ML20215N503
| ML20215N503 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1207, NUDOCS 8611060142 | |
| Download: ML20215N503 (414) | |
Text
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- NUREG-1207 O ,
. _ - . . - . . - _ _ . - __ _ -_-. __ . .i Technical Specifications Seabrook Station, Unit 1 .
Docket No. 50-443 Appendix "A" to License No. NPF-56 lssued by th.e U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1986 p"rc o e i O 8611060142 DR ADOCK 05000443 861031 PDR
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NUREG-1207 Technical SpecsTwmhs Seabrook Station, Unit 1 Docket No. 50-443 Appendix "A" to License No. NPF-56 issued by the U.S. Nuclear Regulator ( Commission Office of Nuclear Reactor Regulation
' October 1906
- v. ..u9 Y....k.
O E l l INDEX B O O
INDEX
/ _.\
1.0 DEFINITIONS SECTION PAGE 1.1 ACTI0N......................... .............................. 1 1.2 ACTUATION LOGIC TEST.......................................... 1-1 1.3 ' ANALOG CHANNEL OPERATIONAL TEST............................... 1-1 1.4 AXIAL FLUX DIFFERENCE......................................... 1-1 1.5 CHANNEL CALIBRATION........................................... 1-1 1.6 CHANNEL CHECK................................................. 1-1 1.7 CONTAINMENT INTEGRITY......................................... 1-2 1.8 CONTROLLED LEAKAGE............................................ 1-2 1.9 CO R E A LT E R AT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.10 DOSE EQUIVALENT I-131........................................ 1-2 1.11 E - AVERAGE DISINTEGRATION ENERGY............................ 1-2 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.13 FREQUENCY N0TATION........................................... 1-3 1.14 GASEOUS RADWASTE TREATMENT SYSTEM............................ 1-3 1.15 IDENTIFIED LEAKAGE........................................... 1-3 1.16 MASTER-RELAY TEST............................................ 1-3 1.17 MEMBER (S) 0F THE PU8LIC...................................... 1-3 1.18 0FFSITE DOSE CALCULATION MANUAL.............................. 1-4 1.19 OPERABLE - OPERABILITY....................................... 1-4 1.20 OPERATIONAL MODE - M0DE...................................... 1-4 o 1.21 PHYSICS TESTS................................................ 1-4 e N 1.22 PRESSURE B0UNDARY LEAKAGE.................................... 1-4 ks, 1.23 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.24 PURGE - PURGING.............................................. 1-4 1.25 QUADRANT POWER TILT RATI0.................................... 1-5 1.26 RATED THERMAL P0WER.......................................... 1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME.... ....................... 1-5 1.28 REPORTABLE EVENT............................................. 1-5 1.29 CONTAINMENT ENCLOSURE BUILDING INTEGRITY..................... 1-5 1.30 SHUTDOWN MARGIN.............................................. 1-5 1.31 SITE B0VNDARY................................................ 1-5 1.32 SLAVE RELAY TEST............................................. 1-6 1.33 SOLIDIFICATION............................................... 1-6 1.34 SOURCE CHECK................................................. 1-6 1.35 STAGGERED TEST BASIS......................................... 1-6 ! 1.36 THERMAL P0WER................................................ 1-6 1.37 -TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 1.38 UNIDENTIFIED LEAKAGE......................................... 1-6 1.39 UNRESTRICTED AREA............................................ 1-6 1.40 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-7 1.41 VENTING...................................................... 1-7 TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 .0PERATIONAL M0 DES....................................... 1-8 N
\s SEABROOK - UNIT 1 i 1
INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS................................................. 2-1 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRES $URE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0INTS.... 2-4 2.0 BASES 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 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 SETP0INTS............... B 2-3 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T avg Greater Than 200 F................ 3/4 1-1 Shutdown Margin - T avg Less Than or Equal to 200 F....... 3/4 1-3 Moderator Temperature Coefficient.................... ... 3/4 1-4 Minimum Temperature for Criticality...................... 3/4 1-6 O SEABROCK - UNIT 1 ii t
n INDEX LIMITING CONDITI'ONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.1.2 B0 RATION SYSTEMS Flow Paths - Shutdown.................................... 3/4 1-7 Flow Paths - 0perating................................... 3/4 1:8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - 0perating............................... 3/4 1-10 Borated Water Sources - Shutdown......................... 3/4 1-11 Borated Water Sources - 0perating........................ 3/4 1-12 Isolation of Unborated Water Sources - Shutdown.......... 3/4 1-14 3/4.1.3 M0VABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1-15 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0PERABLE FULL-LENGTH R00................... 3/4 1-17 Position Indication Systems - Operating.................. 3/4 1-18 Position Indication System - Shutdown.................... 3/4 1-19 Rod Drop Time............................................ 3/4 1-20 Shutdown Rod Insertion Limit............................. 3/4 1-21 (s Control Rod Insertion. Limits............................. 3/4 1-22 FIGURE ~3.1-1 R00 BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR-LOOP 0PERATION...................................... 3/4 1-23 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 P0WER...................................... 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)..................... q 3/4 2-4 FIGURE 3.2-2 K(Z) - NORMALIZED F q(Z) AS A FUNCTION OF CORE HEIGHT. 3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR................. 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-9 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-10 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 O Q SEABROOK - UNIT 1 iii l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-2 (This table number is not used) TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-9 3/4.3.2 ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-14 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-16 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS.................... ...... 3/4 3-24 TABLE 3.3-5 (This table number is not used) TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-31 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-36 TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS..................................... 3/4 3-37 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-39 Movable Incore Detectors................................. 3/4 3-40 Seismic Instrumentation.................................. 3/4 3-41 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-42 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-43 Meteorological Instrumentation........................... 3/4 3-44 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION...... ..... 3/4 3-45 Remote Shutdown Systtm................................... 3/4 3-46 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM.... ................ .. ..... 3/4 3-47 Accident Monitoring Instrumentation...................... 3/4 3-49 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-50 TABLE 3.3-11 (This table number is not used)...................... 3/4 3-53 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-55 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-56 SEABROOK - UNIT 1 iv
i INDEX I
'_ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 4.3-5 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-58 Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-60 TABLE 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/4 3-61 TABLE 4.3-6 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING
- INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-64 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-67 3/4.4 REACTOR COOLANT SYSTEM' 3/4.4.1 REACTOR COOLANT LOOPS AND C0OLANT CIRCULATION Startup and Power Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-4 Cold Shutdown - Loops Filled............................. 3/4 4-6 m Cold Shutdown - Loops Not Filled......................... 3/4 4-7 3/4.4.2 SAFETY VALVES Shutdown................................................. 3/4 4-8 0perating................................................ 3/4 4-9 3/4.4.3 PRESSURIZER.............................................. 3/4 4-10 3/4.4.4 RELIEF VALVES...........,................................ 3/4 4-11 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-13 TABLE-4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-20 i Ope ra ti o na l Le a kage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-21 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-24 3/4.4.7 CHEMISTRY..................... .......................... 3/4 4-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY....... . .............................. 3/4 4-27
. s SEABROOK - UNIT 1 v
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-28 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Genera 1.................................................. 3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY................................. 3/4 4-31 FIGURE 3.4-3 REACTOR C0OLANT-SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY.......... ...................... 3/4 4-32 Pressurizer.............................................. 3/4 4-33 Overpressure Protection Systems.......................... 3/4 4-34 FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETP0INTS........... 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM ~ VENTS............................. 3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation................ 3/4 5-1 3/4 5-3
~
Shutdown................................................. 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350 F.... 3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350 F................... 3/4 5-8 avg ECCS Subsystems - T Equal To or Less Than 200 F....... 3/4 5-10 avg 3/4.5.4 REFUELING WATER STORAGE TANK.. .......................... 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity............... .................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 O SEABROOK - UNIT 1 vi
m INDEX fv LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS............ 3/4 6-5 Containment Air Locks.................................... 3/4 6-7 Internal Pressure.......... ............................. 3/4 6-9 A i r Temp e ratu re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-10 Containment Vessel Structural Integrity.................. 3/4 6-11 Containment Ventilation System........................... 3/4 6-12 3/4.6.2 DEPRESSURIZATION AND C0OLING SYSTEMS Containment Spray System................................. 3/4 6-14 Spray Additive System.................................... 3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES............................. 3/4 6-16 3/4.6.4 COMBUSTIBLE' GAS CONTROL Hydrogen Monitors........................................ 3/4 6-18 Electric Hydrogen Recombiners............................ 3/4 6-19 Hydrogen Mixing System................................... 3/4 6-20 (D i - 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING G Containment Enclosure Emergency Air Cleanup System....... 3/4 6-21 Containment Enclosure Building Integrity................. 3/4 6-24 Containment Enclosure Building Structural Integrity...... 3/4 6-25 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................ 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 0PERATION...................................... 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-2 Auxiliary Feedwater System............................... 3/4 7-3 Condensate Storage Tank.................................. 3/4 7-6 Specific Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY C0OLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-8 Main Steam Line Isolation Valves......................... 3/4 7-9 O Atmospheric Relief Valves................................ 3/4 7-10 SEABROOK - UNIT 1 vii
INDEX LIMITING CONDITIONS'FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-11 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM................... 3/4 7-12 3/4.7.4 SERVICE WATER SYSTEM..................................... 3/4 7-13 3/4.7.5 ULTIMATE HEAT SINK....................................... 3/4 7-14 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM..................... 3/4 7-16 3/4.7.7 SNUBBERS................................................. 3/4 7-18 3/4.7.8 SEALED SOURCE CONTAMINATION.............................. 3/4 7-19 3/4.7.9 (This specification number is not used).................. 3/4 7-21 3/4.7.10 AREA TEMPERATURE M0NITORING.............................. '3/4 7-22 TABLE 3.7-3 AREA TEMPERATURE M0NITORING........................... 3/4 7-23 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-10 Shutdown................................................. 3/4 8-11 3/4.8.2 D.C. SOURCES 0perating................................................ 3/4 8-12 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-14 Shutdown................................................. 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-16 Shutdown................................................. 3/4 8-18 Trip Circuit for Inverter I-2A........................... 3/4 8-19 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. Circuits Inside Primary Containment................. 3/4 8-20 Containment Penetration Conductor Overcurrent Protective Devices..................................... 3/4 8-21 Motor-Operated Valves Thermal Overload Protection........ 3/4 8-24 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 SEABROOK - UNIT 1 viii
n INDEX ( I v' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 COMMUNICATIONS........................................... 3/4 9-5 3/4.9.6 REFUELING MACHINE........................................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS. . . . . . . . . . . . . . . . . . 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-8 Low Water Level.......................................... 3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST 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 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM...... 3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 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 3/4.10.4 REACTOR COOLANT L00PS.................................... 3/4 10-4 O- 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 Dose........................... ......................... 3/4 11-2 Liquid Radwaste Treatment System......................... 3/4 11-3 Liquid Holdup Tanks.... .............. .................. 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................ 3/4 11-5 Dose - Noble Gases.............................-...... ... 3/4 11-6 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form............................. 3/4 11-7 Gaseous Radwaste Treatment System........................ 3/4 11-8 Explosive Gas Mixture - System........................... 3/4 11-9 Gas Storage Tanks....................................... 3/4 11-10 3/4.11.3 SOLID RADI0 ACTIVE WASTES..... ................... ....... 3/4 11-11 3/4.11.4 TOTAL 00SE.............................. ................ 3/4 11-12 3/4.12 RADIOLOGICAL ENVIROMMENTAL MONITORING O 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 SEABROOK - UNIT 1 ix i
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-3 3/4.12.3 INTERLABORATORY COMPARIS0N PR0 GRAM....................... 3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 l 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R........................ B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-3 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-4 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 3/4.3.4 TURBINE OVERSPEED PR0TECTION.............................. B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND C0OLANT CIRCULATION............. B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-1 3/4.4.3 PRESSURIZER.............................................. B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-3 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-7 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE.................................. B 3/4 4-9 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT NDT FOR REACTOR VESSELS EXPOSED TO 550 F............ B 3/4 4-10 SEABROOK - UNIT 1 x
INDEX (y BASES SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL T0VGHNESS.......................... 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 4-16 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 3/4.5.5 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND C0OLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-3 3/4.6.5_ CONTAINMENT ENCLOSURE BUILDING............................ B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 V 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM.................... B 3/4 7-3. 3/4.7.4 SERVICE WATER SYSTEM...................................... B 3/4 7-3 , 3/4.7.5 ULTIMATE HEAT SINK........................................ B 3/4 7-3 1 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM...................... B 3/4 7-3 1 3/4.7.7 SNUBBERS.................................................. B 3/4 7-4 3/4.7.8 SEALED SOURCE CONTAMINATION............................... B 3/4 7-5 l 3/4.7.9 (This specification number is not used)................... B 3/4 7 1 3/4.7.10 AREA TEMPERATURE MONITORING............................... B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS l
~3 /4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and i B 3/4 8-1 ONSITE POWER DISTRIBUTION...............................
