FINAL DRAFT OF TECHNICAL SPECIFICAT10NS 8906070247 890605 PDR ADOCK 05000250 P PDC

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INDEX DEFINITIONS PAGE le 0 DEFINITIONS' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-0

1. 1 ACTION.............

1.2 ACTUATION LOGIC TEST..............

1.3 ANALOG CHANNEL OPERATIONAL TEST...

1.4 AXIAL FLUX DIFFERENCE...

1. 5 CHANNEL CALIBRATION.................................. -... ~....

1. 6 CHANNEL CHECK.................................................

l. 7 CONTAINMENT INTEGRITY......................................... 1-2

l. 8 CONTROLLED LEAKAGE............................................ 1-2

1. 9 CORE ALTERATION................... 1-2 1.10 DOSE E(UIVALENT I-131............ ~ ~ ~ ~ ~ 1-2

1. 11 E-AVERAGE DISINTEGRATION ENERGY.- 1-2

1. 12 FREQUENCY NOTATION.;............. 1-3

.1.13 GAS DECAY TANK SYSTEM............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-3

l. 14 IDENTIFIED LEAKAGE...,........... 1-3 1.15 MEMBER(S) OF THE PUBLIC.......... 1-3

1. 16 OFFSITE DOSE CALCULATION MANUAL.. 1-3 1.17 OPERABLE - OPERABILITY........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-4 1.18 OPERATIONAL MODE - MODE.......... 1-4 1.19 PHYSICS TESTS.......:............

1.20 PRESSURE BOUNDARY LEAKAGE........ 1-4 1.21 PROCESS CONTROL PROGRAM...................................... 1-4 1.22 PURGE - PURGING............................................-. 1-4 1.23 QUADRANT POWER TILT RATIO........ 1-5 1.24 RATED THERMAL POWER.............. 1-5 1.25 REPORTABLE EVENT............................................. 1-5 1.26 SHUTDOWN MARGIN.............................................. 1-5 1.27 SITE BOUNDARY................................................ 1-5 0 TURKEY POINT " UNITS 3 & 4 AMENDMENT NOS. ANO FEB 28 19s9

INOEX DEFINITIONS PAGE

1. 28 SOLIDIFICATION............................................, .. 1-5

1. 29 S OURCE CHECKe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'v ~ ~ 1-5

1. 30 STAGGERED TEST BASIS......................................... 1-5

1. 31 THERMAL POWER.....'............... 1-6

l. 32 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6

1. 33 UNIDENTIFIED LEAKAGE.......... 1-6

1. 34 UNRESTRICTED AREA...................... 1-6

1. 35 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-6

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l. 36 VENTINGo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 TABLE 1. 1 FREQUENCY NOTATION................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 TABLE 1.2 OPERATIONAL MODES................. 1-8 TURKEY POINT - UNITS 3 8L 4 AMENDMENT NOS. AND PPB R 8 1SBQ

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INOEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PAGE

2. 1 SAFETY LIMITS 2.1.1 REACTOR CORE..................................,...........'.. 2-1

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2. 1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION PAGE

2. 1 SAFETY LIMITS 2.1.1 REACTOR CORE..................................,...,......... 8 2"1

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE..................'........... B 2-2

2. 2 LIMITING SAFETY SYSTEM SETTINGS

2. 2. 1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS............... B 2-3 TURKEY POINT - UNITS 3 8L 4 AMENDMENT NOS. AND

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INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4. 0 APPLICABILITY............................................... 3/4 0-1 3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORATION CONTROL Shutdown Margin - T Greater Than 200 F.............. 3/4 1-1 FIGURE 3. 1-1 REQUIRED SHUTDOWN MARGIN VERSUS REACTOR COOLANT BORON CONCENTRATION..........................-...... 3/4 1"3 Shutdown Margin - T Less Than or Equal to 200'F..... 3/4 1-4 Moderator Temperature Coefficient...................... 3/4 1-5 Minimum Temperature for Criticality.................... 3/4 1-7 3/4. 1. 2 BORATION SYSTEMS Flow Path - Shutdown................................ 3/4'1-8 Flow Paths - Operating.............................. 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-11 Borated Water Source - Shutdown.......................... 3/4 1-12.

Borated Water Sources - Operating........................ 3/4 1-14 H eat Tracing............................................. 3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES G roup Height.....................i.. ~ ~ . ~ ~ ~ ~ .~. ~ ~ .. ~ ~ 3/4 1-17 TABLE 3. 1-1 ACCIDENT ANALYSES'EQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD........... 3/4 1-19 Position Indication Systems - Operating............. 3/4 1-20 TABLE 4. 1-1 ROD POSITION INDICATOR SURVEILLANCE REQUIREMENTS. 3/4 1-22 Position Indication System - Shutdown............... 3/4 1-23

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R od Dr op Tlmeo i ~ ~ o ~ . ~ o ~ ~ ~ ~ ~ . ~ ~ ~ i. o. o ~ o o ~ . ~ o ~ ~ o ~ ~ ~ ~ . o 3/4 1-24 Shutdown Rod Insertion Limit........................ 3/4 1-25 Control Rod Insertion Limits............................. 3/4 1"26 FIGURE 3.1"2 ROD BANK INSERTION LIMITS'VERSUS THERMAL POWER THREE-LOOP OPERATION................................. 3/4 1-27 TURKEY POINT - UNITS 3 8c 4 AMENDMENT NOS. AND

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INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS PAGE 3/4. 2 POWER DISTRIBUTION LIMITS 3/4. 2. 1 AXIAL FLUX DIFFERENCE............................. 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER........................... 3/4 2-3 3/4. 2. 2 HEAT FLUX HOT CHANNEL FACTOR.......... ~ ~ ~ ~ ~ ~ ~ 3/4 2-4 FIGURE 3.2-2 K(Z) " NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT. 3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.......... 3/4 2-11 3/4. 2. 4 QUADRANT POWER TILT RATIO.......................'.. 3/4 2-13 3/4. 2. 5 DNB PARAMETERS.............; ...................... 3/4 2-16 3/4.3 INSTRUMENTATION 3/4. 3. 1 REACTOR TRIP SYSTEM INSTRUMENTATION..... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 4. 3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..;.."...................................... 3/4 3"8 3/4.3.2

~ ~ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-13 TABLE 3.3-2

~ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3" 14 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS........................... 3/4 3-23 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3"29 3/4. 3. 3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-35 0 TURKEY POINT - UNITS 3 8[ 4 AMENDMENT NOS. AND FEB 28 19sQ

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'LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREHENTS SECTION PAGE TABLE 3-3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS..................................... 3/4 3"36 TABLE 4. 3" 3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.................. 3/4 3-39

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Movable Incore Detectors.............................. 3/4 3-40 Accident Monitoring Instrumentation................... 3/4 3-41 TABLE 3. 3-5 ACCIDENT MONITORING INSTRUMENTATION................ 3/4 3-42 TABLE 4. 3-4 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS....................................'...... 3/4 3-46 Fire Detection Instrumentation........................ 3/4 3-47 TABLE 3. 3-6 FIRE DETECTION INSTRUMENTS.......................;.... 3/4 3 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-50 TABLE 3. 3-7 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-51 TABLE 4. 3-5 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-53

.Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-54 TABLE 3. 3-8 RADIOACTIVE GASEOUS, EFFLUENT HONITORING INSTRUMENTATION.......................................... 3/4 3-55 TABLE 4. 3-6 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-58 TURKEY POINT - UNITS 3 8 4 vl AMENDMENT NOS. AND 5 )989 FPg g FKB>>

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS PAGE 3/4. 4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation.............................. 3/4 4-1 H ot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops Filled............................. 3/4 4-5 Cold Shutdown - Loops Not Filled......................... 3/4 4"6

. 3/4.4.2 SAFETY VALVES

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S hutdown,,. 3/4 4-7 0 perat~ng.........................................;... 3/4 4 3/4.4.3 P RESSURIZER.............................................. 3/4 4-9 3/4.4.4 RELIfF VALVES............................................ 3/4 4-10 3/4.4. 5 STfAM GENERATORS......................................... 3/4 4-11 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTEO DURING INSERVICE INSPECTION............................. 3/4 4-16 TABLE 4.4-2 STEAM

~ GENERATOR TUBE INSPECTION....................... 3/4 4-17 3/4.4.6 .REACTOR COOLANT SYSTEM LEAKAGE

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Leakage Detection Systems................................ 3/4 4.-18 0 peratsonal Leakage...................................... 3/4 4-19 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-22 3/4.4.7 CHEMISTRY.......;........................................ 3/4 4-23 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-24 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE R EQUIREMENTS............................................. 3/4 4-25 3/4.4.8 SPfCIFIC ACTIVITY........................................ 3/4 4-26 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PfRCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/gram DOSE EQUIVALENT I-131.................................... 3/4 4-27

'TABLE 4.4-4 REACTOR COOLANT SPECIPIC ACTIVITY SAMPLE AND ANALYSIS P ROGRAM.................................................. 3/4 4-28 TURKEY POINT - UNITS 3 8E 4 Vl i AMENDMENT NOS. AND FEB R 8 1989

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System. 3/4 4-30 FIGURE 3.4-2 TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (60'F/hr)

APPLICABLE UP TO 20 EFPY....... 3/4 4-31 FIGURE 3.4-3 TURKEY POINT UNITS 384 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (lOO~F/hr)

APPLICABLE UP TO 20 EFPY........... ~ .......,.......... 3/4 4-32 FIGURE 3.4-4 TURKEY POINT UNITS 354 REACTOR COOLANT SYSTEM COOLDOWN LIMITATION (100 F/hr)

APPLICABLE UP TO 20 EFPY................. 3/4 4-33 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE . 3/4 4-34 Pressurizer............. 3/4 4-35 Overpressure Mitigating

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Systems...................... 3/4 4-36 3/4.4. 10 STRUCTURAL INTEGRITY................................. 3/4 4-38 3/4.4. 11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5. 1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350 F.... 3/4 5-3 avg FIGURE 3.5-1 RHR PUMP CURVE. 3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - T LESS THAN 350 F. 3/4 5"8 3/4.5.4 REFUELING WATER STORAGE TANK. ........................... 3/4 5-9

. TURKEY POINT - UNITS 3 8( 4 vi 1 1 AMENDMENT NOS. AND MAY 0 5 1989

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INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS 0 SECTION PAGE 3/4. 6 CONTAINMENT SYSTEMS ATMOSPHERIC TYPE CONTAINMENT 3/4. 6. 1 PRIMARY CONTAINMENT Containment Integrity.:.................................. 3/4 6"1 t

C ontainment Leakage...................................... 3/4 6-2 Containment Air Locks.................................... 3/4 6"4 Internal Pressure........................................ 3/4 6-6 A ir Temperature........................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ 3/4 6-7 Containment Structural Integrity....... 3/4 6-8 Containment Ventilation System......... 3/4 6-11 3/4.6. 2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ l ~ ~ ~ ~ ~ ~ 3/4 6-12 Emergency Containment Cooling System... 3/4 6-14 3/4.6. 3 EMERGENCY CONTAINMENT FILTERING SYSTEM 3/4 6-15 3/4. 6. 4 ~

CONTAINMENT ISOLATION VALVES........... 3/4 6-17 3/4. 6. 5 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................

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3/4 6-19 3/4. 6. 6 POST ACCIDENT CONTAINMENT VENT SYSTEH.......... 3/4 6-20 TURKEY POINT - UNITS 3 5 4 ix AMENDMENT NOS. AND FEB 88 1S89

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7. 1 TURBINE CYCLE S afety Valves......................... 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION........ 3/4 7"2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP..................... 3/4 7-2 AUxiliary Feedwater System............................ 3/4 7"3 TABLE 3.7-3 AUXILIARY FEEDWATER SYSTEM OPERABILITY............. 3/4 7-5 Condensate Storage Tank....................... 3/4 7-6 Specific Activity............................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7 TABLE 4. 7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.......................... 3/4 7-9 Main Steam Line Isolation Valves..... 3/4 7-10 Standby Feedwater System............. 3/4 7-11 3/4.7. 2 COMPONENT COOLING WATER SYSTEM....... 3/4 7-12 3/4.7. 3 INTAKE COOLING WATER SYSTEM.............................. 3/4 7-14 3/4.7. 4 ULTIMATE HEAT SINK...................... -... -. - - - - - "

~ ~ ~ ~ 3/4 7-15 3/4. 7. 5 CONTROL ROOH VENTILATION SYSTEH.......................... 3/4 7-16 3/4. 7. 6 S NUBBERS................................................. 3/4 7-18 3/4. 7. 7 SEALED SOURCE CONTAMINATION.............................. 3/4 7-22 TURKEY POINT - UNITS 3 8 4 X AMENDMENT NOS. AND FEB X8 198$

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.7.8 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System. 3/4 7-24'/4 Spray and/or Sprinkler Systems. 7-27 Fire Hose Stations............. 3/4 7-29 TABLE 3.7-4 FIRE HOSE STATIONS... 3/4 7-30 Fire Hydrants and Hydrant Hose Houses.. 3/4 7-31 TABLE 3.7-5 FIRE HYDRANTS . 3/4 7-32

.3/4. 7. 9 FIRE RATED ASSEMBLIES 3/4 7-33 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8. 1 A. C. SOURCES Operating.'ABLE 3/4 8-1 4.8-1 DIESEL GENERATOR TEST SCHEDULE . 3/4 8-7 S hutdowno ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-8 3/4. 8. 2

~ ~ D. C. SOURCES 0 perating. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-9 TABLE 3.8-1 BATTERY CHARGER ALLOWABLE OUT-OF-SERVICE TIMES........ 3/4 8-11 TABLE '4. 8-2 BATTERY SURVEILLANCE REQUIREMENTS 3/4 8-13 S hutdown................................................. 3/4 8-14 3/4. 8. 3 ONSITE POWER DISTRIBUTION Operating. 3/4 8-15 Shutdown. 3/4 8" 17 TURKEY POINT - UNITS 3 8( 4 Xi AMENDMENT NOS. AND pQv p g 'f009

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INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREHENTS PAGE 3/4.9 REFUELING OPERATIONS 3/4.9. 1 BORON CONCENTRATION...................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION........................... . .. ......... 3/4 9-2 3/4.9.3 DECAY TIME............................................... 3/4 9"3 3/4.9. 4 CONTAINMENT BUILDING PENETRATIONS....................... 3/4 9-4 3/4.9. 5 COMMUNICATIONS................................,.......... 3/4 9-5 3/4.9. 6 MANIPULATOR CRANE........................................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS................... 3/4 9-7 3/4.9.8 RESIDUAL HEAT REHOVAL AND COOLANT CIRCULATION h

H 1gh Water Level......................................... 3/4 '9-8 Low Water Level..............".---."'""""."" "" 3/4 9-9 TURKEY POINT - UNITS 3 8c 4 xii AMENDMENT NOS. AND PPB t. ~ $ 89

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS PAGE 3/4. 9. 9 CONTAINMENT VENTILATION ISOLATION SYSTEM..............; .. 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL............................. 3/4 9"11 3/4.9 ~ 11 WATER LEVEL - STORAGE POOL .............................. 3/4 9"12 3/4.9.12 HANDLING OF SPENT FUEL CASK.............................. 3/4 9"1'3 3/4. 9. 13 RADIATION MONITORING..........................'........... 3/4 9-14 3/4.9. 14 SPENT FUEL STORAGE....................................... 3/4 9-15 TABLE 3.9.1 SPENT FUEL BURNUP REQUIREMENTS FOR STORAGE IN REGION II OF THE SPENT FUEL PIT.......................... 3/4 9-16 3/4. 10 SPECIAL TEST EXCEPTIONS e

I 3/4. 10. 1 SHUTDOWN MARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4. 10. 3 PHYSICS TESTS............................................ 3/4 10-3 t

3/4.10.4 (This specification number is not used).................. 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN.................... 3/4 10-5 3/4. 11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS C oncentration............................................ 3/4 11-1 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS P ROGRAMo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-2 D ose....................'................................. 3/4 11-5 Liquid Radwaste Treatment System......................... 3/4 11-6

'TURKEY POINT " UNITS 3 4 4 X111 AMENDMENT NOS. AND FEB 2s 19ag

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INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS PAGE 3/4. 11. 2 GASEOUS EFFLUENTS 0 ose Rate................................................ 3/4 11-7 TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS P ROG RAM o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-8 Dose - Nob1e Gases........................................ 3/4 11" 12 Dose - Iodine-131, Iodine-133, Tri tium, and Radioactive Material in Particu1ate Form...... 3/4 11"13 Gaseous Radwaste Treatment System. 3/4 11-14 Exp1osive Gas Mixture............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-15 Gas Decay Tanks................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ 3/4 11-16 3/4. 11. 3 SOLID RADIOACTIVE WASTES... ~..... ~ .. ~ c-. ~ ~ .. ~ . ~ ~ ~ ~ ~ ~ ~ ~, 3/4 11-17 3/4. 11.4 TOTAL DOSE............................................... 3/4 11-18 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12. I MONITORING PROGRAM....................................... 3/4 12-1 TABLE 3. 12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS. FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL.SAMPLES................................. 3/4 12-7 TABLE 4. 12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSI So ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 12-8 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-11 3/4. 12. 3 INTERLABORATORY COMPARISON PROGRAM....................... 3/4 12-13 TURKEY POINT - UNITS 3 8 4 xi v AMENDMENT NOS. AND

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INOEX BASES SECTION PAGE 3/4. 0 APPLICABILITY...............................................' 3/4 0-1 3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORATION CONTROL.......................................... B 3/4 1-1 3/4. 1. 2 BORATION SYSTEMS. B 3/4 1-2 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-4 3L4.2 POWER DISTRIBUTION LIMITS.-....-..-.-..................;.... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 FIGURE B 3/4.2-1 TYPICAL'NDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER............................................. B 3/4 2-3 3/4. 2. 2 and 3/4:2. 3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-4 3/4.2. 4 QUADRANT POWER TILT RATIO............................'... B 3/4 2-8 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-8 3'/4. 3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION................. B 3/4 3-1 3/4.3. 3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 TURKEY POINT - UNITS 3 8c 4 XV AMENDMENT NOS. AND l:EB 2,8 1989

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ENOEX BASES 0 3/4.4 REACTOR COOLANT SYSTEM PAGE 3/4.4.1 REACTOR COOLANT LOOPS ANO COOLANT CIRCULATION............. B 3/4 4"1 3/4. 4. 2 SAFETY VALVES............................................. B 3/4 4-2 3/4.4. 3 PRESSURIZER......................:....... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-3 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-4 3 S4.4.7 CHEMISTRY............................................;.... B 3/4 4 3/4.4. 8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4. 9 PRESSURE/TEMPERATURE LIMITS............................... B 3/4 4-7 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS - UNIT 3................. B 3/4 4-10 TABLE B 3/4.4-2 REACTOR VESSEL TOUGHNESS - UNIT 4. B 3/4 4-11 3/4.4. 10 STRUCTURAL INTEGRITY..................... B 3/4 4-16 3/4.4. 11 REACTOR COOLANT SYSTEM VENTS............................. B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4. 5. 1 ACCUMULATORS.....;........................................ B 3/4 5-1 3/4. 5.2 and 3/4. 5. 3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 0 TURKEY POINT - UNITS 3 8E 4 XV1 AMENOMENT NOS. ANO FEB 2S jggg

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INDEX SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIHARY CONTAINMENT...........'.......:.................... 8 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... 8 3/4 6-3 3/4.6.3 EMERGENCY CONTAINHENT FILTERING SYSTEM.................... B 3/4 6-3 3/4.6.4 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6"4 3/4. 6. 5 HYDROGEN MONITORS......................................... 8 3/4 6-4 3/4. 6. 6 POST ACCIDENT CONTAINMENT VENT SYSTEM..................... 8 3/4 6-4 e

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TURKEY POINT - UNITS 3 EE 4 xvH AMENDMENT NOS. AND FF.B R8 1SSS

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7"1 3/4.7.2 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-5 3/4.7.3 INTAKE COOLING WATER SYSTEM............................... B 3/4 7-5 3/4. 7. 4 ULTIMATE HEAT SINK........................................ B 3/4 7-5 3/4. 7. 5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................. B 3/4 7-6 374. 7. 6 S NUBBERS.................................................. B 3/4 7-6 3/4. 7. 7 SEALED SOURCE CONTAMINATION............................... B 3/4 7-7 SYSTEMS..................................

t 3/4. 7. 8 FIRE SUPPRESSION B 3/4 7-8 3/4.7. 9 FIRE RATED ASSEMBLIES.................................. B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8. 1, 3/4. 8. 2, and 3/4.8. 3 A.C. SOURCES, D. C. SOURCES, and ONSITE POWER DISTRIBUTION.................................. B 3/4 8-1 TURKEY POINT - UNITS 3 8c 4 xviii AMENDMENT NOS. AND FFB S8 1988

I if f

INDEX BASES PAGE 3/4. 9 REFUELING OPERATIONS 3/4. 9. 1 BORON CONCENTRATION................................,...... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ 8 3/4 9-1 3/4. 9. 4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9" 1 3/4. 9. 5 COMMUNICATIONS............................... B 3/4 9-1 3C4 9.6 MANIPULATOR CRANE ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ "" ~ ~ ~ ~ -"---- ~ "~ ~ ~ - ~ ' ~ ~ ~ B 3/4 9-2.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS................... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL ANO COOLANT CIRCULATION............. B 3/4 9-2 t 3/4.9.9 3/4. 9.12 3/4. 9. 13 3'/4.9.14 CONTAINMENT VENTILATION ISOLATION 3/4. 9. 10 and 3/4. 9. 11 STORAGE WATER LEVEL HANDLING OF SPENT FUEL SYSTEM..................

REACTOR VESSEL ANO POOL...................................

CASK....................

RADIATION MONITORING......................................

SPENT FUEL STORAGE........................................

B B

B 3/4 9-2 3/4 9-3 3/4 9-3 3/4 9-3 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4. 10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... 8 3/4 10-1 3/4.10.3 PHYSICS TESTS................,.......;..................... B 3/4 10-1 3/4. 10.4 (This specification number is not used)................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN..................... B 3/4 10-1 TURKEY POINT - UNITS 3 8E 4 Xix AMENDMENT NOS. AND FEB as ums 4 "44 ~ -%P'ITphptlc'g lf%}pvff '

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INDEX SECTION PAGE 3/4. 11 RADIOACTIVE EFFLUENTS 3/4. 11. 1 LItlUID EFFLUENTS.................... B 3/4 11-1 3/4. 11.2 GASEOUS EFFLUENTS................... B 3/4 11"3 3/4. 11. 3 SOLID RADIOACTIVE WASTES................................ B 3/4 11-6 3/4. 11.4 TOTAL DOSE.............................................. B 3/4 11-6 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12. 1 MONITORING PROGRAM.................. B 3/4 12-1 3/4.12.2 LAND USE CENSUS..................... B 3/4 12-1 ST4. 12. 3 INTERLABORATORY COMPARISON PROGRAM.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 12"2 TURKEY POINT " UNITS 3 8 4 XX AMENDMENT NOS. AND FPB R8 1989

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INOEX DESIGN FEATURES PAGE

5. 1 SITE

5. 1. 1 EXCLUSION AREA...:.......................................... 5-1 5.1.2 LOW POPULATION ZONE......................................... 5-1 5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LI(UID EFFLUENTS.................... 5-1

5. 2 CONTAINMENT 5.2. 1 CONFIGURATION............................................... 5-1

5. 2. 2 DESIGN PRESSURE AND TEMPERATURE............................. 5" 1 FIGURE 5.1-1 SITE AREA MAP........................................ 5-2 EIQURE 5.1-2 PLANT AREA MAP..................................;.... 5-3.

5. 3 REACTOR CORE 5.3. 1 FUEL ASSEMBLIES............................................. 5-4 t 5.3.2 CONTROL ROD ASSEMBLIES......................................

5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................

5 .4. 2 VOLUME......................................................

5; 5 METEOROLOGICAL TOWER LOCATION.....................:...........

5-4 5-4 5-4 5-4

5. 6 FUEL STORAGE 5.6. 1 CRITICALITY................................................. 5" 5 5 .6.2 DRAINAGE.................................................... 5-6 5 .6.3 CAPACITY.................................................... 5-6 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-6

.TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.................. 5-'7 TURKEY POINT - UNITS 3 4 4 '" XX1 AMENDMENT NOS. AND

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INOEX ADMINISTRATIVE CONTROLS SECTION PAGE

6. 1 RESPONSIBILITY.................. 6-1

6. 2 ORGANIZATION............ 6-1 6.2.1 ONSITE AND OFFSITE ORGANIZATION........................... 6" 1

6. 2. 2 PLANT STAFF ....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-2 TABLE 6.2"1 MINIMUM SHIFT CREW COMPOSITION...................... 6-4

6. 2. 3 SHIFT TECHNICAL ADVISOR................................... 6-5 6& FACILITY STAFF VALIF ICATIONS 6-5 6 .4 TRAINING.................................................... 6-5 6.5 REVIEW AND AUDIT.;........ 6-5 F unCtlOneeeeee ~ e.......e.eee..e....ee..ee..e.ee.eeeeee ~ eee 6-5 C omposltlon...............................................

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6-6 A lternates................................................ 6" 6 M t Frequency......;..................................

eetlng 6-6 UOrume ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-6 R esponslbllltles.............

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6-6 R ecords................................ 6-8 TURKEY POINT - UNITS 3 8a 4 XX11 AMENDMENT NOS. AND FEB a 8 typal

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INDEX ADMINISTRATIVE CONTROLS SECTION 6.5.2 COMPANY NUCLEAR REVIEW BOARD F UnCtl On ~ ~ ~ ~ ~ ~ io t~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 C OmPOS1

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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 6-8 Alternates............. ~ . 6-8 C onsultants............................................... 6-9 M t Frequency.........................................

eetlng 6-9 uorum.................................................... 6-9 R evlew..................... .............................. 6-9 Ad A~4 t

Udl ts ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-10 R ecordst ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t~~~~~~~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ 6-11 6.5.3 TECHNICAL REVIEW ANO CONTROL t tl J ~ ~g ~

ACtl Vl eS ~ ~ ~ ~ ~ ~ ~ p ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-11 6.6 REPORTABLE EVENT ACTION..................................... 6-12

6. 7 SAFETY LIMIT VIOLATION...................................... 6-12 6.8 PROCEDURES AND PROGRAMS..................................... 6-13

6. 9 REPORTING RE UIREMENTS...................................... 6-15

6. 9. 1 ROUTINE REPORTS........................................... 6-15 S tartup Report..................................... 6-15 Annual Reports..................................... 6-16 Annual Radiological Environmental Operating Report. 6"17 Semiannual Radioactive Effluent Release Report..... 6-18 Monthly Operating Report.......................;.......... 6-19 Peaking Factor Limit Report............................... 6-19 6.9.2 SPECIAL REPORTS................. 6-20 6-20 TURKEY POINT " UNITS 3 & 4 XX111 AMENDMENT NOS. ANO FEB 28 g8g

INDEX ADMINISTRATIVE CONTROLS SECTION

6. 11 RADIATION PROTECTION PROGRAM..............................; 6-21

6. 12 HIGH RADIATION AREA..'...................................... 6-22

6. 13 PROCESS CONTROL PROGRAM (PCP).............................. 6"23

6. 14 OFFSITE DOSE CALCULATION MANUAL ODCM ..................... 6-23

6. 15 MAJOR CHANGES TO LI UID GASEOUS AND SOLID RADWASTE TREATMENT SYSTEMS................................. 6-24 TURKEY POINT " UNITS 3 Ec 4 XX1 V AMENDMENT NOS. AND FEB ia >ea9

SECTION 1. 0 DEFINITIONS TURKEY POINT - UNITS 3 4 4 1-0 AMENOMENT NOS. ANO

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1. 0 OEF INITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications:

ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

ANALOG CHANNEL OPERATIONAL TEST 1% An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the setpoints are within the required range and accuracy.

AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other and/or status derived from independent instrument channels 'ndications measuring the same parameter.

TURKEY POINT " UNITS 3 8E 4 NENOMENT NOS. AND

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DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations'equired to be closed during accident conditions are either:

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6"1 of Specification 3.6.4.

b. The equipment hatch is closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and t

e. The sealing mechanism associated with each penetration (e. g., welds, bellows,'r 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

DOSE E UIVALENT I"131 1.10 DOSE E(UIVALENT I-131 shall be that concentration of I-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test R 8 " -'Ibl E-7 fNRCgg1 yGid 1.109,~.

E - AVERAGE DISINTEGRATION ENERGY t, Os~her, l'flan

1. 11 Z shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample isotopes, other than iodines, with half lives greater than 30 minutes, making up at least 95 percent of the total non-iodine activity in the coolant.

TURKEY POINT - UNITS 3 5 4 1"2 AMENDMENT NOS. AND

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DEFINITIONS FREINOENOT NOTATION

l. 12 The FRE(UENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GAS DECAY TANK SYSTEM

l. 13 A GAS DECAY TANK SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MEMBER S OF THE PUBLIC 1.15 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, vendors or members of the Armed Forces using property located within the SITE BOUNDARY. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recre-ational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL

1. 16 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

TURKEY POINT - UNITS 3 84 4 1-3 AMENDMENT NOS. AND N

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DEFINITIONS OPERABLE - OPERABILITY 1.17 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device t'o perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE " MODE

1. 18 An OPERATIONAL MODE (i. e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS

1. 19 PHYSICS TESTS shall be those tests performed to measure the fundamental ggclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 13.5 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.20 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM 1.21 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to-assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.

PURGE - PURGING

1. 22 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

TURKEY POINT - UNITS 3 4 4 1-4 AMENDMENT NOS. AND cm 2 v / j

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DEFINITIONS UADRANT POWER TILT RATIO 1.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2200 MWt.

REPORTABLE EVENT 1.25 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

MITDOWN MARGIN 1.26 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical.from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted. except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.27 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

ll SOLIDIFICATION 1.28 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1.29 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.30 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified, test interval into n equal subintervals, and

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

TURKEY POINT " UNITS 3 8L 4 1-5 AMENDMENT NOS. AND

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4

DEFINITIONS THERMAL POWER 1.31 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.32 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 1.33 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UHRESTRICTED AREA 1.34 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee .for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.35 A VENTILATION EXHAUST, TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or .HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.36 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

0 TURKEY POINT - UNITS 3 4 4 -. AMENDMENT NOS. AND

0 I

TABLE 1.1 FRE UENCY NOTATION NOTATION ~FRE UENCY At least once per 12 hours.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

At least once per 7 days.

At least once per 31 days.

At least once per 92 days.

SA At least once per 184 days.

At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

Completed prior to each release.