3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 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 1 3/4.9.3 DECAY TIME..............................................., B 3/4 9-1 l 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................................ B 3/4 9-1 3/4.9.6 REFUELING MACHINE......................................... B 3/4 9-1 (' 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING................ B 3/4 9-2 SEABROOK - UNIT 1 xi
INDEX BASES SECTION PAGE 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............. B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM............ B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L............................................ B 3/4 9-2 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM....... B 3/4 9-2 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.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... -B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS .3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASE0US EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................ B 3/4 11-5 3/4.11.4 TOTAL D0SF.............................................. B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARIS0N PR0 GRAM...................... B 3/4 12-2 5.0 DESIGN FEATURES 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1. 2 LOW POPULATION Z0NE......................................... 5-1 5.1. 3 MAPS DEFINING UNRESTRICTED AREAS AND SITE B0UNDARY FOR RADIOACTIVE GASE0US AND LIQUID EFFLUENTS.................... 5-1 FIGURE 5.1-1 SITE AND EXCLUSION AREA B0VNDARY..................... 5-3 FIGURE 5.1-2 LOW POPULATION 20NE.................................. 5-5 FIGURE 5.1-3 LIQUID EFFLUENT DISCHARGE LOCATION................... 5-7 5.2 CONTAINMENT 5.2.1 CONFIGURATION........................ ....... .............. 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-9 O SEABROOK - UNIT 1 xii
n q , INDEX
-[G) 1 5.0 DESIGN FEATURES s /
SECTION PA3E 5.3 REACTOR CORE -> 5.3.1 FUEL ASSEMBLIES............................................. 5-9
-5.3.2 CONTROL R0D ASSEMBLIES...................................... 5-9 j 5.4 REACTOR COOLANT SYSTEM
+- 5.4.1 DESIGN PRESSURE AND TEMPERATURE.......?.........'.I.......... 5-9 5.4.2 V0LUME........................................',.............. ' 5~9 -
~,
5.5 METEOROLOGICAL TOWER L0 CATION................................~L . 5-9 s . i 5.6 FUEL STORAGE ' t 5.6.1: CRITICALITY................................................. 5-10 5.6.2 DRAINAGE.................................................... 5-10 ! 5.6.3 CAPACITY.................................................... 5-10 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-10
- . -~s
( TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.................. 5-11' 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY.............................................. 6-1 6.2 ORGANIZATION................................................ 6-1 1 6.2.1 0FFSITE................................................... 6-1 6.2.2. STATION STAFF............................................. 6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION............................... 6-2 i. 4
- FIGURE 6.2-2 STATION ORGANIZATION............................... 6-3
! TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION...................... 6-4 6.2.3' INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)
-Function........................................ ......... 6-5 j= Composition............................................... 6-5 Rest.onsibilities.......................................... 6-5 i Records...............................:................... 6-5 i
6.2.4 SHIFT TECHNICAL-ADVIS0R............................'....... . 6-5 s 6.3 TRAINING............................ .................. .... 6-5 i i j-n - x t SEABROOK - UNIT-1 . xiii s 4
++ -sy- ----em-g-- e -*m -=,,w-,---, ,- --o-,-,- y- , ace,-, - ,. w $e 4 mr w a = ~ r -n =*-=-e wew,--w---*, -
.--e v,~,wm -me,,,ww, -~w---+wm e- ---~+---~w -,w,-
6.0 ADMINISTRATIVE CONTROLS SECTION PAGE 6.4 REVIEW AND AUDIT............................................ 6-6 6.4.1 STATION OPERATION REVIEW COMMITTEE (50RC) Function.................................................. 6-6 Composition............................................... 6-6 Alternates................................................ 6-6 Meeting Frequency........... ............................. 6-6 Quorum.................. ................................. 6-6 Responsibilities.......................................... 6-6 Records................................................... 6-8 6.4.2 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC) Function.................................................. 6-8 Composition............................................... 6-8 Alternates................................................ 6-8 Consultants............................................... 6-8 Meeting Frequency......................................... 6-9 Quorum.................................................... 6-9 Review.................................................... 6-9 Audits................................... ................ 6-9 Records................................................... 6-11 6.5 REPORTABLE EVENT ACTI0N..................................... 6-11 6.6 SAFETY LIMIT VIOLATION...................................... 6-11 6.7 PROCEDURES AND PR0 GRAMS..................................... 6-12 6.8 REPORTING REQUIREMENTS 6.8.1 ROUTINE REP 0RTS.......................... ................ 6-14 Startup Report...................................... ..... 6-14 Annual Reports............................................ 6-15 Annual Radiological Environmental Operating Report........ 6-15 Semiannual Radioactive Effluent Release Report..... ...... 6-17 Monthly Operating Reports................................. 6-18 Radial Peaki ng Factor Limit Report. . . . . . . . . . . . . . . . . . . . . . . . 6-18 6.8.2 SPECIAL REP 0RTS........................................... 6-19 6.9 RECORD RETENTION.................................... ...... 6-19 6.10 RADIATION PROTECTION PR0 GRAM............................... 6-20 0 SEABROOK - UNIT 1 xiv
._...p. _
i INDEX' . 6.0 ADMINISTRATIVE CONTROLS f F SECTION PAGE 6.11' HIGH RADIATION AREA............................'............ 6-20 6.12 PROCESS-CONTROL PROGRAM (PCP).............................. 6-21 6.13 0FFSITE DOSE CALCULATION MANUAL (00CM)..................... 6-22 < 6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID >
?
RADWASTE TREATMENT SY5TEMS................................. 6-23 o 4 I , i l l l l i l l l l l I l i
- l
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i v SEABROOK - UNIT 1 xy w e <,,,,--nn,g---,~,,,.,g,- _ . , m--.y w--mmm-------,,---- y ., , - ,- - , , _e-- ---ner-,mrm--.---w---m---,,
1.0 DEFINITIONS ,/ 6 i .LI 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. 3 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 us 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 i including the sensors and alarm, interlock and/or trip functions and may be ' performed by any series of sequential, overlapping, or total channel steps l 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
' indications and/or status derived from independent instrument channels measuring.the same parameter.
v) SEABROOK - UNIT 1 1-1
l DEFINITIONS CONTAINMENT INTEGRITY g. O 1.7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required 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.
~
- b. All equipment hatches are 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
- e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-ring 3) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. CORE ALTERATION 1.9 CORE ALTERATION 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 ALTERATION shall not preclude completion of movement of a component to a safe conservative position. DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT 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 NRC Regulatory Guide 1.109, Revision 1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I." E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample with half-lives greater than 10 minutes. SEABROOK - UNIT 1 1-2
DEFINITIONS O ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel _ to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting.and sequence loading delays where applicable. FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.. GASE0US RADWASTE TREATMENT SYSTEM 1.14 A GASE0US RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases 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 q 1.15 IDENTIFIED LEAKAGE shall be: (V 4
- 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. 1 MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
MEMBER (S)'0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees i of the licensee, its contractors, or vendors. Also excluded from this category 1 are persons who enter the site to service equipment or to make deliveries. I This category does include persons who use portions of the site for recre-O) i
\.J ational, occupational, or other purposes not associated with the plant.
l SEABROOK - UNIT 1 1-3
DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) sha.1 contain in Part A the radiclogical effluent sampling and analysis program and radiological environ-mental monitoring program. Part B of the 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. OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILIT'/ 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 to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - MODE 1.20 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 , 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. PRESSURE B0UNDARY LEAKAGE 1.22 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 1.23 The PROCESS CONTROL PROGRAM (FCP) 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 wastr will be accomplished in such a way as to assure compliance with 10 CFR P r.s 20, 61, and 71 and Federal and State Regulations, burial ground iequrements, and other require-ments governing the disposal of radioactive waste. PURGE - PURGING 1.24 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. SEABROOK - UNIT 1 1-4
DEFINITIONS ( (/ QUADRANT POWER TILT RATIO 1.25 -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 1.26 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the-channel sensor until loss of stationary gripper coil voltage. REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. CONTAINMENT ENCLOSURE BUILDING INTEGRITY kJ 1.29 CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall exist when:
- a. Each door in each acce.s opening is closed except when the access opening is being used for normal transit entry and exit,
- b. The Containment Enclosure Filtration System is OPERABLE, and
- c. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
SHUTDOWN MARGIN j 1.30 SHUTOOWN 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 I worth which is assumed to be fully withdrawn. SITE BOUNDARY 1.31 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. J SEABROOK - UNIT 1 1-5
DEFINITIONS SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. SOLIDIFICATION 1.33 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. SOURCE CHECK 1.34 A SOURCE CHECK shall be the qualitative assessement of channel response when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.35 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.
THERMAL POWER 1.36 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE OPERATIONAL TEST 1.37 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 1.38 UNIDENTIFIED LEAKAGE shall be all luakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. UNRESTRICTED AREA 1.39 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY acce.s 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 B0UNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. SEABROOK - UNIT 1 1-6
'OEFINITIONS VENTILATION EXHAUST TREATMENT' SYSTEM 1.40 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and j installed to reduce gaseous radi_oiodine 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 particu-lates 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 1.41 VENTING shall be the controlled process of discharging air cr 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 provided or required during VENTING. Vent, used in system names, does not imply. a VENTING process. v SEABROOK - UNIT 1 1-7
TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours. D At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. N.A. Not applicable. P Completed prior to each release. TABLE 1.2 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, k THERMAL POWER
- TEMPERATURE eff
- 1. POWER OPERATION > 0.99 > 5% > 350 F
- 2. STARTUP i 0.99 -< 5% 5 350 F
- 3. HOT STANDBY < 0.99 0 5 350 F
- 4. HOT SHUTDOWN < 0.99 0 350 F > T > 200 F avg
- 5. COLD SHUTDOWN < 0.99 0 < 200 F
- 6. REFUELING ** < 0.95 0 5140F
- Excluding decay heat.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
O SEABROOK - UNIT 1 1-8
A..M-4%- 74A- m. _ ____ , _ , , , _ m4 a m J= J ___ ._..2,_ l i 4 l 4 I i
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SECTION 2.0 SAFETY LIMITS 4 AND ' LIMITING SAFETY SYSTEM SETTINGS 4 I l 1 .i I 4 i l 4 i i
p 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS - \)-
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2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the. highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 for four-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 hour, and comply with the require-ments of Specification 6.6. 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 H0T STANDBY with the Reactor Coolant System pressure within its limit within 1 hour, and comply with the requirements of Specification 6.6. 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 6.6. O SEABROOK - UNIT 1 2-1
680 UNACCEPTABLE OPERATION 660 % %
$00pS N %
PS w 640 i
~%
l ! N, i
% 1 i :
e i m w% w
~g 600
' l
%h.m 3
X
\_A ACCEPTABLE 5e0 OPERATION
' \
g 560 540 520 0.0 0.20 0.40 0.60 0.80 1.00 1.20 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY ' LIMIT - FOUR LOOPS IN OPERATION O 1 SEABROOK - UNIT 1 2-2
m SAFETY LIMITS AND 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 Value column of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or Q 2. 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.
Equation 2.2-1 Z + R + S < TA Where: Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance)'of Table 2.2-1 for the affected channel. a I SEABROOK - UNIT 1 2-3
TABLE 2.2-1 M REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0INTS ii; E SENSOR i si! TOTAL ERROR ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE i FUNCTIONAL UNIT Manual Reactor Trip N.A. N.A. N.A. N.A. N.A. E 1.
~ 2. Power Range, Neutron Flux
- a. High Setpoint 7.5 4.56 0 $109% of RTP* $111.1% of RTP*
- b. Low Setpoint 8.3 4.56 0 525% of RTP* $27.1% of RTP*
- 3. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with High Positive Rate i time constant i time constant 12 seconds 12 seconds l
- 4. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with m High Negative Rate i time constant i time constant i 12 seconds 12 seconds
- 5. Intermediate Range, 17.'O 8.41 0 $25% of RTP* $31.1% of RTP*
Neutron Flux
- 6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.6 x 10s cps
- 7. Overtemperature AT 6.5 3.31 1.04** See Note 1 See Note 2
+0.47**
- 8. Overpower AT 4.8 1.43 0.12 See Note 3 See Note 4
- 9. Pressurizer Pressure - Low 3.1 0.71 1.69 11945 psig 11,935 psig
- 10. Pressurizer Pressure - High 3.1 0.71 1.69 52385 psig $2,395 psig
*RTP = RATED THERMAL POWER
**The sensor error for T avg is 1.04 and the sensor error for Pressurizer Pressure is 0.47. "As measured" sensor errors may be used in lieu of either or both of these values, which then must be summed to deter-mine the overtemperature AT total channel value for S.
9 9 9
O ~r S-u TABLE 2.2-1 (Continued) M REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS is B SENSOR S TOTAL ERROR
, FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE E 11. Pressurizer Water Level - High 8.0 2.18 1.82 192% of instrument $93.8% of instrument
-Q . span span g 12. Reactor Coolant Flow - Low 2.5 1.87 0.6 193% of loop 189.4% of. loop design flow * ' design flow *
- 13. Steam Generator Water 17.0 15.28 1.76 >21.6% of narrow >15.9% of narrow Level Low - Low range instrument range inst'rument span span
- 14. Undervoltage - Reactor 15.0 1.39 0 110,200 volts -19,822 volts Coolant Pumps
- 15. Uaderfrequency - Reactor 2.9 0 0 255.5 Hz 155.3 Hz s, Coolant Pumps d,
Turbine Trip 16.
- a. Low Fluid Oil Pressure N.A. N.A. N.A. 1500 psig 1450 psig
- b. Turbine Stop Valve N.A. N.A. N.A. 21% open 11% open Closure
- 17. Safety Injection Input N.A. N.A. N.A. N.A. N.A.
from ESF
- Loop design flow = 95,700 gpm
TABLE 2.2-1 (Continurd) I REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E 8 SENSOR TOTAL ERROR
' TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)
C 5
- 18. Reactor Trip System Interlocks
- a. Intermediate Range N.A. N.A. N.A. 21 x 10 10 amp 16 x 10 11 amp Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- 1) P-10 input N.A. N.A. N.A. 510% of RTP* $12.1% of RTP*
- 2) P-13 input N.A. N.A. N.A. 510% RTP* Turbine $12.3% RTP* Turbine m Impulse Pressure Impulse Pressure di Equivalent Equivalent
- c. Power Range Neutron N.A. N.A. N.A. 550% of RTP* 152.1% of RTP*
Flux, P-8
- d. Power Range Neutron N.A. N.A. N.A. 120% of RTP* 122.1% of RTP*
Flux, P-9
- e. Power Range Neutron N.A. N.A. N.A 210% of RTP* 27.9% of RTP*
Flux, P-10
- f. Turbine Impulse Chamber N.A. N.A. N.A. 510% RTP* Turbine $12.3% RTP* Turbine Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent
- 19. Reactor Trip Breakers N.A. N.A. N.A N.A. N.A.
- 20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A.
Logic
*RTP = RATED THERMAL POWER O O #
f' em O t U TABLE 2.2-1 (Continued) y TABLE NOTATIONS m y NOTE 1: OVERTEMPERATURE AT AT y, (y 7sg) $ ATg {K 2 -K 2 [T (y 7sg) - T'] + K3(P - P') -~f t (AI)} 5 --e - Where: AT = Measured AT by RTD Manifold Instrumentation; l y
= Lead-lag compensator on measured AT; It, T2 = Time constants utilized in lead-lag compensator for AT, T1 18s, T2 5 3 s; yf 3
= Lag compensator on measured AT; m T3 = Time constants utilized in the lag compensator for AT,-T3 = 0 s; AT, = Indicated AT at RATED THERMAL POWER; K1 = 1.0995; K2 = 0.0112/ F; I + b 1*I b
= The function generated by the lead-lag compensator for T -avg 5 dynamic compensation;
=
T4, Is Time constants utilized in the lead-lag compensator for T,yg, T4 1 33 s, Ts i 4 s; T = Average temperature, F;
=
1 + TsS Lag compensator on measured T,yg;
=
Ts Time constant utilized in the measured T,yg lag compensator, Ts = 0 s;
TABLE 2.2-1 (Continued) m TABLE NOTATIONS j
- o 8
7: NOTE 1: (Continued) T' at RATED THERMAL POWER); 5 588.5 F (Nominal T avg g Z K3 = 0.000519/psig; P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure); 5 = Laplace transform operator, s 1; and f2 (AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains tc be selected based on measured instrument response during plant startup tests so that: 9bbetween - 35% and + 8%, ft (AI) = 0, where q t "d 9 are percent RATED THERMAL l } (1) For q t b POWER in the top and bottom halves of the core respectively, and qt *9 b is total THEMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q t -9 b exceeds - 35%, the AT Trip Setpoint shall l be automatically reduced by 1.09% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of qt 9bexceeds + 8%, the AT Trip Setpoint shall be automatically reduced by 1.00% of its value at RATED THERMAL POWER. NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.0% of AT span. O O O
) \
0 . j , TABLE 2.2-1 (Continued) l [ TABLE NOTATIONS (Continued) E o i 7 NOTE 3: OVERPOWER AT
) fU AT (y Ts3) 1 AT, M4 -K 3 (1 73) (y 7s3) T - Ks U (y T6 3) - T9 - f '(AI)}
2 I ~ I Where: AT = As defined in Note 1, 1 = As defined in Note 1, ! ti,12 = As defined in Note 1, 1 1 1 = As defined in Note 1, 1+T 3S , i ! '? T3
= As defined in Note 1, AT, = As defined in Note 1, K4 = 1.09, t
) K3 = 0.02/ F for increasing average temperature and 0 for decreasing average j temperature, y{7 73
=
The function generated by the rate-lag compensator for-T,yg dynamic , comnensation,
=
T7 Time constants utilized in the rate-lag compensator for T,yg, 17 > 10 s, 1 = As defined in Note 1, 7 Ts3
= As defined in Note 1,
! Is
TABLE 2.2-1 (Continued) vs 9 TABLE NOTATIONS (Continued) E 8
- NOTE 3: (Continued)
E Ks
= 0.00128/ F for T > T" and Ks = 0 for T $ T", ~ T = As defined in Note 1, 't'.
T" = Indicated T ag at RATED THERMAL POWER (Calibration temperature for AT' instrumentation, 5 588.5 F), S = As defined in Note 1, and f 2(AI) = 0 for all AI. 7 NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than a; 3.4% of AT span. l l 1 l 9 O O
1 n- + s- 4 _ -- -,,sa_- _ --e a 4 - - , 2-.a - - _ w w .- . _ _ - - - I i. BASES FOR SECTION 2.0 i i SAFETY LIMITS AND i LIMITING SAFETY SYSTEM SETTINGS i Y l b d i f I i l i f u....___.__ -- ---- _-_,___ ______ _,
O NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. O O
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 that 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. Operation 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 through the W-3 (R-Grid) correlation. T.he W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. 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. g The minimum value of the DNBR during steady-state operation, normal This (V ) operational transients, and anticipated transients is limited to 1.30. value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. 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 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
These curves are based on an enthalpy hot channel factor, F H, f 1.55 and a reference cosine with a peak of 1,55 for axial power shape. An allowance is included for an increase in F H at reduced power based on the expression: F g = 1.55 [1+ 0.2 (1-P)] 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, assuming the -axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip. t When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent ( ; with core Safety Limits. x._/ SEABROOK - UNIT 1 B 2-1 1
SAFETY LIMITS BASES 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, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants, which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is, therefore, consistent with the design criteria and associated Code requirements. The entire RCS is hydrotested at 125% (3110 psig) of design pressure to demonstrate integrity prior to initial operation. O O SEABROOK - UNIT 1 B 2-2
2.2 LIMITING SAFETY SYSTEM SETTINGS
$n) v 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 aild 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. The Setpoint for a Reactor Trip System or interlock function is-considered to be adjusted consistent with the nominal value when the "as-measured" Setpoint is within the band allowed for calibration accuracy.
~
To accommodate the instrument drift assumed to 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 Setpoints less conservative than the Trip Set-point but within the Allowable Value is acceptable, since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY ~of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as-measured" deviation from the specified calibration Q point for rack and sensor components in conjunction with a statistical combin-ation of the other uncertainties of the instrumentation to measure the process 5 ky variable and the uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor,'and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as-measured" devia-tion, in percent span, for the affected channel from.the specified Trip Set-point. S or Sensor Error is either the "as-measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. The methodology to derive the Trip Setpoints is based on' combining all i of the uncertainties in the channels. Inherent to the determination of the l Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of'the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. s
/
SEABROOK - UNIT 1 B 2-3 1
i LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) The 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 manitors 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 ass;med in the safety analysis to enhance the overall reliability of the Reactor Trip Systam. 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 Systera. Manual Reactor Trip ( The Reactor Trip System includes manual Reactor trip capability. Power Range, 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 Setpoiat 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 protection during power operations to mitigate the consequences of a reactivity excursion from all power levels. The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL PCWER) and is automatically reinstated below the P-10 Setpoint. Power Range, Neutron Flux, High Rates The Fower Range Positive Rate trip provides p"otection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power. The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power, a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than 1.30. O SE/.BRC0K - UNIT 1 B 2-4
\ '
\. !'
'( , ,,
\.
,e s LIMITING SAFETY S S EM SETTINGS '
i-(v) BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) Intermediate and Source Range, Neutron Flux '
, 1 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 redun' dant protection to the Low Setfoint trip of the Power Range, Neutron Flux channels. The Source Range channelite will initiate a Reactor trip at about 105 countspersecondunless' manual 19 blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent.to approximately 251 of RATED THERMAL F0WER unless' manually blocked when P-10 becomes active. ' Overtemperature AT I The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to piping transit delays from the ccre to the-temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pre.ssure trips. The Set-( point 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 distribution. With s normal axial power distribution, this Reactor trip limit is always below the , core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than s design,.as indicated by the difference between t'op and bottom power range nuclear detectors, the Reactop trip.is automatically reduced according to the . notations in Table 2.2-1. ' ~ t- : Overpower AT - s s 3 The Overpower AT trip provides assurance of fuel integrity (e.g., no !s fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range,for Overtemperature AT trip, and provides a backup to the High Neutron, Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors. to ensure that the allowable heat genera i tion rate (kW/ft) is not exceedea. The Overpower AT trip provides protection to mitigate the consequences cf various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."
)
\
'~
s , s SEABROOK - UNIT 1 .B 2-5
LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) 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 that 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 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7. The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure. Pressurizer Water Level The Pressurizer High Water Level 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 approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, the Pressurizer High Water Level trip is automatically reinstated by P-7. Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consdquences 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 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% 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 50% 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 ?-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 Emergency Feedwater System. SEABROOK - UNIT 1 B 2-6 L
i
$i i
LIMITING SAFETY SYSTEM SETTINGS V BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)- Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced conlant flow. The specified Setpoints assure a Reactor trip' signal is generated 3 before the Low Flow Trip Setpoint is reached. Time delays are incorporated in } the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump ous circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set I so that the time required for a signal to reach the Reactor trip breakers after'the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second. On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse enamber pressure at approximately 10% of full power equivalent); and on increasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are reinstated automatically by P-7. l Q Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor i trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 20% of RATED THERMAL POWER); and on increasing power, the Reactor trip from the Turbine trip is reinstated automatically by P-9. Safety Injection Input from ESF- i If a Reactor trip has not already bden 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 that initiate a Safety Injection signal are shown in Table 3.3-3. Reactor. Trip System 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). On dec-reasing power, Source Range Level trips are automatically reactivated and high voltage is restored.
SEABROOK - UNIT 1 B 2-7
LIMITING SAFETY SYSTEM SETTINGS BASES O 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) Reactor Trip System Interlocks (Continued) P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure, and pressurizer high level. On decreasing power, the above listed trips are automatically blocked. P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above trip. P-9 On-increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip. P-10 On increasing 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. Provides input to P-7. P-13 Provides input to P-7. O SEABROOK - UNIT 1 B 2-8
g_. .. O i SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND O SURVEILLANCE REQUIREMENTS O
3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ( 3 Q) 3/4.0 APPLICABILITY LIMITING CONDITION 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 is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
- a. At least HOT STANDBY within the next 6 hours,
- b. At least HOT SHUTDOWN within the following 6 hours, and
- c. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time j limits as measured from the time of failure to meet the Limiting Condition for l Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not applicable in MODE 5 or 6. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. 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. C\ t v) SEABROOK - UNIT 1 3/4 0-1
APPLICABILITY SURVEILLANCE REQUIREMENTS 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. A maximum allowable extension not to exceed 25% of the surveillance interval, but
- b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated _in the individual specifications. 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 Surveillance Requirement (s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice 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 Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i);
O SEABROOK - UNIT 1 3/4 0-2
APPLICABILITY '( O G SURVEILLANCE REQUIREMENTS 4.0.5 (Continued)
- b. Surveillance intervals specified in Section XI of the ASME Boiler and. Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by 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 inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or evcry 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 g- required frequencies for performing inservice inspection and. testing g activities; V
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
l l l O ) SEABROOK - UNIT 1 3/4 0-3 l
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - T GREATER THAN 200 F avg LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN for four-loop operation shall be greater than or equal to 3.8% Ak/k in MODES 1,' 2, and 3 and 1.3% Ak/k in MODE 4. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than the limiting value, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limiting value: V a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours 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 kgff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the' limits of Specification 3.1.3.6;
- c. When in MODE 2 with k eff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation.above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and i
*See Special Test Exceptions Specification 3.10.1. 'SEABROOK - UNIT 1 3/4 1-1 l
REACTIVITY CONTROL SYSTEMS BORATION CONTROL SHUTDOWN MARTIN - T GREATER THAN 200 F avg SURVEILLANCE REQUIREMENTS 4.1.1.1.1 (Continued)
- e. When in MODE 3 or 4, at least once per 24 hours 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 i
- 6) Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within~t 1% Ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.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 each fuel loading. O SEABROOK - UNIT 1 3/4 1-2
REACTIVITY CONTROL SYSTEMS BORATION CONTROL SHUTDOWN MARGIN - T,yg 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 1.2% Ak/k.- Additionally, the Reactor Coolant System boron concentration shall be, greater than or equal to 2000 ppm boron when the reactor coolant loops are in a drained condition. APPLICABILITY: MODE 5. j ACTION: With the SHUTDOWN MARGIN less .than 1.2% Ak/k or the Reactor Coolant System boron concentration less than 2000 ppm boron, immediately initiate and continue boration at greater. than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN and boron concentration are restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The. SHUTDOWN MARGIN and boron concentration shall be determined to be greater.than or equal to 1.2% Ak/k:
- a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours 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 hours 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.
Ch U SEABROOK - UNIT 1 3/4 1-3
REACTIVITY CONTROL SYSTEMS B0 RATION CONTROL MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:
- a. Less positive than 0 Ak/k/ F for the all rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition; and
- b. Less negative than -4.2 x 10 4 Ak/k/ F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.
APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only**. Specification 3.1.1.3b. - MODES 1, 2, and 3 only**. ACTION:
- a. With the MTC more positive than the limit of Specification 3.1.1.3a.
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 than 0 Ak/k/ F within 24 hours or be in HOT STANDBY within the next 6 hours.
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 6.8.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
- b. With the MTC more negative than the limit of Specification 3.1.1.3b.
above, be in HOT SHUTDOWN within 12 hours.
*With k eff greater than or equal to 1.
- See Special Test Exceptions Specification 3.10.3.
SEABROOK - UNIT 1 3/4 1-4
REACTIVITY CONTROL SYSTEMS f . BORATION CONTROL-MODERATOR TEMPERATURE COEFFICIENT SURVEILLANCE REQUIREMENTS l 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 5% of~
RATED THERMAL POWER, after each fuel loading; and
- b. The MTC shall be measured at any THERMAL POWER and compared to
-3.3 x 10 4 Ak/k/*F (all rods withdrawn, RATED THERMAL POWER condi-tion) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.3 x 10 4 Ak/k/ F, the MTC shall be remeasured, and compared to the E0L MTC limit of Specification 3.1.1.3b., at least once per 14 EFPD during the remainder of the fuel cycle.
l i l ! SEABROOK - UNIT 1 3/4 1-5
REACTIVITY CONTROL SYSTEMS BORATION CONTROL MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (T*V9) shall be greater than or equal to 551 F. APPLICABILITY: MODES 1 and 2* **. ACTION: With a Reactor Coolant System operating loop temperature (Tavg) less than 551 F, restore T 3yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 551 F:
- a. tiithin 15 minutes prior to achieving reactor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T avg is less than 561 F with the T avg-Tref Deviation Alarm not reset.
l l l l t l
*With k eff greater than or equal to 1. O
! **See Special Test Exceptions Specification 3.10.3. I SEABROOK - UNIT 1 3/4 1-6 l
,,q REACTIVITY CONTROL SYSTEMS f
v
) 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boren 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 tanks via either a boric acid l transfer pump or a gravity feed connection and a charging pump to l
' the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a. 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 3.1.2.5b. is OPERABLE.
APPLICABILITY: MODES 4, 5, and 6. ACTION: (N
/"
V) 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 REQUIREMENTS 4.1.2.l At least one of the above required flow paths shall be demonstrated OPERABLE 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 other-wise secured in position, is in its correct position. O SEABROOK - UNIT 1 3/4 1-7
REACTIVITY CONTROL SYSTEMS 80 RATION SYSTEMS. FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:
- a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS), and
- b. Two flow paths from the refueling water storage tank via charging pumps to the RCS.
APPLICABILITY: MODES 1, 2, and 3* ACTION: With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection f1'w paths to the RCS to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200 F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths 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;
- b. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal; and
- c. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.
- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first.
SEABROOK - UNIT 1 3/4 1-8
r3 REACTIVITY CONTROL SYSTEMS i ' V )- B0 RATION SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION-3.' 1. 2. 3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power. source. APPLICABILITY: MODES 4,.5, and 6.
. ACTION:
With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency' power source, suspend all operations involving CORE ALTERATIONS or. positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow,.that a differential pressure across the pump i of greater than.or equal to 2480 psid is developed when tested pursuant to ,V Specification 4.0.5. 4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position within 4 hours after entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS' cold legs decreas-ing below 325 F, whichever comes first, and at least once per 31 days there-after, except when the reactor vessel head is removed.
*An inoperable pump'may b_e energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured-in the closed position.