TURKEY POINT - UNITS 3 8( 4 1-7 AMENOHENT NOS. ANO

V al

TABLE 1.2 OPERATIONAL MODES REACTIVITY I RATED AVERAGE COOLANT MODE CONDITION K ff THERMAL POWER" TEMPERATURE

1. POWER OPERATION > 0.99 > 5X > 350'F

2. STARTUP > 0.99,, < 5X > 350 F

3. HOT STANDBY < 0.99 > 350 F

4. HOT SHUTDOWN < 0.99 350 F > T

> 200 F

5. COLD SHUTDOWN < 0.99 < 200 F

6. REFUELING"" < 0.95 < 140 F "Excluding decay heat.

""Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

TURKEY POINT - UNITS 3 8c 4 1" 8 AMENDMENT NOS. AND

0 2.0 SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS

2. 1 SAFETY LIMITS REACTOR CORE

2. l. 1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg ) shall not exceed the 1'imits shown in Figure 2. 1"1, for 3 loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.7. 1.

REACTOR COOLANT SYSTEM PRESSURE

2. 1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANOBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6. 7.

1 l.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7. 1.

TURKEY POINT " UNITS 3 8c 4 2-1 AMENDMENT NOS. ANO

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675 665 UNACC PTABLE 24 PSIA FERA TION 655 2250 P 635 2000 PSIA 625 615 sos 595 CCEPTABi E 585 OPERATION 575 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 POSER (FRACTION OF HOMINAL)

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION TURKEY POINT - UNITS 3 8L 4 2-2 AMENDMENT NOS. AND F,"B z s;-.-eo

P kyar

675 665 2400 PSIA UNACCEPTABLE 655 OPERATION 2250 PSIA 645 635 2000 PSIA 625 I 1825 PSIA 615 605 595 575 00 0.1 0.2 0.3 0.4 0.5 0.6 0.7 08 0.9 1.0 1.1 1.2 POWER (FRACTION OF NOMINAL)

FIGURE 2.1-1 REACTOR CORE SAFETY'IMIT

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SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2. 1 The Reactor Trip System Instrumentation and Interlock Setpoints shall.

be set consistent with the Trip Setpoint values shown in Table 2'.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

a~ With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2"1, adjust the setpoint consistent with the Trip setpoint value.

b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3. 1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

TURKEY POINT " UNITS 3 & 4 2" 3 AMENOMENT NOS.'ND n

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TABLE 2.2-1

~ M I R7 REACTOR TRIP SYSTEH INSTRUHENTATION TRIP SETPOINTS ITI CD FUNCTIONAL UNIT TRIP SETPOINT ALLOMABLE VALUE 8

1. Hanual Reactor Trip N.A. N.A.

2. Power Range, Neutron Flux

a. High Setpoint <109X of RTP*" <[ ]X of RTP**

C/l

b. Low Setpoint <25X of RTP** <[ ]X of RTP**

3. Intermediate Range, <25X of RTP~* <[ ]X of RTP*"

Neutron Flux

4. Source Range, Neutron Flux <10s cps <[ ] x 10s cps

5. Overtemperature hT See Note 1

6. Overpower hT See Note 3

7. Pressurizer Pressure-Low >1835 psig >[ ] psig

8. Pressurizer Pressure-High <2385 psig <[ ] psig

9. Pressurizer Water Level-High <92X of instrument'span <[ ]X of instrument span

10. Reactor Coolant Flow-Low >90X of loop >[ ]X of loop design flow" Gesign flow*

m 11. Steam Generator Mater >15X of narrow range >[ ]X of narrow range instrument Level Low-Low instrument span span oop es>gn f ow = 89,500 gpm

"*RTP = RATED THERHAL POMER CXl CCl

TABLE 2. 2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m

C) FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE ¹

12. Steam/Feedwater Flow Feed Flow Feed Flow Mismatch <0.64 x 10 lb/hr <[ ] x 10 lb/hr Coincident With below steam flow Below steam flow Steam Generator Water >15K of narrow <[ ]X of narrow range instrument Level-Low range instrument span span

13. Undervoltage - 4. 16 kV >2496 vol ts- ] vol ts-Busses A and B each bus each bus

14. Underfrequency - Trip of Reactor >56.1 Hz >[ ] Hz Coolant Pump Breaker(s) Open

15. Turbine Trip

a. Auto Stop Oil Pressure >45 psig >[ ] psig

b. Turbine Stop Valve Fully Closed """ Fully Closed """

Closure

16. Safety Injection Input N. A. N.A.

from ESF g 17. Reactor Trip System Interlocks

a. Intermediate Range >1 x 10- amp >[ ] amp Neutron Flux, P-6 L>mst swstch 1s set when Turbine Stop Valves are fully closed.

0 7%

TABLE . i Continued C

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 8 CD

b. Low Power Reactor Trips Block, P-7 I

1) P-10 input <10K of RTP*" <[ ]X of RTP**

2) Turbine First Stage <lOX Turbine Power <[ ]X Turbine Power Pressure

c. Power Range Neutron <45K of RTP** <[ ]X of RTP"*

Flux, P-8

d. Power Range Neutron >10K of RTP** >[ ]X of RTP**

Flux, P-10

18. Reactor Coolant Pump N.A. N.A.

I CJl Breaker Position Trip

19. Reactor Trip Breakers N.A. N.A.

20. Automatic Trip and Inter lock N.A. N.A.

Logic CD CD CA

- RTP = RATEO THERMAL POWER C'> ~

fp

('

TABLE Continued TABLE NOTATIONS NOTE 1: OVERTEMPERATURE KT(

1 + r S 4T

) <IIT (K o

-K ~

(1 + TIES)

[T(

1 + tgS

)- T']+ K(P-P') f (4I))

Where: 4T Measured 4T by RTD Instrumentation; 1

+ g~ Lag compensator on measured 4T; Time constants utilized in the lag compensator for 4T, ~8-s.

4T Indicated 4T at RATED THERMAL POWER~ g+RTt Rey~Se O'I < =~.~ S~

Kg 1.095; K2 0.0107/ F;

~1+ t S The function generated by the lead-lag compensator for T 1 + TIES dynamic compensation; i2s Ta Time- constants utiTized in the lead-lag compensator for T , x2 = 25s, ta = 3 s; Average tempei ature, F;

+ Lag compensator on measured lg Tav Time constant utilized in the measured Tav lag compensator v 574.2 F (Nominal T at RATED THERMAL POWER);

~ +gQ rp .~f(~e=2..5 3 '+

Kg 0.000453/psig; p Pressurizer pressure, psig;

lg I ~

iq,

TABLE . 2-1 (Continued I

Kl TABLE NOTATIONS Continued)'l pc m

NOTE 1: (Continued) 2235 psig (Nominal RCS operating pressure);

Laplace transform operator, s-',

and f (hI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For'qt - qb between - 14K and + lOX, f .(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qb exceeds - 14X, the hT Trip Setpoint shall be automatically reduced by 2.0X of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of q - qb exceeds + lOX, the hT Trip Setpoint shall be automatically reduced by 3.5X of its value at RATED THERMAL POWER.

NOTE 2: (This note number is not used.)

m C7 m

C)

CA GJ

0 TABLE . -1 Continued TABLE NOTATIONS Continued CI C

I NOTE 3: OVERPOWER (1 +

4T TIES)

<AT fK o

-K ~

(1 + xsS) (1 + rgS)

T-K [T (1 + vgS)

T"]-f(AI)j Where: 4T As defined in Note 1,

. 1

+ rg As defined in Note 1, As defined in Note 1, 4T As defined in Note 1, Kg 1. 09, Ks 0.02/ F for increasing average temperature and 0 for decreasing average temperature, x S

~+tg The function generated by the rate-lag compensator for T dynamic compensation, Ts Time constants utilized in the rate-lag compensator for T

, xs = 10 s, m

1 C) 1+ tg As defined in Note 1, m

C)

As defined in Note 1, Vl

TABLE . -1 Continued I

TABLE NOTATIONS Continued '

NOTE 3: (Continued)

I K6 0.00068/'F for T > T" and K6 = 0 for T < T",

T As defined in Note 1, T

II Indicated Tavg at RATED THERMAL POWER (Calibration temperature for hT instrumentation, < 574.2 F),

S As defined in Note 1, and f (hI) = As defined in Note 1.

I I

NOTE 4: (This note number is not used.)

CD CD 8 If no allowable valve is specified as indicated by f ], the trip set point shall also be the al l owab1 e value.

CA 0 ~

BASES FOR 4

SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS NOTE The BASES contained in succeeding pages summarize the reasons'or the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.

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2. 1 SAFETY LIMITS BASES

2. 1. 1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Oper ation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relationship has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distribu-tions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability with 95 percent confidence that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

The curves of Figure 2. 1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is equal to the enthalpy of 'saturated liquid.

N These curves are based on an enthalpy hot channel factor, F~, of 1.62 and and a reference cosine with a pe~k of 1.55 for axial power shape. An allowance is included for an increase in F~ at reduced power based on the expression:

F~ < 1.62 [1+ 0.3 (1-P)j Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion limit assuming the axial power imbalance is within the limits of the f (GI) function of the Overtemperature trip. When the axial power imbalance is not within. the tolerance, the axial power imbalance effect on the Over-temperature bT trips will reduce the setpoints to provide protection consistent with core Safety Limits.

TURKEY POINT - UNITS 3 4 4 B 2-1 AMENDMENT NOS. AND

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SAFETY LIMITS BASES

2. l. 1 REACTOR CORE (Continued)

Fuel rod bowing reduces the values of DNB ratio (DNBR). The penalties are calculated pursuant to "Fuel Rod Bow Evaluation," WCAP-8691-P-A Revision 1 (Proprietary) and WCAP-8692 Revision 1 (Non-Proprietary). The restrictions of the Core Thermal Hydraulic Safety Limits assure that an amount of DNBR margin greater than or equal to the above penalties is retained to offset the rod bow DNBR penalty.

2. 1. 2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of llOX (2735 psig) of design pressure. The RCS piping, valves and fittings are designed to ANSI B31. 1 which permits a maximum transient pressure of 120K -of design pressure af 2485 psig. The Safety Limit of 2735 psig is therefore more conservative than the ANSI B31. 1 design criteria and consistent with associated ASME Code requirements.

The entire RCS is hydrotested at 125K (3107 psig) of design pressure, to demonstrate integrity prior to initial operation.

TURKEY POINT - UNITS 3 8( 4 B 2-2 AMENDMENT NOS. AND lHAY C 5 1989

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2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES

2. 2. 1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2. 2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their safety limits during. normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents.

To accommodate the instrument drift that may occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable. If no value is listed in the Allowable column, the setpoint value is the limiting setting.

The methodology to derive the Trip Setpoints includes an allowance for instrument uncertainties. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instru-mentation utilized in these'hannels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

'he various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Tri The Reactor Trip System includes manual Reactor trip capability.

TURKEY POINT - UNITS 3 L 4 B 2-3 AMENDMENT NOS. AND fÃl~Y 0 5 1989

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II LIMITING SAFETY SYSTEM SETTINGS BASES Power Ran e Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protectio'n'uring power operations for all power levels to mitigate the consequences-of a reactivity excursion whi,ch may be too rapid for the temperature and pressure protective trips.

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10K of RATED THERMAL'POWER) and is automatically reinstated below the P-10 Setpoint.

Intermediate and Source Ran e Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip

~ ~ ~ ~ ~

~

of the Power Range, Neutron Flux channels. The Source Range channels will

~

initiate a Reactor trip at about 10 counts per second unless manually blocked

~ ~ ~

The Intermediate Range channels will initiate a

~ ~ ~ ~ ~

when P"6 becomes active. ~

Reactor trip at a current level equivalent to approximately 25K of RATED

~ ~

THERMAL POWER unless manually blocked when P-10 becomes active. No 'credit is taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional to enhance the overall iahik'e capability at the specified trip settings is required by this specification Reactor Prate ~tion S stem.

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" '" +~~~8"4+~<<e~~+~~4>~ o" ~~"~~~+ ~

Overtem erature lhtT ~ +~~ +~cP er4+ Go~pc<+4&~$ A re'~ oe/~~ Kzm sec o&z rs j~dtacect Hnwre n cwei>>Sen<<heC'rare . oess +w o e'>see~<<'rs ogre <<oesreerr Overtempe rovl des tio to prevent ON5 for ~'>

' '/'he all combinations of pressure, power, coo ant temperature, an r i distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure

, trips. The setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2. 1"l. If axial peaks are

~

greater than design, as indicated by the difference between top and bottom

~

power range nuclear detectors, the Reactor trip is automatically reduced

~ ~

according to the notations in Table 2.2-1. ~

~

~ ~

TURKEY POINT - UNITS 3 81 4 B 2-4 AMENDMENT NOS. ANO FEB 2 8 ~n8g

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LIMITING SAFETY SYSTEM SETTINGS BASES Over ower hT The Overpower 4T trip prevents power density anywhere in the core from exceeding 118K of the design power density. This provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1X cladding strain) under all possible overpower conditions, limits the, requi red range for Over-temperature ET trip, and provides a backup to the High Neutron Flux trip. The setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.

Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately .10K of RATED THERMAL'OWER with turbine first stage pressure at approximately 10K of full power equivalent); and on increasing power, automatically reinstated by P-7.

/

The High Setpoint trip functions in conjunction with the pressurizer safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer Water Level-High trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10K of RATED THERMAL POWER with a turbine first stage pressure at approximately 10K of full power equivalent); and on increasing powe~, auto-matically reinstated by P-7.

Reactor Coolant Flow The Reactor Coolant Flow-Low trip provides core protection to prevent ONB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

~ ~

On increasing power above P-7 (a power level of approximately 10K of RATED THERMAL POWER or a turbine first stage pressure at approximately 10K TURKEY POINT " UNITS 3 & 4 B 2"5 AMENDMENT NOS. AND FEB 2 8 wnen I 'gg

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LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow (Continued) of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 45% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater. The specified setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Water Level-Low trip is not used in the transient and accident analyses but is

~ ~

included in Table 2. 2-1 to ensure the functional capability of the specified

~

trip settings and thereby enhance the overall reliability of the Reactor Trip This trip is redundant to the Steam .Generator Water Level Low-Low

~ ~

System. ~

trip. The Steam/Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to 0.64 x 10 lbs/hour. The Steam Generator Water Level-Low portion of the trip is activated when the water level drops below 15%, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor coolant System and steam generators is minimited.

Undervolta e - 4.16 kV Bus A and B Tri s The 4.16 kV. Bus A and B Undervoltage trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified setpoint assures a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients.

The delay is set so that the time required for a signal to reach the Reactor both of trip breakers .following the trip of at least one undervoltage relay indecreasing the associated Units 4. 16 kV busses shall not exceed 1.3 seconds. On TURKEY POINT - UNITS 3 8 4 B 2-6 AMENDMENT NOS. AND

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LIMITING SAFETY SYSTEM SETTINGS BASES Undervolta e and - 4. 16 kV Bus A and B Tri s (Continued) power the Undervoltage Bus trips are automatically blocked by P-7 (a power level of approximately 10K of RATED THERMAL POWER with a turbine first stage pressure at approximately 10K of full power equivalent) and on increasing power, reinstated automatically by P-7.

Turbine Tri A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor Trip from the Turbine trip is automatically blocked by P-7 (a power level of approximately lOX of RATED THERMAL POWER with a turbine first stage pressure at approximately lOX of full power equivalent); and on increasing power, reinstated automatically by P-7.

Safet In 'ection In ut from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

Reactor Coolant Pum Breaker Position Tri The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB. The open/close position trips assure a reactor trip signal is generated before the low flow trip setpoint is reached. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.

Above P-7 (a power level of approximately lOX of RATED THERMAL POWER or a turbine first stage pressure at approximately 10K of full power equivalent) an automatic reactor trip will occur if more than one reactor coolant pump breaker is opened. Above P-8 (a power level of approximately 45K of RATED THERMAL POWER) an automatic reactor trip will occur if one reactor coolant pump breaker is opened. On decreasing power between P-8 and P-7, an automatic reactor trip will occur if more than one reactor coolant pump breaker is opened and below P-7 the trip function is automatically blocked.

Underfrequency sensors are also installed on the 4.16 kV busses to detect underfrequency and initiate breaker trip on underfrequency. The underfrequency trip setpoints preserve the coast down energy of the reactor coolant pumps, in case of a grid frequency decrease so DNB does not occur.

TURKEY POINT " UNITS 3 4 4 B 2-7 AMENDMENT NOS. AND

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LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Tri S stem Interlocks The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip) and deenergizes the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, more than one reactor coolant pump breaker open, reactor coolant pump bus undervoltage and underfrequency, Turbine trip, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are auto" matically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops, and one or more reactor coolant pump breakers open. On decreasing power, the P"8 interlock automatically blocks the trip on low flow in one coolant loop or one coolant pump breaker 'open.

P-10 On incr easing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated.

P-10 also provides input to P-7. The trip setpoint on increasing power shall be > 10K and the reset point shall be less than or equal to 10K.

TURKEY POINT - UNITS 3 8c 4 B 2-8 AMENOMENT NOS. ANO

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS 3/4. 0 APPLICABILITY LIMITING CONDITIONS FOR OPERATION

3. 0. 1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation's not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit, as applicable, in:

a. At least HOT STANDBY within the next 6 hours,

b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

This specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

TURKEY POINT - UNITS 3 8( 4 3/4 0-1 AMENDMENT NOS. AND MAY 0 5 1989

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APPLICABILITY LIMITING CONOITIONS FOR OPERATION Continued 3.0.5 Limiting Conditions for Operation including the associated ACTION requirements shall apply to each unit individually unless otherwise indicated as follows:

a. Whenever the Limiting Conditions for Operation refers to systems or components which are shared by both units, the ACTION requirements will apply to both units simultaneously.

b. Whenever the Limiting Conditions for Operation applies to only one unit, this will be identified in the APPLICABILITY section of the specification; and C. Whenever certain portions of a specification contain operating parameters, Setpoints, etc., which are different for each unit, this will be identified in parentheses, footnotes or body of the requirement.

TURKEY POINT " UNITS 3 8 4 3/4 0-2 AMENOMENT NOS. ANO Fc8 g g p~g

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APPLICABILITY SURVEILLANCE RE UIREMENTS I

4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

4. 0. 2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25K of the surveillance interval.

4. 0. 3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4. 0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .Surveillance

,Requirements do not have to be performed on inoperable equipment.

4. 0. 4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveill.ance Requirement(s) associated with a Limiting Condition for Operation has been performed within the stated surveillance

~ ~ ~ ~

~

interval or as otherwise specified. This provision shall not prevent passage

~ ~ ~ ~ ~

through or to OPERATIONAL MODES as required to comply with ACTION requirements.

~

Surveillance Requirements for inservice inspection and testing of

~ ~ ~ ~ ~

4.0.5 ASME Code Class 1, 2, and 3 components shall be applicable, as follows:

Inser vice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50,55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

TURKEY POINT " UNITS 3 8 4 3/4 0-3 AMENDMENT NOS. AND FiB 28 '--

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APPLICABILITY SURVEILLANCE RE UIREMENTS CONTINUED 1

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing ins ection and testin activities activities Week y At east once per days Monthly At least once per 31 days quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days C. The provisions of Specification 4.0. 2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e. Nothing in the ASME Boiler and Pressure Vessel Code shall'e construed to supersede the requirements of any Technical Specification.

4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3. 0. 5 for individual specifications or whenever certain portions of a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes or body of the requirement.

TURKEY POINT - UNITS 3 8E 4 3/4 0-4 AMENDMENT NOS. AND

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3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.1.

~ ~ 1 BORATION CONTROL SHUTDOWN MARGIN - T GREATER THAN 2000F LIMITING CONDITION FOR OPERATION

3. 1. l. 1 The SHUTDOWN MARGIN shall be greater than or equal to the applicable value shown in Figure 3.1-1.

APPLICABILITY: MODES 1, 2", 3, and 4.

ACTION:

With the SHUTDOWN MARGIN less than the applicable value shown in Figur -1, immediately initiate and continue boration at greater than or equal t gpm of a solution containing greater than or, equal to 20,000 ppm boron equivalent until. the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS

4. 1. 1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the applicable value shown in Figure 3.1-1:

a~ Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b. When in MODE 1 or MODE 2 with Keffff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. in MODE 2 with K less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality b) f$ erifying that the predicted critical control When rbd position is within the limits of Specification 3. 1.3.6;

d. Prior to initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification

4. l. l. 1. 1e. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and See Special Test Exceptions Specification 3. 10. 1.

TURKEY POINT - UNITS 3 & 4 3/4 1"1 AMENDMENT NOS. AND FEB 28 tg)

'g es f Al II 4, ~

'l1

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. When in HODE 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and,

6) Samarium concentration.

4. l. 1.1.2 When in MODE 1 or 2, the overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1X hk/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4. 1. 1. l.le, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after TURKEY POINT - UNITS 3 8a 4 3/4 1-2 AMENDMENT NOS. AND FEB 2 S lngg

2.0 (0,1.77) 500 l750) $ 00 500 K5 MS% GKDG8kTN PM Figure 3.1-1 Required Shutdown Margin vs Reactor Coolant Boron Concentration TURKEY POINT - UNITS 3 4 4 3/4 1-3 AMENDHENT NOS. AND

~<

J ly

'I

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% 3 a>.

II ~

2.0 (0.0,1.77) i.5 hC AC RATION CI R

K 1.0 (750,1

'.5 CD 0.0 0

RCS BORON CONCENTRATION (PPM)

FIGURE 3.1 1 REQUIRED SHUTDOWN MARGIN vs RCS BORON CONCENTRATION TURKEY POINT UNITS 3 8c 4

0 1

I',

4

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN

- T LESS THAN OR EQUAL TO 200 F LIMITING CONDITION FOR OPERATION

3. 1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1X. hk/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than 1X hk/k, immediately initiate and continue boration at greater than or equal to 4 gpm of a solution containing greater than or equal to 20,000 ppm boron or equivalent until the'required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1X ak/k:

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12'ours thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the. immovable or untrippable control rod(s); and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

TURKEY POINT - UNITS 3 8 4 3/4 1"4 AMENDMENT NOS. AND FQ g l

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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION

3. 1. 1.3 The moderator temperature coefficient (MTC) shall be:

Less positive than or equal to 5.0 x 10- b,k/k/ F'for all rods withdrawn, beginning of the cycle life (BOL), hot zero THERMAL POWER (HZP) conditions; and

b. Less positive than or equal to 5.0 x 10-s hk/k/ F from HZP to 70K RATE THERMAL POWER condition; and

c. Less positive than or equal to 5.0 x 10-s b,k/k/4F from 70K RATED THERMAL POWER decreasing linearly to less positive than or equal to 0 hk/k/4F at 100K RATED THERMAL POWER conditions; and

d. Less negative than -3.5 x 10 ~ b,k/k/ F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3. I. 1.3a, b and c. - MODES 1 and 2" only"".

Specification 3.1.1.3d. - MODES 1, 2 and 3 only"".

0 With the MTC more positive than the limit of Specification 3.1.1.3a, b or c above, operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive or equal to limits described in 3. 1.1.3a, b and c above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3. 1. 3. 6;

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

3. A Special Report is prepared and submitted to the Commission, pursuant to Specification 6. 9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the preditted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

"With than or equal to l.

K ff greater

""See S'pecial Test Exceptions Specification 3. 10. 3.

TURKEY POINT -,UNITS 3 & 4 3/4 1-5 AMENDMENT NOS. AND

~~ 188

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REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

b. With the MTC more negative than the limit of Specification 3: 1.1.3d.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS

4. 1. 1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

'a ~ The MTC shall be measured and compared to the BOL limit of Specifi-cation 3. 1.1.3a., above, prior to initial operation above 5X of RATED THERMAL POWER, after each fuel loading; and

b. The MTC shall be measured at any THERMAL POWER and compared to

-3.0 x 10-~ b,k/k/ F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates the MTC is more negative than'-3.0 x 10-~ 4k/k/ F, the MTC shall be 0 C.

remeasured, and compared to the EOL MTC limit of Specification

3. 1.1.3d., at least once per 14 EFPD during the remainder of the fuel cycle.

Perform design calculation to verify conformance to Specifications 3.1.1.3b and c.

TURKEY POINT - UNITS 3 5 4 AMENDMENT NOS. AND FEB 28 )g8y

IL F

~ II

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 0

3. 1. 1.4 The Reactor Coolant System lowest operating loop temperature (T )

shall be greater than or equal to 541 F.

APPLICABILITY: MODES 1 and 2" "".

ACTION:

With a Reactor Coolant System operating loop temperature (T ) less than 541 F, restore T avg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE RE UIREMENTS

4. 1.1.4 The Reactor Coolant System temperature (T ) shall be determined to be greater than or equal to 541 F:

a. Within 15 m'inutes prior to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T is less than 547'F with the T -T Deviation Alarm not reset.

"With Keff greater than or equal to 1.

~

""See Special Test Exceptions Specification 3. 10.3.

~ ~ ~ ~

~ ~ ~

TURKEY POINT - UNITS 3 8a 4 3/4 1" 7 AMENDMENT NOS. AND FEB 2 8 ]gsg

l

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0

REACTIVITY CONTROL SYSTEMS 3/4. 1. 2 BORATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:

a. A flow path from the boric acid storage tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification 3. 1.2.4a. is OPERABLE, or

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3. 1.2.4b. is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE UIREMENTS

4. 1.2. 1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is greater than or equal to 145'F when a flow path from the boric acid tanks is used, and

b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

TURKEY POINT - UNITS 3 & 4 3/4 1-8 AMENDMENT NOS. AND NAY 0 5 leeg

I' 1,

I L 4

$ 1

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 The following boron injection flow paths shall be OPERABLE:

a. The source path from a boric acid storage tank via a boric acid transfer pump to the charging pump suction", and

b. At least one of the two source paths from the refueling water storage tank to the charging pump suction; and,

c. The flow path from the charging pump discharge to the Reactor Coolant System via the regenerative heat exchanger.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

'a 0 With no boration source path from a boric acid storage tank OPERABLE,

l. Demonstrate the OPERABILITY of the second source path from the refueling water storage tank to the charging pump suction by verifying the flow path valve alignment; and

2. Restore the boration source path from a boric acid'torage tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least IX a k/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the boration source path from a boric acid storage tank to OPERABLE status within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With only one boration source path OPERABLE or the regenerative heat exchanger flow path to the RCS inoperable, restore the required flow paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1X 4 k/k at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two boration source paths to OPERABLE status within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

C. With the boration source path from a boric acid storage tank and the charging pump discharge path via the regenerative heat exchanger inoperable, within one hour initiate boration to a SHUTDOWN MARGIN equivalent to 1X b, k/k at 200 F and go to COLD SHUTDOWN as soon as possible within the limitations of the boration and pressurizer level control functions of the CVCS.

flow required in Specification 3.1.2.2.a above shall be isolated from

~ ~ ~ ~ ~

The ~ ~ ~ ~

the other unit. ~

TURKEY POINT - UNITS 3 8( 4 AMENDMENT NOS. AND FEB 2 S )gag

0 REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS

4. 1. 2. 2 The above required flow paths shall be demonstrated OPERABLE:

At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks 'is greater than or equal to 145'F when it is a required water source;

b. At least once per 31 days by verifying that each valve (manual,

.power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; C. At least once per 18 months by verifying that the flow path required by Specification 3. 1. 2. 2a. and c. delivers at least 4 gpm to the RCS.

TURKEY POINT - UNITS 3 8 4 3/4 1-10 AMENOMENT NOS. AND FEB 2S )g..g

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REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.3 At least two charging pumps with independent power supplies shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With two charging pumps OPERABLE and powered from a common power supply, restore at least two ch'arging pumps from independent power supplies to OPERABLE status within 7 days or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least lX b, k/k at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps from independent power supplies to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With only one charging pump OPERABLE, restore any two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 2X b, k/k at 200'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore any two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

e C. The provisions of Specification 3:0.4 are not applicable to provided the 7 day limit of ACTION a is not exceeded.

ACTION a, SURVEILLANCE RE UIREMENTS

4. l. 2.3.1 The required charging pumps shall be demonstrated OPERABLE by testing pursuant to Specification 4.0.5 The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 and 4.

TURKEY POINT - UNITS 3 8( 4 3/4 1-11 AMENDMENT NOS. AND FPp g8 L,

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REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE " SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1.2.4. As a minimum, one of the following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

1) A minimum indicated borated water volume of 500 gallons,

2) A boron concentration between 20,000 ppm and 22,500 ppm, and

3) A minimum solution temperature of 145 F.

b. The refueling water storage tank (RWST) with:

1) A minimum indicated borated water volume of 20,000 gallons,

2) A minimum boron concentration of 1950 ppm, and I~

3) A minimum solution temperature of 39 F.

APPLICABILITY: MODES. 5 and 6.

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE UIREMENTS

4. 1. 2.4 The above required borated water source shall be demonstrated OPERABLE:

At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Verifying the indicated borated water volume, and

3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.

TURKEY POINT - UNITS 3 8 4 3/4 1-12 AMENDMENT NOS. AND FEB g8 ),.

II I

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued)

b. By verifying the RWST t'emperature is above its limit whenever the outside air temperature is less than 39'F at the following frequencies:

1) Within one hour when the outside temperature is below 39 F for 23 consecutive hours, and

2) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the outside temperature is below 390F.

TURKEY POINT - UNITS 3 8c 4 3/4 1-13 AMENDMENT NOS. AND PpQ 2 8 l"69

l~

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REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION

3. 1.2.5 The following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

1) A minimum indicated bor ated water volume of 3080 gallons,

2) A boron concentration between 20,000 ppm and 22,500 ppm, and

3) A minimum solution temperature of 145 F;

b. The refueling water storage tank (RWST) with:

1) A minimum indicated borated water volume of 320,000 gallons,

2) A minimum boron concentration of 1950 ppm,

3) A minimum solution temperature of 39~F, and

4) A maximum solution temperature of 100'F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a4 With the required Boric Acid Storage System inoperable verify that the RWST is OPERABLE; restore the system to OPERABLE status within

.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least iX hk/k at 200 F; restore the Boric Acid Storage System to OPERABLE status within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in COLO SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TURKEY POINT - UNITS 3 4 4 3/4 1-14 AMENDMENT NOS. AND FEB 28 t..-gg

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS

4. 1.2.5 Each borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration in the water,

2) Verifying the indicated borated water volume of the water source, and

3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated- water.

b. By verifying the RWST temperature is within limits whenever the outside air temperature is less than 39 F or greater than 100'F at the following frequencies:

1) Within one hour upon the outside temperature exceeding its limit for 23 consecutive hours, and

2) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while the outside temperature exceeds its limits.

TURKEY POINT " UNITS 3 8E 4 3/4 1"15 AMENOMENT NOS. AND

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REACTIVITY CONTROL SYSTEMS HEAT TRACING LIMITING CONDITION FOR OPERATION

3. 1.2.6 At least two independent channels of heat tracing shall. be OPERABLE for the boric acid storage tank and for the heat traced portions of the associated flow paths required by Specification 3. 1. 2. 2.