, \ SEABROOK - UNIT 1 3/4 1-9
REACTIVITY CONTROL SYSTEMS B0 RATION SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3.* ACTION: With only one charging pump OPERABLE, restore at least two charging punips to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200 F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2480 psid is developed when tested pursuant to Specification 4.0.5. RThe provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first. SEABROOK - UNIT 1 3/4 1-10
REACTIVITY CONTROL SYSTEMS-J BORATION SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources sh'all be OPERABLE:
~
- a. .A Boric Acid Storage System with:
1). A minimum contained borated water volume of.6,500 gallons,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65*F.
b'. The refueling' water' storage' tank (RWST) dith:
- 1) A minimum contained borated water volume of 24,500 gallons,
.2) A minimum boron concer.tration of 2000 ppm, and
- 3) A minimum solution temperature of 50*F.
APPLICABILITY: MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
-SURVEILLANCE-REQUIREMENTS-4.1.2.5 The above required borated water source.shall be demonstrated OPERABLE:
- a. At least once per 7 days.by:
-1) Verifying the boron concentration'of the water,
- 2) Verifying the contained borated water volume, and
- 3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST' temperature.
SEABROOK - UNIT 1 3/4 1-11
REACTIVITY CONTROL SYSTEMS B0 RATION SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2:
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water volume of 22,000 gallons,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A~ minimum solution temperature of 65 F.
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained borated water volume of 477,000 gallons,
- 2) A minimum boron concentration of 2000 ppm,
- 3) A minimum solution temperature of 50 F, and
- 4) A maximum solution temperature of 98 F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water. sources, restore the system to OPERABLE status within '72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200 F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours.
- b. With the RWST inoperable, restore the tank to OPERABLE status withih . hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
O SEABROOK - UNIT 1 3/4 1-12
REACTIVITY CONTROL SYSTEMS O BORATION SYSTEMS
.{)
BORATED WATER SOURCES - OPERATING i SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE: l a. At least once'per 7 days-by:
- 1) Verifying the boron concentration in the water, i
2)~ Verifying the contained 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. At least once per 24 hours by verifying the RWST temperature.
N i k i i c: I 4 4 I SEABROOK - UNIT 1 3/4 1-13
REACTIVITY CONTROL SYSTEMS BORON SYSTEMS ISOLATION OF UNB0 RATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 Provisions to isolate the Reactor Coolant System from unborated water sources shall be OPERABLE with:
- a. The Baron Thermal Regeneration System (BTRS) isolated from the Reactor Coolant System, and
- b. The Reactor Makeup Systems inoperable except for the capability of delivering up to the capacity of one Reactor Makeup Water pump to the Reactor Coolant System.
APPLICABILITY: MODES 4, 5, and 6 ACTION: With the requirements of the above specification not satisfied immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and, if within 1 hour the required SHUTDOWN MARGIN is not verified, initiate and continue boration at greater than or equal to 30 gpm of a solu-tion containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored and the isolation provisions are restored to OPERABLE. SURVEILLANCE REQUIREMENTS 4.1.2.7 The provisions to isolate the Reactor Coolant System from unborated water sources shall be determined to be OPERABLE at least once per 31 days by:
- a. Verifying that at least the BTRS outlet valve, CS-V-302, or the BTRS moderating heat exchanger outlet valve, CS-V-305, is closed and locked closed, and
- b. Verifying that power is removed from at least one of the Reactor Makeup Water pumps, RMW-P-16A or RMW-P-168.
O SEABROOK - UNIT 1 3/4 1-14
REACTIVITY CONTROL SYSTEMS
/
( I 3/4.1.3 MOVABLE CONTROL ASSEMBLIES
.V GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control' rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter
' demand position.
APPLICABILITY: ~ MODES 1* and 2*. Ar. TION:
- a. With'one or more full-length rods inoperable because of being immov-l able as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours.
- b. With one full-length rod trippable but inoperable due to causes l other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that
('~'s within 1 hour: (s_,) 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure 3.1-1. 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:
a) 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; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours; l l l /N k_,) *See Special Test Exceptions Specifications 3.10.2 and 3.10.3. SEABROOK - UNIT 1 3/4 1-15
REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 ACTION b.3 (Continued) c) A power distribution map is obtained from the movable incore detectors and Fq(Z) and F H are verified to be within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.
- c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue prov-ded that:
- 1. Within 1 hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and
- 2. The inoperable rods are restored to OPERABLE status within 72 hours.
- d. With more than one rod misaligned from its group step counter demand height by more than 12 steps (indicated position), be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS I- 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours, except during time intervals when the rod position deviation monitor is inoperable; then verify the group positions at least once per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one l direction at least once per 31 days. O SEABROOK - UNIT 1 3/4 1-16
I l l I l TABLE 3.1-1 l
,-,~ , ,- ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH RdD Rod Cluster Control Assembly _ Insertion Characteristics l Rod Cluster Control Assembly Misalignment l- 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) ; 1 k a 't 4 i e SEABROOK - UNIT 1 3/4 1-17
REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within i 12 steps. APPLICABILITY: MODES I and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable, either:
- 1. Determine the position of the nonindicating rod (s) indirectly by the movable incore detectors at least once per 8 hours and immediately after any motion of the nonindicating 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 50% of RATED THERMAL POWER within 8 hours.
- b. With a maximum of one demand position indicator per bank inoperable, either:
- 1. Verify that all digital rod position indicators for the affected bank are OPERABLE and 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 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours, except during time intervals when the rod position deviation monitor is inoperable; then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours. O SEABROOK - UNIT 1 3/4 1-18
REACTIVITY CONTROL SYSTEMS -p MOVABLE CONTROL ASSEMBLIES ( POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully -
~
inserted. APPLICABILITY: MODES 3* **, 4* **, and 5* **. ACTION: With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers. SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agreo with the demand position indicators within 12 steps when exercised over g ) the full range of rod travel at least once per 18 months. v l l
.l l
l 1
*With the Reactor Trip System breakers in the closed position.
**See Special Test Exceptions Specification 3.10.5.
SEABROOK - UNIT 1 3/4 1-19
REACTIVITY CONTROL SYSTEMS MOVABLE C0t: TROL ASSEMBLIES ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
- a. T avg f r each loop greater than or equal to 551 F, and
- b. All reactor cochnt pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION: With 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 proceedirg to MODE 1 or 2. SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods measurement prior to reactor. criticality: shall be demonstrated through O
- 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 that could affect the drop time of those specific rods, and
- c. At least once per 18 months.
l l l l 1 O l SEABROOK - UNIT 1 3/4 1-20
~~s . REACTIVITY CONTROL SYSTEMS s- MOVABLE CONTROL ASSEMBLIES SHUTDOWN R0D INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.~3.5 All shutdown rods shall be fully withdrawn.
APPLICABILITY: MODES 1* and 2* **. ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either:
- a. Fully withdraw the rod, or
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 (A) Each shutdown rod shall be determined to be fully withdrawn:
- a. Within 15 minutes prior to withdrawal of any' rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
i 4 l [ \ *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
**With k eff g.'e ter than or equal to 1.
SEABROOK - UNIT 1 3/4 1-21
REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES CONTROL R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1. APPLICABILITY: MODES 1* and 2* **. ACTION: With the control banks inserted beyond the above insertion limits, except for-surveillance testing pursuant to Specification 4.1.3.1.2:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the above figure, or
- c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS O 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours, except during time intervals when_the rod insertion limit monitor is inoperable; then verify the individual rod positions at least once per 4 hours.
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With k eff greater than or equal to 1.
SEABROOK - UNIT 1 3/4 1-22
G 228 (0.30,228) (0.844,228)-
/- . /
200 '
/ BANK B ,
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/ l- 1 I 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 l /O I i ) R0D BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR-LOOP OPERATION SEABROOK - UNIT 1 3/4 1-23 L .
n 3/4.2 POWER DISTRIBUTION LIMITS-I l (/ 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target bands (flux difference units) about the target flux difference:
- a. 1 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU;
- b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTU; and
- c. + 3%, -12% for each subsequent cycle.
~
The indicated AFD may deviate outside the above required target band at greater than or equal to'50% but less than 90% 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 hour during the previous.24 hours. The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour during the previous'24 hours. APPLICABILITY: MODE 1, above 15% 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 90% of RATED THERMAL POWER, within 15 minutes either:
- 1. Restore the indicat- ' " r) to within the target band limits, or
- 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
- b. With the indicated AFD outside of the above required target band for more than l' hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:
- 1. THERMAL POWER to let, than 50% of RATED THERMAL POWER within 30 minutes, and
- 2. The Power Range Neutron Flux * ** - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
*See Special Test Exceptions Specification 3.10.2.
** Surveillance testing of the Power Range Neutron Flux Channel may be performed V 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 hours' SEABROOK - UNIT 1 3/4 2-1
i POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION
3.2.1 ACTION
(Ccntinued)
- c. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target 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 50% 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 15% and 50% of RATED THERMAL f0WER.
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.
**(Coatinued) operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
SEABROOK - UNIT 1 3/4 2-2
7.- s i
\ '\wl 120 alJ I dE5 g
58 o % Q-m N 100 U,NACCEP,rABLE OPERATION UNACCEPTABLE OPERATION i i l
' (-11,90)
(11,9 0) 5 / \ O )( )(
' 80
_J
/
if
\
C / T
\
z g [ ACCEPTABLE OPERATION
\
UJ h 60 \
) h t*\
( W L
%/
i q l-31,50) (31,5 0) a g 40 o N _ 20 1 1 0
-50 -40 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE ( AI)%
FIGURE 3.2-1 ( ) AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF 'O' RATED THERMAL POWER fEABR00K - UNIT 1 3/4 2-3 l
I POWER DISTRIBUTION LIllITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 Fq (Z) shall be limited by the following relationships: F0 (Z) 5 2.32 K(Z) for P > 0.5 P Fq (Z) 5 (4.64) K(Z) for P $ 0.5 THERMAL POWER , and Where: P = RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: MODE 1. ACTION: With F (Z) exceeding its limit: q
- a. Reduce THERMAL POWER at least 1% for each 1% qF (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Set-points have been reduced at least 1% for each 1% F q(Z) exceeds the limit, and
- b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be-increased, provided Fq(Z) is demonstrated through incore mapping to be within its limit.
O SEABROOK - UNIT 1 3/4 2-4
I C )N
\
1.2 (6.0,1.0) 1.0
, (10.0.2.94) 1
\
_ 0.8 ,
\
y U l L_ a LL Q Lu (12.0,0 65) CN b 0.6 ! i _J q) I a - - (r O
.2
' O.4 S
M C.2 - 0.0 2 4 6 8 10 12 CORE HEIGHT (FT) FIGURE 3.2-2 8
)
("'N K(Z) - NORMALIZED F q(Z) AS A FUNCTION OF CORE HEIGHT SEABROOK - UNIT 1 3/4 2-5
i i POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Fg SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F xy shall be evaluated to determine if F (Z) is within its limit by: 9
- a. Using the movable incore detectors'to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER,
- b. Increasing the measured F xy component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, C
- c. Comparing the F computed (F*Y) obtained in Specification 4.2.2.2b.,
above,-to: *Y
- 1) The F limits for RATED THERMAL POWER (F ) for the appropriate xy measured core planes given in Specification 4.2.2.2e. and f. ,
below, and
- 2) The relationship:
F =FRTP [1+0.2(1-P)], x L is the limit for fractional THERMAL POWER operation Where F P express d as a function of F and P is the fraction of RATED THERMAL POWER at which F xy was measured.
- d. Remeasuring F xy according to the following schedule:
- 1) When F is greater than the F limit for the appropriate measured core plane but less than the F relationship, additional P
power distribution maps shall be taken dF comparea to F and F either: xy a) Within 24 hours after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F was last determined, or
*Y b) At least once per 31 Effective Full-Power Days (EFPD),
whichever occurs first. O SEABROOK - UNIT 1 3/4 2-6 I - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
POWER DISTRIBUTION LIMITS (
.v/ -HEATFLUXHOTCHANNELFACTOR-Fg SURVEILLANCE REQUIREMENTS 4.2.2.2d. (Continued)
- 2) When the F is less than or equal. to the F xRTP limit for the appropriate measured core plane, additional power distribution maps shall be taken and F compared-to F and F at least x
once per 31 EFPD.
- e. The F RTP xy limits for RATED THERMAL POWER (Fxy ) shall be provided for all core planes containing Bank "D" control rods and a;'. unrodded core planes in a Radial Peaking Factor Limit Report per Specifica-tion 6.8.1.6;
- f. The F xy limits of Specification 4.2.2.2e., above,.are not applicable in the following core planes. regions as measured in percent of core height from the bottom of the fuel:
- 1) Lower core region from 0 to 15%, inclusive,
- 2) ' Upper core region from 85 Lo 100%, inclusive,
- 3) Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%,
and 74.9 1 2%, inclusive, and
- 4) Core plane regions within i 2% of core height (i 2.88 inches) about the bank demand position of the Bank "D" control rods.
- g. With F exceeding Fx , the effects of Fxy n Fq (Z) shall be evaluated to determine if F 9 (Z) is within its limits.
4.2.2.3 When Fq (Z) is measured for other than F xy determinations, an overall measured qF (Z) shall be obtained-from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. SEABROOK - UNIT 1 3/4 2-7
POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 F H shall be less than 1.49 [1.0 + 0.2 (1-P)]. Where: P= THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1. ~ { ACTION: _ With F H exceeding its limit:
- a. Within 2 hours reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.
- b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by ACTION a.,
above; THERMAL POWER may then be increased, provided F H is demonstrated through incore mapping to be within its limit. SURVEILLANCE REQUIREMENTS o 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 F shall be demonstrated to be within its limit prior to operation H above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:
- a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER.
- b. Using the measured value of F H which does not include an allowance for measurement uncertainty.
O SEABROOK - UNIT 1 3/4 2-8
POWER DISTRIBUTION LIMITS
/ \ / 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *. ACTION: With the QUADRANT POWER TILT RATIO determined to exceed 1.02:
- a. Within 2 hours reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% 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 hours.
- b. Within 24 hours and every 7 days thereafter, verify that F q(Z) (by F
yy evaluation) and F H ure within their limits by performing Surveil-lance Requirements 4.2.2.2 and 4.2.3.2. THERMAL POWER and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3. O v SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and
- b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by.using the movable incore: detectors to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours by either:
- a. Using the four pairs of symmetric thimble locations or
- b. Using the movable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.
t O t
*See Special Test Exceptions Specification 3.10.2.
SEABROOK - UNIT 1 3/4 2-9
l l POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the the following limits:
- a. Reactor Coolant System Tavg, < 594.3 F
- b. Pressurizer Pressure, > 2205 psig*
- c. Reactor Coolant System Flow, >_ 391,000 gpm**
APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours. Additionally, Reactor Coolant System flow shall be demonstrated to be within its limit prior to operation above 75% of RATED THERMAL POWER after each fuel loading. The provisions of Specifica-tion 4.0.4 are not applicable for the verification that Reactor Coolant System flow is within its limit. 4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months. 4.2.5.3 The RCS total flow rate shall be determined by precision heat balance measurements at least once per 18 months.
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%
of RATED THERMAL POWER.
** Includes a 2.1% flow measurement uncertainty.
SEABROOK - UNIT 1 3/4 2-10
3/4.3 INSTRUMENTATION 3_/a.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System inst'rumentation 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. SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reac. tor 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 Requirements specified in Table 4.3-1. 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the
" Total No. of Channels" column of Table 3.3-1.
l l SEABROOK - UNIT 1 3/4 3-1
TABLE 3.3-1 M g REACTOR TRIP SYSTEM INSTRUMENTATION 8 g MINIMUM
, TOTAL N0. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Z 1. Manual Reactor Trip 2 1 2 1, 2 1 l - 2 1 2 3*, 4*, 5* 10
- 2. Power Range, Neutron Flux
- a. High Setpoint 4 2 3 1, 2 2#
- b. Low Setpoint 4 2 3 1###, 2 2#
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2#
High Positive Rate l l 4. Power Range, Neutron Flux, 4 2 3 1, 2 2# l High Negative Rate l M
- 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 w
E 6. Source Range, Neutron Flux
- a. Startup 2 1 2 2## 4
- b. Shutdown 2 0 1 3,4,5 5
- c. Shutdown 2 1 2 3* , 4* , 5* 10
- 7. Overtemperature AT 4 2 3 1, 2 6#
- 8. Overpower AT 4 2 3 1, 2 6#
- 9. Pressurizer Pressure--Low 4 2 3 1** 6# (1)
- 10. Pressurizer Pressure--High 4 2 3 1, 2 6# (1)
- 11. Pressurizer Water Level--High 3 2 2 1** 7#
A y- . _. Y' TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION E 8 ^ MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS 'TO TRIP OPERABLE MODES ACTION Z 12. Reactor Coolant Flow--Low ~ a. Single Loop (Above P-8) 3/ loop 2/ loop ~in 2/ loop in 1 7# any oper- each oper-ating loop ating loop
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 7#
below P-8) two oper- each oper-ating loops ating loop
- 13. Steam Generator Water 4/stm. gen.
2/stm. gen. 3/stm. gen. 1, 2 : 6# (1) Level--Low-Low 'in any oper- each oper-R ating stm. ating stm. gen. . gen.
- 14. Undervoltage--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus. 1** 6#
Pumps
- 15. Underfrequency--Reactor Coolant 4-2/ bus 2-1/ bus 2 on one bus 1** 6#
Pumps
- 16. Turbine Trip
- a. Low Fluid Oil Pressure 3 2 2 1*** 7#
- b. Turbine Stop Valve Closure 4 4 4 1*** 11#
- 17. Safety Injection Input from ESF 2 1 2 1, 2 9
- 18. Reactor Trip System Interlocks
- a. Intermediate Range Neutron Flux, P-6 2 1 2 2## 8
TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION E 8 ^ MINIMUM TOTAL N0. CHANNELS CHANNELS APPLICABLE i FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION l E Q 18. b. Low Power Reactor ~ Trips Block, P-7 P-10 Input 4 2 3 1 8 or P-13 Input 2 1 2 1 8
- c. Power Range Neutron Flux. P-8 4 2 3 1 8
- d. Power Range Neutron 4 2 3 1 8 Flux, P-9 R
- e. Power Range Neutron Y Flux, P-10 4 2 3 1 8
+
- f. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8
- 19. Reactor Trip Breakers 2 1 2 1, 2 9, 12 2 1 2 3*, 4*, 5* 10
- 20. Automatic Trip and Interlock 2 1 2 1, 2 9 Logic 2 1 2 3*, 4*, 5* 10 e O O
(' j_s TABLE 3.3-1 (Continued) h
/ TABLE NOTATIONS
*When the Rea:: tor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.
** Trip function automatically blocked or bypassed below the P .7 (At Power)
Setpoint.
*** Trip function automatically blocked below the P-9 (Reactor Trip / Turbine Trip Interlock) Setpoint.
'#The provisions of Specification 3.0.4 are not applicable.
##Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
~
(1) These channels also provide inputs to ESFAS. Comply with applicable MODES and ACTION statements of Specification 3.3.2 for any portion of the channel required to be OPERABLE by Specification 3.3.2. ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel l to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6~ hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours,
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel.may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1, and
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the .
QUADRANT POWER TILT RATIO is monitored at least ~once per ) 12 hours per Specification 4.2.4.2. > v . SEABROOK - UNIT 1 3/4 3-5
TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 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 Setpoint, and
- b. Above the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% ef RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER. ACTION 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, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers, suspend all operations involving positive reactivity changes and verify that valve RMW-V31 is closed and secured in position within the next hour. ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours, and
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1.
ACTION 7 - 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 OPERr.- TIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours. ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by cbservation 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. O SEABROOK - UNIT 1 3/4 3-6
-s TABLE 3.3-1 (Continued)
N,_/ ACTION STATEMENTS (Continued) ACTION' 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY ~ within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION 11~- With the number of OPERABLE channels less than the' Total Number l of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours. ACTION 12 - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within-48 hours or declare the breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE n status. S l l J I SEABROOK - UNIT 1 3/4 3-7 t
i i TABLE 3.3-2 l w i 9' en (This table number is not used) m i O 1 O 7 1 i i c a., z i 4 l I i i W I N ! 4 i W '! I i co , l 4 i 2 O O O
/- p g.
TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
- =
8 TRIP l 7 ANALOG ACTUATING MODES FOR-c CHANNEL DEVICE WHICH se CHANNEL CHANNEL OPERATIONAL' OPERATIONAL ACTUATION SURVEILLANCE Z FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED w
- 1. Manual Reactor Trip N.A. N.A. N.A. R(13) N.A. 1, 2, 3*,.4*, 5*
- 2. Power Range, Neutron Flux
- a. High Setpoint S D(2, 4), Q(16) N.A. ~N.A. 1, 2 M(3, 4),
Q(4, 6), R(4, 5)
- b. Low Setpoint S R(4) S/U(1) N.A. N.A. 1***, 2
- 3. Power Range, Neutron Flux, N.A. R(4) Q(16) N.A. N.A. 1, 2 y High Positive Rate a
y 4. Power Range, Neutron Flux, N.A. R(4) Q(16) N.A. N.A. 1, 2 e High Negative Rate
- 5. Intermediate Range, S R(4, 5) S/U(1) N.A. N.A. 1***, 2 Neutron Flux
- 6. Source Range, Neutron Flux S R(4,5) S/U(1),Q(9,16) N.A. N.A. 2**, 3-, 4, 5
- 7. Overtemperature AT S R(12) Q(16) N.A. N.A. 1, 2
- 8. Overpower AT S R Q(16) N.A. N.A. 1, 2
- 9. Pressurizer Pressure--Low S R Q(16,17) N.A. N.A. 1
- 10. Pressurizer Pressure--High S R Q(16,17) N.A. N.A. 1, 2
- 11. Pressurizer Water Level--High S R Q(16) N.A. N.A. 1
- 12. Reactor Coolant Flow--Low S R Q(16) N.A. N.A. 1
TABLE 4.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E TRIP 8
- ANALOG ACTUATING MODES FOR
' CHANNEL DEVICE WHICH CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE E CHANNEL CALIBRATION TEST TEST LOGIC TEST IS REGUIRED Z FUNCTIONAL UNIT CHECK
~ N.A. 1, 2
- 13. Steam Generator Water Level-- S R Q(16, 17) N.A.
Low-Low Undervoltage - Reactor Coolant N.A. R N.A. Q(16) N.A. 1 14. Pumps N.A. N.A. Q(16) N.A. 1
- 15. Underfrequency - Reactor R Coolant Pumps t' 16. Turbine Trip a S/U(1, 10) N.A. 1
- a. Low Fluid Oil Pressure N.A. R N.A.
y S/U(1, 10) N.A. 1 E; b. Turbine Stop Valve N.A. R N.A. Closure N.A. N.A. R N.A. 1, 2
- 17. Safety Injection Input from N.A.
ESF
- 18. Reactor Trip System Interlocks
- a. Intermediate Range 2**
Neutron Flux, P-6 N.A. R(4) R N.A. N.A.
- b. Low Power Reactor N.A.
Trips Block, P-7 N.A. R(4) R N.A. I
- c. Power Range Neutron N.A. R(4) R N.A. N.A. I Flux, P-8
- d. Power Range Neutron N.A. R(4) R N.A. N.A. 1 Flux, P-9 O O O
i l O O TABLE 4.3-1 (Continued) , a N
- g; REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 1 5
) j R TRIP ;
, ANALOG ACTUATING MODES FOR i e CHANNEL DEVICE- WHICH I 5 H
CHANNEL CHANNEL OPERATIONAL OPERATIONAL. ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS' REQUIRED w j Reactor Trip System Interlocks (Continued) l e. Power Range Neutron Flux, P-10 N.A. R(4) R N.A. N . A'. 1, 2 l, i
- f. Turbine Impulse Chamber Pressure, P-13 N.A. R R N.A. N.A. 1
- 19. Reactor Trip-Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, 2, 3*,
- 4*, 5*
i R !
- 20. Automatic Trip and Interlock N.A. N.A. N. A. N.A. M(7)' 1, 2 , 3 * , .
4 y Logic 4*, 5*
- w I 21. Reactor Trip Bypass Breaker N.A. N.A. N.A. M(14),R(15) N.A. 1, 2, 3*,
j 4*, 5* t i i i } I i i l l } i
TABLE 4.3-1 (Continued) TABLE NOTATIONS
*0nly if the Reactor Trip System breakers happen to be 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 31 days. (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. 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 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. 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) Initial plateau curves shall be measured for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. 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. (8) (Not used) (9) Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. (10) Setpoint verification is not applicable. (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers. O SEABROOK - UNIT 1 3/4 3-12
TABLE 4.3-1 (Continued) h TABLE NOTATIONS (Continued) (12) Verify the RTD bypass loops flow rate. (13) The TRIP ACTUATING DEVICE ~0PERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip' circuit (s). (14) local manual shunt trip prior to placing breaker in service. (Or for plants that do not actuate the shunt trip attachment of the bypass breakers on a manual reactor trip): Remote manual undervoltage trip when breaker placed in service. (15) Automatic undervoltage trip. (16) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS. (17) These channels also provide inputs to ESFAS. Comply with the applicable MODES and surveillance frequencies of Specification 4.3.2.1 for any por-tion of the channel required to be OPERABLE by Specification 3.3.2. O O O SEABROOK - UNIT 1 3/4 3-13
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System-(ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints. set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip 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 3.3-4, adjust the Setpoint consistent with the Trip Setpoint-value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3-4, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + R + S < TA Where: Z = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.
- c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
O SEABROOK - UNIT 1 3/4 3-14
2. INSTRUMENTATION ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 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. 4.3.2.2- The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" column of Table 3.3-3. -Q L) 1 SEABROOK - UNIT 1 3/4 3-15
TABLE 3.3-3 N g; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E E
' MINIMUM TOTAL N0. CHANNELS CHANNELS APPLICABLE
_E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w - 1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan / Filter R Actuation, and Latching [. Relay). M a. Manual Initiation 2 1 2 1,2,3,4 17
- b. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays
- c. Containment 3 2 2 1,2,3 14*
Pressure--Hi-1
- d. Pressurizer 4 2 3 1, 2, 3# 18*
Pressure--Low
- e. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 14*
Pressure--Low any steam line e O O
TABLE 3.3-3 (Continued) B; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8 R
, MINIMUM c- TOTAL NO. CHANNELS CHANNELS APPLICABLE 2 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
~
- 2. Containment Spray
- a. Manual Initiation 2 1 with- _
2 1,2,3,4 '17 2 coincident switches.
- b. Automatic Actuation 2 1 2 1,2,3,4 13-Logic and Actuation Relays
- c. Containment Pressure-- 4 2 3 1,2,3 15 R Hi-3 y 3. Containment Isolation C
- a. Phase "A" Isolation
- 1) Manual Initiation 2 1 2 1,2,3,4 17
- 2) Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays
- 3) Safety Injection See. Item 1. above for all Safety Injection initiating' functions and requirements.
- b. Phase "B" Isolation
- 1) Manual Initiation. 2 1 with 2 1,2,3,4- 17 2 coincident
-switches
TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8 si?
' MINIMUM TOTAL N0. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 4 - 3. Containment Isolation (continued)
- b. Phase "B" Isolation (continued)
- 2) Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays
- 3) Containment 4 2 3 1,2,3 15 Pressure--Hi-3 R
- c. Containment Ventilation Isolation 3cn 1) Manual Initiation 2 1 2 1,2,3,4 16
- 2) Automatic Actuation 2 1 2 1,2,3,4 16 Logic and Actuation Relays
- 3) Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
- 4) Containment On Line 2 1 2 1,2,3,4 16 Purge Radioactivity-High
- 4. Steam Line Isolation
- a. Manual Initiation
- 1) Individual 1/ steam line 1/ steam line 1/ operating 1,2,3 23 steam line
- 2) System 2 1 2 1,2,3 21 e G #
O O O TABLE 3.-3-3 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E MINIMUM i R TOTAL NO. CHANNELS CHANNELS APPLICABLE-s FUNCTIONAL UNIT - 0F CHANNELS TO TRIP OPERABLE MODES ACTION
- 4. Steam Line Isolation (continued)
~
- b. Automatic Actuati~on 2 1 2 1,2,3 20 Logic and Actuation Relays
- c. Containment Pressure-- -3 2 2 1,2,3 14*
Hi-2 ,
- d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 14*
Pressure-Low any steam line R
- e. Steam Generator Pressure -' Negative 3/ steam line 2/ steam line 2/ steam line 3** 14" Rate--High any steam
[ e line-
- 5. Turbine Trip
- a. Automatic Actuation 2 1 2 1, 2 22 Logic and Actuation-Relays
- b. Steam Generator 4/stm. gen. 2/stm. gen.
3/stm. gen. 1, 2 18* Water Level-- High-High (P-14)
- 6. Feedwater Isolation
- a. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 18*
Level--High-High (P-14)
- b. Low RCS T Coincident 4 2 3 1, 2 18 avg with Reactor Trip
TABLE 3.3-3 (Continued). ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION x TOTAL NO. CHANNELS HINIMUM CHANNELS APPLICABLE
' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C $ c. Safety Injection See Item 1. above for all Safety Injection initiating functions l and requirements.