APPLICABILITY: MODES 1, 2, 3 and 4 MODES 5 and 6 (when the boric acid storage tank is the borated water source per Specification 3. 1.2.4)

ACTION:

MODES 1 2 3 and 4 With only one channel of heat tracing on either the boric acid storage tank or er the heat traced portion of an associated flow path OPERABLE, operation may continue for up to 30 days provided the tank and flow path temperatures are verified to be greater than or equal to 145 F at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6 With only one channel of heat tracing on either the boric acid storage tank or

~

on the heat traced portion of an associated flow path OPERABLE, operations involving CORE ALTERATIONS or positive reactivity additions may continue for

~ ~

up to 30 days provided the tank and flow path temperatures are verified to be greater than or equal to 145'F at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise, suspend all activities involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE UIREMENTS

4. 1.2.6 Each heat tracing channel for the boric acid storage tank and associated flow path required by Specification 3. 1. 2. 2 shall be demonstrated OPERABLE:

a. At least once per 31 days by energizing each heat tracing channel, and

b. At least once per 7 days by verifying the tank and flow path temperatures to be greater than or equal to 145'F. The tank temperature shall be determined by measurement. The flow path temperature shall be determined by either measurement or recirculation flow until establishment of equilibrium temperatures within the tank.

T'URKEY POINT " UNITS 3 & 4 3/4 1" 16 AMENDMENT NOS. ANO

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REACTIVITY CONTROL SYSTEMS 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION

3. 1.3. 1 All full length (shutdown and control) rods shall be OPERABLE and positioned within + 12 steps (Analog Rod Position Indication) of the group step counter demand position within one hour after rod motion.

APPLICABILIT'Y: MODES 1* and 2" ACTION:

a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3. 1. 1. 1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT.

STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With more than one full length rod inoperable or misaligned from the group step counter demand position by more than + 12 steps (Analog Rod Position Indication), be in HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s.

C. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand position by more than + 12 steps (Analog Rod Position Indicatioh), POWER OPERATION may continue provided that within one hour either:

The rod is restored to OPERABLE status within the above alignment requirements, or

2. remainder of the rods in the bank with the inoperable rod are T'he aligned to within + 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1. 3. 6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

ee Specia est Exceptions 3. 10.2 and 3. 10.3.

TURKEY POINT - UNITS 3 & 4 3/4 1" 17 AMENDMENT NOS. AND FEB g 8,lg.'

r ll 4s lv kgi 4

0

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION Continued a) The THERMAL POWER level is reduced to less than or equal to 75K of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the power range neutron -flux high trip setpoint is reduced to less than or equal to 85K of RATED THERMAL POWER. THERMAL POWER shall be maintained less than or equal to 75K of RATED THERMAL POWER until compliance with ACTIONS 3. 1. 3. 1. c. 3. c and 3. 1. 3. 1. c. 3. d below are demonstrated, and b) The SHUTDOWN MARGIN requirement of Specification 3. 1. 1. 1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and c) A power distribution map is obtained from the movable incore detectors and F~(Z) and-F are verified to be within their limits within 72 hours, and d) A reevaluation of each accident analysis of Table 3. 1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of, these accidents remain valid for the duration of operation under these conditions.

4.1.3.1.1 The position of each full length rod shall be determined to be within + 12 steps (Analog Rod Position Indication) of the group step counter demand position at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (allowing for one hour thermal soak after rod motion) except during time invervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1.3. 1.2 Each full length rod not fully inserted in the core shall be deter-mined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

TURKEY POINT " UNITS 3 8( 4 3/4 1"18 AMENDMENT NOS. AND

TABLE 3.1-1 ACCIDENT ANALYSES RE UIRING REEVALUATION IN H V N AN INOPERABL ULL"L NG H ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident)

Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

TURKEY POINT - UNITS 3 8( 4 3/4 1-19 AMENDMENT NOS. AND FEB 28 1S g

0 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS " OPERATING LIMITING CONDITION FOR OPERATION

3. 1.'3. 2 The Analog Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the respective actual and demanded shutdown and control rod positions as follows:

a. Analog rod position indicators, within one hour after rod motion (allowance for thermal soak);

All Shutdown Banks: + 12 steps of the group demand counters for w>th rawa ranges of 0-30 steps and 200-228 steps.

Control Bank A and B: + 12 steps of the group demand counters for w>th rawa ranges o 0-30 steps and 200-228 steps.

Control Banks C and D: + 12 steps of the group demand counters for wsthdrawa range o "228 steps.

b. Group demand counters; + 2 steps.

APPLICABILITY: MODES 1 and'. ~

With a maximum of one analog rod position indicator per bank inoperable

~ ~

a.

either:

l. Determine the position of the non-indicating rod(s) indirectly'y the movable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and within one hour after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL POWER to less than 75K of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b, With a maximum of one demand position indicator per bank inoperable either:

l. Verify that all analog rod position indicators for the affected bank are OPERABLE an'd that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

2. Reduce THERMAL POWER to less than 75K of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

TURKEY POINT - UNITS -3 8E 4 3/4 1-20 AMENDMENT NOS. AND FPB >8 yg

lk '

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS

4. 1.3.2. 1 Each analog rod position indicator shall be determined to be OPERABLE by verifying that the Oemand Position Indication System and the Anal'og Rod Posi-tion Indication System agree within 12 steps (allowing for one hour thermal soak after rod motion) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Oeviation Monitor is inoperable, then compare the Oemand Posi-tion Indication System and the Analog Rod Position Indication System at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. 1. 3. 2. 2 Each of the above required analog rod position indicator(s) shall be determined to be OPERABLE by performance of a CHANNEL CHECK, CHANNEL CALIBRA-TION and ANALOG CHANNEL OPERATIONAL TEST performed in accordance with Table 4.1-1.

TURKEY POINT - UNITS 3 5 4 3/4 1" 21 AMENOMENT NOS. AND

TABLE 4.1-1 ROD POSITION INDICATOR SURVEILLANCE RE UIREMENTS Check Calibration 0 erational Test Individual Rod Position Demand Position N/A TURKEY POINT - UNITS 3 8( 4 3/4 1-22 AMENDMENT NOS. AND FEB 28 .-<

Iver/

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REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1.3.3 The group step counter demand position indicator shall be OPERABLE and capable of determining within f 2 steps the demand position for each shut-down and control rod not fully inserted.

APPLICABILITY: MODES 3"¹, 4"¹, and 5"¹ ACTION:

With less than the above required group step counter demand position indica-tor(s) OPERABLE, open the reactor trip system breakers.

SURVEILLANCE RE UIREMENTS

4. 1.3.3.1 a ~ ~ ~ Each of the above required group step counter demand position indi-cator(s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days.

4. 1.3.3.2 A CHANNEL CHECK CALIBRATION AND ANALOG CHANNEL OPERATiONAL TEST shall be performed per Table 4.1-1.

"With the Reactor TripSystem breakers in the closed position.

~

¹See Special Test Exceptions Specification 3. 10.4.

~

TURKEY POINT - UNITS 3 8c 4 3/4 1"23 AMENDMENT NOS. AND FEB 28 i."""

A i

N'

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REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod dr'op time from the fully withdrawn position shall be less than or equal to 2.4 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T greater than or equal to 541 F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

1>"th the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE RE UIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

a ~ For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and C. At least once per 18 months.

'TURKEY POINT - UNITS 3 8a 4 3/4 1-24 AMENDMENT NOS. AND

1$

'I K

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION

3. 1.3.5 All shutdown rods shall be fully withdrawn.

APPLICABILITY: MODES 1" and 2"0 .

ACTION:

With a maximum of on'e shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

e 1

SURVEILLANCE RE UIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or 0 during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exceptions Specifications 3. 10.2 and 3. 10.3.

Alith Keff greater than or equal to 1.0 ff TURKEY POINT - UNITS 3 8c 4 3/4 1-25 AMENDMENT NOS., AND FEB 28 )g))

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3. 1-2.

APPLICABILITY: MODES 1" and 2"¹ ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4. 1. 3. 1. 2 either:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED. THERMAL POWER which is allowed by the bank posi-tion using the above figures, or

c. Be in at least HOT STANDBY within 6 hours.

SURVEILLANCE RE UIREMENTS 4, 1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

e "See Special Test Exceptions

¹With K Specifications ff greater than or equal to 1.0

3. 10.2 and 3. 10.3.

TURKEY POINT - UNITS 3 5 4 3/4 1-26 AMENDMENT NOS. AND MAY c 5 198S

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(>2,0) 40 50 60 70 80 90 100

+ RATED THERMAL POWER FIGURE 3.1 2 ROD BANK INSERTION LIMITS va x THERMAL POWER TURKEY POINT UNITS 3 8c 4

Ct 14 4

3/4. 2 POWER DISTRIBUTION LIMITS 3 /4.2.1 AXIAL FLUX DIFFERENCE IMITING CONDITION FOR OPERATION 3.2. 1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a + 5X target band (flux difference units) about the target flux difference.

The indicated AFD may deviate outside the above required target band at greater than or equal to 50K but less than 90K of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~he indicated AFD may deviate outside the above required target band at greater

'han 15K but less than 50K of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15K of RATED THERMAL POWER."

ACTION:

a. With the 'indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal-to 90K of RATED THERMAL POWER, within 15 minutes either:

Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL POWER to less than 90K of RATED THERMAL POWER.

b. With the indicated AFD outside of the above re uired tar et band fo r more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90K but equal to or greater than 50K of RATED THERMAL POWER:

1. Reduce THERMAL POWER to less than 50K of RATED THERMAL POWER within 30 minutes, and

- 2. Reduce the Power Range Neutron Flux" ~" - High Trip Setpoints to less than or equal to 55K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c.'ith the more indicated AFD outside than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative of the above required target band for penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50K but greater

See Special Test Exceptions Specification 3.10. 2.

t ""Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1. 1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

TURKEY POINT - UNITS 3 4 4 3/4 2-1 AMENDMENT NOS. AND FEB 2 s t,".g

4 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued CTION Continued than 15K of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal .to or greater than 50K of RATED THERMAL POWER until the indicated AFD is within the above required target band.

SURVEILLANCE RE UIREMENTS 4.2. l. 1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15K of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the alarm used to monitor the AFD is OPERABLE, and

2) At least once per hour for the first 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after restoring the alarm used to monitor the AFD to OPERABLE status."

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the alarm used to monitor the

. AFD is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

'.2.1.2 The indicated AFD shall be considered outside of its target band when wo or more OPERABLE excore channels are indicating the AFD t'o be outside the arget band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

'a ~ One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50K of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15K and 50K of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

"Performance of a functional test to demonstrate OPERABILITY of the alarm used to monitor the AFD may be substituted for this requirement.

TURKEY POINT - UNITS 3 8L 4 3/4 2-2 AMENDMENT NOS. AND

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100 11,90) (11,90) 90 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 80 70 60 ACCEPTABLE OPERATION a >0 31,50)

M 40 20 10

-50 -40 -X -20 -10 0 10 20 FLUX DIFFERENCE (B1)m FIGURE 3.2 1 AXIAL FLUX DIFFERENCE LIMITS vs. x THERMAL POWER TURKEY POINT UNITS 3 dt 4

<r s

C1

, h h t

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)

IMITING CONDITION FOR OPERATION P

3.2.2 F (Z) shall be limited by the following relationships:.

FO(Z) < [F ) X [K(Z)) for P > 0.5 FO(Z) < [F ] X [K(Z)) for P < 0.5 where: LF ] = 2.32 limit P = Thermal Power ate erma ower

[F ] = The Measured Value, and K(Z) is the function obtained from Figure 3. 2-2 for a given core height location.

PPLICABILITY: MODE 1 ACTION:

With the.-measured value of F (Z) exceeding its limit:

M L

a. Reduce THERMAL POWER at least 2X for each 1X F~(Z) exceeds F~(Z) within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ;. subsequent POWER OPERATION may proceed provided the Overpower Delta-T Trip Setpoints (value of K~)

have been reduced at least 2X for each lX F (Z) exceeds the F

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced power limit required by ACTION a., above; THERMAL POWER may then be increased provided F~(Z) is demonstrated through incore mapping to be within its limit.

TURKEY POINT - UNITS 3 8L 4 3/4 2-4 AMENDMENT NOS. AND FEB 2 S 19g9

0 4

'1

FIGURE 3.2-2 K(Z) NORMALIZED F((Z) AS A FUNCTION OF CORE HEIGHT TURKEY POINT - UNITS 3 4 4 3/4 2-5 AMENDMENT NOS. AND Fgg g8 tg

(6.0,1.0) 1.0 0.9 0.8 a p7 Ch LJ (1 2.0,0.647) 0.6 0.5 I

p4 0.3 0.2 0.1 0.0 0 3 4 5 6 7 8 9 10 11 12 CORE HEIGHT (FT.)

FIGURE 3.2 2 K{Z} NORMALIZED F0{2} AS A FUNCIION OF CORE HEIGHT TURKEY POINT UNITS 3 8c 4

'11 l"q f

1g gl l

Kg LI

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS

.2.2.

~ ~ 1 If [F ] as

~

predicted by approved physics calculations is greater than

~ ~

~

[F~] and P is greater than PT" as defined in 4.2. 2. 2, F~(Z) shall be

~ ~

evaluated by MIDS (Specification 4. 2. 2. 2), BASE LOAD (Specification 4. 2. 2. 3) or RADIAL BURNDOWN (Specification 4.2.2.4) to determine p =

if F~ is within its limit [F~] Predicted F~).

If [F~] p , is less than [F~] L or P is less than PT, F~(Z) shall be evaluated to determine if F~(Z) is within its limit as follows:

a. .Using the movable incore detectors to obtain power distribution map at any THERMAL POWER greater than 5X of RATED THERMAL POWER.

b. Increasing the measured F~(Z) component of the power distribution map by 3X to account for manufacturing tolerances and further increasing the value by 5X to account for measurement uncertainties.

Verifying that the requirements of Specification 3.2.2 are satisfied.

c. Fq(Z) < Fq(Z)

Where F (Z) is the measured F (Z) increased by the allowance for manu-facturing tolerances and measurement uncertainty and F (Z) is the F limit defined in 3.2.2.

~ ~ ~

~

d.~ Measuring F~(Z) according to the following schedule:

~

1. Prior to exceeding 75K of RATED THERMAL POWER,"" after refueling, 2.'t least once per 31 Effective Full Power Days.

e. With the relationship specified in Specification 4.2.2. 1. c above not being satisfied:

1) Calculate the percent F (Z) exceeds its limit by the following expression:

F (Z) X 100 for P > 0.5

[Fq] X K(Z)/P FM(Z) - 1 X 100 for P < 0.5

[F0] X K(Z)/0. 5 PT

= Reactor power level at which predicted F~ would exceed its limit.

During 'power. escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained.

TURKEY POINT - UNITS 3 8 4 3/4 2-6 AMENDMENT NOS. AND

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POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued

2) The following action shall be taken:

a) Comply with the requirements of Specification 3.2. 2 for F (Z) exceeding its limit by the percent cal'culated Q

above.

4. 2. 2. 2 MIDS Operation is permitted at power above PT where PT equals the ratio of [F ]

divided by [F ] if the following Augmented Surveillance (Movable Incore letection System, MIDS) requirements are satisfied:

a. The axial power distribution shall 'be measured by MIDS when required such that the limit of [FQ] /P times Figure 3.2.2 is not exceeded.

F.(Z) is the normalized axial power distribution from thimble at j j core elevation (Z).

1. If F.(Z) exceeds [F .(Z)]

as defined in the bases by < 4X, immediately reduce thermal power one percent for every percent by which [F (Z)] is exceeded.

2. If F (Z) exceeds [F (Z)] by > 4X immediately reduce. thermal power below PT. Corrective action to reduce F (Z) below the limit will permit return to thermal power not to exceed current P

"" as defined in the bases.

L

b. F.(Z) shall be determined to be within limits by using MIDS to monitor the thimbles required per Specification 4.2.2.2. c at the following frequencies.

l. At least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. Immediately following and as a minimum at 2, 4 and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the events listed below and every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

1) Raising the thermal power above PT, or

2) Movement of control-bank 0 more than an accumulated total of 15 steps in any one direction..

C. MIDS shall be operable when the thermal power exceeds PT with:

1. At least two thimbles available for which R. and ~ as defined in the bases have been determined.

"[F (Z)]s is the al.arm setpoint for MIDS.

~

j~

"P is reactor thermal power expressed as a fraction of the Rated Thermal

~ ~

'Power that is used to calculate [F.(Z)] . ~

TURKEY POINT - UNITS 3 8 4 3/4 2-7 AMENDMENT NOS. AND FEB 2 8 )gag

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POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued

2. At least two movable detectors avail'able for mapping F.(Z).

3. The continued accuracy and representativeness of the selected thimbles shall be verified by using the most recent flux map to update the R for each selected thimble. The flux map must be updated at least once per 31 effective full power days.

where:

R = Total peaking factor from a full flux map ratioed to the axial peaking factor in a selected thimble.

j - The thimble location selected for monitoring.

4.2.2.3 Base Load Base Load operation is permitted at powers above PT if the following require-ments are satisfied:

a ~ Either of the following preconditions for Base Load operation must be satisfied..

1. For entering Base Load operation with power less than PT, a) Maintain THERMAL POWER between PT/1.05 and PT for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b) Maintain the AFD (Delta-I) to within a + 2X or + 3X target band for at least 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months per 24-hour period.

c) After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months have elapsed, take a full core flux map to determine F (Z) unless a valid full core flux map was taken within the time period specified in 4.2.2.ld.

d) Calculate PBL per 4.2.2.3b.

2. For entering Base Load operation with power greater than PT, a) Maintain THERMAL POWER between P and the power limit determined in 4.2.2.2 for at learnt 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and maintain Augmented Surveillance requirements of 4. 2. 2. 2 during this period.

b) Maintain the AFD (Delta-I) to within a + 2X or a 3X target band for at least 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months per 24-hour period,

'URKEY POINT - UNITS 3 6 4 3/4 2-8 "AMENDMENT NOS. 'ND Y 0~ >889

L 4'$

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'4 0

il 4

i

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued c) After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months have elapsed, take a full core flux map to determine F (Z) unless a valid full core flux map was taken within the time period specified in 4.2. 2. ld.

d) Calculate PBL per 4. 2. 2.3b.

ll

b. Base Load operation is permitted provided:

1. THERMAL POWER is maintained between PT and PBL or between PT and 100K (whichever is'most limiting).

2. AFD (Delta-I) is maintained within a 2 2X or 2 3X target band.

3. Full core flux maps are taken at least once per 31 effective Full Power Days.

n are defined as:

BL T PBL F] XKZ FO(Z) X W(Z) BL X 1.09

= L P PT [Fq] /[Fq]

where: F~(Z) is the measured F~(Z) with no allowance for manufactur-ing tolerances og measurement uncertainty. For the purpose of this 5eci fi cat i on [F((Z)]shal 1 be obtained between el evati ons bounded by 10K and 90K of .the active core height. [F ] is the F limit.

K(Z) is given in Figure 3.2-2. W(Z) BL is the cycle dependent function that accounts for limited power distribution transients encountered during base load operation.

The function is given in the Peaking Factor Limit Report as per Specification 6.9.1.6. The 9X uncertainty factor accounts for manufacturing tolerance, measurement error, rod bow and any burnup and power dependent peaking factor increases.

C. During Base Load operation, if the THERMAL POWER is decreased below 4.2.2.3.a shall be. satisfied before PT, then the conditions of re-entering Base Load operation.

d. If any of the conditions of 4.2.2.3b are not maintained, reduce THERMAL POWER to less than or equal to PT, or, within 15 minutes initiate the Augmented Surveillance (MIDS) requirements of 4.2.2.2.

TURKEY POINT - UNITS 3 5 4 3/4 2-9 AMENDMENT NOS. AND F~8 ~S

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POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued

~ 2. 2. 4

~ ~ RADIAL BURNDOWN Operation is permitted at powers above PT if the following Radial Burndown conditions are satisfied:

a. Radial Burndown operation is restricted to use at powers between PT and PRB or PT and 1.00 (whichever is most limiting). The maximum relative power permitted under Radial Burndown operation, PRB, is equal to the minimum value of the ratio of [F~(Z)]/[F~(Z)]RB Meas.

where: [F~(Z)]RB = [Fx (Z)]Ma Meas. x Fz(Z) x 1.09 and Meas

[F (Z)] is equal to [F ] x K(Z).

b. A full core flux map to determine [Fx (Z)]M Meas. shall be taken within the time period specified in Section 4.2. 2.1d.2. For the pur-pose of the specification, [Fx (Z)]M Meas. shall be obtained between the elevations bounded by 10K and 90K of -the active core height.

C. The function F z (Z), provided in the Peaking Factor Limit Report (6.9.1.6), is determined analytically and accounts for the most per-turbed axial power shapes which can occur under axial power distribu-tion control. The uncertainty factor of 9X accounts for manufacturing tolerances, measurement error, rod bow, and any burnup dependent peak-ing factor increases.

d. Radial Burndown operation may be utilized at powers between PT and PRB, or, PT and 1.00 (whichever is most limiting) provided that the AFD (Delta-I) is within a 5X of the target axial offset.

e., If the requirements of Section 4.2.2.4d are not maintained, then the power shall be reduced to less than or equal to PT, or within 15 minutes Augmented Surveillance of hot channel factors shall be initiated if the power is above PT.

4. 2. 2. 5 When F (Z) is measured for reasons other than meeting the requirements of Specifications 4.2.2.1, 4.2.2;2, 4.2.2.3 or 4.2.2.4 an overall measured F (Z) shall be obtained from a power distribution map and increased by 3X to I.

account for manufacturing tolerances and further increased by 5X to account for measurement uncertainty.

TURKEY POINT " UNITS 3 8c 4 3/4 2"10 AMENDMENT NOS. AND F/8 g8

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POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR IMITING CONDITION FOR OPERATION 3.2.3 F<H shall be limited by the following relationship:

F~H < 1.62 f1.0 + 0.3 (1-P)],

Where:

THERMAL POWER H MAL W APPLICABILITY; MODE 1.

ACTION:

With F~ exceeding its,limit:

'a 0 Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore F~ to within the above limit, or

2. Reduce THERMAL POWER to less than 5(C of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limit, verify through incore flux mapping that F~ has been restored to within the above limit, or reduce THERMAL POWER to less than SX of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed provided that F~ is demonstrated, through incore flux mapping, to be within the limit of acceptable operation prior to exceeding the following THERMAL POWER levels:

l. A nominal 50K of RATED THERMAL POWER,

2. A nominal 75K of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95K of RATED THERMAL POWER.

TURKEY POINT - UNITS 3 & 4 3/4 2"11 AMENDMENT NOS. AND FEB ZS

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Tl, 7r'

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS

.2.3.1

~ ~ ~ The provisions of Specification 4.0.4 are not applicable.

N N 4.2.3.2 When a measurement of F<H is taken, the measured F>H shall be increased by 4X to account for measurement error.

4. 2. 3. 3 This corrected F~ shall be determined to be within its limit through incore flux mapping:

a. Prior to operation above 75K of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

TURKEY POINT - UNITS 3 8L 4 3/4 2-12 AMENDMENT NOS. AND FE'8 g8 )g

POWER DISTRIBUTION LIMITS 3/4.2.4 UADRANT POWER TILT RATIO IMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50K of RATED THERMAL POWER".

ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50K of RATED THERMAL POWER.

2. Within 2 hours either:

a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3X from RATED THERMAL POWER for each 1X of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50K of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than, or equal to 55K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50K of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 bours or until verified acceptable at 95K or greater RATED THERMAL POWER.

See Special Test Exceptions Specification 3.10.2.

TURKEY POINT - UNITS 3 8L 4 3/4 2-13 AMENDMENT 'NOS. AND FEB 28 19'

I' I 4 t

I I,

POWER DISTRIBUTION LIMITS

'MITING CONDITION FOR OPERATION Continued CTION Continued

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to mi sal i gnment of ei ther a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 5(C of RATED THERMAL POWER.

2. Reduce THERMAL POWER at least 3X from RATED THERMAL POWER for each 1X of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30.minutes;

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50K. of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50K of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95K or greater RATED THERMAL POWER.

C. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50'f RATED THERMAL .

POWER.

TURKEY POINT - UNITS 3 8L 4 3/4 2-14 AMENDMENT NOS. AND FE8 g8

cr l

II Il

+~4 j~

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION Continued)

CTION (Continued

2. Reduce THERMAL POWER to less than 50K of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50K of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95K or greater RATED THERMAL POWER.

d. The provisions of Specification 3. 0. 4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above

~

50K of RATED THERMAL POWER by:

Calculating the ratio at least once per 7 days when the Power Range

~ ~

a.~

Upper Detector High Flux Deviation and Power Range Lower Detector

~

High Flux Deviation Alarms are OPERABLE, and

~

b. Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when either alarm is inoperable. i 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75K of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained either from two sets of four symmetric thimble locations or full-core flux map, or by incore thermocouple map is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.4.3 If the QUADRANT POWER TILT RATIO is not within its limit within of 3.2.2 and 3.2.3 are within,their 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the POWER DISTRIBUTION LIMITS limits, a Special Report in accordance with 6.9.2 shall be submitted within 30 days including an evaluation of the cause of the discrepancy.

TURKEY POINT - UNETS 3 5 4 3/4 2-15 AMENDMENT NOS. AND FEB z8 )gg

L3 I

8~

t f~

t l

t

POWER DISTRIBUTION LIMITS 3/4. 2. 5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 1

3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a. Reactor Coolant System T

b. Pressurizer Pressure > 2209 psig*, and

c. Reactor Coolant System Flow >277,900 gpm APPLICABILITY: MODE 1.

ACTION:

. With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.5. 1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. 2. 5. 2 The RCS flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS flow rate shall be demonstrated by measurement once per 18 months.

Limit not applicable during either a THERMAL POWER ramp in excess of 5X of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10K of RATED THERMAL POWER.

TURKEY POINT - UNITS 3 &,4 3/4 2-16 AMENDMENT NOS, AND NAY co )ggg

(I 3/4. 3 INSTRUMENTATION 3/4.3. 1 REACTOR TRIP SYSTEM INSTRUMENTATION MITIHG CONDITION FOR OPERATION

3. 3. 1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be .OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

JRVEILLANCE RE UIREMENTS 4.3. l. 1 Each reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirement specified in T451e 4. 3-1.

1 TURKEY POINT - UNITS 3 8 4 3/4 3-1 AMENDMENT NOS. AND EB z8 t88g

A I TABLE I

REACTOR TRIP SYSTEM INSTRUMENTATION HINIHUM

'OTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Hanual Reactor Trip 1, 2 3A 4A 5*

2. Power Range, Neutron Flux

a. High Setpoint 1, 2

b. Low Setpoint 188, 2

3. Intermediate Range, Neutron Flux 180, 2 Source Range, Neutron Flux

a. Startup 1 28 b.. Shutdown"" 0 3, 4, 5

c. Shutdown 1 3A 4A'*

Overtemperature 4T 1, 2

6. Overpower hT 1, 2

7. Pressurizer Pressure-.Low ~

2 (Above P-7)

8. Pressurizer Pressure High 1, 2 6

9. Pressurizer Water Level High (Above P-7)

10. Reactor Coolant Flow--Low

a. Single Loop (Above P-8) 3/loop 2/loop 2/1 oop 1

b. Two Loops (Above P-7 3/loop 2/loop 2/loop 1 and below P-8)

C7

0 l

V

TABLE 3. 3-1 ~>eed REACTOR TRIP SYSTEM INSTRUMENTATION m

a MINIMUM Cl CHANNELS APPLICABLE M TOTAL NO. CHANNELS FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MOOES ACTION ll. Steam Generator Water 3/stm. gen. 2/stm. gen. 2/stm. gen. 1, 2 Level Low-Low

12. Steam Generator Water Level-- 2 stm. gen. 1 stm. gen. 1 stm. gen. 1, 2 Low Coincident With Steam/ level and level coin- level and Feedwater Flow Mismatch 2 stm./feed- cident with 2 stm./feed-water flow 1 stm./feed- water flow mismatch in water flow mismatch in each stm. gen. mismatch in same stm. gen.

same stm. or 2 stm. gen.

gen. level and 1 stm./feedwater flow mismatch in same stm.

gen.-

13. Undervoltage

4. 16 KV Busses 2/bus 1/bus on 2/bus 12 A and 8 (Above P-7) both busses

14. Underfrequency-Trip of Reactor 2/bus 1 to trip 2/bus Coolant Pump Breaker(s) Open RCPs*~*

m (Above P-7)

C7

15. Turbine Trip (Above P-7)

a. Autostop Oil Pressure 12

b. Turbine Stop Valve Closure 12

f

~4

TABLE 3. 3-1 ~ued I

REACTOR TRIP SYSTEM INSTRUMENTATION HINIHUH TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Safety Injection Input from ESF 1, 2

17. Reactor Trip System In terlocks

a. Intermediate Rang e Neutron Flux, P-6

b. Low Power Reactor Trips Block, P-7 P-10 Input or Turbine First Stage Pressure C.

d.

Power Range Neutron Flux, P-8 Power Range Neutron

ll Flux, P-10 1, 2

18. Reactor Coolant Pump Breaker Position Trip

a. Above P-8 1/breaker 1/breaker B. Above P-7 and below P-8 1/breaker 1/breaker

19. Reactor Trip Breakers 1, 2 8, 10 3* 4A'A 9

20. Automatic Trip and Interlock 1, 2 Logic 3* 4 III 5*

C)

f, S'

P I

6

TABLE 3. 3-1 Continued TABLE NOTATION "When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.

"*When the Reactor Trip System breakers are in the open position, one or both of

'he backup NIS instrumentation channels may be used to satisfy .this require-ment. For backup NIS testing requirements, see Specification 3/4.3.3.3, ACCIDENT MONITORING.

"*"Reactor Coolant Pump breaker A is tripped by underfrequency sensor UF-3A1 (UF-4A1) or UF-381(UF-4Bl). Reactor Coolant Pump breakers B and C are tripped by underfrequency sensor UF-3A2(UF-4A2) or UF-3B2(UF-4B2).

YBelow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

¹¹Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed e provided the following conditions are satisfied:

a. The inoperable channel is placed 'in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other channels per per Specifica-tion 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75K of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85K of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored per Specification 4. 2.4. 2.

TURKEY POINT - UNITS 3 5 4 3/4 3-5 AMENDMENT NOS. AND

hei TABLE 3.3-1 Continued ACTION STATEMENTS Continued

'TION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 S'etpoint, and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10K of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above lOX of RATED THERMAL POWER.

eCTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes and verify compliance with the SHUTDOWN MARGIN requirements of Specification 3. 1.1. 1 or 3. 1. 1. 2, as appli-cable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6- With the number of OPERABLE channels one less'than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 7 " With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 8- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however,'ne channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1. 1, provided the other channel is OPERABLE.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers within the next hour.