- 7. Emergency Feedwater
- a. Manual Initiation (1) Motor driven pump 1 1 1 1,2,3 21 (2) Turbine driven pump 2 1 2 1,2,3 21
- b. Automatic Actuation Logic 2 1 2 1,2,3 20 and Actuation Relays m c. Stm. Gen. Water Level--
g low-Low [ Start Motor-Driven Pump 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 18* o and Start Turbine - Driven Pump
- d. Safety Injection Start Motor-Driven Pump See Item 1. above for all Safety Injection initiating functions and and Turbine-Driven Pump requirements.
- e. Loss-of-Offsite Power Start Motor-Driven Pump and Turbine- See Item 9 for Loss-of-Offsite Power initiating functions and Driven Pump requirements.
l 8. Automatic Switchover to Containment Sump
- a. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation l Relays e 9 9
,a
/m ,
,s t
'w ;
e TABLE 3.3-3 (Continued) M g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8 g MINIMUM
, TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- b. RWST Level--Low-Low 4 2 3 1,2,3,4 18*
Coincident With: Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements. i 9. Loss of Power (Start Emergency Feedwater)
- a. 4.16 kV Bus E5 and E6- 2/ bus 2/ bus 1/ bus 1,2,3,4 18*
Loss of Voltage
! b. 4.16 kV Bus E5 and E6-m Degraded Voltage 2/ bus 2/ bus 1/ bus 1,2,3,4 18*
} Coincident with SI
, See Item 1. above for all Safety Injection initiating functions g, and requirements.
- 10. Engineered Safety Features Actuation System Interlocks
- a. Pressurizer Pressure, 3 2 2 1,2,3 19 P-11
- b. Reactor Trip, P-4 2 2 2 1,2,3 21 l c. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 18*
Level, P-14
TABLE 3.3-3 (Continued) TABLE NOTATIONS
*The provisions of Specification 3.0.4 are not applicable. # Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.
- Trip function automatically blocked above P-11 and may be blocked below P-ll when Safety Injection on low steam line pressure is not blocked.
ACTION STATEMENTS ACTION 13 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least H0T STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. ACTION 14 - 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 provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours for surveillance testing per Speci fication 4. 3. 2.1. ACTION 16 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed. 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 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: O SEABROOK - UNIT 1 3/4 3-22
1: l l
,m TABLE 3.3-3 (Continued)
ACTION STATEMENTS (Continued)
- a. The inoperable channel.is placed in the tripped condition within 1 hour, and
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.2.1.
ACTION 19 - With less than the Minimum Number of Channels OPERABLE, within 1 hour 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 20 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours 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 Total h V Number of Channels, restore the inoperable channel to OPERABLE status within.48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. ACTION 22 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours 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 Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoper-able and take the ACTION required by Specification 3.7.1.5. ( x SEABROOK - UNIT 1 3/4 3-23
TABLE 3.3-4 h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E 8 ^ SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) (S) TRIP SETPOINT ALLOWABLE VALUE Z_ O ~ 1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan / Filter Actuation, and Latching Relay). g a. Manual Initiation N.A. N.A. N.A. N.A. N.A. Y b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
- c. Containment Pressure--Hi-1 4.2 0.71 1.67 5 4.3 psig 5 5.3 psig
- d. Pressurizer Pressure--Low 13.1 10.71 1.69 2 1875 psig 1 1840 psig
- e. Steam Line Pressure--Low 13.1 10.71 1.63 2 585 psig 1 568 psig*
- 2. Containment Spray
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Containment Pressure--Hi-3 3.0 0.71 1.67 5 18.0 psig $ 18.7 psig G G G
s TABLE 3.3-4 (Continued) _ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS x 8
" SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE Z 3. Containment Isolation
- a. Phase "A" Isolation
- 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- 3) Safety Injection. See Item-1. above for all Safety Injection Trip Setpoints and Allowable Values.
y b. Phase "B" Isolation { 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actuation N.A. N.A. N.A. N.A. N.A.
Logic and Actuation Relays
- 3) Containment Pressure-- 3.0 0.71 1.67 5 18.0 psig 1 18.7 psig Hi-3
- c. Containment Ventilation Isolation
- 1) Manual Initiation- N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actuation N.A. N.A. N.A. N.A. N.A.
Logic and Actuation Relays
- 3) _ Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and
~
Allowable Values.
- 4) Containment On'Line Purge N.A. N.A. N.A. 52x N.A.
Radioactivity-High Background
TABLE 3.3-4 (Crntinued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E SENSOR E TOTAL ERROR e FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE E 3 4. Steam Line Isolation
- a. Manual Initiation (System) N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N A.
and Actuation Relays
- c. Containment Pressure--Hi-2 5.2 0.71 1.67 54.3 psig <5.3 psig
- d. Steam Line Pressure--Low 13.1 10.71 1.63 1585 psig 1568 psig*
- e. Steam Generator Pressure - 3.0 0.5 0 <100 psi (123 psi **
g Negative Rate--High x-y 5. Turbine Trip
- a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
Actuation Relays
- b. Steam Generator Water 4.0 2.18 1.76 <86.0% of <87.2% of narrow Level--High-High (P-14) narrow range range instrument instrument span.
span.
- 6. Feedwater Isolation
- a. Steam Generator Water 4.0 2.18 1.76 <86.0% of <87.2% of narrow Level--Hi-Hi-(P-14) narrow range range instrument instrument span.
span. -
- b. Low RCS T avg Coincident 4.6 1.12 1.38 1564 F 1561.2 F with Reactor Trip
- c. Safety Injection N.A. N.A. N.A. N.A. N.A.
O O O
, -~s ,e 7 m
i ) ( ( )
\J LJ Q ,/
TABLE 3.3-4 (Continued) M g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8 R SENSOR TOTAL ERROR [ FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP ~SETPOINT ALLOWABLE VALUE z Z 7. Emergency Feedwater
- a. Manual Initiation (1) Motor driven pump N.A. N.A. N.A. N.A. N.A.
(2) Turbine driven pump N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Steam Generator Water 17.0 15.28 1.76 > 17.0% of > 15.9% of narrow Level--Low-Low iiarrow rarge range instrument
, Start Motor-Driven Pump instrument span. g and Start Turbine-Driven span. w Pump $ d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Start Motor-Driven Pump Allowable Values. and Turbine-Driven Pump
- e. Loss-of-Offsite Power See Item 9. for Loss-of-Offsite Power Setpoints and Allowable Values.
Start Motor-Driven Pump and Turbine-Driven Pump
- 8. Automatic Switchover to Containment Sump
- a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- b. RWST Level--Low-Low 2.75 1.0 1.8 2122,525 gals. 1121,609 gals.
Coincident With Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
TABLE 3.3-4 (Continued) M g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E R SENSOR
, TOTAL ERROR c FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 5
- 9. Loss of Power (Start Emergency Feedwater)
- a. 4.16 kV Bus E5 and E6 N.A. N.A. N.A. > 2975 > 2908 volts Loss of Voltage volts with Uith a < 1.315 a < 1.20 second time second time delay.
delay.
- b. 4.16 kV Bus E5 and E6 N.A. N.A. N.A. > 3933 volts > 3902 volts Degraded Voltage Uith a < 10 Gith a < 10.96 second time second time delay. delay.
y Coincident with:
- Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and
<y Allowable Values. $ 10. Engineered Safety Features Actuation System Interlocks
- a. Pressurizer Pressure, P-11 N.A. N.A. N.A. $ 1950 psig 5 1960 psig
- b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.
- c. Steam Generator Water Level, See Item 5. above for all Steam Generator Water Level Trip P-14 Setpoints and Allowable Values.
O G G
1 4 g~s -TABLE 3.3-4 (Continued) l \s / TABLE NOTATIONS e
- Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are I CHANNEL CALIBRATION shall ensure )
l > 50 seconds and T2 < 5 seconds. l l that these time constants are adjusted to these values.
**The time constant utilized in the rate-lag controller for Steam Line Pressure-
', Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value. i i l l i l
! I i
i 4 i
! t ! SEABROOK - UNIT 1 3/4 3-29 i ;
i ! l
a ,__,__a. -
. - -- - , , ...a-as ----.,-,-eme- _. -, m ,---sa-e - as.,- .. . wa - - -a - m-TABLE 3.3-5 (This table number is not used) i l
l l O l r l O SEABROOK - UNIT 1 3/4 3-30 l l l
> O O TABLE 4.3-2 u,
9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION o E SURVEILLANCE REQUIREMENTS o TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH U CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
- FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 1. Safety Infection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan / Filter Actuation, and Latching
{ Relay). y a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
% b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
- c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 Hi-1 -
- d. Pressurizer Pressure S R M N.A. N.A. N.A. N.A. 1,2,3 Low
- e. Steae Line S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low
- 2. Containment Spray
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays C. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 Hi-3 C
TABLE 4.3-2 (C ntinued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E SURVEILLANCE REQUIREMENTS 8 ^ TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH O CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 3. Containment Isolation
- a. Phase "A" Isolation
- 1) Manual Initiativo N.A. N.A. N.A. R N.A. N.A. N.A 1,2,3,4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
$ 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. w a ~
- b. Phase "B" Isolation
- 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A 1,2,3,4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic Actuation Relays
- 3) Containment S R H N.A. N.A. N.A. N.A. 1,2,3 Pressure-Hi-3
- c. Containment Ventilation Isolation
- 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
- 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- 4) Containment On Line 5 R M(2) N.A. N.A. N.A. N.A. 1, 2, 3, 4 Purge Radioactivity-High 9 9 9
(~
\
O 'N
/
]-
TABLE 4.3-2 (Continued) El ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 35 $URVEILLANCE REQUIREMENTS E SE TRIP
. ANALOG ACTUATING MODES c: CHANNEL DEVICE MASTER SLAVE FOR WHICH 35 CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY. RELAY SURVEILLANCE-FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
-[
- 4. Steam Line Isolation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 (System)
- b. Automatic Actuation N.A. N.A N.A N.A. M(1) M(1) Q 1,2,3 Logic and Actuation Relays
- c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 Hi-2 u, d. Steam Line S R M N.A. N.A. N.A. N.A. 1,2,3 3 Pressure-Low u, e. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 3 Ja Negative Rate-High w
- 5. Turbine Trip
- a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2 Logic and Actuation Relays
- b. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1, 2 Level-High-High (P-14)
- 6. Feedwater Isolation
- a. Steam Generator Water S R M i!. A. N.A. N.A. N.A. 1, 2 Level--High-High (P-14)
- b. Low RCS T avg Coincident S R M N.A. N.A. N.A. N.A. 1, 2 with Reactor Trip
- c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- 7. Emergency Feedwater
- a. Manual Initiation
- 1) Motor-driven pump N.A. N.A. N.A. R- N.A. N.A. N.A. 1,2,3
- 2) Turbine-driven pump N.A. N.A. N.A. R N.A. N.A. N.A. 1,.2, 3
TABLE 4.3-2 (Continued) M g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8 SURVEILLANCE REQUIREMENTS E TRIP c ANALOG ACTUATING MODES 5 ~ CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 7. Emergency Feedwater (Continued)
- b. Automatic Actuatian N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3 and Actuation Relays
- c. Steam Generator Water S R M N.A. N.A. N.A N.A 1,2,3 Level-Low-Low, Start Motor-Driven Pump and Turbine-Driven Pump w
1 d. Safety Injection, Start See Item 1. above for all Safety Injection Surveillance Requirements, m Motor-Driven Pump and a a Turbine-Driven Pump
- e. Loss-of-Of fsite Power See Item 9. for all Loss of-Offsite Power Surveillance Requirements.
Start Motor-Driven Pump and Turbine-Driven Pump
- 8. Automatic Switchover to Containment Sump
- a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
- b. RWST Level-Low-Lov S R M N.A. N.A. N.A. N.A 1,2,3,4 Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
O O G
O O TABLE 4.3-2 (Continued) g g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS 1 sE TRIP
- ANALOG ACTUATING MODES c
1 2 CHANNEL DEVICE MASTER SLAVE FOR WHICH ! U CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE i ** FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST . LOGIC TEST TEST TEST IS REQUIRED- ) 9. Loss of Power (Start Emergency Feedwater) I j a. 4.16 kV Bus E5 and N.A. R N.A M N.A. N.A. N.A. 1,2,3,4 j E6 Loss of Voltage
- b. 4.16 kV Bus E5 and N.A. R. N.A.' M N.A. N.A. N.A. 1,2,3,4 E6 Degraded Voltage Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements'
[ 10. Engineered Safety a Features Actuation j
$ System Interlocks
) a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1,2,3 ] Pressure, P-11
- b. Reactor Trip, P-4 N.A. N.A N.A. N.A. R N.A. N.A. 1,2,3 _
i c. Steam Generator S R M N.A. M(1) M(1) Q 1,2,3 ) Water Level, P-14 1 TABLE NOTATION (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (2) A DIGITAL CHANNEL OPERATIONAL TEST will be performed on this instrumentation. ) 1 i
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-6. ACTION:
- a. With a radiation monitoring channel Alarm / Trip Setpoint for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 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 DIGITAL CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3. O SEABROOK - UNIT 1 3/4 3-36
f% 7% O
/
TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS E MINIMUM 8 CHANNELS CHANNELS APPLICABLE ALARM / TRIP FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION 7 c- 1. Containment z Z a. Containment - Post LOCA - 1 2 All 5 10 R/h 27 Area Monitor
- b. RCS Leakage Detection
- 1) Particulate Radioactivity N.A. 1 1,2,3,4 N.A. 26
- 2) Gaseous Radioactivity N.A. I 1,2,3,4 N.A. 26
- 2. Containment Ventilation Isolation
- a. On Line Purge Monitor 1 2 1,2,3,4
- 23
- b. Manipulator Crane Area Monitor 1 2 5, 6 ** 23 t' 3. Main Steam Line 1/ steam line 1/ steam 1,2,3,4 N.A. 27 line T
w
- 4. Fuel Storage Pool Areas
" a. Fuel Storage Building Exhaust Monitor N . ,*. . 1 *** **** 25
- 5. Control Room Isolation
- a. Air Intake-Radiation Level
- 1) East Air Intake 1/ intake 2/ intake All 24
- 2) West Air Intake 1/ intake 2/ intake All 24
- 6. Primary Component Cooling Water
- a. Loop A 1 1 All 52x 28
Background
- b. Loop B 1 1 All 52x 28
Background
TABLE NOTATIONS
- Two times background; purge rate will be verified to ensure compliance with Specification 3.11.2.1 requirements.
** Two times background or 15 mR/hr, whichever is greater.
*** With irradiated fuel in the fuel storage pool areas.
**** Two times background or 100 CPM, whichever is greater.
TABLE 3.3-6 (Continued) ACTION STATEMENTS ACTION 23 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment ventilation isolation valves are maintained closed. ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation. ACTION 25 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool areas. ACTION 26 - Must satisfy the ACTION requirement for Specification 3.4.6.1. ACTION 27 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours, 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 6.8.2 within 14 days following the event outlining the actions taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
ACTION 28 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, collect grab samples daily from the Primary Component Cooling Water System and the Service Water System and analyze the radioactivity until the inoperable Channel (s) is restored to OPERABLE status. O SEABROOK - UNIT 1 3/4 3-38
. - ~ .
\_ 'x s 's_ .,/
TABLE 4.3-3 v, RADIATION MONITORING INSTRUMENTATION FOR PLANT 9 OPERATIONS SURVEILLANCE REQUIREMENTS E 8 DIGITAL CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE E FUNCTIONAL UNIT- CHECK CALIBRATION TEST IS REQUIRED
- 1. Containment
- a. Containment - Post LOCA -
Area Monitor S R M All
- b. RCS Leakage Detection 4
- 1) Particulate Radio- S R M 1,2,3,4 activity
- 2) Gaseous Radioactivity S R M 1,2,3,4
- 2. Containment Ventilation Isolation
- a. On Line Purge Monitor S R M 1,2,3,4 -
- R' b. Manipulator Crane Area S R M 5, 6
, Monitor w J, 3. Main Steam Line S R M 1,2,3,4 e
- 4. Fuel Storage Pool Areas
- a. Radioactivity-High-Gaseous Radioactivity S R M *
- 5. Control Room Isolation
- a. Air Intake Radiation Level
- 1) East Air Intake S R M All
- 2) West Air Intake S R M All
- 6. Primary Component. Cooling Water
- a. Loop A S R M All
- b. Loop B S R M All TABLE NOTATIONS
- With irradiated fuel in the fuel storage pool areas.
INSTRUMENTATION MONITORING INSTRUMENTATON MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:
- a. At least 75% of the detector thimbles,
- b. A minimum of two detector thiables 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 F3g, pq(Z) and Fxy.
ACTION: With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of l Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS (Plant procedures are used to determine that the Movable Incore Detection System is OPERABLE.) O SEABROOK - UNIT 1 3/4 3-40
INSTRUMENTATION MONITORING INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING' CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be 0PERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Re~ port to the Commission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS rh 4.3.3.3.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALI-BRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4. 4.3.3.3.2 Each of the above required seismic monitoring' instruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 30 days following the seismic event. Data shall be retrieved from actuated
-instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 14 days describing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety.
O - (v) SEABROOK - UNIT 1 3/4 3-41
TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Time-History Accelerographs
- a. 1-SM-XT-6700 Free Field East Cont. lg 1*
Room Air Intake
- b. 1-SM-XT-6701 Containment Foundation lg 1*
- c. 1-SM-XT-6710 Cont. Opr. Floor lg 1*
- 2. Triaxial Peak Accelerographs
- a. 1-SM-XR-6702 Reactor Vessel Support 0-20 Hz. 1
- b. 1-SM-XR-6703 Reactor Cool. Piping 0-20 Hz. 1
- c. 1-SM-XR-6704 PCCW Piping 0-20 Hz. 1
- 3. Triaxial Seismic Switches
- a. 1-SM-XS-6700 Free Field N.A. 1*
- b. 1-SM-XS-6701-Containment Foundation N.A. 1*
- c. 1-SM-XS-6709 Containment Foundation 0.025g to 0.25g 1*
- d. 1-SM-XS-6710 Cont. Opr. Floor N.A. 1*
- 4. Triaxial Response-Spectrum Recorders
- a. 1-SM-XR-6705 Containment Foundation 1-30 Hz. 1*
- b. 1-SM-XR-6706 SG 11B Support 1-30 Hz. 1 c.1-SM-XR-6707 Prim. Aux. Bldg. 1-30 Hz. 1
- d. 1-SM-XR-6708 Service Water Pump House 1-30 Hz. 1
*With reactor control room indication O
l i SEABROOK - UNIT 1 3/4 3-42
TABLE 4.3-4 t SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL' OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelerographs
- a. 1-SM-XT-6700 Free Field East Cont. M* R SA Room Air Intake
- b. 1-SM-XT-6701 Containment Foundatinn M* R N.A.
- c. 1-SM-XT-6710 Cont. Opr. Floor M* R N.A.
- 2. Triaxial Peak Accelerographs
- a. 1-SM-XR-6702 Reactor Vessel Support N.A. R N.A.
- b. 1-SM-XR-6703 Reactor Cool. Piping N.A. R N.A.
- c. 1-SM-XR-6704 PCCW Piping N.A. R N.A.
- 3. Triaxial Seismic Switches a.'l-SM-XS-6700 Free Field **
M R SA
- b. 1-SM-XS-6701 Containment Foundation ** M R N.A.
h c. 1-SM-XS-6709 Containment Foundation ** M R N.A.
- d. 1-SM-XS-6710 Cont. Opr. Floor **
.M R N.A.
- 4. Triaxial Response-Spectrum Recorders
- a. 1-SM-XR-6705 Containment Foundation ** M#' R N.A.
- b. 1-SM-XR-6706 SG llB Support N.A. R N.A. q
- c. 1-SM-XR-6707' Prim. Aux. Bldg. N.A. R N.A.
- d. 1-SM-XR-6708 Service Water Pump House N.A. R N.A.
J 4
*Except seismic trigger
; **With reactor control room indications.
# CHANNEL CHECK to consist of turning the test / reset switch and verify all {
) lamps illuminate on 1-SM-XR-6705.
xJ SEABROOK - UNIT 1 3/4 3-43 l _ - - - - - - O
INSTRUMENTATION MONITORING INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in. Table 3.3-8 shall be OPERABLE. APPLICABILITY: At'all times. ACTION:
- a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be demonstrated OPERABLE by the performance of:
- a. A Daily CHANNEL CHECK, and
- b. A Semiannual CHANNEL CALIBRATION O
SEABROOK - UNIT 1 3/4 3-44
TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION - OPERABLE
- 1. Wind Speed
- a. Lower Level Nominal Elev. 43 ft 1
- b. Upper Level . Nominal Elev. 209 ft 1
- 2. Wind Direct' ion
- a. Lower-Level Nominal Elev. 43 ft 1
- b. Upper Level Nominal Elev._209 ft 1
.3. Air Temperature - AT
- a. Lower Level Between Elev.'43 ft and 150 ft_ 1
- b. Upper Level Between Elev. 43 ft and 209'ft 1 SEABROOK - UNIT 1- 3/4 3-45
INSTRUMENTATION MONITORING INSTRUMENTATION REMOTE SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.5 The Remote Shutdown System transfer switches, power, controls and monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE remote shutdown monitoring channels less than the Total Number of Channels as required by Table 3.3-9, within 60 days restore the inoperable channel (s)'to OPERABLE status or, pursuant to Specification 6.8.2, submit a Special Report that defines the corrective action to be taken.
- c. With one or more Remote Shutdown System transfer switches, power, or control circuits inoperable, restore the inoperable switch (s)/
circuit (s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel in Table 3.3-9 shall be demonstrated OPERABLE:
- a. Every 31 days by performance of a CHANNEL CHECK, and
- b. Every 18 months by performance of a CHANNEL CALIBRATION.
4.3.3.5.2 Each Remote Shutdown System transfer switch, power and control circuit listed in Table 3.3-9, including the actuated components, shall be demonstrated OPERABLE at least once per 18 months. ,
?
O SEABROOK - UNIT 1 3/4 3-46 L
O O O TABLE 3.3-9 m REMOTE SHUTDOWN SYSTEM m E TOTAL NO. MINIMUM E OF CHANNELS S INSTRUMENT LOCATION CHANNELS OPERABLE c- 1. Intermediate Range Neutron Flux CP-108 A and B 2 1 i'i
- 2. Source Range Neutron Flux CP-108 A and B 2 1
- 3. Reactor Coolant Temperature -
Wide Range for Loops 1 and 4
- a. T CP-108 A 2 2 c
- b. T CP-108 B 2 2 H
- 4. Pressurizer Pressure CP-108 A and B 2 2
. 5. Pressurizer Level CP-108 A and B 2 2
- 6. Steam Generator Pressure CP-108 A and B 1/stm. gen. 1/stm. gen.
- 7. Steam Generator Water Level CP-108 A and B 1/stm. gen. 1/stm. gen.
w 8. Steam Generator-Emergency Feedwater i Flow Rate CP-108 A and B 1/stm. gen. 1/stm. gen. w 9. Boric Acid Tank Level CP-108 A and B 1/ tank 1/ tank LOCATION
- b. TRANSFER SWITCHES / CONTROL CIRCUITS
- 1. Emergency Feedwater Pump Steam Supply Valves MS-V-393/127 CP-108 A
- 2. Emergency Feedwater Pump Steam Supply Valves MS-V-394/128 CP-108 B
- 3. Emergency Feedwater Pump Steam Supply Valves MS-V-395 CP-108 A and B
- 4. Emergency Fe'edwater Pump FW-P-378 BUS 6 SWGR
- 5. Emergency Feedwater Recirculation Valve FW-V-346 CP-108 A
- 6. Emergency Feedwater Recirculation Valve FW-V-347 CP-108 8
- 7. SG A EFW Control Valve FW-FV-4214 A CP-108 A
- 8. SG A EFW Control Valve FW-FV-4214 B CP-108 B
- 9. SG B EFW Control Valve FW-FV-4224 A CP-108 A
- 10. SG B EFW Control Valve FW-FV-4224 B CP-108 8
- 11. SG C EFW Control Valve FW-FV-4234 A CP-108 A
- 12. SG C EFW Control Valve FW-FV-4234 B CP-108 B
- 13. SG D EFW Control Valve FW-FV-4244 A CP-108 A
- 14. SG D EFW Control Valve FW-FV-4244 B CP-108 8
- 15. SG A Atmospheric Relief Valve MS-PV-3001 CP-108 A
- 16. SG B Atmospheric Relief Valve MS-PV-3002 CP-108 8
- 17. SG C Atmospheric Relief Valve MS-PV-3003 CP-108 A
TABLE 3.3-9 (Continued) p REMOTE SHUTDOWN SYSTEM 3; TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION
- 18. SG D Atmospheric Relief Valve MS-PV-3004 CP-108 B E 19. MS Isolation Valves MS-V-86/88/90/92 Z CP-108 A
- 20. MS Isolation Valves MS-V-86/88/90/92 CP-108 B
- 21. Pressurizer Heaters, Group A CP-108 A
- 22. Pressurizer Heaters, Group B CP-108 B
- 23. Charging Pump CS-P-2A BUS.5 SWGR
- 24. Charging Pump CS-P-28 BUS 6 SWGR
- 25. Charging Pump Suction from RWST CS-LCV-112D CP-108 A
- 26. Charging Pump Suction from RWST CS-LCV-112E CP-108 8
- 27. Pressurizer Relief Valve (PORV) RC-PCV-456A CP-108 A
- 28. Pressurizer Relief Valve (PORV) RC-PCV-456B CP-108 8
- 29. PORV Block Valve RC-V-122 CP-108 A
- 30. PORV Block Valve RC-V-124 CP-108 B R
- 31. High Pressure Injection SI-V-138 CP-108 A
- 32. High Pressure Injection SI-V-139 CP-108 8 Y 33. VCT Discharge Isolation Valve CS-LCV-1128 CP-108 A
$ 34. VCT Discharge Isolation Valve CS-LCV-112C CP-108 B e -- G #
INSTRUMENTATION (mY V MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 -The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status.within 7 days,.or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- b. . With the number of OPERABLE accident monitoring instrumentation channels except the containment atmosphere-high range radiation mcnitor, less than the Minimum Channels OPERABLE requirements of (N Table 3.3-10, restore the inoperable channel (s) to OPERABLE status l within 48 hours or be in at least HOT STANDBY within the next
'N 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- c. With the number of OPERABLE channels for the containment Post-LOCA high range area monitor less than required by the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s), within 72 hours, and either restore the
-inoperable channel (s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specifi -
cation 6.8.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE: . a. Every 31 days by performance of a CHANNEL CHECK, and
- b. Every 18 months by performance of a CHANNEL CALIBRATION.
V SEABROOK - UNIT 1 3/4 3-49.
TABLE 3.3-10 g ACCIDENT MONITORING INSTRUMENTATION g TOTAL MINIMUM NO. OF CHANNELS c INSTRUMENT CHANNELS OPERABLE 5 H 1. Containment Pressure
- a. Normal Range 2 1
- b. Extended Range 2 1 2.
Reactor Coolant Outlet Temperature - TH0T (Wide Range) 4 2 Temperature 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 4 2
- 4. Reactor Coolant Pressure - Wide Range 2 w 1 A 5. Pressurizer Water Level 2 1
- 6. Steam Generator Pressure 2/ steam generator 1/ steam generator
- 7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator
- 8. Steam Generator Water Level - Wide Range 1/ steam generator 1/ steam generator
- 9. Refueling Water Storage Tank Water Level 2 1
- 10. Reactor Coolant System Subcooling Margin Monitor 2 1
- 11. Containment Building Water Level 2 1
- 12. Core Exit Thermocouples 4/ core quadrant 2/ core quadrant
- 13. Containment Post-LOCA Area Monitor 2 1 e 9 8
i l i TABLE 3.3-10 (Continued)' g ACCIDENT-MONITORING INSTRUMENTATION I 1 5 o TOTAL MINIMUM i
- CHANNELS' NO. OF i INSTRUMENT CHANNELS OPERABLE
[ I E i + 14. Intermediate Range Neutron Flux 2 1
- w
- 15. Intermediate Range Neutron Flux Rate 2 1
)
- 16. Containment Isolation Valve Position * ~2/ Penetration 1/ Penetration i
i 17. Containment Enclosure Negative' Pressure 2 1
- 18. Condensate Storage Tank Water Level ** 2 1 ,.
l
- 19. Reactor Vessel Level Indication System 2 1 l N a 20. Containment Hydrogen Concentration 2 1 i
} T l E !