ACTION 10- With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 8. The breaker shall not be bypassed while one of the diverse trip features is i'noperable except for the time required for performing mainten-ance to restore the breaker to OPERABLE status.

TURKEY POINT " UNITS 3 8 4 3/4 3-6 AMENDMENT NOS. AND

gl "l

Lg 1f

TABLE 3. 3-1 Continued ACTION STATEMENTS Continued ACTION ll - With

~

the Channels number of OPERABLE OPERABLE channels one less than the Minimum requirement; be in at least HOT STANDBY within 6 hours. ~

ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ACTUATION LOGIC TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

TURKEY POINT - UNITS 3 4 4 3/4 3-7 AMENDMENT NOS. AND

0 lg

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST EEIEUE E .

1. Manual Reactor Trip N.A. N.A. N.A. R(11) N.A. ] 2 3* 4* 5A

2. Power Range, Neutron Flux

a. High Setpoint D(2, 4), N.A. N.A. 1, 2 M(3,'4)',

q(4,'),'(4) 1*%'k

b. Low Setpoint R(4) N.A. N.A.

3. Intermediate Range, R(4) S/U(1),M N.A. N.A. ] 4'AA Neutron Flux

4. Source Range, Neutron Flux S R(4) S/U(1),M(9) N.A. N.A. 3 4 5

5. Overtemperature hT

'(12)

N. A. N.A. 1 2

e. Overpower BT N.A. N.A. 1 2

7. Pressurizer Pressure Low S M N.A. N.A.

8. Pressurizer Pressure--High S N.A. N.A. 1, 2

9. Pressurizer Water Level High S N.A. N.A.

10. Reactor Coolant Flow--Low S N.A. N.A.

ll. Steam Generator Water Level S N.A. N.A. 1 2 Low-Low

t ABLE 4.3-1 I I

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP ANALOG ACTUATING MODES FOR

- CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS RE UIRED

12. Steam Generator Water S R M N.A. N.A. 1, 2 Level Low Coincident with Steam/Feedwater Flow Mismatch

13. Undervoltage - 4. 16 kV N.A. R N.A. N.A. N.A.

Busses A and B

14. Underfrequency - Trip of N.A. R N.A. N.A. N.A. 1 Reactor Coolant Pump Breakers(s) Open

15. Turbine Trip

a. Autostop Oil Pressure N.A. N.A. S/U(1, 10) N.A.

b. Turbine Stop Valve Closure N.A. N.A. S/U(1, 10) N.A.

16. Safety Injection Input from ESF N.A. N.A. .N.A. N.A. 1, 2

17. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6 N.A. R(4) N.A. N.A.

b. Low Power Reactor Trips Block, P-7 N.A. R(4) M(e) N.A. N.A.

(includes P-10 input and Turbine First Stage Pressure)

c. Power Range Neutron Flux, P-8 N.A. R(4) M(e) N.A. N.A.

REACTOR t BLE 4.3-1 I

I TRIP SYSTEH INSTRUMENTATION SURVEILLANCE TRIP RE UIREHENTS ACTUATING NODES FOR ANALOG CHANNEL OEV ICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS RE UIRED

17. Reactor Trip System Interlocks (Continued)

d. Power Range Neutron Flux, P-10 N.A. R(4) H(8) N.A. N. A. 1, 2

18. Reactor Coolant Pump N.A. N.A. .N.A. N.A.

Breaker Position Trip 19.'eactor Trip Breaker N.A. N.A. N.A. H(7, ll) N.A. 3A 4A 5A

20. Automatic Trip and Inter-lock Logic N.A. N.A. N.A. N.A. H(7,14) 1,. 2, 3*, 4", 5*

21. Reactor Trip Bypass Breaker N.A. N.A. N.A. H(13),R(15) N.A. 1, 2, 3*, 4", 5*

A I'7

TABLE 4. 3"1 (Continued)

TABLE NOTATIONS "When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

""Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

~

"""Below P-10 (Low Setpoint Power 'Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 7 days.

,(2) Comparison of calorimetric to excore power indication above 15K of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2X. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15K of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3X. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) This table Notation number is not used.

(6) Incore-Excore Calibration, above 75'f RATED THERMAL POWER (RTP).

coincides with sustained If operation the quarterly surveillance requirement between 30K and 75K of RTP, calibration shall be performed at this lower power level. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

With power greater than or equal to the Interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive annunciator window.

Monthly surveillance in MODES 3", 4", an'd 5" shall also include verifica-tion that permissive P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive annunciato~ window.

Monthly surveillance shall include verification of the High Flux at Shut-down Alarm Setpoint of 1/2 decade above the existing count rate.

(10) Setpoint verification is not applicable.

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the OPERABILITY of the undervoltage and shunt trip attachment of the Reactor Trip Breakers.

TURKEY POINT - UNITS 3 8 4 3/4 3-11 'MENDMENT NOS. AND FEB > 8 1;.,

TABLE 4. 3" 1 Continued TABLE NOTATIONS

12) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(13) Remote manual undervoltage trip when breaker placed in service.

(14) Interlock Logic Test shall consist of verifying that the interlock is in its required state by observing the permissive annunciator window.

(15) Automatic undervoltage trip.

TURKEY POINT - UNITS 3 8 4 3/4 3-12 AMENDMENT NOS. AND

I

'1<

II'f

~C ~g ~ +

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MITING CONDITION FOR OPERATION

3. 3. 2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-2 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-3.

APPLICABILITY: As shown in Table 3.3-2.

ACTION:

a0 With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conserv'ative than the value shown in the Allowable Value column of Table 3.3-3, adjust the Setpoint consistent with the Trip Setpoint value.

b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Allowable Value column of Table 3.3-3, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-2 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-2.

SURVEILLANCE RE UIREMENTS 4.3.2. 1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

TURKEY POINT - UNITS 3 4 4 3/4 3-13 AMENDMENT NOS. AND Fc3 2 8 l4r'gg

4 1 TABLE 3.3-2 I

I ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUHENTATION

'l7 CD HINIHUH TOTAL NO. CHA(NELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE HODES ACTION 1.. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Isolation, Start Diesel Generators, Containment Cooling Fans, Contain-ment Filter Fans, Start Sequencer, Component Cooling Mater, Start Auxiliary Feedwater and Intake Cooling Water).

a. Hanual Initiation '2 1, 2, 3, 4 . 17

b. Automatic Actuation 1, 2, 3, 4 14 Logic and Actuation Relays C. Containment 1, 2, 3 15 Pressure-High

d. Pressurizer 1, 2, 38 15 Pressure - Low

e. High Differential 3/steam line 2/steam line 2/steam 1, 2, 3" 15 CD Vl Pressure Between in any steam line the Steam Line line Header and any Steam Line

J)

J I I l'

k

I TOTAL NO.

E 3. 3-2 CHANNELS Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUMENTATION HINIMUH CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE HODES ACTION

f. Steam Line Flow High 2/steam line 1/steam line 1/steam line 1,2;3* 15 Coincident with: in any two in any two steam lines steam lines Steam Generator Pressure--Low 1/steam 1/steam line 1/steam 1, 2, 3* 15 generator in any two generator steam lines in'any two steam lines or T Low 1/loop 1/loop in any two loops 1/loop in any two loops 1, 2, 3" 15

2. Containment Spray

a. Automatic Actuation 1, 2, 3, 4 14 Logic and Actuation Relays

b. Containment Pressure 2 1, 2, 3 15 High-High Coincident with:

Containment Pressure High 1> 2> 3 15

3. Containment Isol ati on a0 Phase "A" Isolation

1) Hanual Initiation 2 1, 2, 3, 4 17

2) Automatic Actuation 2 1, 2, 3, 4 14 Logic and Actuation Relays

A p 4

't i+

E 3.3-2

~

Continued)

ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION HINIHUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

3. Containment Isolation (Continued)

3) Safety Injection See Item l. above for all Safety Injection initiating functions and requirements. (Manual S.I. initiation will not initiate Phase A I sol ation. )

b. Phase "B" Isolation

1) Hanual Initiation 2 . 2 (Both 1, 2, 3, 4 17 buttons must be pushed simultaneously to actuate)

2) Automatic 1,2,3,4 14 Actuation Logic and Actuation Relays

3) Containment 3 2 1, 2, 3 15 Pressure High-High Coincident with:

Containment Pressure High 3 1, 2, 3 15 C. Containment Ventil ation Isolation

1) Containment Isola- See Items 3.a.l and 3.b.l above for all Manual Containment tion Hanual Phase A Ventilation functions and requirements.

or Phase B

h g P'

. 3.3-2 (Continued

~

ENGINEEREO SAFETY FEATURE ACTUATION SATEH INSTRUHENTATION HINIHUH C) TOTAL NO. CHANNELS CHANNELS APPLICABLE

,FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE HODES ACTION

3. Containment Isolation (Continued)

I C/l

2) Automatic Actuation 2 1, 2, 3, 4 16 Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection initiating functions requirements.

4) Containment 2N 1, 2, 3, 4 16 Radioactivity-High CAJ

4. Steam Line Isolation I

a. Hanual Ini tiation 1/operating 1/operating 1/operating 1, 2, 3 (individual) steam line steam line steam line 2',

Automatic 1, 2, 3 20 Actuation Logic and Actuation Relays C7 m

C. Containment Pressure High-High Coincident with:

Containment Pressure 2,

1, 2, 3'5 3

15 C3 C/l High CC)

Cn

~k 3.3-2 Continued

Ã7 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION m

HINIHUH C) TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I

4. Steam Line Isolation (Continued)

d. Steam Line Flow High 2/steam line 1/steam line 1/steam line 1, 2, 3 15 CA Coincident with:

Steam Generator Pressure Low 1/steam 1/steam 1/steam 1, 2, 3 15 generator generator generator in any two in any two steam lines steam lines or T Low 1/loop 1/loop in 1/loop in 1 2 3 15 any two any two.

loops loops Col I

CQ 5. Feedwater Isolation Automatic Actua- 1, 2 22 tion Logic and Actuation Relays

b. Safety- Injection See Item 1. above for all Safety Injection initiating functions and requilements.

6. Auxi l iary Feedwaterkf8

a. Automatic Actua- 1, 2, 3 20 Cl C/l tion Logic and Actuation Relays .

l'g

)

~g If<

C II ~

rk

T 3. 3-2 Continued ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION yc m

MINIMUM C) TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION.

6. Auxiliary FeedwaterN8 (Continued)

b. Stm. Gen. Mater Level 3/steam 2/steam 2/steam 1, 2, 3 15 Low-Low generator generator generator in any steam generator

c. Safety Injection See Item l. above for all Safety Injection initiating functions and requirements.

d. Bus Stripping 1/bus 1/bus 1/bus 1 2 3-. 23

e. Trip of All Hain Feed-water Pumps Breakers 1/breakeI (1/breaker) (1/breaker) 1, 2 23

/operating /operating pump pump

7. Loss of Power aO 4. 16 kV Busses A and B 2/bus 2/bus 2/bus 1;2,3,4 18 (Loss of Voltage)

C)

b. 480 V Load Centers 2 per load 2 on any 2 per load 1, 2, 3, 4 18 m center load center center 3A, 3B, 3C, 30 and .

4A, 4B, 4C, 40 (2 instantaneous relays per load center)

Degraded Voltage Coincident with: See Item 1. above for all Safety Injection initiating functions I

Safety Injection and requirements Co Oo 4C>

ID

0 I'

3. 3-2

~

~ Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUHENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

7. Loss of Power (Continued)

c. 480 V Load Centers 2 per load 2onany 2per load. 1,2,3,4 18 3A, 3B, 3C, 3D and center . load center center 4A, 4B, 4C 4D (2 inverse time relays per load center) Degraded Voltage

8. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure 1, 2, 3 19 b Tav Low 1 2.3 19

9. Control Room Isolation

a. Automatic Actuation 1, 2, 3, 4,6** 16 Logic and Actuation Relays

b. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

C. Containment Radio- 2 1 1 1, 2, 3, 4,6"" 16 activityHigh

d. Containment Isolation 2 1, 2, 3, 4 17 Manual Phase A or Phase B

e. Control Room Air Al l 24 Intake Radiation Level

0

~I

'))V))

'1 0

il III

TABLE 3.3-2 Continued TABLE NOTATION

¹Trip function may be blocked in this MODE below the Pressurizer Pressure

~ ~

Interlock Setpoint of 2000 psig.

~

¹¹Channels are for particulate radioactivity and for gaseous radioactivity.

¹¹¹Auxiliary feedwater manual initiation is included in Specification 3.7. 1.2.

"Trip function may be blocked in this MODE, below the T Low Interlock Setpoint.

"*Only during CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION STATEMENTS ACTION 14- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2. 1, provided the other channel is OPERABLE.

ACTION 15- With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST or TRIP ACTUATING DEVICE OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 16- With less than the Minimum Channels OPERABLE requirement, comply with the ACTION statement requirements of Specifica-tion 3.3.3.1 Item la of Table 3.3-4.

ACTION 17- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TURKEY POINT - UNITS 3 8 4 3/4 3-21 AMENDMENT NOS. AND FEaS )gag

)h 4

9 h

"Fh Y"

TABLE 3.3"2 Continued TABLE NOTATION (Continued)

TION 18- With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 19- With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specifica-tion 3.0.3. I nCTION 20 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for

, surveillance testing per Specification 4.3.2. 1 provided the other channel is OPERABLE.

ACTION 21- With the number of OPERABLE channels one less than the of Channels, restore the inoperable channel to OPERABLE Total'umber status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare, the associated valve inoper-able and take the ACTION required by Specification 3.7.1.5.

"CTION 22- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2. 1 provided the other channel is OPERABLE.

ACTION 23- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, comply with Specification 3.0.3.

ACTION 24- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate the control room Emergency Ventilation System and initiate operation of the Control Room Emergency Ventilation System in the recirculation mode.

TURKEY POINT - UNITS 3 5 4 3/4 3-22 AMENDMENT NOS. AND FEB 2 s tssg

-~ 18gg

cp l

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM UM N S N TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE¹

1. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Isolation, Start Diesel Generators, Containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water)

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N:A.

c. Containment Pressure High <6 psig <[ ] psig

d. Pressurizer Pressure Low >1715 psig- >[ ] psig

e. High Differential Pressure <150 psi <[ ] psi Between the Steam Line Header and any Steam Line.

f. Steam Line Flow High <A function defined [

as follows: A hp corresponding to 0.64 x 10e lbs/hr at OX load increa-ing linearly to a Lip corresponding to 3.84 x 10e lbs/hr at full load.

Coincident with: >600 psig >[ ] psig Steam Generator Pressure Low or T Low >531 F [ ]oF

'. Containment Spray

a. Automatic Actuation Logic N. A. N.A.

and Actuation Relays

b. Containment Pressure High- <30.0 psig <[ ] psig High Coincident wi.th:

Containment Pressure High <6.0 psig ] psig TURKEY POINT " UNITS 3 8L 4 3/4 3-23 AMENDMENT NOS. AND FEB 2S )g9

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INS RUMEN A ON P 0 N TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE¹

3. Containment Isolation

a. Phase "A" Isolation I

1) Manual Initiation N.A. N.A.

2) Automatic Actuation Logic N.A. N.A.

and Actuation Relays

3) Safety Injection See Item 1 above for all Safety Injection Trip Setpoints and Allowable Values.

b. Phase "B" Isolation

1) Manual Initiation N.A. N.A.

2) Automatic Actuation Logic N.A. N.A.

and Actuation Relays

3) Containment Pressure <30.0 psig <[ ]'sig High-High Coincident with:

Containment Pressure High <6.0 psig ] psig C. Containment Ventilation Isolation

1) Containment Isolation N.A. N.A.

Manual Phase A or Phase B

2) Automatic Actuation Logic N.A. N.A.

and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Al 1 owabl e Values.

4) Containment Radio- Parti cul ate (R-ll) [

activity--High (1) <6.1 x 10s CPM Gaseous (R-12)

See (2)

TURKEY POINT - UNITS 3 8 4 3/4 3-24 AMENDMENT NOS. AND

(7 lg

(,

Q

TABLE 3.3"3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM NS UM N A N P N

'TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUEO

4. Steam Line Isolation

a. Manual Initiation N.A. N.A

b. Automatic Actuation Logic N. A. N. A.

and Actuation Relays

c. Containment Pressure--High- <30.0 psig <[ ] psig High Coincident with:

Containment Pressure--High <6.0 psig <[ ] psig

f. Steam Line Flow High <A function defined [ ]

as follows: A Lg corresponding to 0.64 x 10e lbs/hr at GX load increa-ing linearly to a hp corresponding to t

3.84 x 10 lbs/hr at full load.

Coincident with: >600 psig >[ ] psig Steam Line Pressure--Low or T Low >531'F >L ]'F

5. Feedwater I so 1 ati on

a. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

b. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

6. Auxiliary Feedwater (3)

a. Automatic Actuation Logic N.A. N. A.

and Actuation Relays

b. Steam Generator Water >>15K of narrow >[ ]X of narrow Level--Low-Low range instrument range instrument span. span.

See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

TURKEY POINT - UNITS 3 8 4 3/4 3-25 AMENDMENT NOS. AND

~B Ss ].-.,q

gi

~

'I~

4 J

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUM NTATION TRI S 0 NTS TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUEO

6. Auxiliary Feedwater (Continued)

d. Bus Stripping See Item 7. below for all Bus Stripping Setpoints and Allowable Values.

e. Trip of All Main Feedwater N. A. N. A.

Pump Breakers.

7. Loss of Power

a. 4.16 kV Busses A and B N.A. N. A.

(Loss of Voltage)

b. 480V Load Centers (Instantaneous Relays)

Degraded Voltage Load Center 3A 436V+5V (10 sec + 1 sec delay) 3B 416V+5V (10 sec + 1 sec delay) 3C 417V+5V (10 sec 2' sec delay) 3D 428V+5V (10 sec + 1 sec delay) 4A 415V+5V (10 sec 2 1 sec delay) 4B 414V15V (10 sec + 1 sec delay) 401V+5V (10 sec + 1 sec delay) 4D 403V+SV (10 sec + 1 sec delay)

Coincident with:

Safety Injection See Item l. above for all Safety Injection Trip Setpoints and Allowable Values.

'TURKEY POINT - UNITS 3 8 4 3/4 3-26 AMENDMENT NOS. AND

('<i' 5 lggg

I lg 1f

~ J III

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM NS TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE¹

7. Loss of Power (Continued)

c. 480V Load Centers (Inverse Time Relays)

Degraded Voltage Load Center 3A 419Vt5V(60 sec +30 sec delay) 3B 426Va5V(60 sec 130 sec delay) 3C 427VR5V(60 sec 230 sec delay) 3D 436Vi5V(60 sec 230 sec delay)'27VR5V(60 sec 130 sec delay) 4B 424V15V(60 sec 130 sec delay) 4C 413VR5V(60 sec 130 sec delay) 4D 412V25V(60 sec 230 [

sec delay)

8. Engineering Safety Features Actuation System Interlocks

a. Pr essurizer Pressure <2000 psig <[ ] psig

b. T Low >531 F [ ]

9. Control Room Isolation

a. Automatic Actuation Logic and Actuation Relays N. A. N.A.

b. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

TURKEY POINT - UNITS 3 5 4 3/4 3-27 AMENDMENT NOS. AND

I l

'k<

'E

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM N RUM NTA ON IP S POINTS

'TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE¹

9. Control Room Isolation (Continued)

c. Containment Radioactivity Par ticulate (R-11) [

High (1) <6.1 x 10s CPM Gaseous (R-12)

See (2)

d. Containment Isolation N.A. N.A.

Manual Phase A or Phase B

e. Air Intake Radiation Level < 2 mR/hr z.83 ~R/kr TABLE NOTATIONS

'1) Either the particulate or gaseous channel in the OPERABLE status will satisfy this

'PM, LCO.

(2) Containment Gaseous Monitor Setpoint =

( F )

Where F Actual Pur e Flow Design Purge Flow (35,000 CFM)

Setpoint may vary according to current plant conditions provided that the release rate does not exceed allowable limits provided in Specification 3.11.2.1.

(3) Auxiliary feedwater manual initiation is included in Specification 3.7.1.2.

¹If no allowable value is specified so indicated by [ ], the trip setpoint shall also be the allowable value.

TURKEY POINT - UNITS 3 & 4 3/4 3-28 AMENDMENT NOS. AND 4@4 FEQ gg

'3 j5 4

I ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E C TRIP ANALOG ACTUATING NODES CHANNEL DEVICE FOR MHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST¹ IS RE UIRED

1. Safety Injection (Reactor Trip, Turbine Trip, Feed-water Isolation, Control Room Isolation, Start Diesel Generators, Contain-ment .Cooling Fans, Contain-ment Filter Fans, Start Sequencer, Component Cooling Mater, Start Auxiliary Feed-water and Intake Cooling Water)

a. Manual -Initiation N.A. N.A. N.A. N.A. 1 2'3

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) 1, 2, 3(4)

Logic and Actuation C.

Relays Containment Pressure High N.A. N.A. ~ P.A. 1, 2, 3

d. Pressurizet Pressure S R V(6) N.A. 1, 2, 3(4)

Low

e. High Differential H(6) N.A. N.A. 1,,2, 3(4)

C) Pressure Between the Cil Steam Line Header and any Steam Line

f. Steam Line Flow High S V(6) N.A. N.A. 1, 2, 3(4)

Coincident with:

Steam Generator Pressure--Low S H(6) N.A. N.A. 1, 2, 3(4) or T Low S v(e) N.A. N.A. 1, 2, 3(4)

~ g' C:J

TABLE (Continued)

I ENGINEERED SAFETY FEATURE ACTUATION SYSTEH JNSTRUHENTATION m S R A C RE U RE E C) TRIP M

ANALOG ACTUATING MODES I CHANNEL DEVICE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TElY LIIBIC lESTN 15 REISUlREII CA CAP 2. Containment Spray Qo

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) 1, 2, 3, 4 Logic and Actuation Relays

b. Containment Pressure N.A. N.A. M(l) 1, 2, 3 High-High Coincident with:

Containment Pressure-" N.A. N.A. M(1) 1, 2, 3 High

3. Containment Isolation

a. Phase "A" Isolation

1) Manual Initiation N.A. N.A. N.A. N.A. 1, 2, 3, 4 m 2) Automatic Actua- N.A. N.A. N.A. N.A. M(1) 1, 2, 3, 4 C7 tion Logic and m Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

b. Phase "B" Isolation

1) Manual Initiation N.A. N.A. N.A. N.A. 1, 2, 3, 4

2) Automatic Actua- N.A. N.A. N.A. N.A. M(1) 1, 2, 3, 4 tion Logic and Actuation Relays

Qp g4 I,I

TABLE (Continued)

I ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP ANALOG ACTUATING MODES CHANNEL DEVICE FOR MHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST)7 IS RE UIRED

3. Containment Isolation (Continued)

3) Containment N.A. N.A. 1 2 3 Pressure High-High Coincident with: Containment Pressure High N.A. N.A. M(1) 1, 2, 3

c. Containment Venti-lation Isolation

1) Containment N.A. N.A. N.A. N.A. 1, 2, 3, 4 Isolation Manual Phase A or Phase B H.A.

2) Automatic Actua- N.A. N.A. -

N.A. N.A.

tion Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. =

4) Containment Radio- N.A. N.A. 1,'2,3,4 activityHigh

4. Steam Line Isolation

a. Manual Initiation N.A. N.A. N.A. N.A. 1, 2, 3

b. Automatic Actuation N.A. N.A. N.A. ~ N.A. M(1) 1, 2, 3(4)

Logic and Actuation Relays

4

.$ F tw

TABLE (Continued)

I I

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUHENTATION S RYE LAC RE UIRHENS C)

TRIP ANALOG ACTUATING HODES CHANNEL DEVICE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TESTS IS RE UIRED 5 pp,~ 'N 8 z&p.k(okf (co t.~ucg

c. Containment Pressure N.A. R N.A. R H(1) 1, 2, 3 High-High Coincident with:

Containment Pressure N.A. N.A. 1, 2, 3 High High

d. Steam Line Flow Coincident with:

Steam Generator Pressure--Low or S(4)

S(4)

H(6)

'.A. N.A.

N.A.

1,2,3 1, 2, 3 T Low S(4) H(6) N.A. N.A. 1, 2, 3

5. Feedwater Isolation

a. Automatic Actuation N.A. N.A. N.A. N.A. 1, 2 Logic and Actuation Relays m

CD b. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

6. Auxiliary Feedwater (2)

a. Automatic Actuation N.A. N.A. N.A. N.A. 1, 2, 3 Logic and Actuation Relays

b. Steam Generator N.A. N.A. 1, 2, 3 Water Level Low-Low Oy

<<e

t I

TABLE . (Continued)

I ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

~

m RV C TRIP ANALOG ACTUATING NODES I CHANNEL DEVICE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TESTS IS RE UI RED CA Crd 6. Auxiliary Feedwater (Continued)

Qo

c. Safety Injection See Item l. above for all Safety Injection Surveillance Requirements.

d. Bus Stripping N.A. N.A. N.A. 1 2 3

e. Trip of All Hain N.A. N.A. N.A. N.A. 1, 2 Feedwater Pump Breakers.

7. Loss of Power

a. 4.16 kV Busses A N.A. N.A. N.A. 1, 2, 3, 4 and B (Loss of Voltage)

b. 480V Load Centers N.A. N.A. 1, 2, 3, 4 3A,3B,3C,3D and m 4A,4B,4C,40 (Instantaneous m

. Relays) Degraded Voltage C) tA Coincident with:

Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

c. 480V Load Centers S N.A. H(1) N.A. 1, 2, 3, 4 3A,3B,3C,3D and 4A,4B,4C,40 (Inverse Time Relays) Degraded Voltage

r;

( ~

k,

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION S VEILLANCE RE UIREME TS TRIP CD ANALOG ACTUATING MODES CHANNEL DEVICE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST L IC TE TN I5 REISUIRED

8. Engineering Safety Features Actuation System Interlocks

a. Pressurizer Pressure N.A. M(6) N.A. N.A. 1, 2, 3(4)

b. T --Low N. A. M(6) N.A. N.A. 1, 2, 3(4)

9. Control Room Isolation

a. Automatic Actuation N.A. N.A. N.A. N.A. N.A. (3)

Logic and Actuation Relays

b. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

C. Containment S R M N. A. N.A. (5)

Radioactivity--High

d. Containment Isolation N.A. N.A. N.A. N~ A. 1, 2, 3, 4 Manual Phase A or Phase B

e. Control Room Air N.A. N.A. Al 1 Intake Radiation Level CD TABLE NOTATIONS (1) Each-trai'n shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(2 uxi3~~ 'e ater manual initiation is included in S ecificat CD C/l Applicable in MO S 1,, 3, 4 or dunng CORE ALTE A containment or in the s ent fuel pool.

or movement of irradiated fuel wVKsn (4) e provsssons o pecs )ca son .. are not applicable for entering Mode 3, provided that the applicable surveillances are completed within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from entering Mode 3.

CD (5) Applicable in MODES 1, 2, 3, 4 or during CORE ALTERATIONS or movement of irradiated fuel within the containment.

(6) Test of alarm function not required when alarm locked in.

At east once per 18 months each Actuation Logic Test shall include energization of each relay and verification of OPERABILITY of each relay,

INSTRUMENTATION 3/4. 3. 3 MONITORING INSTRUMENTATION AOIATION MONITORING FOR PLANT OPERATIONS LIMITING CONOITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-4 shall be OPERABLE with their Alarm/Trip Setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-4.

ACTION:

aO With a radiation monitoring channel Alarm/Trip Setpoint for plant operations exceeding the value shown in Table 3.3-4, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Tabl.e 3.3-4.

C. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 3. 3. 1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MOOES and at the frequencies shown in Table 4.3-3.

TURKEY POINT - UNITS 3 4 4 3/4 3-35 AMENOMENT NOS. ANO

7 Y

'\

p

'Ir

RADIATION MONITORING INSTRUHENTATION FOR P/ANT OPERATIONS CHANNELS CHANNELS APPLICABLE ALARH/TRIP FUNCTIONAL UNIT TO TRIP/ALARH OPERABLE HODES SETPOINT ACTION

1. Containment

a. Containment Atmosphere 1 Al l" Particulate 26 for HODES 1, 2, 3, 4 Radioactivity-High <6. lxlOsCPH or (Particulate or t'aseous Gaseous (See Note -1.)) See Note 2. 27 HODES 5 AND 6

b. RCS Leakage Detection N.A. 1, 2, 3, 4 N.A. 26 Particulate Radio-activity or Gaseous Radioactivity

2. Spent Fuel Storage Pool Areas

a. Unit 3 Radioactivity - 1 <5.5x10-2 pCi 28 High Gaseous CC

.b. Unit 4 Radioactivity- 1 <2.8xl0-2 ~Ci 28 High CC Gaseous'.

(SPING) or

<1.0x106CPH PRHS)

C ro oem adiation Level i~mt,e~el 2 Al 1 < m~~r 23

N C,~

J 4/1 a,

TABLE 3. 3-4 Continued TABLE NOTATIONS During CORE ALTERATIONS or movement of irradiated fuel within the containment comply with Specification 3/4.9.13.

With irradiated fuel in the spent fuel pits.

Unit 4 Spent Fuel Pool Area is monitored by Plant Vent radioactivity instrumentation.

Note 1 Either the particulate or gaseous channel in the OPERABLE status will satisfy this LCO.

Containment Gaseous Monitor Setpoint 3.2 x 10~

Note 2 CPM,

( F )

Actual Pur e Flow es>gn urge ow Setpoint may vary according to current plant conditions provided that the release rate does not exceed allowable limits provided in Specification 3.11.2.1.

ACTION STATEMENTS ACTION 26- In MODES 1. thru 4: With both the Particulate and Gaseous Radioactivity. Monitoring Systems inoperable, operation may continue for up to 7 days provided:

1) A Containment sump level monitoring system is OPERABLE,

2) Appropriate grab samples are obtained and analyzed at

~

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

3) A Reactor Coolant System water inventory balance is performed at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during steady state operation except when operating in shutdown cooling mode, and

4) Containment Purge, Exhaust and Instrument Air Bleed Valves are maintained closed.

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (ACTION 27 applies in MODES 5 and 6).

TURKEY POINT - UNITS 3 4 4 3/4 3-37 AMENDMENT NOS. AND FEB 2 8 >989

4 yL

'I I

TABLE 3. 3-4 Continued ACTION STATEMENTS (Continued)

TION 27 " In MODES 5 or 6 (except during CORE ALTERATION or movement of irradiated fuel within the containment): With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement perform the following:

1) Obtain and analyze appropriate grab samples at least once .

per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2) Monitor containment atmosphere with area radiation monitors.

Otherwise, isolate all penetrations that provide direct access from the containment atmosphere to the outside atmosphere.

During CORE ALTERATION or movement of irradiated fuel within the containment: With the number of OPERABLE Channels. less than the Minimum Channels OPERABLE requirements, comply with ACTION statement requirements of Specification 3.9.9 and 3.9. 13.

ACTION 28- With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, immediately suspend operations in the Spent Fuel Pool area involving spent fuel manipulations.

ACTION 29- With the'pumber of OPERABLE channels one ss than the Minimum, Channels OPERABLE requiremh t,.within 1 houq isolate the Control Roap Emergency Vents ation, System and initiate operation og the Control Roo Emergency Venti~laat .on System in mode.

~~recircul ation TURKEY POINT - UNITS 3 8L 4 3/4 3-38 AMENDMENT NOS. AND

pk 1

r

4. 3-3 I

I RADIATION HONITORING INSTRUMENTATION FOR PLANT R ON R H N a

CD ANALOG CHANNEL MODES FOR WICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS RE UIRED CA

1. Containment

a. Containment Atmosphere Al 1 Radiaoctivity High

b. RCS, Leakage Detection

1) Particulate Radio- R 1, 2, 3, 4 activity

2) Gaseous Radioactivity 1, 2, 3, 4

2. Spent Fuel Pool Areas

a. Unit 3 Radioactivity High Gaseous R

b. Unit 4 (Plant Vent)

Radioactivity High CD Gaseous'SPING m and PRHS)

C ro RoopA ntake R A l-.

CD lA adi ati o~~ve1 TABLE NOTATIONS

" with irradiated fuel in the fuel storage pool areas.

8 Unit 4 Spent Fuel Pool Area is monitored by Plant Vent radioactivity instrumentation.

OO OCt3 C. 1 lQ

/'

)ll tl lt+

yE>> ~ a h

['I I

INSTRUMENTATION MOVABLE INCORE DETECTORS IMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a ~ At least 16 detector thimbles wh'en used for recalibration and check of the Excore Neutron Flux Detection System and monitoring the gUANDRANT POWER TILT RATIO", and at least 38 detector thimbles when used for monitoring F~, F~(Z) and F (Z).

b. A minimum of two detector thimbles per core quadrant, and

c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO", or N

c. Measurement of .F~, F~(Z) and Fx (Z).

AUCTION:

With the Movable Incore Detection System inoperable, do not use the system

~

for the above applicable monitoring or calibration functions.

~

The provisions of Specification 3.0.3 are not applicable.

~

SURVEILLANCE RE UIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO", or N

c. Measurement of F~, F~(Z) and Fx (Z).

Exception to the 16 detector thimble requirement of monitoring the QUADRANT POWER TILT RATIO is acceptable when performing Specification 4.2.4.2 using two sets of four symmetric thimbles.

TURKEY POINT - UNITS 3 & 4 3/4 3-40 AMENDHENT NOS. AND FED 2 8 jgg9

'h 1'

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION MITING CONDITION FOR OPERATION

3. 3. 3. 3 The accident monitoring instrumentation channels shown in Table 3. 3-5 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-5.

ACTION:

a. As shown in Table 3.3-5.

b. The provisions of Specification 3.0.4 are not applicable to ACTIONS in Table 3.3-5 that require a shutdown.

SURVEILLANCE RE UIREMENTS 4.3.3.3 Each accident monitoring instrumentation channel shall be demonstrated "PERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the requencies shown in Table 4.3-4.

TURKEY POINT - UNITS 3 8 4 3/4 3-41 AMENDMENT NOS. AND FEB ~8 >Sag

1 J)

I~

3LE 3.3-5 I

ACCIDENT HONITOR ING INSTRUHEETATION TOTAL HINIHUH APPLI-NO. OF CHANNELS CABLE INSTRUMENT CHANNELS OPERABLE NODES ACTIONS

1. Containment Pressure (Mide Range) 1, 2, 3 31, 32

2. Containment Pressure (Narrow Range) 1, 2, 3 36

3. Reactor Coolant Outlet Temperature 2-2 Detectors 1-2 Detectors 1, 2, 3 31, 32 THOT (Wide Range) per Channel per Channel Reactor Coolant Inlet Temperature 2-2 Detectors 1-2 Detectors 1, 2, 3 31, 32 TCOLO (Wide Range) per Channel per Channel

5. Reactor Coolant Pressure - Mide Range 1, 2, 3 31, 32

6. Pressurizer Mater Level 1, 2, 3 31, 32

7. Auxiliary Feedwater Flow Rate 2/steam 1/steam 1, 2, 3 31, 32 generator generator

8. Reactor Coolant System Subcooling Hargin 2(2) 1(2) 1, 2, 3 31, 32 Honitor

9. PORV Position Indicator (Primary Detector) 1/valve 1/valve 1, 2, 3 33

10. PORV Block Valve Position Indicator 1/valve 1/valve 1, 2, 3 33

11. Safety Valve Position Indicator (Primary 1/valve 1/valve 1 2 3 32 Detector)

12. Containment Water Level (Narrow Range) 1, 2, 3 36

13. Containment Water Level (Wide Range) 1, 2, 3 31, 32

$1 Fl,

0 TABLE 3.3-5 (Continued)

ACCIDENT HONITORING INSTRUMENTATION a

C)

TOTAL MINIHUH APPLI-NO. OF CHANNELS CABLE INSTRUHENT CHANNELS OPERABLE HODES ACTIONS C: 14. In Core Thermocouples (Core Exit Thermo- 4/core 2/core 1, 2, 3 31, 32 couples) quadrant C/l 4J 15. Containment High Range Area Radiation 1, 2, 3 34 go

16. Reactor Vessel Level Monitoring 2(1) 1, 2, 3 37, 38 System

17. Neutron Flux, Backup NIS (Wide Range) 1, 2, 3 31, 32 CrJ

18. Containment Hydrogen Monitors 1, 2 35 40 I 19. High Range-Noble Gas Effluent Monitors

a. Plant Vent Exhaust ALL 34

b. Unit 3-Spent Fuel Pit Exhaust ALL 34

c. Condenser Air Ejectors 1, 2, 3 34

d. Main Steam Lines 1, 2, 3 34

20. RWST Water Level 2 1 1,2,3 31, 32

21. Stean Generator Mater Level (Narrow Range) 2/stm. gen. 1/stm. gen. 1, 2, 3 31, 32 Containment Isolation Valve Position Indication* 1/valve 1/val ve 1, 2, 3 39 1ABLE N01A1IONS C)

C/7 A channel is eight sensors in a probe. A channel is OPERABLE if a minimum of four sensors are OPERABLE.

2. Inputs to this instrument are from instrument items 3, 4, 5 and 14 of this Table.

" Applicable for containment isolation valve position indication designated as post-accident monitoring instru-mentation (containment isolation valves which receive containment isolation Phase A, Phase B, or containment ventilation isolation signals).

l

= 1t

, I U

A'

TABLE 3. 3-5 Continued ACTION STATEMENTS CTION 31 With the number of OPERABLE accident monitoring instrumentation channel(s) less than the Total Number of Channels either restore the inoperable channel(s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 32 With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 33 Close the associated block valve pn open its circuit breaker.

ACTION 34 With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s),

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) Either restore the inoperable channel(s) to OPERABLE status within 7 days of the event,.or

2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

ACTION 35 With one or both hydrogen monitor(s) inoperable, comply with Action Requirements of Specification 3.6.5.

ACTION 36 With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channel OPERABLE, either restore the inoperable channel to OPERABLE status within 30 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 37 With the number of OPERABLE chanhels one less than the Total Number of Channels, restore the system to OPERABLE status within 7 days. If repairs are not feasible without shutting down, prepare and submit a Special Report to the Commission pursuant to Specification 6. 9. 2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

TURKEY POINT - UNITS 3 8 4 3/4 3"44 AMENDMENT NOS. AND 2 d lggg

) ~

a

~,

TABLE 3. 3-5 Continued ACTION STATEMENTS ACTION 38 With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirements, restore the inoperable within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . If repairs are not channel(s)'o OPERABLE status feasible without shutting down:

1. Initiate an alternate method of monitoring the reactor vessel inventory; and

2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9. 2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and

3. Restore at least one channel to OPERABLE status at the next scheduled refueling.

ACTION 39 With the number OPERABLE channels less than the Minimum Channels OP LE re irement, comply with the provisions of Specificati 3. 4. r an ino erable containment isolation valve.

pop I t LO~

aHe~~ee nne~~g (p,q.

9 /

~~ >t v'4t'iv'e. Zo<Tm)g pOS i t<Vm lndigq4,g~

vlsuat aksevvati'og) ~~yt~

Qgv r ~ g~d POSE<< We (Hoylt-aha.

nc((5) ~ tk gp I ou 5 Gf'URKEY POINT - UNITS 3 5.4 3/4 3-45 AHENDHENT NOS. AND

/AY 05 1989

TABLE 4.3-4 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATIQH

1. Containment Pressure (Wide Range) M R

2. Containment Pressure (Narrow Range)

3. Reactor Coolant Outlet Temperature " THOT (Wide Range)

4. Reactor Coolant Inlet Temperature -

TCOLD

{Wide Range)

5. Reactor Coolant Pressure - Wide Range

6. Pressurizer Water Level

7. Auxiliary Feedwater Flow Rate

8. Reactor Coolant System Subcooling Margin Monitor M

9. PORV Position Indicator {Primary Detector) M

10. PORV Block Valve Position Indicator M

11. Safety Valve Position Indicator (Primary Detector) M

12. Containment Water Level {Narrow Range)

13. Containment Water Level (Wide Range)

14. In Core Thermocouples (Core Exit Thermocouples)

15. Containment - High Range Area Radiation Monitor

16. Reactor Vessel Level Monitoring System

17. Neutron Flux, Backup NIS (Wide Range)

18. Containment Hydrogen Monitor

19. High Range - Noble Gas Effluent Monitors

a. Plant Vent Exhaust

b. Unit 3 - Spent Fuel Pit Exhaust

c. Condenser Air Ejectors

d. Main Steam Lines

20. RWST Mater Level

21. Steam Generator Water Level {Narrow Range) 2382;. ontainment Isolation Valve Position Indication M R "Acceptable criteria for calibration are provided in Table II.F.1-3 of NUREG-0737.

TURKEY POINT - UNITS 3 4 4 3/4 3-46 'AMENDMENT NOS. AND AWAY 05 t988

$z I

+ik y 'yg ll

INSTRUMENTATION FIRE DETECTION INSTRUMENTATION MITING CONDITION FOR OPERATION 3.3.3.4 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-6 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the fire detection instrument is required .to be OPERABLE.

ACTION:

a. With any, but not more than one-half the total in any fire zone, Function A fire detection instruments shown in Table 3. 3-6 inoperable, restore the inoperable instrument(s) to OPERABLE status within 14 days or within the next 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the zone(s) with the inoperable instrument(s) at least once per hour, unless the instrument(s) is located inside the containment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5).

b. With more than one-half of the Function A fire detection instrument in any fire zone shown in Table 3.3-6 inoperable, or with any Function B fire detection instruments shown in Table 3.3-6

~ ~ ~ ~

inoperable, or with any two or more adjacent fire detection

~

instruments shown in Table 3.3-11 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

~ ~ ~ ~

~

establish a fire watch patrol to inspect the zone(s) with the

~ ~

~

inoperable instrument(s) is located inside the containment, then

~ ~ ~ ~ ~

inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or monitor the containment air temp ature at least once per hour at the locations listed in Speci 'tion 4.6.1.5).

/ra~( once @

d p'. The provisions o Specifications 3.0.3 are not applicable. guur >,.~~ f4e

, p1an+ inkruivenf I'O)

SURVEILLANCE RE UIR NTS TRIP 4.3.3.4.1 Each o the above re uired fire detection instruments which are accessible during operation sh 1 be demonstrated OPERABLE at least once per 6 months by performance of a ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLO SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

4.3. 3. 4. 2 The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instru~m~)al3. e demonstrated OPERABLE at leapt onsceyer 6 months.

Qj/A Ke. Fi're ouaG fua"frro noF85fagig)e7 ~/' /3'm'cue/ op

'the /ur/aine area, reofore 4e ~re aefcia afro/ sui'tt,'n one pour or revcctr< and ~ a 5Pecja/"' /C'oIoor'f" fe 4e TURKEY POINT - UNITS 3 & 4 3/4 3 47 NENOMENl RR. AiID'EB as >"=>9

V

'I I'I jP

TABLE 3.3"6 FIRE 'ETECTION INSTRUMENTS 0 SN AL U MN TOTAL NUMBER INSTRUMENT LOCATION OF INSTRUMENTS H LAM MO FIRE ZONE AREA ~xy)" ~x> y J . 7x7yy' Aux. Bldg. Corridor E. (2/0) 5 Drain/Laundry/Shower Tank Room 10'hem.

(2/0) 9 Laundry/Chemical Drain Tank Room (1/0.)

10 " Pipeway (ll/0) 11- Unit 3 RHR Heat Exchanger Room (5/0)>>~

12- RHR Pump 3A Room (2/0 ) EEEAEEE 13- RHR Pump 3B Room (2/0) EEEEEEEEE 14- Unit 4 RHR Heat Exchanger Room (5/P)EEE>>

15- RHR Pump 4A Room (2/P)EEE>>

16- RHR Pump 4B Room (2/0)EEE>>

19- Unit 3 W Elect Penet Room (5/P)EEEEEEA 20' Unit 3 S Elect Penet Room (11/0) 21- Instrument Shop (2/0) 22- Radioactive Laboratory (2/0) 26- Unit 4 N Elect Penet Room (8/0) 27- Unit 4 W Elect Penet Room (6/0) EEE>>

30- Unit 4 Piping and Valve Room (4/P )

40- Unit 3 Piping and Valve. Room (4/P)EEE>>

Unit 4 Charging Pump Room (0/4) (3/0)

Unit 4 Component Cooling Water Area (0/4) (5/2)>>EEE Unit 3 Component Cooling Water Area (0/4) (4/2) 'AEEE'k 55- Unit 3 Charging Pump Room (0/4) (3/0) 58- Aux Bldg Corridor, El. (18/0) 59- 4 Containment Electrical 18'nit (10/0)

Penet. Area""

60 Unit 3 Containment Electrical (16/0)

Penet. Area""

61- Reactor Control Rod Eqpmt Room - Unit 4 (4/0) 62- Computer Room (11/0) 63- Reactor Control Rod Eqpmt Room' Unit 3 (4/0) 67 " 4160V Switchgear 4B (10/0) .

68- 4160V Switchgear 4A (6/O) 70- 4160V Switchgear 3B (1o/0) 71- 4160V Switchgear 3A (6/0) 72- Emergency Diesel B (O/3 (1/0) (1/0)

.73- Emergency Diesel A (O/3 (1/0) (1/0) 74 " Emergency Day Tank Room B (1/1) 75- Emergency Day Tank Room A (1/1) 76- Unit 4 Turbine Lube Oil Reservoir (1/0) 79A- North-South Breezeway (0/6) (4/0) 81- Unit 4 Main Transformer (1/0)

Unit 4 Aux Transformer Area (1/0) a~e- Unit 3 and 4 Aux Feedwater Pump Area (3/0)

(DC Enclosure Bldg)

TURKEY POINT - UNITS 3 8E 4 3/4 3-48 AMENDMENT NOS. AND FEB 2 8 1S89

0 t

I J, H 7i 4y

TABLE 3. 3-6 Continued FIRE OETECTION INSTRUMENTS N L U M N TOTAL NUMBER INSTRUMENT LOCATION OF INSTRUMENTS FIRE ZONE AREA ~xy)" ~xi'MOK LAM

~xi'7

- Unit 3 Aux Trans former Area (1/0) 93 - 480V Load Center 4A and 4B (1/0) 94 - 480V Load Center 4C and 40 (2/o) 95 - 480V Load Center 3A and 3B (1/o) 96 - 480V Load Center 3C and 30 (2/0) 7 - Mechanical Equipment Room (1/0) 98 - Cable Spreading Room (16/15) 101- RPI Inverter and MG Sets (1/o) 102- Battery Rack 4B (1/0) 103- Battery Rack 3A (1/0) 104- RPI Inverter and MG Sets (2/0)

$68- Control Room (1/0) (17/0) 108A- Train A Inverters (3/4)

'108B- Train B Inverters (4/4) 109- Battery Rack 4A (1/0) 110- Battery Rack 3B (1/0) 113- Unit 4 Feedwater Platform (2/0)ALA

.116- Unit 3 Feedwater Platform (2/0)AAA

'.19- Unit 4 Intake Cooling Water Pump Area (4/0)AAA 20- Unit 3 Intake Cooling Water Pump Area (g/0) nnn 32- Control R

/pr /eire/ of ge 7orkin8.

TABLE NOTATIONS Q/p+ (0/AQ (y/Ag~ )

(x/y): x is number of Function A (early warning fire dectection and notification only) instruments.

y is number of Function B (actuation of Fire Suppression Systems and early warning fire detection and notification) instruments..

The fire detection instruments located within the containment are not required to be operable during the performance of Type A Containment Leakage Rate Test.

Installed to meet the requirements of 10 CFR Part 50, Appendix R,

-~ Section III.G.

A fire ~otol parol All 4 e~tgl,sgk ko in5pe~~ 4+ I It Teot level oI 4e 7vrLneAr~,e,qe. eeet 4ovr.

TURKEY POINT " UNITS 3 8 4 3/4 3"49 AMENDMENT NOS.'ND 2 8 1S89

'i 0

P J1 4

'k'l

'w'

INSTRUMENTATION RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION IMITING CONDITION FOR OPERATION 3.3.3.5- The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3"7 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Specification 3. 11. l. 1 are not exceeded. The Alarm/

Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL-(ODCM).

PPLICABILITY: At all times, except as indicated in Table 3.3-7.

ACTION:

'a 0 With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above spe-cification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-7. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9. 1.4 why this inoperability was not corrected in a timely manner.
C. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE UIREMENTS 4.3.3.5 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATION'AL TEST at the frequencies shown in Table 4.3-5.
TURKEY POINT - UNITS 3 & 4 3/4 3-50 AMENDMENT NOS. AND FEB 28 19S9
P 4
t l
11 ii '
TABLE 3.3-7
'ADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release
a. Liquid Radwaste Effluent Line 35
b. Steam Generator Blowdown Effluent Line 36
2. Flow Rate Measurement Devices
a. Liquid Radwaste Effluent Line 1* 37
b. Steam Generator Blowdown Effluent Line 1""/Steam 37 generator "Applicable during liquid effluent releases.
""Applicable during blowdown operations.
TURKEY POINT - UNITS 3 8( 4 3/4 3-51 AMENDMENT NOS. AND MAY 0 5 1989
i F'
TABLE 3. 3-7 Continued TABLE NOTATION ACTION 35- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiating a release:
a. At least two independent samples are analyzed in accordance with Specification 4. 11. 1.1. 1, and
b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 36- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross (beta or gamma) radioactivity at a lower limit of detection of no more than 1 x 10-~ microcuries/ml or analyzed isotopically (Gamma at a lower limit of detection of at least 5 x-10-~ microcurie m:
a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram OOSE E(UIVALENT I"131, or
b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 micro-curies/gram OOSE E(UIVALENT I-131.
ACTION 37- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves may be used to estimate flow.
TURKEY POINT - UNITS 3 8 4 3/4 3-52 AMENOMENT NOS. AND
/
FE8 2 S >g88
4' 1
h
RADIOACTIVE LI UID EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE RE UIREHENTS ANALOG CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUHENT CHECK CHECK CALIBRATION TEST Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release
a. Liquid Radwaste Effluents Line R(2)*
b. Steam Generator Blowdown Effluent Line R(2)
2. Flow Rate Heasurment Devices
a. Liquid Radwaste Effluent Line D(3) N.A.
b. Steam Generator Blowdown Effluent Lines D(3) N.A.
"Channel calibration frequency shall be at least once per 18 months.
TABLE NOTATIONS (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measures levels above the Alarm/Trip Setpoint.
(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.
P (h.
o'
INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Specification 3. 11.2. 1 and 3. 11.2.5 are not exceeded.
The Alarm/Trip Setpoints of these channels meeting Specification 3. 11.2. 1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
APPLICABILITY: As shown in Table 3.3-8 ACTION:
a~ With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable or change the setpoint so it is acceptably conservative.
b. With les's than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-8: Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9. 1.4 why this inoperability was not corrected in a timely manner.
C. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE UIREMENTS
4. 3. 3. 6 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-6.
,TURKEY POINT - UNITS 3 & 4 3/4 3-54 AMENDMENT NOS. AND MAY 0 5 1S89
I T 3. 3-8 I
I RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION HINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION I
C 1. GAS DECAY TANK STEH
'a 0 Noble s ctivity Honitor-Provi ng Alarm and Automatic Termination of Release (Plant Vent Honitor) 45
b. Effluent System Flow Rate Measuring Device
2. WASTE GAS DISPOSAL SYSTEH (Explosive Gas Honitoring System)
a. Hydrogen and Oxygen Honitors 49
3. Condenser Air Ejector Vent System
a. Noble Gas Activity Honitor (SPING or PRMS) 47
b. Iodine .Sampler 48 m
C7 C. Particulate Sampler m
d. Efflent A
System Flow Rate Heasur ng Device 1 46
e. Sampler Flow Rate Heasuring Device 1 46
i TABLE 3.3 ~ontinued I
I Al RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUHENTATION m
tD HINIHUH CHANNELS I INSTRUMENT OPERABLE APPLICABILITY ACTION
4. Plant Vent System (Include Unit 4's Spent Fuel Pool)
a. Noble Gas Activity Honitor (SPING 47 or PRHS)
b. Iodine Sample>
C. Particulate Sampler 48
d. fflpnt ystem Flow Rate Heasuring 46
~l
e. Sampler Flow Rate He supping Dev ce 46
5. Unit 3 Spent Fuel Pit Building Vent Noble Gas Activity Honitor
b. Iodine Sampler 48 m C. Particulate Sampler 48 ED m d Sampler Flow Rate Heasuring Device 46 tD EA CD
a n It, I,
til s
QF
TABLE 3. 3-8 (Continued) 0 At all times.
TABLE NOTATION
During GAS DECAY TANK SYSTEM operation.
Applies during MODE 1, 2, 3 and 4.
¹¹ Applies during MODE 1, 2, 3 and 4 when primary to secondary leakage is detected as indicated by condenser air ejector noble gas activity monitor.
ACTION 45- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:
a. At least two independent samples of the tank's contents are analyzed, and
.b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 46- With the.,number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , ACTION 47- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
ACTION 48- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 4. 11-2 and analyzed at least weekly.
ACTION 49- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the GAS DECAY TANK SYSTEM may continue provided that grab samples are collected and analyzed for hydrogen and oxygen concentration at least a) once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during degassing operations, and b) once per day during other operations.
TURKEY POINT - UNITS 3 5 4 3/4 3-57 AMENDMENT NOS. AND MAY 0 5 Iss9
1 4.3-6 l 1 I
7C RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE RE UIREHENTS m
D CD ANALOG CHANNEL MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE IS INSTRUMENT CHECK CHECK CALIBRATION TEST RE UIRED Cil 1. GAS DECAY TANK STEH Noble as Activity Honitor-Provi ng Alarm and Automatic Termination of Release (Plant Vent Honitor) R(3)
b. Effluent System Flow Rate CAR Heasuring Device N.A. N.A.
Col 2. GAS DECAY TANK SYSTEH (Explosive I
CJl Gas Honitoring System)
a. Hydrogen and Oxygen Monitors 0 N.A. g(4,5)
3. Condenser Air Ejector Vent System
a. Noble Gas Activity Honitor (SPING or PRHS) R(3) O(2) m Effluent CD b. Iodine Sampler N.A. N.A. N.A.
c. Particulate Sampler N.A. N.A. N. A.
CD Vl d. System Flow Rate Measuring Device N.A. N. A.
f(
'I 4~
I/i'>
,1'
TABLE 4. Continued I I
Xl Q ~ PC RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUHENTATION SURVEILLANCE RE UIREHENTS CD ANALOG CHANNEL MODES FOR WHICH I CHANNEL SOURCE CHANNEL OP E RATIONAL SURVEILLANCE IS INSTRUMENT CHECK CHECK CALIBRATION TEST R EIREIII R II
3. Condenser Air Ejector Vent System (Continued)
Qo
e. Sample Flow Rate Heasuring Device N.A. N.A.
Plant Vent System (Include Unit.
4's Spent Fuel Pool)
a. Noble Gas Activity Hohitor (SPING or PRHS) S(2)
b. Iodine Sampler N.A. N.A. N.A.
C. Pa lyte Sampler M N.A. N.A. N.A.
d. ffident S stem Flow Rate asuri Device 0 N.A. (6) N.A.
e. Sampler Flow Rate Me ur png CD Device D N.A. (6) N.A.
CD Cil CCR a.
CC
TABLE 4.3-6 (Continued)
I C:
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CD ANALOG CHANNEL MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE- IS INSTRUMENT CHECK CHECK CALIBRATION TEST RE UIRED
5. Unit 3 Spent Fuel Pit Building Vent
a. Noble Gas Activity Monitor R(3) S(2)
b. Iodine Sampler N.A: N.A. N.A.
C. Par ticul ate Sampler N.A. N.A. N.A.
d. Sampler Flow Rate Measuring Device D N.A. N.A.
I CJl CD TABLE NOTATION At all times.
During GAS DECAY TANK SYSTEM operation.
Applies during MODE 1, 2, 3 and 4.
¹¹ Applies during MODE 1, 2, 3 and 4 when primary to secondary leakage is detected as indicated by condenser air ejector noble gas activity monitor.
m C) (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measur levels above the Alarm/Trip Setpoint. J.
CD (2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that if the instrument indicates measured levels above the Alarm Setpoint, alarm, annunciation occurs in the control room (for PRMS only) and in the computer room (for SPING only).
CD (3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that paIticipate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
f r
TABLE 4.3-6 (Continued)
TABLE NOTATIONS Continued
) The CHANNEL CALIBRATION shall
~ ~ ~
include the use of standard gas samples containing a nominal.~
a. One volume percent hydrogen, balance nitrogen, and
b. Four volume percent hydrogen, balance nitrogen.
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
a. One volume percent oxygen, balance nitrogen, and
b. Four volume percent oxygen, balance nitrogen.
(6) CHANNEL CALIBRATION frequency shall be at least once per 18 months.
e TURKEY POINT - UNITS 3 & 4 3/4 3"61 AMENOHENT NOS. AND
~EB 4 s 1989
'I g>>
'J]
3/4.4 REACTOR COOLANT SYSTEM
/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION TARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION
3. 4. l. 1 All reactor coolant loops shall be in operation.
APPLICABILITY: MODES 1 and 2.
h iCTION:
With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SUREVEILLANCE RE UIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in opera-tion and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
TURKEY POINT - UNITS 3 8( 4 3/4 4"1 AMENDMENT NOS. AND FEB 28 1sSS
4't ig g't>
K
~
'1
REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4. 1.2 All of the reactor coolant loops listed below shall be OPERABLE with all reactor coolant loops in operation when the Reactor Trip breakers are closed and two reactor coolant loops listed below shall be OPERABLE with at least one reactor coolant loop in operation when the Reactor Trip breakers are open:"
a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,
b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump, and
c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump.
ARPLICABILITY: MODE 3 ACTION:
'a 0 With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
b. With less than three reactor coolant loop in operation and the Reactor Trip breakers in the closed position, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor Trip breakers.
C. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.
SURVEILLANCE RE UIREMENTS 4.4. 1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4,4. 1.2.2 The required steam generators shall be determined OPERABLE by verify-ing secondary side water level to be greater than or equal to 10K at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
.4.4. 1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
t "All reactor coolant (1) no
,TURKEY POINT "
pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:
operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.
UNITS 3 8L 4 3/4 4-2 AMENDMENT NOS. AND FEB R 8 1989
0 q'3
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REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4. 1.3 At least two of the loops listed below shall be OPERABL'E and at least one of these loops shall be in operation:"
a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,"*
b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,""
c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,""
d. RHR Loop A, and
e. RHR Loop B.
APPLICABILITY: MODE 4.
ACTION:
0 '.
With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
With no loop in operation, suspend all operations involving a reduc-tion in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation.
"All reactor coolant pumps, and RHR pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
""A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 275 F unless the secondary water temperature of each steam generator is less than 50'F above each of. the Reactor Coolant System cold leg temperatures.
TURKEY POINT " UNITS 3 5 4 3/4 4-3 AMENDMENT NOS. AND FEB 2 8 lio9
]It l'
REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS 4.4. 1.3.
~ ~ ~ ~ 1 The required reactor coolant pump(s), if not in operation, shall be once per 7 days by verifying correct breaker alignments and
~
determined OPERABLE indicated power availability.
4.4. 1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10K at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4. 1.3.3 At least one reactor coolant or RHR loop shall be verifie'd in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
0 TURKEY POINT " UNITS 3 & 4 3/4 4-4 AMENDMENT NOS. AND FEB 2 8 1989
't I
jcg.
))t 4'.,
REACTOR COOLANT SYSTEM COLD SHUTDOWN
- LOOPS FILLED IMITING CONDITION FOR OPERATION 3.4. 1.4. 1 At least one residual heat removal (RHR) loop shall be OPERABLE and operation", and either:
IA
a. One additional RHR loop shall be OPERABLE"", or'.
The secondary side water level of at least two steam generators shall be greater than 10K.
APPLICABILITY: MODE 5 with reactor coolant loops filled""".
ACTION:
'a 0 With one of the RHR loops inoperable or with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.
b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE RE UIREMENTS
4. 4. 1. 4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4. 1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
"The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
""One RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing t
provided the other RHR loop is OPERABLE.
"""A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 2750F unless the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.
TURKEY POINT - UNITS 3 8 4 3/4 4"5 AMENDMENT NOS. AND fjg Cv FEB 2 8 1988
P J t
REACTOR COOLANT SYSTEM COLO SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4. 1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE" and at least one RHR loop shall be in operation.""
APPLICABILITY: MODE 5 with reactor coolant loops not filled.
ACTION:
a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
t SURVEILLANCE RE UIREMENTS 4.4. 1.4.2 At least one RHR loop shall be determined to circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
be in operation and
One RHR loop may be inoperable for up to 2 hours .for surveillance testing provided the other RHR loop is OPERABLE.
""The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.
TURKEY POINT - UNITS 3 8( 4 3/4 4-6 AMENDMENT NOS. AND NAY 0 5 1989
ys h
'l $ =
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES 1 .<UTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE" with a lift setting of 2485 psig t 1X.*"
APPLICABILITY: MODES 4 and 5.
ACTION:
With no pressurizer Code safety valve OPERABLE, immediately suspend all opera-tions involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.
4.4.2. 1 No additional requirements other than those required by Specification 4.0.5.
While in MODE 5, an equivalent size vent pathway may be used provided that the vent pathway is not isolated or sealed.
The lift setting pressure shall correspond to ambient conditions of the valve e at nominal operating temperature and pressure.
TURKEY POINT UNITS 3 8E 4 3/4 4-7 AMENDMENT NOS. AND FEB R 8 1989
lt t ~
Qt V,
P
REACTOR COOLANT SYSTEM OPERATING IMITING CONDITION FOR OPERATION 3.4 2.2
~ All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig + lX."
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
lith one pressurizer Code safety valve inoperable, either restore the inoper-able valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY:
within 6 hours and in at least HOT SHUTDOMN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS t 4.4.2.2 No additional requirements other than those required by pecification 4.0.5.
"The at lift setting
~
~
pressure shall correspond to ambient conditions of the valve
~
nominal operating temperature and pressure.
TURKEY POINT - UNITS 3 8( 4 3/4 4-8 AMENDMENT NOS. AND fF.B 2 8 1989
A, I'f
REACTOR COOLANT SYSTEM 3/4. 4. 3
~ ~ PRESSURIZER LIMITING CONDITION FOR OPERATION 4
3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 92K of indicated level, and at least two groups of pressurizer heaters each having a capacity of at least 125 kW and capable of being supplied by emergency power.
APPLICABILITY: MODES 1, 2, and 3.
<CTIQN:
'a 4 With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the follow-ing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the'following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS 4.4.3. 1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4.3. 2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at least once per 92 days.
4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at'east once per 18 months by demonstrating the capabil-ity to power the heaters from the emergency power.
TURKEY POINT - UNITS 3 8( 4 3/4 4"9 AMENDMENT NOS. AND gpss 28 1':SB
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Ar 4
REACTOR COOLANT SYSTEM 3/4. 4. 4 RELIEF VALVES IMITING CONDITION FOR OPERATION 3.4.4 Each power-operated relief valve (P RV ) b ock valve shall 'be OPERABLE.
V APPLICABILITY: MODES 1, 2 and 3.
ACTION:
'a 0 With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve(s) to OPERABLE status, or close the block valve(s) and remove power from the block valve(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. The provisions of Specification 3.0.4 are not applicable.
URVEILLANCE RE UIREMENTS 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements. of Specification 3.4.4 or is closed to provide an isolation function.
TURKEY POINT " UNITS 3 8 4 3/4 4-10 AMENDMENT NOS. AND FEB 2 8 1989
IP g,
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS IMITING CONDITION FOR OPERATION 3.4. 5 Each steam generator shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more steam generators inoperablq, restore the inoperable genera-
.or(s) to OPERABLE status prior to increasing T above 200'F.
SURVEILLANCE RE UIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
4.4.5. 1 Steam Generator Sam le Selection and Ins ection -.Each steam, generator vshall be east the 4.4.5.2 determined minimum 0
number Steam Generator A L of during steam Tube Sam shutdown by se ecting and inspecting at generators specified in Table 4.4-1.
le Selection and Ins ection - The steam generator tube m>nsmum samp e size, snspectson resu t c ass> ication, and the corresponding action required shall be 'as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3X of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50K of the tubes inspected shall be from these critical areas;
b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
TURKEY POINT - UNITS 3 8c 4 3/4 4-11 AMENDMENT NOS. AND FEB 28 1989
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REACTOR COOLANT SYSTEM STEAM GENERATORS URVEILLANCE RE UIREMENTS (Continued)
1) All nonplugged tubes that previously had detectable wall pene-trations (greater than 20K),
2) Tubes in those areas where experience has indicated potential problems, and
3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
C. The tubes selected as the second and third samples in the inservice inspection may be less than a full tube inspection by concentrating (selecting at least 50K of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Ins ection Results C-1 Less than 5X of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1X of the total tubes inspected are defective, or between 5X and 10K of the total tubes inspected are degraded tubes.
C-3 More than 10K of the total tubes inspected 'are degraded tubes or more than 1X of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10K) further wall penetrations to be included in the above percentage calculations.
TURKEY POINT " UNITS 3 8 4 3/4 4-12 AMENOMENT NOS. AND FEg 28 l989
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REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE RE UIREMENTS Continued 4.4.5.3 Ins ection Fre uencies - The above required inservice i,nspections of steam generator tubes sha e performed at the following frequencies:
a0 The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months following replace-ment of steam generators. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
b. If the results of the inservice inspection of a steam generator con-ducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency sliall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specifica-tion 4.4.5.3a; the interval may then be extended to a maximum of once per 40 months; and C. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4"2 during the shutdown subsequent to any of the following conditions:
Primary-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or
2) A seismic occurrence greater than the Operating 8asis Earthquake, or
3) A loss-of-coolant accident resulting in rapid depressurization of the primary system, or
4) A main steam line or feedwater line break resulting in rapid depressurization of the affected steam generator.
TURKEY POINT - UNITS 3 8 4 3/4 4-13 AMENOMENT NOS. AND
'I FEB 2 8 1989
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REACTOR COOLANT SYSTEM STEAM GENERATOR SURVEILLANCE RE UIREMENTS (Continued) 4.4.5.4 Acce tance Criteria a~ As used in this specification:
1) Im erfection means an exception to the dimensions, finish or contour o a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20K of the nominal tube wall thickness, if detectable, may be con-sidered as imperfections;
2) Oe radation means a service-induced cracking, wastage, wear or genera corrosion occurring on either inside or outside of a tube;
3) Oe raded Tube means a tube containing imperfections greater .
t an or equa to 20K of the nominal wall thickness caused by degradation;
4) X Oe radation means the percentage of the tube wall thickness a ected or removed by degradation;
5) Oefect means an imperfection of such severity that it exceeds tte pTugging iimit. A tube containing a defect is defective.
6) Plu in Limit means the imperfection depth at or beyond which t e tu e s a be removed from service because it may become unserviceable prior to the next inspection and is equal to 40K of the nominal tube wall thickness;
7) Unserviceable describes the condition of ff ti a tube if it leaks t t 1i or f t1 g gh g rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above;
8) Tube Ins ection means an inspection of the steam generator tube rom t e point of entry (hot leg side) completely around the U-bend to the top support of the cold leg, or from the point of entry (cold leg side) completely around the U-bend and to the bottom of the hot leg; and TURKEY POINT - UNITS 3 5 4 3/4 4-14 AMENOMENT NOS. AND FFB 2 8 $ 989
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REACTOR COOLANT SYSTEM STEAM GENERATOR SURVEILLANCE RE UIREMENTS Continued
9) Preservice Ins ection means an inspection of the .full length of eac tube 1n each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 ~Re orts
'a 4 Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1) Number and extent of tubes inspected,
2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
3) Identification of tubes plugged.
C. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to 10 CFR Part 50.72 and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
TURKEY POINT - UNITS 3 8 4 3/4 '4-15 AMENDMENT NOS. AND FEB 2 8 1989
4';e TABLE 4.4-'1 MINIMUM NUMBER OF STEAM GENERATORS TO BE N P T OU NG S V C N C ON Preservice Inspection No of Unit No. Steam Generators First Inservice Inspection per One'es Three Al 1 Three Two Second 8 Subsequent Inservice Inspections One~
Table Notation The inservice inspection may be limited to one steam generator on a rotat-ing schedule encompassing 9X of the tubes if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating condi-tions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
I
2. The other steam generator,not inspected during the first inservice inspec-tion shall be inspected. The third and subsequent inspections should follow the instruction described in 1 above.
TURKEY POINT - UNITS 3 8 4 3/4 4-16 AMENDMENT NOS. ANO NAY 6 5 i."83
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STEAM GENE TUBE INSPECTION SAMPLE 1st SAMPLE INSPECTION 2nd SAMPLE INSPECTION 3rd SAMPLE INSPECTION SIZE o Result Action Required Result Action Required Result Action Required CD A minimum C-1 N/A N/A N/A None N/A of S Tubes per S.G. C-2 Plug defective tubes C-1 None N/A N/A and inspect addi-tional 2S tubes in C-2 Plug defective tubes C-1 None this S.G. inspect additional 4S tubes in this S.G. C-2 Plug defective tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for N/A N/A C-3 result of first sample C-3 Inspect all tubes All other None N/A N/A in this S.G. plug S.G.s are defective tubes and C-1 inspect 2S tubes in each other S.G. Some S.G.s Perform action for N/A N/A C-2 but no C-2 result of second Notification to NRC additional sample pursuant to Sec- S.G.s are tion 4.4.5.5c. C-3 CD tA Additional Inspect all tubes in N/A N/A S.G. is each S.G. and plug C-3 defective tubes.
Notification to NRC pursuant to Section 4.4.5.5C.
S = -X Where n is the number of steam generators inspected during an inspection.
n
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION .
3.4.6. 1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
a. The Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and
b. A Containment Sump Level Monitoring System.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a. With both the Particulate and Gaseous Radioactivity Monitoring Systems inoperable, operation may'continue for up to 7 days provided:
1} A Containment Sump Level Monitoring System is OPERABLE;
2) Appropriate grab samples are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;
3) A Reactor Coolant System water inventory balance is performed at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during steady state operation except when operating in shutdown cooling mode; and
4) Containment Purge, Exhaust and Instrument Air Bleed valves are maintained closed.
Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With no Containment Sump Level Monitoring System operable, restore at least one Containment Sump Level Monitoring System to OPERABLE status within 7 days, or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.4.6. 1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
a. Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
b. ~
~
Containment Sump Level Monitoring System-performance of CHANNEL CALIBRATION at least once per 18 months.
TURKEY POINT - UNITS 3 8 4 3/4 4-18 AMENDMENT NOS. AND FEB 28 1988
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION
3. 4. 6. 2 Reactor Coolant System leakage shall be limited to:
a. No PRESSURE BOUNDARY LEAKAGE,
b. 1 GPM UNIDENTIFIED LEAKAGE,
c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,
d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
e. Leakage as specified in Table 3.4-1 up to a maximum of 5 GPM at a Reactor Coolant System pressure of 2235 a 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1."
APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With any Reactor Coolant System leakage greater than any .one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
C. With any Reactor Coolant System Pressure Isolation Valve leakage greater than allowed by 3.4.6.2.e above operation may continue provided:
1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months verify that at least two valves in each high pressure line having a non-functional valve are in:, and remain in that mode corresponding to the isolated condition, i.e.,
manual valves shall be locked in the closed position; motor operated valves shall'e placed in the closed position and power supplies deenergized. Follow applicable ACTION statement for the affected system, and "Test pressures less than 2235 psig are allowed. Minimum differential test pressure shall not be less than 150 psid. Observed leakage shall be adjusted for the, actual test pressure up to 2235 psig assuming .the leakage to be directly proportional to pressure differential to the one-half power.
~ ~
TURKEY POINT - UNITS 3 8E 4 3/4 4-19 AMENDMENT NOS. AND FF.B 2 8 1989
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE IMITING CONDITION FOR OPERATION (Continued
2. The leakage" from the remaining isolation valves in each high pressure line having a valve not meeting the criteria of Table 3.4-1, as listed in Table 3.4-1, shall be determined and recorded daily. The positions of the other valves located in the high pressure line having the leaking valve shall be recorded daily unless they are manual valves located inside containment.
Otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
d. With any Reactor Coolant System Pressure Isolation Valve leakage greater than 5 gpm, reduce leakage to below 5 gpm within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
a. Monitoring the containment atmosphere gaseous or particulate radio-activi'ty monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Monitoring the containment sump level at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Performance of a Reactor Coolant System water inventory balance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving steady-state operation"" and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during steady-state operation, except that not more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months shall elapse between any two successive inventory balances; and
d. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours.
4.4.6 '.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4"1 shall be demonstrated OPERABLE by verifying leakage" to be within its limit:
At least once per 18 months.
b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, and C. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.
o satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage crite>ia.
"RCS average coolant temperature being changed by less than 5 F/hour.
TURKEY POINT - UNITS 3 8c 4 3/4 4-20 AMENDMENT NOS. AND PPB 28 1989
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE
.IMITING CONDITION FOR OPERATION Continued)
d. Following valve actuation due to automatic or manual action or flow through the valve:
1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and
2. Prior to entering Mode 2 by verifying leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
TURKEY POINT - UNITS 3 8 4 3/4 4-21 AMENDMENT NOS. AND gpss R 8 1989
TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION Unit 3 Unit 4 High-Head Safety Injection Check Valves 3-874A 4-874A Loop A, hot 3-875A 4-875A leg leg'old 3-873A 4-873A cold leg 3-874B 4-8748 Loop B, hot leg 3-875B 4"875B cold leg 3-8738 4-873B cold leg 3" 875C 4-875C Loop C, cold leg 3"873C 4"873C cold leg Residual Heat Removal Line Check Valves 3-876-A 4-876A Loop A,'cold leg 4-876E 3-876B 4-876B Loop B, cold leg 3-876D 4"8760 3-876C 4-876C Loop C, col d 1 eg 3-876E MOV3-750 MOV4"750 Loop A, hot leg to RHR MOV3"751 MOV4-751 Loop A, hot leg to RHR ACCEPTABLE LEAKAGE LIMITS Leakage rates less than or equal to 1.0 gpm are considered acceptable.
2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable provided that the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between previously measured leakage rate and the maximum permis-sible rate of 5.0 gpm by SO% or greater.
3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between previously measured leakage rate and the maximum permissible rate of 5.0 gpm by 50K or greater.
L'eakage rates greater than 5. 0 gpm are considered unacceptable.
0 TURKEY POINT " UNITS 3 & 4 3/4 4-22 AMENOMENT NOS. AND pp9 2 8 1989
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REACTOR COOLANT SYSTEM 3/4. 4. 7
~ ~ CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained 'within the limits specified in Table 3.4-2.
APPLICABILITY: At al 1 times.
ACTION:
BODES 1, 2, 3 and 4:
a/ With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and
b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within-6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
/
At All'Other Times:
With the conc entration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to pro-ceeding to MODE 4.
SURVEILLANCE RE UIREMENTS 4.'.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.
TURKEY POINT - UNITS 3 8 4 3/4 4-23 AMENDMENT NOS.'ND FEB 2 8 1989
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TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen" < 0.10 ppm < 1.00 ppm Chloride"" < 0.15 ppm < 1.50 ppm Fluoride** < 0.15 ppm < 1.50 ppm Lsmst not applicable with average reactor coolant temperature less than or equal to 250'F.
Not required when reactor is defueled and RCS forced circulation is u'navai lable.
TURKEY POINT - UNITS 3 8( 4 3/4 4-24 AMENDMENT NOS. AND NAY Oa i.
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TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE RE UIREMENTS PARAMETER SAMPLE AND ANALYSIS FRE UENCY Dissolved Oxygen" 't least 5 times per week not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples Chloride"" At least 5 times per week not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples Fluoride** At least 5 times per week not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples Not required w>th average reactor coolant temperature less than or equal to 250 F.
0 *"Not required unavailable.
when reactor is defueled and RCS forced circulation is TURKEY POINT - UNITS 3 8( 4 3/4 4-25 AMENDMENT NOS. AND
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REACTOR COOLANT SYSTEM 3/4.4. 8 SPECIFIC ACTIVITY IMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:
a. Less than or equal to 1.0 microcurie per gram DOSE E(UIVALENT I-131, and
b. Less than or equal to 100/E microcuries per gram of gross radioactivity.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.'CTION:
4 5QPES 1, 2 and 3":
a~ With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE E(UIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with average reactor coolant temperature less than 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and
b. With the specific activity of the reactor coolant greater than 100/E microcurie per gram, be in at least HOT STANDBY with average reactor coolant temperature less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
MODES 1, 2, 3, 4, and 5:
With the specific activity of the reactor coolant greater than 1 micro-curie per gram DOSE E(UIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements. of Item 6.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.
SURVEILLANCE RE UIREMENTS 4,4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4. 4-4. \
"With the average reactor coolant temperature greater than or equal to 500'F.
TURKEY POINT - UNITS 3 4 4 3/4 4-26 AMENDMENT NOS. AND
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0 40 IO 00 '1 tEIlEMJ OF RATES FIGURE 3.4-1 t DOSE E(UIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSU PERCENT OF RATED THERMAL POMER MITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 pCi/gram DOSE E(UIVALENT I-131.
TURKEY POINT - UNITS 3 8 4 3/4 4-27 AMENDMENT NOS. AND FEB 28 I89
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> RATED THERMAL POWER t DOSE EQUIVALENT TURKEY POINT UNITS I 131 FIGURE 3.4 1 (WITH THE REACTOR COOLANT SPECIFIC 3 0 4 ACTMlY ~
Rx COOLANT SPECIFIC ACTIVITY LIMIT vs x THERMAL POWER 1.0uCi/gran 00SE EQUIVALENT 1-131.)
REACTOR COOLANT e I SPECIFIC ACTIVITY SAHPLE AND ANALYSIS PROGRAH TYPE OF HEASUREHENT SAHPLE AND ANALYSIS HODES IN WHICH SAHPLE ANY ANALYSIS FRE UENCY AND ANALYSIS RE UIRED
1. Gross Radioactivity At least once per 1, 2, 3, 4 Determination 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
2. Tritium Activity 1 per 7 days. 1, 2, 3, 4 Determination
3. Isotopic Analysis for 1 per 14 days.
DOSE EQUIVALENT I-131 Concentration Radiochemical Isotopic Honthly 1, 2, 3, 4 Determination Including Gaseous Activity I
CO
5. Radiochemical for E 1 per 6 months*
Determination
6. Isotopic Analysis for a) Once per 4 hours, 1¹, 2¹, 3¹, 4¹, 58 Iodine Including I-131, whenever the I-133, and I-135 specific activity exceeds 1 pCi/gram DOSE E(UIVALENT m I-131 or 100/E C3 pCi/gram of gross m radioactivity, and b) One sample between 1 2 3 C)
CA 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months follow-ing a THERMAL POWER change exceeding 15K of the RATED THERHAL CD POWER within a 1-hour period.
Table 4.4-4 (Continued TABLE NOTATIONS
Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.
¹Until the specific activity of the Reactor Coolant System is restored within its limits.
TURKEY POINT - UNITS 3'8L 4 3/4 4-29 AMENDMENT NOS. ANO Fpg 2 8 1989
P4
I i@'
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION j
3.4.9. 1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 and 3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
a. A maximum heatup of 100'F in any 1-hour period,
b. A maximum cooldown of 100'F in any 1-hour period, and
c. A maximum temperature change of less than or equal to 5 F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves.
APPLICABILITY: At al 1 times.
ACTION:
Mith any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next and pressure to less than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.4.9. 1. 1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4. 9. 1. 2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.
TURKEY POINT - UNITS 3 8( 4 3/4 4-30 AMENDMENT NOS. AND NIA'( 0 5 ]988
1
$4 t)h h
gf, 4
F y 1
PROPERTY SASS C TROLLING MATERIAL:CIRCUMFERENTJAI, WELD l RTg)T:I0"F I
/ ~
200.A'ATERIAL ERIOlh 20 EFPY = 2525.F SERVIC IITHDT 0'/<THICKNESS
=
HEATUP R ES:UP TO 60"F/HR RTNDT 3/4 THICKNESS 2250 ST LQIT l750 CRITlCALITY iQrr eASEO ON SSERVKE HYDRO-Pae STATK TEST NtA TABLE ERATSN TEMPERA TINE 080"F) FOR g usa THE SERVICE PERI00 UP TO 20 EFPY) g ew ACCEPTABLE 750 OPERATION IP RATES 9 70 0"
250 AC PT E
,0
/
0 50 m0 50 200 250 300 %0 MATEO TEMPERATNE SELR FIGURE 3.4"2 TURKEY POINT UNITS 3 8E 4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (60'F/hr) - APPLICAB UP TO 20 EFPY TURKEY POINT - UNITS 3 8( 4 3/4 4-31 AMENDMENT NOS. AND FEB 28 1989
4
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MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD INITIAL RTNDT lo'F SERVICE PERIOD: 20 EFPY RT NDT I/4 THICKNESS 252.5 F HEATUP RATES: UP TO 60 F/HR RT NDT
~ 3/4 THICKNESS = 200.4 F NOTE: NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS.
2500 LEAK TEST LIMIT 2250 1750 C9 CRITICALI7(
w 15OO LIMIT (BASED ON INSERVICE HYDRO CCEPTAB OP ATION STATIC TEST w 1250 T MPE ATURE 380'F FOR CD E S RVICE PERIOD UP TO 1000 20 EFPY)
CD z 750 P RATES U ACCEPTABLE 60 oF HR OPERATION ACCEPTABLE OPERATION 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE ('F)
FIGURE 3.4 2 REACTOR COOLANT HEATUP LIMITATIONS ((60'F/HRj TURKEY POINT UNITS 3 8c 4
t V
P '"
fJ sL 4
I I
1p
~k Yg S
NTERIAL PROPERTY BASIS CO OLUNG QATERIAI 'ICUMFERENTIAL IELO INIT RTg)T. Kf'F SERVKE P OD:20 EFPY RTH0T I'/<T CKHESS
= 2529'F HEATUP RATE UP TO $0 F/HR RTg)T 14 THICKNESS = 200.A TEST L CRITICALITY LNIT SASEO ON SERVICE HYDRO-STATIC TEST Pam TEMPERATE IWA TABLE TSN 680"F) FOR TfK SERVICE g @50 PERIOD tF TO 20 EFPO N00 g ACCEPTABLE RATION 750 AT TO ACCEPTABLE 50 $0 50 200 250 300 350 400 450 MEATEO TBFERATNE SELR FIGURE 3.4-3 pg TURKEY POINT UNITS 3 & 4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (100'F/hr) - APPLICAB POX'URKEY UP TO 20 EFPY POINT " UNITS 3 & 4 3/4 4-32 AMENDMENT NOS. AND FEB 2 8 1S89
I j
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MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD INITIAL RT NDT
'0'F SERVICE PERIOD: 20 EFPY RT NPT
~ 1/4 THICKNESS = 252.5 F HEATUP RATES: UP TO 100 F/HR RT NOT
~ 3/4 THICKNESS 200.4 F NOTE: NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS.
25X il LEAK TEST LIMIT 2250 CRITICALI71'IMIT (BASED ON 1750 INSERVICE HYDRO-STATIC TEST TEMPERATURE w 1500 UNAC TAB (380'F) FOR ON THE SERVICE wN 1250 PERIOD UP TO 20 EFPY Ch ACCEPTABLE OPERATION HEATUP RATES UP TO 100'F HR ACCEPTABLE OPERATION 0 50 100 150 200 250 XO 350 400 450 500 INDICATED TEMPERATURE ('F)
FIGURE 34 3 REACTOR COOLANT HEATUP LIMITATIONS (~100 FlHR)
TURKEY POINT UNITS 3 h 4
N S
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A" N
K P
41 I&
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MATERlAL PROPERTY SASS e ROLUNG MATERlAL!ClRCUMFEREttTIAL RTg)T cMf F IELO SERYIC COOLDOW ERIOlh20 EFPY ATES: O'O $ 0
'T~~
f/HR . RTg)T el/TRlCRRESS =
03/4 THlCK S =
P.f P 200A"F 2250 2000 l750 lNACCEPTABL t250 CCEPTABLE OPERATlON 00 750 RATES S0 0 50 200 250 . 300 350 e0 MEATED TSPERA7lK SECB FIGURE 3.4-4 /e-s mB TURKEY POINT UNITS 3 8 4 p+)
REACTOR COOLANT SYSTEM COOLDOMN LIMITATIONS (100 F/hr) - APPL LE UP TO 20 EFPY TURKEY POINT - UNITS 3 8( 4 3/4 4-33 AMENDMENT NOS. AND FEB R 8 19S9
4 Ci "IP>>
tg g 0 ~ 'I X+
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MATERIAL PROPERTY BASIS INITIAL RTNDT '0 CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD F RTNDT 1/4 THICKNESS = 252.5 F SERVICE PERIOD: 20 EFPY RTNDT 3/4 THICKNESS 200.4~F COOLDOWN RATES: UP TO 100'F/HR NOTE: NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS.
2250 1750 Cf)
CL LJ 1500 C/l UNACCEPTABLE UJ OPERATION 1000 COOL DOWN
'F ACCEPTABLE HR OPERATION 0
20 40 60 100 0
0 50 100 150 200 250, 300 ZO 400 450 500 INDICATED TEMPERATURE ('F)
FIGURE 3.4 4 REACTOR COOLANT COOLDOWN LIMITATIONS (0 100'F/HR)
TURKEY POINT UNITS 3 8c 4,,
TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM
- WITHDRAWAL SCHEDULE UNIT 3 CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME
0. 49 Standby 30'0 Specimen withdrawn at 12 years
0. 34 Standby 500 0.34 33 years 150 0.49 Standby 230 0. 34 Standby UNIT 4 CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME 30 0. 49 Standby 290 0. 79 24 years 40 0. 34 Standby 50o 0. 34 Standby 150 0. 49 Standby 230 0. 34 Standby TURKEY POINT - UNITS 3 8c 4 3/4 4-34 AMENDMENT NOS. AND NAY 05 Isa'
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REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION
3. 4. 9. 2 The pressurizer temperature shall be limited to:
a. A maximum heatup of 100'F in any 1-hour period,
b. A maximum cooldown of 200'F in any 1-hour period, and
c. A maximum spray water temperature differential of 320'F.
APPLICABILITY: At all times.
ACTION:
With the pressurizer temperature limits in excess of any. of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig with-in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4,4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.
TURKEY POINT " UNITS 3 8 4 3/4 4-35 AMENDMENT NOS. AND
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REACTOR COOLANT SYSTEM OVERPRESSURE MITIGATING SYSTEMS LIMITING CONDITION FOR OPERATION 3 ~ 4.9.3 The high pressure safety injection flow paths to the Reactor Coolant System (RCS) shall be isolated, and below an RCS average coolant temperature of 275'F at least one of the following Overpressure Hitigating Systems shall be OPERABLE:
a) Two power-operated t 15 relief valves (PORVs) with a lift setting of 415 psig, or b) The RCS depressurized with a RCS vent of greater than or equal to 2.20 square inches.
APPLICABILITY: MODES 4, 5 and 6 with the reactor vessel head on.
AC-TION:
ao With the high pressure safety injection flow paths to the RCS unisolated, restore isolation of these flow paths within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and
b. In HODE 4 with RCS average coolant temperature less than or equal to 275 F, and in MODE 5 or in Mode 6 with the reactor vessel head on:
1. With one PORV inoperable, perform at least one of the following within the next 7 days:
a) Restore the inoperable PORV to OPERABLE status, or b) Depressurize and vent the RCS through at least a 2.20 square inc", or M8nw c) Depressurize and maintain a RCS vent through at least one open PORV and open associated block valve.
2. With both PORVs'inoperable, depressurize and vent the RCS through at least a 2.20 square inch vent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
3. In the event either the PORVs or a 2.20 square inch vent is used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or RCS vent(s) on the transient, and any corrective action necessary to prevent recurrence.
0 TURKEY POINT - UNITS 3 8 4 w
3/4 4-36 AMENDMENT NOS. AND FEB 28 1g%
'I J
g
REACTOR COOLANT SYSTEM
. OVERPRESSURE MITIGATING SYSTEMS SURVEILLANCE RE UIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
a o Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actua-tion channel, but excluding valve operation, within 31 days prior to enteri'ng a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and
c. Verifying the PORV block valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used, for overpr essure protection.
d. While the PORVs are required to be OPERABLE, the backup air supply shall be verified OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
4.4.9.3.2 The 2.20 square inch vent shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months " when the vent(s) is being used for overpressure protection. F 4.4.9.3.3 Verify the high pressure injection flow path to the RCS .is isolated at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by closed valves with power removed or by locked closed manual valves.
"Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
0 TURKEY POINT - UNITS 3 & 4 3/4 4-37 AMENDMENT NOS. AND
cI
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REACTOR COOLANT SYSTEM 3/4.4.~ ~ 10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION
3. 4. 10 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4. 10.
APPLICABILITY: Al 1 MODES
'CTION:
a ~ With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.
b. With the str'uctural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant 0 ~
System temperature above 200'F.
With the structural integrity of any ASRE Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.
SURVEILLANCE RE UIREMENTS
~
.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1. 14, Revision 1, August 1975.
TURKEY POINT - UNITS 3 4 4 3/4 4-38 AMENDMENT NOS. AND r
FEB R8 $ 8s
~4 "g ~
~F J
+iw I
REACTOR COOLANT SYSTEM 3/4.4. 11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolant System vent path consisting of at least two vent valves in series and powered from emergency busses shall be OPERABLE and closed at each of the following locations:
a. Reactor vessel head, and
b. Pressurizer steam space APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one of'he above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the follow-ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With both Reactor Coolant System vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actua-tors of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.4. 11 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:
a. Verifying all manual isolation valves in each vent path are locked in the open position,
'I
b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and C. Verifying flow through the Reactor Coolant System vent paths during venting.
TURKEY POINT - UNITS 3 8 4 3/4 4-39 AMENDMENT NOS. AND FEB 2 8 1S88
ill (
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EMERGENCY CORE COOLING SYSTEMS 3/4. 5. 1 ACCUMULATORS IMITING CONDITION FOR OPERATION 3.5. 1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:
'a 0 The isolation valve open and its circuit breaker open, t
b. An indicated borated water volume of between 6545 and 6665 gallons, c ~ A boron concentration of between 1950 and 2350 ppm,
d. A nitrogen cover-pressure of between 600 and 675 psig, and
e. A water level and pressure channel OPERABLE.
APPLICABILITY: MODES 1, 2, and 3".
RCTION:
'a 0 With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE t
status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to. less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS 4.5.1. 1 Each accumulator shall be demonstrated OPERABLE:
At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:
1) Verifying the indicated borated water volume and nitrogen cover-pressure in the tanks, and
2) Verifying that each accumulator isolation valve is open by control room indication (power may be restored to the valve operator to perform this surveillance i.f redundant indicator is inoperable).
"Pressurizer pressure above 1000 psig.
TURKEY POINT - UNITS 3 8 4 3/4 5"1 AMENDMENT NOS. AND FEB 28 f989
0 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued
b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1X of tank volume by verifying the boron concentration of the solution in the water-filled accumulator; C. At least once per 31 days:
When the RCS pressure is above 1000 psig, by verifying that the power to the isolation valve operator is disconnected by a locked open breaker, and
2) Each accumulator water level and pressure channel shall be demonstrated OPERABLE by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and
d. At least once per 18 months:
1) Each accumulator water level and pressure channel, shall be demonstrated OPERABLE by the performance of a CHANNEL CALIBRATION, and
2) Each accumulator check valve shall be checked for operability.
0 TURKEY POINT - UNITS 3 8( 4 3/4 5-2 AMENDMENT NOS. AND Fpa 2 8 >sss
11
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EMERGENCY CORE COOLING SYSTEMS
/4.5.2 ECCS SUBS STEMS - T GREATER THAN OR EQUAL TO 350 F LIMITING CONDITION FOR OPERATION 3.5.2 The following Emergency Core Cooling System (ECCS) equipment and flow paths shall be OPERABLE:
a. Four OPERABLE Safety Injection (SI) pumps with discharge aligned to the RCS cold legs,
b. Two OPERABLE RHR heat exchangers,
c. Two OPERABLE RHR pumps with discharge aligned to the RCS cold legs,
d. An OPERABLE flow path capable of taking suction from the refueling water storage tank as defined in Specification 3.5.4, and
e. Two OPERABLE flow paths capable of taking suction from the containment sump.
APPLICABILITY: MODES 1, 2, and 3".
ACTION:
a ~ With any one of the required ECCS components or flow paths inoperable, except for inoperable Safety Injection Pump(s), restore the inoperable component or flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. In the event the ECCS is actuated and injects water in the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6. 9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date since M~~T<<r~ I l990.
C. With one Safety Injection pump inoper , res ore the pump to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
This ACTION applies to both units simultaneously.
d. With two Safety Injection Pumps inoperable, restore one of the two inoperable pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This ACTION applies to both units simultaneously.
The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the Safety Injection flow paths isolated pursuant to Specifica" tion 3.4.9.3 provided that the Safety Injection flow paths are restored to
~ ~ ~
~
exceeding 380 F. Safety Injection flow paths may
~ ~ ~
OPERABLE status prior to T ~
avg
~
be isolated when T avg is less than 380 F.
~
TURKEY POINT " UNITS 3 8i 4 3/4 5-3 AMENDMENT NOS. , AND FEB 28 108
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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS component and flow path shall be demonstrated OPERABLE:
a ~ At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying by control room indication that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position 864A and B Supply from RWST to ECCS Open 862A and B RWST Supply to RHR pumps Open 863A and B RHR Recirculation Closed 866A and B H.H.S.I. to Hot Legs Closed HCV-758" RHR HX Outlet Open To permit temporary operation of these valves for surveillance or maintenance purposes, power may be restored to these valves for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. At least once per 31 days by:
1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping,
2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, and
3) Verifying that each RHR Pump develops the indicated differential pressure applicable to the operating conditions in accordance with Figure 3. 5-1 when tested pursuant to Specification 4. 0. 5.
C. At least once per 92 days by:
1) Verifying that each SI pump develops the indicated differential pressure applicable to the operating conditions when tested pursuant to Specification 4.0.5.
SI pump > 1126 psid at a metered flowrate > 300 gpm (normal alignment or Unit 4 SI pumps aligned to Unit 3 RWST), or
> 1156 psid at a metered flowrate > 280 gpm
/Unit 3 SI pumps aligned to Unit 4 RWST).
"Air Supply to HCV-758 shall be verified shut off and sealed closed once per 31 days.
TURKEY POINT - UNITS 3 84 4 3/4 5-4 AMENDMENT NOS. AND
I P
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SELECTED POINTS ON CURVE 290 0 277 280 250'00 277 277 1000 275 1500 270 270 265 251 J000 2Q QM0 21$
260 3600 205 3750 200 E
250 g ACCEPTABLE OPERATION PJ 240 UNACCEPTABLE OPERATION 220 210 200 2000 RHR PUMP FLOII (GPM)
Figure 3. 5-1 P~j P RHR Pump Curve TURKEY POINT - UNITS 3 8 4 3/4 5-5 AMENOMENT NOS. ANO II"AY "5 IBS9
5 ta y
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SELECTED POINTS ON THE CURVE 290 FLOW 0 277 280 250 277 500 277 1000 275 1500 270 270 2000 265 2500 251 3000 233 3500 213 260 3600 208 3750 200 o 250 ACCEPTABLE OPERATION 240 UNACCEPTABLE OPERATION 220 210 1000 RHR PUMP FLOW (GPM)
FIGURE 3.5 1 RESIDUAL HEAT REMOVAL PUMP MINIMUM ACCEPTABLE PERFORMANCE CURVE TURKEY POINT UNITS 3 h 4
0
~ 4
EMERGENCY CORE COOLING SYSTEMS
d. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc. ) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and
2) Of the 'areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is t
established.
e. At least once per 18 months by:
Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that with a simulated .or actual Reactor Coolant System pressure signal greater than or equal to 525 psig the interlocks cause the valves to automatically close and prevent the valves from being opened, and.
2) Verifying correct interlock action to ensure that the RWST is isolated from the RHR System during RHR System operation and to ensure that the RHR System cannot be pressurized from the Reactor Co'olant System unless the above RWST Isolation Valves are closed.
3) A visual inspection of the containment sump and verifying that the suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc. ) show no evidence of structural distress or abnormal corrosion.
f.'t least once per 18 months, during shutdown, by:
Verifying that each automatic valve in the flow path actuates to its correct position on Safety Injection actuation test signal, and
2) that each of the following pumps start automatically
'erifying upon receipt of a Safety Injection actuation test signal:
a) Safety Injection pump, and b) RHR pump.
TURKEY POINT - UNITS 3 8( 4 3/4 5-6 AMENOMENT NOS. ANO MAY 0 6 1399
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS
g. By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
1) Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS components are required to be OPERABLE, and
2) At least once per 18 months.
RHR S stem a ve um er HCV-"-758 MOV""-872 FCV-""605
. TURKEY POINT - UNITS 3 8( 4 3/4 5-7 AMENDMENT NOS. 'ND MAY G 5 i:~~
EMERGENCY CORE COOLING SYSTEMS 3/4. 5.3 ECCS SUBSYSTEMS " T LESS THAN 350 F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, the following ECCS components and flow path shall be OPERABLE:
a. One OPERABLE RHR heat exchanger,
b. One OPERABLE RHR pump, and
c. An OPERABLE flow path capable of (1) taking suction from the refueling water storage tank upon being manually realigned and (2) transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4.
ACTION:
a. With no OPERABLE ECCS flow path from the refueling water storage tank, restore at least one ECCS flow path to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. With either the residual heat removal heat exchanger or RHR pump inoperable, restore the components to OPERABLE status or maintain the Reactor Coolant System T less than 350'F by use of alternate heat removal methods'.
In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date since" I> ($ /0, SURVEILLANCE RE UIREMENTS 4.5.3 The above ECCS components shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.
TURKEY POINT - UNITS 3 8( 4 3/4 5-8 AMENDMENT NOS. AND MAY 05 iS8Ci
I
g
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 For single Unit operation, one refueling water storge tank (RWST) shall be OPERABLE or for dual Unit operation two RWSTs shall be OPERABLE with:
a. A minimum indicated borated water volume of 320,000 gallons per RWST,
b. A minimum boron concentration of 1950 ppm of boron,
c. A minimum solution temperature of 39 F, and
d. A maximum solution temperature of 100'F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With less than the required number of RWST(s) OPERABLE, restore the tank(s) to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT.STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within th6 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS
a. At least once per 7 days by:
1) Verifying the indicated borated water volume in the tank, and E
2) Verifying the boron concentration of the water.
b. By verifying the RWST temperature is within limits whenever the outside air temperature is less than 39 F or greater than 100'F at the following frequencies:
1) Within one hour upon the outside temperature exceeding its limit for consecutive 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months , and
2) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while the outside temperature exceeds its limit.
TURKEY POINT - UNITS 3 & 4 3/4 5-9 AMENDMENT NOS. AND NAY G b .:z;.
%If 0
gt
'I
t 3/4.6 3/4. 6. 1 CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained."
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.6. 1. 1 CONTAINMENT INTEGRITY shall be demonstrated:
a ~ At least once per 31 days by verifying that all penetrations"" not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions;
b. By verifying that each containment air lock is in compliance with the requirements of Specification 3. 6. 1. 3; and C. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure not less than 50 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6. 1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L .
"Exception may be taken under Administrative Controls for opening of valves
, and airlocks necessary to perform surveillance, testing requirements and/or corrective maintenance. In addition, Specification 3.6.4 shall be complied with.
""Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
TURKEY POINT - UNITS 3 & 4 3/4 6-1 AMENDMENT NOS. AND A~>AY 06 lg89
gi
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CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6. 1.2 Containment leakage rates shall be limited to:
a. An overall integrated leakage rate of less than or equal to L ,
0,25K by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P , 49.9 psig.
b. A combined leakage rate of less than 0.60 L for all penetrations and valves subject to Type B and C tests, when pressurized to p ~
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With either the measured overall integrated containment leakage rate exceeding 0.75 L or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L , restore the overall integrated leakage rate to less than 0.75 L and the combined leakage rate for
~
11 penetrations subject to Type B and C tests to less than 0.60 L -prior to
~ ~
ncreasing the Reactor Coolant System temperature above 200'F.
~
SURVEILLANCE RE UIREMENTS 4.6. 1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:
a~ Three Type A tests (Overall Integrated Containment Leakage Rate) t shall be conducted at 40 10 month intervals during shutdown at a pressure not less than P , 49.9 psig, during each 10-year service The third test of each set shall be conducted during the a'eriod.
shutdown for the 10-year plant inservice inspection; I
TURKEY POINT " UNITS 3 8c 4 3/4 6-2 AMENDMENT NOS. AND FEB 2 8 1989
I'4 4
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CONTAINMENT SYSTEMS URVEILLANCE RE UIREMENTS Continued
b. If any'periodic Type A test fails to meet 0.75 L the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L ,
a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L at which time the above test schedule may be resumed;
c. The accuracy of each Type A test shall be verified by a supplemental test which:
1) Confirms the accuracy of the test by verifying that the supple-mental test result, L , is in accordance with the following equation:
I
- ( + ) I 0 2 where Lam is the measured Type A test leakage and L is the superimposed leak;
2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and
3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment during the supplemental. test be limited to between 0.75 L and 1.25 L ;
d. Type B and C tests shall be conducted with gas at a pressure not less than P , 49.9 psig, at intervals no greater than 24 months except tests involving:-
a'or
1) Air locks,
2) Purge supply and exhaust isolation valves, and
3) Equipment access opening which shall be tested at least once every 12,months and after each use.
e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6. 1.3;
f. Purge supply and exhaust .isolation valves seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.2, as applicable;
g. The provisions of Specification 4.0.2 are not applicable.
TURKEY POINT " UNITS 3 8 4 3/4 6"3 AMENDMENT NOS. AND gpss R8 1989 FEB E ~ ~..;.
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CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6. 1.3 Each containment air lock shall be OPERABLE with:
a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, or during the performance of containment air lock surveillance and/or testing requirements, then at least one air lock door shall be closed, and
b. An overall air lock leakage 'rate of less than or equal to 0.05 L at P , 49. 9 psig.
APPLICABILITY: MODES 1, 2, 3, and 4.
4 A%ION:
a. With one containment air lock door inoperable:
1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door- to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPEPABLE air lock door closed;
2. Operation may then continue until performance of the next required overall 'air lock leakage test provided that the 31 days; OPi4i5L~
tO bE q<<
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e a rIE.a.+
ia v~,F>e3
3. Otherwise, be in at le ws sn hours and in COLD SHUTDOWN within the following 30 hours.
b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air 1ock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
TURKEY POINT - UNITS 3 8E 4 3/4 6"4 AMENDMENT NOS. AND Ffa 8 8 1s89
e 4'
CONTAINMENT SYSTEMS JRVEILLANCE RE UIREMENTS 4.6. 1.3 Each containment air lock shall be demonstrated OPERABLE:
a 0 Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that the seals have not been damaged and have seated properly by vacuum testing the volume between the door seals in accordance with approved plant procedures.
b. By conducting overall air lock leakage tests at not less than 50 psig,'nd verifying the overall air lock leakage rate is within its limit at least once per 6 months."
C. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
0 "The provisions of Specification 4.0.2 are not applicable.
TURKEY POINT - UNITS 3 8 4 3/4 6-5 AMENDMENT NOS. AND Fpp g8 >989
0 CONTAINMENT SYSTEMS INTERNAL PRESSURE IMITING CONDITION FOR OPERATION
3. 6. 1.4 Primary containment internal pressure shall be maintain'ed between -2 and +3 psig.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: k I
.ith the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.6. 1.4 The primary containment internal pressure shal'1 be determined to be within the limits .at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
TURKEY POINT - UNITS 3 8( 4 3/4 6-6 AMENDMENT NOS. 'ND PPB t8 1989
1, t q'l
%II 1'V5
CONTAINMENT SYSTEMS IR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6. 1.5 Primary containment average air temperature shall not exceed 125'F and shall not exceed 120'F by more than 336 equivalent hours" during a calendar year.
APPLICABILITY: MODES 1, 2, 3, and 4.
CTION:
With the containment average air temperature greater than 125 F or greater than 120'F for more than 336 equivalent hours" during a calendar year, reduce the average air temperature to within the applicable limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
"URVEILLANCE RE UIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arith-metical average of the temperatures at the following locations and shall be determined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:
A roximate Location
a. O'zimuth 58 feet elevation
b. 120 Azimuth 58 feet elevation
c. 240'zimuth 58 feet elevation "Equivalent hours are determined from actual hours using the time-temperature relationships that support the environmental qualification requirements of 10 CFR 50.49.
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TURKEY POINT - UNITS 3 5 4 3/4 6-7 AMENDMENT NOS. AND FEB 28 1989
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CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4. 6. l. 6.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a~ With more than one tendon (not including exempted" tendons) with an observed lift-off force between the predicted lower limit and 90K of the predicted lower limit or with one tendon below 90K of the predicted lower limit,. restore the tendon(s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specification 6. 9. 2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With one exempted" tendon with an observed lift-off force at the accessible end below 86'f the predicted lower limit, restore the tendon to the, required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
C. With any abnormal degradation of the structural integrity other than ACTION a. at a level below the acceptance criteria of Specifications 4.6. 1.6. 1, 4.6. 1.6.2 and 4.6. 1.6.3, restore the containment to the required level of integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Specification 6. 9. 2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.6. 1.6. 1 Containment Tendons. The containment tendons'tructural integrity shall be demonstrated every fifth year from the date of the initial structural integrity test. The tendons'tructural integrity shall be demonstrated by:
"Exempted in accordance with IWL-2521. 1(a). Lift-off forces observed at the accessible end below 90K of the predicted lower limit shall be reported to the Commission for information only.
TURKEY POINT - UNITS 3 8c 4 3/4 6"8 AMENDMENT NOS. AND MAY 0 o l989
4~
K'
t CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS a.
Continued)
Determining that a random but representative sample"* of at least 12 tendons (3 dome, 4 vertical, and 5 hoop) each have an observed lift-off force within predicted limits for each. For each subsequent inspection one tendon from each group may be kept unchanged to develop a history and to correlate the observed data. If the observed lift-off force of any one tendon (not including exempted" tendons) in the original sample population lies between the predicted -lower limit and 90K of the predicted lower limit, two tendons, one on each side of this tendon should be checked for their lift-off forces. If both of these adjacent tendons are found to be within their predicted limits, all three tendons should be restored to the required level of integrity. This single deficiency may be considered unique and acceptable.
b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (dome, vertical, and hoop).
A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and determining that over the entire length of the removed wire or strand that:
1) The tendon wires or strands are free of corrosion, cracks, and damage,
2) There are no changes in the presence or physical appearance of the sheathing filler-grease, and
3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) for at least three wire or strand samples (one from each end and one at mid-length) cut from each removed wire or strand. Failure of any one of the wire or strand samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment structure.
C. Performing tendon retensioning of those tendons detensioned for inspection to their observed lift-off force with a tolerance limit of +6, -OX. During retensioning of these tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating force. If the elongation corresponding to a specific load differs by more than 5X from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages; "Exempted in accordance with IWL-2521.1(a). Lift-off forces observed at the accessible end below 90K of the predicted lower limit shall be reported to the Commission for information only.
After the process of randomly selecting tendons is performed, inaccessible tendons may be exempted in accordance with IWL-2521. 1(a). Substitute tendons shall be selected that 'are located as close 'as possible to the exempted tendons.
The accessible end of exempted tendons shall have the lift-off force measured.
TURKEY POINT - UNITS 3 5 4 3/4 6-9 AMENDMENT NOS. AND MAY 0 ~ 1889
II I ~
I g~1 4'I i4-
CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued
d. Assuring that the observed lift-off force for each bendo exce ds the minimum required lift o-ff force. Required liftjoff be calculated individually for each surveillance te don prior o the forces ha11 beginning of each surveillance, and should consider s f tors as:
1) Prestressing history;
2) Friction losses; and
3) Time-dependent losses (creep, shrinkage, relaxation), considering time elapsed from prestressing.
e. Verifying the OPERABILITY of the sheathing filler grease by:
1) Minimum grease coverage exists for the different parts of the anchorage system, and
2) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.
4. 6. 1. 6. 2 End Anchora es and Ad 'acent Concrete Surfaces. The structural integrity o t e end anchorages o a ten ons >nspecte pursuant to Specifi-cation 4.6. 1.6. 1 and the adjacent concrete surfaces shall be demonstrated by determining through visual inspection that no unacceptable levels of corrosion exist on the end anchorages and no unacceptable cracking exists in the concrete adjacent to the end anchorages. Determination of acceptance levels shall be by engineering evaluation of the areas in question. If unacceptable conditions are found, the tendons inspected during the previous surveillance shall be examined to determine whether the corrosion levels or concrete cracking have increased since the previous surveillance. Inspection of adjacent concrete surfaces shall be performed concurrently with the containment tendon surveillance (Technical Specification 4.6. 1.6. 1).
4.6. 1.6.3 Containment Surfaces. In accordance with 10 CFR 50, Appendix J.
Section V. , a visua inspect)on of the accessible interior and exterior surfaces of the containment, including the liner plate, shall be performed during the shutdown for (but prior to) each Type A containment leakage rate test (Technical Specification 4.6. 1.2. 1). The purpose of this inspection shall be to identify any evidence of structural deterioration which may affect containment structural integrity or leaktightness. The visual inspection shall be general in nature; its intent shall be to detect gross areas of widespread cracking, spalling, gouging, rust, weld degradation, or grease leakage. The visual examination may include the utilization of binoculars or other optical devices. Corrective actions taken, and recording of structural deterioration and corrective actions, shall be in accordance with 10 CFR 50, Appendix J, Section V. A. Records of previous inspections shall be reviewed to verify no apparent changes in appearance. The first inspection performed will form the baseline for future surveillances.
TURKEY POINT - UNITS 3 8t 4 3/4 6" 10 AMENDMENT NOS.
I",lAY 0 ~, 1989
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t CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION 3.6. 1.7 Each FOR OPERATION containment purge supply and exhaust isolation valve shall be OPERABLE and:
a. The containment purge supply and exhaust isolation valves shall be sealed closed to the maximum extent practicable but may be open for purge system operation for pressure control, for environmental conditions control, for ALARA and respirable air quality considera-tions for personnel entry and for surveillance tests that require the valve to be open.
b. The purge supply and exhaust isolation valves shall not be opened wider than 33 or 30 degrees, respectively (90 degrees is fully open).
APPLICABILITY: MODES 1, 2) 3, AND 4.
ACTION:
a. With a containment purge supply and/or exhaust isolation valve(s) open for reasons other than. given in 3.6. 1.7.a above, close the open valve(s) or isolate the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the
'following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With a containment purge supply'and/or exhaust isolation valve(s) having a measured leakage rate exceeding the limits of Specification 4.6. 1.7.2, restore the inoperable valve(s) to OPERABLE status or isolate the penetrations such that the'measured leakage, rate does not exceed the limits of Specification 4.6. 1.7.2 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.6. 1.7. 1 Each containment purge supply and exhaust isolation valve shall be verified to be sealed closed or open in accordance with Specification 3.6. 1.7.a at least once per 31 days.
4.6. 1.7.2 At least once per 6 months, each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 La when pressurized to P .
4.6. 1.7.3 At least once per 18 months, the mechanical stop on each containment purge supply and exhaust isolation valve shall be verified to be in place and that the valves will open no more than 33 or 30 degrees, respectively.
TURKEY POINT - UNITS 3 8( 4 3/4 6"11 AMENDMENT NOS. 'ND l~'(AY 0 o 1989
/4 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2. 1 Two i,ndependent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a. With one Containment Spray System inoperable restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,. or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With two Containment Spray Systems inoperable restore at least one Spray System to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore both Spray Systems to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initial loss or be in at least HOT STANDBY within the neRt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.6.2. 1 Each Containment Spray System shall be demonstrated OPERABLE:
a ~ At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position and that power is available to flow path components that require power for operation;
b. By verifying that on recirculation flow, each pump develops the indicated differential pressure, when tested pursuant to Specification 4.0.5:
Containment Spray Pump 2241.6 psid while aligned in recirculation mode.
TURKEY POINT - UNITS 3 8,4 3/4 6-'12 AMENDMENT NOS. AND NAY os >989
A CONTAINMENT SYSTEMS URVEILLANCE RE UI REMENTS Continued C. At least once per 18 months during shutdown by:
1) Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation test signal, and
2) Verifying that each spray pump starts automatically on a containment spray actuation test signal. The manual isolation valves in the spray lines at the containment shall be locked closed for the performance of these tests.
d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
TURKEY POINT - UNITS 3 & 4 3/4 6" 13 AMENOMENT NOS. ANO PgB E8 1980
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CONTAINMENT SYSTEMS EMERGENCY CONTAINMENT COOLING SYSTEM IMITING CONDITION FOR OPERATION
3. 6. 2. 2 Three emergency containment cooling units shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
'a 0 -With one of the above required emergency containment cooling units inoperable restore the inoperable cooling unit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b. With two or more of the above required emergency containment cooling units inoperable, restore at least two cooling units to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore all of the above required cooling units to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS
a. At least once per 31 days by:
1) Starting each cooler unit from the control room and verifying that each unit motor reaches the nominal operating current for the test conditions and operates for at least 15 minutes, and
2) Verifying a cooling water flow rate of greater than or equal to 2000 gpm to each cooler.
b. At least once per 18 months by verifying that each unit starts automatically on a safety injection (SI) test signal.
TURKEY POINT - UNITS 3 & 4 3/4, 6-14 AMENDMENT NOS. AND ping g8 1989
0 V
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CONTAINMENT SYSTEMS
~4.6.3 EMERGENCY CONTAINMENT FILTERING SYSTEM LIMITING CONDITION FOR OPERATION
3. 6. 3 Three emergency containment filtering units shall be OPER'ABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACT10N:
ith one emergency containment filtering unit inoperable, restore the inoperable
.i lter to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS
4. 6.3 Each emergency containment filtering unit shall be demonstrated OPERABLE:
a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal 0, adsorbers and verifying that the system operates for at least 15 minutes; At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release or (3) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by:
1) Performance of a visual inspection for foreign material and gasket deterioration, and verifying that the filtering unit satisfies the in-place penetration and bypass leakage testing criteria of greater than or equal to 99K removal of 'cceptance
. DOP and halogenated hydrocarbons at the system flow rate of 37,500 cfm 110K;
2) Verifying within 31 days after removal, that a laboratory analy-sis of.a representative carbon sample obtained in accordance with applicable portions of Regulatory Position C.6.b of Regula-tory Guide 1.52, Revision 2, March 1978, and performed in accordance with ANSI N-510-1975, meets the acceptance criteria of greater than 99.9X removal of elemental iodine; and that any charcoal failing to meet this criteria be replaced with charcoal that meets or exceeds the criteria of position C. 6a of Regulatory Guide 1.52, Rev. 2; and
3) Verifying a system flow rate of 37,500 cfm 110K and a pressure drop across the HEPA and charcoal filters of less than 6 inches water gauge during system operation when tested in accordance with ANSI N510-1975 TURKEY POINT - UNITS 3 & 4 3/4 6-15 AMENDMENT NOS. AND Fp,g 28 1s80
4 t'i Ej V
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CONTAINMENT SYSTEMS URVEILLANCE RE UIREMENTS Continued
c. After maintenance affecting flow distribution, by performance of a visual inspection and an air distribution. test at a system flow rate of 37,500 cfm ilOX;
d. At least once per 18 months by:
Verifying that the system starts on a Safety Injection test signal and;
2) Verifying that the filter cooling solenoid valves can be opened by operator action and are opened automatically on a loss of flow signal.
e. After each complete or partial replacement of a HEPA filter bank, by performance of a visual inspection for foreign material and gasket deterioration and by verifying that the filtering unit satisfies the
'f in-place penetration and bypass leakage testing acceptance criteria greater than or equal to 99K removal of OOP test aerosol while operating the system at a flow rate of 37,500 cfm tlOX; and
f. After each complete or partial replacement of a charcoal adsorber bank, by performance of a visual inspection for foreign material and gasket deterioration and by verifying that the filtering unit satisfies the in-place penetration and bypass leakage testing acceptance criteria of greater than or equal to 99K removal of halogenated hydrocarbon while operating the system at a flow rate of 37,500 cfm +10%.
TURKEY POINT - UNITS 3 8 4 3/4 6-16 AMENDMENT NOS. AND ff9 58 1S89
J
CONTAINMENT SYSTEMS 3/4. 6. 4
~ ~ CONTAINMENT ISOLATION VALVES LIMITING CONDITION -FOR OPERATION 3.6.4 Each containment isolation valve shall be OPERABLE with isolation times less than or equal to required isolation times.
APPLICABILITY: MODES 1, 2, 3, and 4.-
ACTION:
"Mith one or more isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
I
a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or
b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic containment isolation valve secured in the isolation position, or C. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange, or 0 Be in at SHUTDO'W least HOT STANDBY within the next within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SURVEILLANCE RE UIREMENTS 4.6.4. 1 The isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.
"CAUTION: The inoperable isolation valve(s) may be part of a system(s).
Isolating the affected penetration(s) may affect the use of the system(s).
Consider the technical specification requirements on the affected system(s) and act accordingly.
TURKEY POINT - UNITS 3 8 4 3/4 6-17 AMENDMENT NOS. AND FPB 28 1$ I9
I'g CONTAINMENT SYSTEMS
'URVEILLANCE RE UIREMENTS Continued 4.6.4.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;
b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and
c. Verifying that on a Containment Ventilation Isolation test signal, each purge, exhaust and instrument air bleed valve actuates to its isolation position.
4.6.4.3 The isolation time of each power operated or automatic valve shall be Betermined to be within its limit when tested pursuant to Specification 4.0.5.
0 TURKEY POINT - UNITS 3 5 4 3/4 6-18 AMENDMENT NOS. AND PPB 2 8 1989
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CONTAINMENT SYSTEMS 3/4. 6. 5 COMBUSTIBLE GAS CONTROL DROGEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.5 Two independent containment hydrogen monitors shall be OPERABLE.
APPLICABILITY: MODES 1 and 2.
'CTION:
a. With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
6. With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS
.6.5.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:
t
a. One volume percent hydrogen, balance nitrogen, and
b. Four volume percent hydrogen,.balance nitrogen.
4.6.5.2 The flow path to each hydrogen monitor shall be demonstrated OPERABLE at least once per 31 days by a system walkdown to verify that each accessible manual, power operated, or automatic valve is in its correct position and that power is available to those components related to the operability of the f 1 owpath.
TURKEY POINT " UNITS 3 & 4 3/4 6-19 AMENDMENT NOS. AND Pgg t. 8 198S
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CONTAINMENT SYSTEMS 3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM IMITING CONDITION FOR OPERATION 3.6.6 A Post Accident Containment Vent System shall be OPERABLE.
MODES 1 and 2. 'PPLICABILITY:
ACTION:
'ith the Post Accident Containment Vent System inoperable, restore the Post
~ccident Containment Vent System to OPERABLE status within 7 days or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS 4.6.6 The Post Accident Containment Vent System shall be demonstrated OPERABLE:
a. At least once per 31 days by demonstrating system flow path operability via a system walkdown to verify that each accessible manual valve is in its correct position.
b. At least once per 18 months or (1) after any structural maintenance of the HEPA filter or charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release in any ventilation zone communicating with the system, or (3) after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation or (4) after replacement of a filter by:
1) A visual inspection of the system for foreign materials and gasket deterioration and verifying that the filter system satisfies the penetration and bypass leakage testing acceptance criter ia of less than 1X for DOP and halogenated hydrocarbon tests conducted at a design flow rate of 55 cfm 210K;
2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample performed in accordance with ANSI N510-1975, meets the methyl iodide removal criteria of greater than or equal to 90K and that any charcoal failing to meet the criteria be replaced with charcoal that meets or exceeds the criteria of Position C.6.a of Regulatory Guide 1.52, Revision 2.
TURKEY POINT - UNITS 3 8E 4 3/4 6-20 AMENDMENT NOS. AND FEB SS 1s89
CONTAINMENT SYSTEMS RVEILLANCE RE UIREMENTS Continued)
C. At least once per 18 months by:
1
1) Verifying that the pressure drop across the combined HEPA filter and charcoal adsorber is less than 6 inches Water Gauge at a flow rate of 55 cfm + 10K,
2) Visual inspection of the system and operation of all valves.
TURKEY POINT - UNITS 3 8 4 3/4 6-21 AMENOMENT NOS. AND FEB 28 1989
3/4.7. PLANT SYSTEMS 3/4. 7. 1 TURBINE CYCLE AFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY: MODES 1, 2, and 3.
aCTION:
With (3) reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the in-operable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.7. 1.1 No additional requirements other than those required by Specification
4. 0. 5.
e TURKEY POINT - UNITS 3 8 4 3/4 7"1 AMENDMENT NOS. AND Fpa g8 198s
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TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABL S AM LIN SAF TY VALVES MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL. POWER 82 54 27 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP .
ORIFICE SIZE VALVE NUMBER LIFT SETTING +1% " S UARE INCHES
~Loo A ~Loo B ~Loo C
1. RV1400 RV1405 RV1410 1085 16 psig'100
2. RV1401 RV1406 RV1411 psig 16
3. RV1402 RV1407 RV1412 1115 psig 16
4. RV1403 RV1408 RV1413 1130 psig 16 "The lift setting pressure'hall correspond to ambient conditions of the valve and pressure.
at nominal operating temperature TURKEY POINT - UNITS 3 & 4 3/4 7-2 AMENDMENT NOS. ANO fgB R 8 1989
I I
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Il 4$
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PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM IMITING CONDITION FOR OPERATION
3. 7. 1.2 Two independent auxiliary feedwater trains including 3 pumps as specified in Table 3.7"3 and associated flowpaths shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3 ACTION:
With one of the two required independent auxiliary feedwater trains inoperable, either restore the inoperable train to an OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or place the affected unit(s) in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months " and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
2) With both required auxiliary feedwater trains inoperable, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either restore both trains to an OPERABLE status, or restore one train to an OPERABLE status and follow ACTION statement 1 above for the other train. If neither train can be restored to an OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , verify the OPERABILITY of both standby feed-water pumps and place the affected unit(s) in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months " and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Otherwise, initiate corrective action to restore at least one auxiliary feedwater train to an OPERABLE status as soon as possible and follow ACTION statement 1 above for the other train.
With a single auxiliary feedwater pump inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify OPERABILITY of two independent auxiliary feedwater trains, or follow ACTION statements 1 or 2 above as applicable. Upon verifica-tion of the OPERABILITY of two independent auxiliary feedwater trains, restore the inoperable auxiliary feedwater pump to an OPERABLE status within 30 days, or place the operating unit(s) in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months " and in HOT SHUTDOWN Within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The provisions of Specification 3.0.4 are not applicable during the 30 day period for the inoperable auxiliary feedwater pump.
SURVEILLANCE RE UIREMENTS 1.7.1.2. 1 The required independent auxiliary feedwater trains shall be demonstrated OPERABLE:
a. At least once per 31 days on a STAGGERED TEST BASIS by:
1) Verifying by control panel indication and visual observation of equipment that each steam turbine-driven pump operates for 15 minutes or greater and develops a flow of greater than or "If this ~
ACTION applies to both units simultaneously, be in at least HOT
~
STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT SHUTDOWN within the following 6 hours:
TURKEY POINT - UNITS 3 8 4 3/4 7-3 AMENDMENT NOS. AND pp9 2 s 1s89
gl 1
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PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE RE UIREMENTS Continued equal to 373 gpm to the entrance of the steam generators. The provisions of Specification 4.0.4 are not applicable for entry into MODES 2 and 3;
2) Verifying by control panel indication and visual observation of equipment that the auxiliary feedwater discharge valves and the steam supply and turbine pressure valves operate as required to deliver the required flow during the pump performance test above;
3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and
4) Verifying that power is available to those components which require power for flow path operability.
b. At least once per 18 months by:
1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of each Auxiliary Feedwater Actuation test signal, and
2) Verifying that each auxiliary feedwater pump receives a start signal as designed automatically upon receipt of each Auxiliary Feedwater Actuation test signal.
1.2.2 An auxiliary feedwater flow path to each steam generator
'.7.
shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 1 by verifying normal flow to each steam generator.
TURKEY POINT - UNITS 3 Ea 4 3/4 7"4 AMENDMENT NOS. AND
~p R8 1989
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TABLE 3.7-3 AUXILIARY FEEDWATER SYSTEM OPERABILITY NIT TRAIN STEAM SUPPLY FLONPATH PUMP DISCHARGE MATER FLOMPATH 3 1 SG 3C via MOV-3-1405' A or C SG 3A via CV-3-2816 or SG 3B via MOV-3-1404 (1) SG 3B via CV-3-2817 SG 3C via CV-3-2818 SG 3A via MOV-3-1403( > or SG 3A via CV-3-2831 MOV-3-1404 (])
B C or SG 3B via SG 3B via CV-3-2832 SG 3C via CV-3-2833 SG 4C via MOV-4-1405( > A or C SG 4A via CV-4-2816 or SG 4B via MOV-4-1404 (1) SG 4B via CV-4-2817 SG 4C via CV-4-2818 SG 4A via MOV-4-1403( > B or C (2) SG 4A via CV-4-2831 or SG 4B via MOV-4-1404 (1) SG 4B via CV-4-2832 SG 4C via CV-4-2833 NOTES:
)Steam admission valves MOV-3-1404 and MOV-4-1404 can be aligned to either
~ ~
train (but not both) to restore OPERABILITY in the event MOV-3-1403 or
~
MOV-3-1405, or MOV-4-1403 or MOV-4-1405 are inoperable.
During single and two unit operation, one pump shall be OPERABLE in each train and the third auxiliary feedwater pump shall be OPERABLE and capable of being powered from, and supplying water to either train, except as noted in ACTION 3 of Technical Specification 3.7. 1.2. The third auxiliary feedwater pump (normally the "C" pump) can be aligned to either train to restore OPERABILITY in the event one of the required pumps is inoperable.
If any local manual realignment of valves is required when operating the auxiliary feedwater pumps, a dedicated individual, who is in communication with the control room, shall be stationed at the auxiliary feedwater pump area. Upon instructions from the control room, this operator would realign the valves in the AFW system train to its normal operational alignment.
TURKEY POINT " UNITS 3 4 4 3/4 7-5 AMENDMENT NOS. AND
H U
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PLANT SYSTEMS CONDENSATE STORAGE TANK IMITING CONDITION FOR OPERATION 3.7. 1.3 The condensate storage tanks (CST) system shall be OPERABLE with:
0 osite Unit in MODES 4 5 or 6 An indicated water volume of 185,000 gallons in either or both condensate storage tanks.
0 osite Unit in MODES 1 2 or 3 An indicated water volume of 370,000 gallons.
APPLICABILITY: MODES 1, 2 and 3.
ACTI ON:
0 osite Unit in MODES 4 5 or 6 With the CST system inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the CST system to OPERABLE status or be in at least HOT STANDBY in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
0 osite Unit in
~ ~
MODES 1 2 or 3
) With the CST system inoperable due'o containing less than 370,000 gallons, but greater than or equal to 185,000 gallons within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the inoperable CST system to OPERABLE status or place one unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours.
2) With the CST system inoperable with less than 185,000 gallons, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the CST system to OPERABLE status or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This ACTION applies to both units simultaneously.
TURKEY POINT - UNITS 3 8 4 3/4 7-6 AMENDMENT NOS.. AND FEB 28 >gas
PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued) 4.7. 1.3 The condensate storage tank (CST) system shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limit when the tank is the supply source for the auxiliary feedwater pumps.
TURKEY POINT " UNITS 3 8L 4 3/4 7-7 AMENOMENT NOS. ANO Pgg RS 589
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PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7. 1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.10 microCurie/gram DOSE E(UIVALENT I-131.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
<lith the specific activity of the Secondary Coolant System greater than 0. 10 microCurie/gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.7. 1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4. 7-1.
TURKEY POINT - UNITS 3 4 4 3/4 7-8 AMENDMENT NOS. AND FpJ3 2 8 1989
Il I'S i'k
TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS ANO ANALYSIS FRE UENCY
1. Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
Determination
2. Isotopic Analysis for DOSE a) Once per 31 days, when-E(UIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10K of the allowable limit for radioiodines.
b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10K of the allowable limit for radioiodines.
TURKEY POINT - UNITS 3 L 4 3/4 7-9 AMENDMENT NOS. AND MAY 0 5 1989
PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES IMITING CONDITION FOR OPERATION
3. 7. 1. 5 Each main steam line i sol ation valve (MSIV) shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
'ODE 1:
With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise be in HOT STANDBY within the'ext 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
MODES 2 and 3:
With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT, STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4. 0.4 are not appli<able for entry into HOOf 3.
TURKEY POINT - UNITS 3 8 4 3/4 7-10 AMENDMENT NOS. AND
.fEB ~ 8 3989
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PLANT SYSTEMS TANDBY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7. 1.6 Two standby feedwater pumps shall be OPERABLE" and at least 60,000 gallons of water (available volume), shall be in the Demineralized Water Storage Tank*".
APPLICABILITY: MODES 1, 2 and 3 ACTION:
a4 With one standby feedwater pump inoperable, restore the inoperable pumps to available status within 30 days or submit a SPECIAL REPORT per 3.7.1.6d.
b. With both standby feedwater pumps inoperable:
1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify the NRC and provide cause for inoperability and plans to restore pump(s) to OPERABLE status and,
2. Submit a SPECIAL REPORT per 3.7.1.6d.
0d.
With less than 60,000 gallons of water in the Demineralized Water Storage Tank restore the available volume to at least 60,000 gallons within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or submit a SPECIAL REPORT per 3.7.1.6d.
If a SPECIAL REPORT is required per the above specifications submit a report describing the cause of the inoperability, action taken and a schedule for restoration within 30 days in accordance with 6. 9.2.
SURVEILLANCE RE UIREMENTS 4.7. 1.6. 1 The Demineralized Water Storage tank water volume shall be deter-zined to be within limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
4.7.1.6.2 At least monthly verify the standby feedwater pumps are OPERABLE by testing in recirculation on a STAGGERED TEST BASIS.
4.7.1.6.3 At least once per 18 months, verify operability of the respective standby feedwater pump by powering from the non-safety grade diesel generators and providing feedwater to the steam generators.
"These pumps do not require plant safety related emergency power sources for operability and the flowpath is normally isolated.
"The Demineralized Water Storage Tank is non-safety grade.
TURKEY POINT - UNITS 3 8( 4 3/4 7-11 AMENDMENT NOS. AND OtD 2 A <oaa
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PLANT SYSTEMS 3/4. 7. 2 COMPONENT COOLING WATER SYSTEM MITING CONDITION FOR OPERATION
3. 7. 2 The Component Cooling Water System (CCW) shall be OPERABLE with:
a. Three CCW pumps, and
b. Two CCW heat exchangers.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
'a With only two CCW pumps with independent power supplies OPERABLE, or-
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restore the inoperable CCW pump to OPERABLE status within 30 days be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification
3. 0.'4 are not applicable.
b. With only one CCW pump OPERABLE or with two CCW pumps OPERABLE but not from independent power supplies, restore two pumps from independent power supplies to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
With less than two CCW heat exchangers OPERABLE, restore two heat exchangers to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.7.2 The Component Cooling Water System (CCW) shall be demonstrated OPERABLE:
a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , by verifying that two heat exchangers and one pump are capable of removing design basis heat loads.
TURKEY POINT " UNITS 3 & 4, 3/4 7-12 AMENDMENT NOS. AND fEB 28 1989
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SURVEILLANCE RE UIREMENTS Continued)
At least once per 31 days by: (1) verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position, and (2) verifying by a performance test the heat exchanger surveillance curves.
C. At least once per 18 months during shutdown, by verifying that:
1) Each automatic valve servicing safety-related equipment actuates to its correct position on a SI test signal, and
2) Each Component Cooling Water System pump starts automatically on a SI test signal.
3) Interlocks required for CCW operability are OPERABLE.
TURKEY POINT - UNITS 3 8 4 3/4 7-13 AMENOMENT NOS. ANO f59 0 8 l988
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PLANT SYSTEMS
/4.7.3
~ ~ INTAKE COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The Intake Cooling Water System (ICW) shall be OPERABLE with:
a. Three ICW pumps, and
b. Two ICW headers.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only two ICW pumps with independent power supplies OPERABLE, restore the inoperable ICW pump to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable.
b. With only one ICW pump OPERABLE or with two ICW pumps OPERABLE but not from independent power supplies, restore two pumps from independent power supplies to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
fJ C. With only one ICW header OPERABLE, restore two headers to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.7.3 The Intake Cooling Water System (ICW) shall be demonstrated OPERABLE:
At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
b. At least once per 18 months during shutdown, by verifying that:
1) Each automatic valve servicing safety-related equipment actuates to its correct position on a SI test signal, and
2) '
Each Intake Cooling Water System pump starts automatically on SI test signal.
3) Interlocks required for system operability are OPERABLE.
TURKEY POINT - UNITS 3 8 4 3/4 7"14 AMENDMENT NOS. AND FEB 28 >sas
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tPLANT SYSTEMS 3/4. 7. 4 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.4 The ultimate heat sink shall be OPERABLE with an average supply water temperature to the Intake Cooling Water System less than or equal to 100 F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
This action shall be applicable to both units simultaneously.
tSURVEILLANCE RE UIREMENTS 4.7.4 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the average supply water temperature" to the Intake Cooling Water System to be within its limit.
Portable monitors may be used to measure the temperature.
TURKEY POINT - UNITS 3 4 4 3/4 7-15 AMENDMENT NOS. AND RES 28 >ggg
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PLANT SYSTEMS 4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.5 The Control Room Emergency Ventilation System shall be OPERABLE.
APPLICABILITY: Al 1 MODES.
ACTION:
lODES 1, 2, 3 and 4:
With the Control Room Emergency Ventilation System inoperable, suspend all movement of fuel in the spent fuel pool and restore the inoperable system to OPERABLE status within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
If this ACTION applies to both units simultaneously, be in HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
MODES 5 and 6:
With the Control Room Emergency Ventilation System inoperable, suspend all operations involving CORE ALTERATIONS, movement of fuel in the spent fuel pool, or positive reactivity changes. This ACTION shall apply to both units simultaneously.
SURVEILLANCE RE UIREMENTS
4. 7. 5 The Control Room Emergency Ventilation System shall be demonstrated OPERABLE; a ~ At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air temperature is less than or equal to 120'F;
b. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes; C. At least once per 18 months or (1) after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system opera-tion, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (3) following operational exposure of the filters to effluents from painting, fire, or chemical release in any ventilation zone communicating with the system, or (4) after complete or partial replacement of a filter bank by:
TURKEY POINT - UNITS 3 8c 4 3/4 7-16 AMENDMENT NOS. AND FpB R 8 198s
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PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued
1) Verifying that the air cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of greater than or equal to 99K OOP and halogenated hydrocarbon removal at a system flow rate of 1000 cfm 110K.
2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per ANSI N510-1975, meets the criteria for methyl iodine removal efficiency of greater than or equal to 90K or the charcoal be replaced with charcoal that meets or exceeds the criteria of position C.6.a. of Regulatory Guide 1.52 (Revision 2), and 3). Verifying by a visual inspection the absence of foreign materials and gasket deterioration.
At least once per 12 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 1000 cfm ilOX; e e. At least once per 18 months by verifying that on a Containment Phase "A" Isolation test signal the system automatically switches into the recirculation mode of operation.
TURKEY POINT - UNITS 3 8 4 3/4 7-17 AMEHOMENT NOS. AND FEB 2 s i.-~
FEB 28 1S89
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PLANT SYSTEMS
/4.7.6 SNUBBERS IMITING CONDITION FOR OPERATION 3.7.6 All snubbers shall be OPERABLE. The only snubbers excluded from the requirements are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are instal]ed would have no adverse effect on any safety-related system.
APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.
ACTION:
Pith one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or re-store the inoperable snubber{s} to OPERABLE status and perform an engineering eval-uation per Specification 4.7.6f. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.
SURVEILLANCE RE UIREMENTS 4.7.6 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice'inspection program in addition to the require-
'~IT ments of Specification 4.0.5..
As used in this specification, type of snubber shall mean snubbers
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of the same design and manufacturer, irrespective of capacity.
~ ~ ~
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b.
~ Visual Ins ections
~ ~ ~
Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The inservice visual inspection of each type of snubber shall be 'irst performed after 4 months but.within 10 months of commencing POWER OPERATION and shall include all snubbers. If all snubbers of each type (on any system} are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection (of that system) shall be performed at the first refueling outage. 'therwise, subsequent visual inspections of a given system shall be performed in accordance with the following schedule:
No. of Inoperable Snubbers of Each Type Subsequent Visual on an s stem er Ins ection Period Ins ection Period" ""
18 months a 2 1
2 12 months 6 months i 25K t 25K 3,4 124 days k 25K t
5,6,7 62 days 4 25K 8 or more 31 days ~ 25K "The inspection interval for each type of snubber (on a given system) shall not be lengthened more than one step at a time unless a generic problem has been identified and corrected; in that event the inspection interval may be length-ened one step the first time and two steps thereafte~
that-if no inoperable snubbers of that type are found (on system).
""The provisions of Specification 4.0.2 are not applicable.
TURKEY POINT - UNITS 3 & 4 3/4 7-18 AMENDMENT NOS. AND
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PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued)
C. Visual Ins ection Acce tance Criteria Visual inspections shall verify*that: (1) there are no visible indi-cations of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that partic-ular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Specification 4. 7. 6e. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers.
d. Functional Tests For each unit during refueling shutdown, a representative sample of snubbers shall be tested using the following sample plan:
1) At least 10K of the total number of safety related snubbers for the respective unit identified by site records shall be func-tionally tested either in-place or in a bench test. For each snubber of a type that does not meet the functional test accep-tance criteria of Specification 4.7.6e, an additional 10K of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been fun'ctionally tested;
2) The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At least 25X of the snubbers in the representative sample shall include snubbers from the following categories; A. Snubbers within 5 feet of heavy equipment (ex. valves, pumps, turbines, motors, etc. )
B. Snubbers within 10 feet of the discharge from a safety relief valve.
3) Snubbers identified by site records as "Especially Difficult to Remove" or in "High Radiation Zones During Shutdown" shall also t
be included in the representative sample."
"Permanent or other exemptions from functional testing for individual snubbers in these categories may be granted by the Commission only if a justifiable basis for exemption is presented and/or snubber life destructive testing was performed to qualify snubber OPERABILITY for all design conditions at either the completion of their fabrication or at a subsequent date.
TURKEY POINT - UNITS 3 8( 4 3/4 7-19 AMENDMENT NOS. AND FEB 2 8 1989
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PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period. If a spare snubber has been installed in place of a failed snubber, then both the failed snubber (if it and'nstalled in another position) and the spare snubber shall is repaired be retested. Test results of these snubbers may not be included for the re-sampling.
e. Mechanical Snubbers Functional Test Acce tance Criteria U
The snubber functional test shall verify that:
1) Activation (restraining action) is achieved with the specified range of velocity or acceleration in both tension and compression;
2) Snubber release rate, where required, is within the specified range in tension and compression,
3) The force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel.
f. Functional Test Failure Anal sis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
If any snubber selected for functional testing either fails to activate or fails to move, i.e., frozen-in-place, the cause will be evaluated under the provisions of 10 CFR Part 2l.
Should the -results of the evaluation indicate that the failure was caused by either manufacturer or design deficiency, further action shall be taken, if needed, based on manufacturer or engineering recommendations.
For the snubber(s) found inoperable, an evaluation shall be performed on the components to which the inoperable snubbers are purpose of this evaluation shall be to determine ifattached.'he the components to which the inoperable snubber(s) are attached were adversely affected by the inoperability of the snubber(s) in order to ensure that the component remains capable of meeting the designed service.
TURKEY POINT " UNITS 3 8 4 3/4 7"20 AMENOMENT NOS. AND FEB S5 gg
h 'i PLANT SYSTEMS SURVEILlANCE RE UIREMENTS Continued
g. Snubber Service Life Monitorin Pro ram A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification 6.10.3m.
Concurrent with the first inservice visual inspection and during refueling shutdown thereafter, the installation and maintenance records for each safety related snubber as identified by site records shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review,
.the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records.
TURKEY POINT - UNITS 3 5 4 3/e 7-ZZ AMENDMENT NOS. ANO FEB 28 1989
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PLANT SYSTEMS 3/4.7.7
~ ~ SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.7 Each sealed source containing radioactive material either in excess of 100 microCuries of beta and/or gamma emitting material or 5 microCuries of alpha emitting material shall be free of greater than or equal to 0.005 microCurie of removable contamination.
APPLICABILITY: At all times.
ACTION:
a 0 With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:
1. Decontaminate and repair the sealed source, or
2. Dispose of the sealed source in accordance with Commission Regulations.
b. The P rovisions of .Specification 3. 0. 3 are not applicable.
SURVEILLANCE RE UIREMENTS 4.7.7. 1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:
a. The licensee, or
b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microCurie per test sample.
4.7.7.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.
a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:
1) With a half-life greater than 30 days (excluding Hydrogen 3),
and
2) In any form other than gas.
TURKEY POINT - UNITS 3 5 4 3/4 7-22 AMENDMENT NOS. AND FEB 28 1s89
PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued
b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and C. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
4.7.7. 3 Reports - A report. shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal f3e presence of greater than or equal to 0.005 microCurie of removable contamination.
4.7.7.4 A complete inventory of licensed radioactive materials in possession shall be maintained current at all times.
e TURKEY POINT - UNITS 3 & 4 3/4 7-23 AMENOMENT NOS. ANO FEB s8 >sss
PLANT SYSTEMS 3/4.7.8 FIRE SUPPRESSION SYSTEMS FIRE WATER SUPPLY AND DISTRIBUTION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 The Fire Water Supply and Distribution System shall be OPERABLE with:
a. At least two fire suppression pumps, one electric driven and one diesel driven with their discharge aligned to the fire suppression header',
b. Two separate water supplies, each with a minimum contained volume of 300,000 gallons, and
c. An OPERABLE flow path capable of taking suction from the Raw Mater Tank I and Raw Mater Tank II and transferring the water through dis-tribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.8.2, 3.7.8.3, and 3 7 8.4.
~ ~
APPLICABILITY: At all times.
ACTION:
a. With one pump and/or one water supply inoperable, restore the inoper-able equipment to OPERABLE status within 7 days or provide an alter-nate backup pump or supply. The provisions of Specification 3.0.3 are not applicable. This action applies to both units simultaneously.
b. Mith the Fire Mater Supply and Distribution System otherwise inoper-able, establish a backup fire water capability within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This action applies to both units simultaneously.
TURKEY POINT - UNITS 3 8 4 3/4 7-24 AMENDMENT NOS. AND FEB 28 )989
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PLANT SYSTEMS e SURVEILLANCE RE UIREMENTS Fire Water Supply Oistribution System shall be demonstrated
4. 7. 8. 1.1 The and OPERABLE'.
At least once per 7 days by verifying the contained water supply volume,
b. At least once per 31 days by starting the electric motor-driven pump and operating it for at least 15 minutes on recirculation flow,
c. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position,
d. At least once per 12 months by performance of a system. flush,
e. At least once per 12 months by cycling each testable valve. in the flow path through at least one complete cycle of full travel,
f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
1) Verifying that each automatic valve in the flow path actuates to its correct position,
2) Verifying that the electric-driven pump develops at least 1880 gpm at a system pressure of 130.2 psig, and the diesel-driven pump develops at least 2350 gpm at a system pressure of 130.2 psig, by verifying 3 points on the pump performance curve.
3) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and
4) Verifying that each fire pump starts sequentially to maintain the Fire Water Supply and Oistribution System pressure greater than or equal to 125 psig.
g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.
'I TURKEY POINT " UNITS 3 & 4 3/4 7-25 AMENOMENT NOS. AND FEB 28 198g
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PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.7.8.1..2 The fire pump diesel engine shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying:
1) The fuel storage tank contains at least 375 gallons of fuel, and
2) The diesel starts from ambient conditions and operates for at least 30 minutes on recirculation flow.
b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-0270-1975 is within the acceptable limits specified in Table 1 of ASTM D975-1977 when checked for viscosity and water and sediment; and
c. At least once per 18 months by subjecting the diesel to an inspec-tion in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.
t4. 7. 8. l. 3 The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:
a. At least once per 7 days by verifying that:
1) The electrolyte level of each battery is above the plates, and
2) The overall battery voltage is greater than or equal to 24 volts.
b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and C. At least once per 18 months by verifying that:
1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and
2) The battery'-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.
TURKEY POINT - UNITS 3 8( 4 3/4 7-26 AMENDMENT NOS. AND FF.B Rs ls89
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PLANT SYSTEMS SPRAY ANO/OR SPRINKLER SYSTEMS LIMITING CONOITION FOR OPERATION 3.7.8.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:
a. Fire Zones 47 and 54 - Component Cooling Water Areas
b. Fire Zones 45 and 55 - Charging Pump Rooms
c. Fire Zones 79A - North - South Breezeway
d. Fire Zones 72, 73, 74 and 75 - Emergency Oiesel Generator and Oay Tank Rooms APPLICABILITY: Whenever equipment protected by the Spray/Sprinkler System is required to be OPERABLE.
ACTION:
ae With one or more of the above required Spray and/or Sprinkler Systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipment. This ACTION'applies to both units simultaneously.
b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE UIREMENTS 4.7.8.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic)'n the flow path is in its correct position,
b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, c., At least once per 18 months:
1) By performing a system functional test which includes simulated automatic actuation of the system, and:
a) Verifying that the automatic valves'n the flow path actuate to their correct positions on a test signal, and P
b) Cycling each valve in the flow path that is not testab1 e plant operation through at least one complete cyc1e 'uring of full travel.
TURKEY POINT - UNITS 3 8( 4 3/4 7-27 AMENDMENT NOS. ANO FEB 28 198g
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PLANT SYSTBlS
2) By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity; and
3) By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.
d. At least once per 3 years by performing an air or water flow test through each open head spray/sprinkler header and verifying each open head spray/sprinkler nozzle is unobstructed.
e TURKEY POINT - UNITS 3 8 4 3/4 7-28 AMENOHENT NOS. AND FEB 2 8 1989
tq 41
'4
FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION L
3.7.8.3 The fire hose stations given in Table 3.7-4 shall be OPERABLE.
APPLICABILITY: Mhenever equipment in the areas protected by the fire hose stations 1s required to be OPERABLE.
ACTION:
Nith one or more of the fire hose stations given in Table 3.7-,4 inoperable, provide an equivalent capacity fire hose from the nearest equivalent OPERABLE water source. The fire hose shall be of a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station, and shall be stored in a roll at the outlet of the OPERABLE water supply. The, above ACTION requirement sha'll be accomplished within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This ACTION applies to both units simultaneously.
b. The provisions of Specification 3. 0.3 are not applicable.
SURVEILLANCE RE UIREMENTS
.7.8.3
~ ~ ~ Each of the fire hose stations given in Table 3.7-4 shall be demonstrated OPERABLE:
At least once per 31 days, by a visual inspection of the fire hose stations accessible during plant operations .to assure all required equipment is at the station.
b. At least once per 12 months, by:
1) Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station,
2) Removing the hose for inspection and re-racking, and
3) Inspecting all gaskets and replacing any degraded gaskets in the couplings.
C. At least once per 3 years, by:
1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and
2) Conducting a hose hydrostatic test at a pressure of 150 psig or.
'at least 50 psig above maximum fire main operating pressure, whichever is greater.
TURKEY POINT " UNITS 3 8 4 3/4 7-29 AMENDMENT NOS. AND I
PE@ R 8 1989
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p, II I'N'
TABLE 3.7"4 FIRE HOSE STATIONS IDENTIFICATION LOCATION FIRE ZONE HS"03-01 EL. 18' East of 4160V SWGR Room on Column 88 HS-03-02 EL. 18' West of 3A Condensate-Pump on Pedestal 87 HS"03-03 EL. 18' Passageway South of SG Feed Pump Room 83 HS-03"04 EL. 30' East of 480V Load Center on Column 105 HS"03"05 EL. 30' South End of Mezzanine Deck 105 HS-03-06 EL. 42' NW End of Turbine Deck 117 HS-03-07 EL. 42' North of 6A HPFW Heater 117 HS-03-08 EL. 42' NW Corner of Entrance to Elevator 79 HS" 04-01 EL. 18' South of 4160V SWGR Room on Column 82 HS-04-02 EL. 18' Passageway South of SG Feed Pump Room 78 HS-04"03 EL. 30' East of 480V Load Center at Stairway 105 HS-04-04 EL. 30' South End of Mezzanine Deck 105 HS-04-05 EL. 42' 'West End of Turbine Deck 117 HS-04-06 EL. 42' East Side of Turbine Deck and North of 117 6A FW Heater HS-04-07 EL. 42' East Side of Turbine Deck and North of 117 I
6B FW Heater HS"04-08 EL. 42' Southwest Corner of Turbine Deck 117 HS-AB-01 EL. 18' East-West Passageway at West End 58 HS-AB"02 EL. 18' East-West Passageway at East End 58 HS"AB-03 EL. 18' North-South Passageway Outside Unit 3 58 Charging Pump Room HS-AB"04 EL. 50' Roof of Unit 3 New Fuel Storage Area 118 HS-AB"05 EL. 50' Roof of Unit 4 New Fuel Storage Area 118 TURKEY POINT - UNITS 3 & 4 3/4 7-30 AMENDMENT NOS. ANO FF.g R8 198aI
PLANT SYSTEMS FIRE HYDRANTS AND HYDRANT HOSE HOUSES LIMITING CONDITION FOR OPERATION 3.7.8.4 The fire hydrants and associated hydrant hose houses given in Table 3.7-5 shall be OPERABLE.
APPLICABILITY: Whenever equipment in the areas protected by the fire hydrants ss required to be OPERABLE.
ACTION:
a ~ With one or more of the fire hydrants or associated hydrant hose houses given in Table 3.7-5 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months have sufficient additional lengths of 2 I/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area(s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise, provide the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. The provisions of Specification 3.0.3 and are not applicable.
SURVEILLANCE RE UIREMENTS
4. 7.8.4 Each of the fire hydrants and associated
~ ~ ~ hydrant hose houses given in Table 3.7-5 shall be demonstrated OPERABLE:
~
~
a. At least once per 31 days, by visual inspection of the hydrant hose house to assure all required equipment is at the hose house,
b. At least once per 6 months by visually inspecting each fire hydrant and verifying that the hydrant is not damaged, and C. At least once per 12 months by:
1) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater,
2) Inspecting all the gaskets and replacing any degraded gaskets in the couplings, and
3) Performing a flow check of each hydrant'to verify its OPERABILITY.
TURKEY POINT - UNITS 3 8 4 3/4 7-31 AMENDMENT NOS. AND fig 28 lsSs
C 1
)
4j
'I
~+
TABLE 3.7"5 FIRE HYDRANTS NUMBER OF IDENTIFICATION FIRE ZONE LOCATION HYDRANTS FH-01 124 NE Corner of Unit 3 near Vehicle Gat'e into RCA FH-06 NA W of Nuclear Maintenance Building FH" 07 86 Unit 3 Transformer Area FH-08 81 Unit 4 Transformer Area FH"09 76 Unit 4 Turbine-Generator Area FH" 10 77 Unit 4 Condensate Storage Tank Area FH-11 FH" 12 Unit 4 New Fuel Storage Area FH-13 123 Refueling Water Storage Area FH-17 NA Nuclear Dry Storage Area FH"16 NA Steam Generator Storage Area TOTAL TURKEY POINT - UNITS 3 8 4 3/4 7-32 AMENDMENT NOS. AND.
FEB 28 1989
tj I, 0 p
4;c r
4J
PLANT SYSTEMS 3/4. 7. 9
~ ~ FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.9 A11 fire rated assemblies (walls, floor/ceilings, fire barrier penetration seals, and other fire barriers) separating safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable, piping, and ventilation duct penetration seals) shall be OPERABLE.
APPLICABILITY: At all times.
ACTION:
a ~ Mith one or more of the above required fire rated assemb1ies and/or sealing devices inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either establish a continuous fire watch on at least one side of the affected assembly,-
or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly and establish an hourly fire watch patrol.
b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE UIREMENTS 4.7.9.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of:
a. The exposed surfaces of each fire rated assembly,
b. Each fire window/fire damper and associated hardware, and
c. At least 10K of each type of sealed penetration. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10K of each type of sealed penetration shall be made. This inspection process shall continue until a 10K sample with no apparent changes in appearance or abnormal degradation is found. Samples shall be selected such that each penetration will be inspected every 15 years.
TURKEY POINT - UNITS 3 8 4 3/4 7-33 AMENDMENT NOS. AND FEB N 8 1989
gP PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued
4. 7.9.2 Each of the above required fire doors shall be verified OPERABLE by inspecting the release and closing mechanism and latches at least once per 6 months, and by verifying:
The OPERABILITY of the fire door supervision system for each electrically supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days,
b. That each locked closed fire door is closed at least once per 7 days, C. That doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and a functional test is performed at least once per 18 months, and
d. That each unlocked fire door without electrical supervision is closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
e TURKEY POINT - UNITS 3 It 4 3/4 7-34 . AMENDMENT NOS. ANO FEB 88 ]98g
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ML17347B114

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