- Applies to penetrations with 2 active valves in series. These valves are moved to'the closed
{ position by automatic signals. j ** Calculated on basis of pressure sensed at suction to the Emergency Feedwater Pumps. l 4 i i i
INSTRUMENTATION MONOTORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 (This specification number is not used.) O O SEABROOK - UNIT 1 3/4 3-52
1 e i
! TABLE 3.3-11 1. ! '(This table number is not used) 4 i
l 1 1 1 1 1 1 I. 4 4 1 ? i l 1 1 ,4 lO i 1 i - i I. J i i !- i i
- ?
I l '.i i i i i 1 : i t f ! l l, - ! i
- SEABROCX - UNIT 1 3/4 3-53 i' !
I i I
INSTRUMENTATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.~8 (This specification number-is not used). O O SEABROOK - UNIT 1 3/4 3-54
, INSTRUMENTATION
/
- k. )\ MONITORING INSTRUMENTATION RADI0 ACTIVE' LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 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 (0DCM). APPLICABILITY: At all times. ACTION:
- a. With a'. radioactive. liquid effluent monitoring instrumentation channe.1 Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels.0PERABLE, take~the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-annual Radioactive' Effluent Release Report pursuant to Specifica-tion 6.8.1.4 why this inoperability was not corrected in a timely manner.
- c. The provisions of Specifications 3.0.3 and 3.0.4, are not applicable.
SURVEILLANCE-REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance-of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-5. ( SEABROOK - UNIT 1 3/4 3-55
TABLE 3.3-12 h RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION E 8
' MINIMUM CHANNELS E INSTRUMENT OPERABLE ACTION 4
~ 1. Radioactivit'y Monitors Providing Alarm and Automatic Termination of Release
- a. Liquid Radwaste Test Tank Discharge 1 29
- b. Steam Generator Blowdown Flash Tank Drain 1* 30
- c. Turbine Building Sumps Effluent Line 1 30
- 2. Flow Rate Measurement Devices R
- a. Liquid Radwaste Test Tank Discharge 1 31 Y
8 b. Steam Generator Blowdown Flash Tank Drain 1* 31
- c. Circulating Water Discharge 1** N.A.
- 3. Radioactivity Monitors Providing Alarm but Not Termination of Release
- a. Primary Component Cooling Water System (In lieu of 1 32 service water monitors)
- 4. Rate of Change Monitor
- a. Primary Component Cooling Water System Head Tank 1 33 (In lieu of service water monitors)
*0nly applicable when steam generator blowdown is directed to the discharge transition structure. ** Pump performance curves generated in place should be used to estimate flow.
O O O
/o s TABLE 3.3-12 (Continued) \ ACTION STATEMENTS ACTION 29 - With the number of channels OPERABLE less than 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 station staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 30 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection of no more than 10 7 microcurie /ml:
- a. At least once per 12 hours when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or V b'. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
ACTION 31-- With the number of channels OPERABLE less than the Minimum' Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curves generated in place may be used to estimate flow. ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity levels in ! the Primary Component Cooling Water System and the Service Water System are determined at least once per 24 hours. ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity level is determined at least once per 12 hours during actual releases. v SEABROOK - UNIT 1 3/4 3-57
TABLE 4.3-5 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS B R
, CHANNEL c CHANNEL SOURCE CHANNEL OPERATIONAL z INSTRUMENT CHECK CHECK CALIBRATION TEST H 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release
- a. Liquid Radwaste Test Tank Discharge D P R(2) P(1)
- b. Steam Generator Blowdown Flash Tank Drain D M R(2) Q(1)
- c. Turbine Building Sumps Effluent Line D M R(2) Q(1) g 2. Flow Rate Measurement Devices Y a. Liquid Radwaste Test Tank Discharge
- D(3) N.A. R N.A.
- b. Steam Generator Blowdown Flash Tank Drain D(3) N.A. R N.A.
- c. Circulating Water Discharge ** N.A. N.A. N.A.
- 3. Radioactivity Monitor Providing Alarm But Not Termination of Release
- a. Primary Component Cooling Water System D M R(2) Q(1)
(In lieu of service water monitors)
- 4. Rate of Change Monitor
- a. Primary Component Cooling Water System D(4) N.A. R N.A.
(In lieu of service water monitors)
^1 solation of the flow path is accomplished by the Waste Test Tank Discharge Pump Trip Circuitry. ** Pump curves may be used to estimate flow.
O O O
p TABLE 4.3-5 (Continued) TABLE NOTATIONS (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that tutomatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured 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) C'HANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. (4) CHANNEL CHECK shall consist of verifying indication of tank level during periods of release. CHANNEL CHECK shall be made'at least once per 24 hours. /~N O SEABROOK - UNIT 1 3/4 3-59
INSTRUMENTATION MONITORING INSTRUMENTATION RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specifications 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 3.3-13. 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.
- b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within 30 days or, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.8.1.4 why this inoperability was not corrected in a timely manner.
- c. The provisions of Specif-ications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-6. O SEABROOK - UNIT 1 3/4 3-60
.- . - - - _ - . . . _ . ..- - .- - - _ - _ ~ . . - - . - - . . --
TABLE 3.3-13 l h RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 8
' 7" MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY . ACTION E
Z 1. RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE GAS
~ MONITORING SYSTEM l 0xygen Monitor (Process) 1 ** 34
- 2. PLANT VENT-WIDE RANGE GAS MONITOR
- a. Noble Gas Activity Monitor 1
- 33 t
- b. Iodine Sampler 1
- 35 R c. Particulate Sampler 1
- 35 y d. Flow Rate Monitor 1
- 32 6
- e. Sampler Flow Rate Monitor 1
- 32
- 3. GASEOUS WASTE PROCESSING SYSTEM t (Providing Alarm and Automatic Termination of Release - RM 6504) .
- a. Noble Gas Activity Monitor (Process) 1
- 33 i
?
o t I i i ,
TABLE 3.3-13 (Continued) RADI0 ACTIVE GASE0VS'EF $ VENT MONITORING INSTRUMENTATION E o
' MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION E
Z 4. # TURBINE GLAND SEAL CONDENSER EXHAUST
~
- a. Iodine Sampler 1 ***
35
- b. Particulate Sampler 1 ***
35
- c. Sampler Flow Rate Indicator 1 ***
32 R
=
8 i l G G #
d
.s TABLE 3.3-13 (Continued)
[ \ N,_,/ TABLE NOTATIONS At all times.
** During GASE0US RADWASTE TREATMENT SYSTEM operation.
*** When the gland seal exhauster is in operation.
# Noble Gas Monitor _for this release point is based on the main condenser air evacuation monitor.
ACTION STATEMENTS' ACTION 32 - With the number of channels OPERABLE less than the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours. ACTION 33- With the number of channels OPERABLE less than the Minimum
~
Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. For RM-6504, RM 6503 may be'used as an alternate. ACTION 34 - With the number of channels OPERABLE less than the Minimum i (~'s 1 Channels OPERABLE requirement, operation of this RADIOACTIVE GAS WASTE SYSTEM may continue provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. ACTION 35 - With the number of channels OPERABLE less than the Minimum d Channels OPERABLE requirement, effluent releases via the affected ; pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in the ODCM. J O SEABROOK - UNIT 1 3/4 3-63 _. _ _ _ . -. . - _ - . . . _ - _ _ = . - . . . - . _ - . . .
TABLE 4.3-6 M g; RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS. 5 x CHANNEL MODES FOR WHICH
- c. CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE z INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED
~ 1. RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE GAS MONITORING SYSTEM 0xygen Monitor D N.A. ** Q(4) M (Process)
- 2. PLANT VENT-WIDE RANGE GAS MONITOR
- a. Noble Gas Activity Monitor D M R(3) Q(2) *
$ b. Iodine Sampler W N.A. N.A. N.A. *
- c. Particulate Sampler W N.A. N.A. N.A. *
- d. Flow Rate Monitor D N.A. R Q**** *
- e. Sampler Flow Rate Monitor D N.A.
- R Q****
O 9 9
O TABLE 4.3-6 (Continued) E g RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E o
- CHANNEL MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE E INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED Z
H 3. GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic Termination of Release)
- a. Noble Gas Activity Monitor D N.A. R(5) Q(1)
(Process)
- 4. TURBINE GLAND SEAL CONDENSER EXHAUST
- a. Iodine Sampler W N.A. N.A. N.A
- b. Particulate Sampler W N.A. N.A. N.A.
$ c. Sampler Flow Rate Indicator D N.A. N.A. N.A.
I TABLE 4.3-6 (Continued) TABLE NOTATIONS At all times. During RADI0 ACTIVE WASTE GAS SYSTEM operation. When the gland seal exhauster is in operation.
**** The CHANNEL OPERATIONAL TEST for the flow rate monitor shall consist of a verification that the Radiation Data Management System (RDMS) indicated flow is consistent with the operational status of the plant.
# Noble Gas Monitor for this release point is based on the main condenser air evacuation monitor.
(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm / Trip Setpoint. (2) The Digital CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm Setpoint. (3) The iniital 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 per-
~
mit calibrating the system over its intended range of energy and measure-ment range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) 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.
(5) 'The CHANNEL CALIBRATION shall be performed using sources of various activities covering the measurement range of the monitor to verify that the response is linear. Sources shall be used to verify the monitor response only for the intended energy range. l O SEABROOK - UNIT 1 3/4 3-66 L
, INSTRUMENTATION
/ s V 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Prot _ection System shall be OPERABLE. APPICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one stop valve or one control valve per high pressure turbine steam line inoperable and/or with one intermediate stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours.
- b. With the above-required Turbine Overspeed Protection System otherwise inoperable, within 6 hours isolate the turbine.from the steam supply.
~
SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable. N^J 4.3.4.2 The above required Turbine Overspeed Protection System shall be-demonstrated OPERABLE:
- a. At least once per 7 days by cycling each of the following valves through at least one complete cycle from the running position:
- 1) Four high pressure turbine stop valves,
- 2) Four high pressure turbine governor valves,
- 3) Four low pressure turbine reheat stop valves, and
- 4) Four low pressure turbine reheat intercept valves,
- b. At. least once per 31 days by direct observation of the movement of-each of the above valves through one complete cycle from the running position,
- c. At .least once per 18 months by performance of a CHANNEL CALIBRATION on the Turbine Overspeed Protection' Systems, and
- d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks, and stems and verifying no unacceptable flaws or excessive corros. ion. If unacceptable flaws or excessive corrosion are found, all other valves of that type shall be inspected.
/m\ V SEABROOK - UNIT 1 3/4 3-67
..2 .-- -
4 d O O l l t a l o D A-w p? A k w 1 (' O J' Mi [ S J
, , , . . n .. ..,---- - - - e w. ew- -e n ,---m_-ww-. --,r---m,me-m~~
.,m. 3/4.4 REACTOR COOLANT SYSTEM
) 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation ~.
APPLICABILITY: MODES 1 and 2.* ACTION: With.less than the above required reactor coolant loops in operation, be in-at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. -f r\
)
*See Special Test Exceptions Specification 3.10.4.
SEABROOK - UNIT 1 3/4 4-1
REACTOR COOLANT SYSTEM REACTORCOOLANTLOOPSANDCOOLANTCIRCULATIbN HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System 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,
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
i APPLICABILITY: MODE 3. ACTION: ' a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour open the Reactor Trip System breakers.
- c. With no reactor coolant loop in operation, suspend all operations 3
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. )
- 1
*All reactor coolant pumps may be deenergized for up to 1 hour 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.
N 3/4 4-2 (pEABR00K-UNIT 1
1 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION HOT STANDBY SURVEILLANCE REQUIREMENTS 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 17% at least once per 12 hours. 4.4.1.2.3 The required' reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. O O
~
SEABROOK - UNIT 1 3/4 4-3
REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE 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. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,**
- e. RHR Loop A, and
- f. RHR Loop B.
APPLICABILITY: MODE 4. ACTION:
- a. 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 hours.
- b. 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 I hour 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 unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold-leg temperatures.
SEABROOK - UNIT 1 3/4 4-4
REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION-HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.-l.3.1.The required reactor coolant pump (s), if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated. power availability. 4.4.1.3.2 The required steam g'enerator(s) shall'be determined OPERABLE by verifying secondary-side water level to be greater than or' equal to 17% at least once per 12 hours. 4.4.1.3.3 At. least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours. 9 O SEABROOK - UNIT 1 3/4 4-5 l I
REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:
- a. One additional.RHR loop shall be OPERABLE **, or
- b. The secondary-side water level of at least two steam generators shall be greater than 17%.
APPLICABILITY: MODE 5 with reactor coolant loops filled ***. ACTION:
- a. With one of the RHR loops inoperable and 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 REQUIREMENTS 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 hours. 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 hours.
*The RHR pump may be deenergized for up to 1 hour 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. **0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
- A reactor coolant pump shall not be started unles's the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold-leg temperatures.
SEABROOK - UNIT 1 3/4 4-6
-REACTOR COOLANT SYSTEM fr '
( l REACTOR COOLANT LOOPS AND COOLANT CIRCULATION COLD 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. SURVEILLANCE REQUIREMENTS O' 4. 4.1.' 4. 2 At least one RHR loop shall be determined.to be in operat' ion and circulating reactor coolant at least once per 12 hours.
*0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
**The RHR pump may be deenergized for up to 1 hour 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.
t :
- l SEABROOK - UNIT 1 3/4 4-7 1
4
%% - , ,~,w.--- , , ,s . . . , , , - . ,v,..-,--,,.--mm- ....,,.--r
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN 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 i 1%.* APPLICABILITY: MODES 4 and 5. ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity' changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by Specifica-tion 4.0.5. O
- The lift setting pressure shall correspond to ambient conditions of the valve at nc ninal operating temperature and pressure.
SEABROOK - UNIT 1 3/4 4-8
REACTOR COOLANT SYSTEM SAFETY VALVES OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig i 1%.* APPLICABILITY: MODES 1, 2, and 3. ACTION: With 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 SHUTOOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specifica-tion 4.0.5. O
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
k SEABROOK - UNIT 1 3/4 4-9 _y... y-,.,-ms- .,-.#m-m--me---ww--em
l l REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 92% of pressurizer level (1656 cubic feet), and at least two groups of pressurizer heaters each having a capacity of at lec.st 150 kW. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY ,
with the Reactor Trip System breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters from the emergency power supply and measuring circuit current at least once per 92 days. O SEABROOK - UNIT 1 3/4 4-10
REACTOR COOLANT SYSTEM
/- \g 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated re. lief valves (PORVs) and their associated block valves shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block
~ valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
b) c. With both PORV(s) inoperable due to causes other than excessive seat V leakage, within I hour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN.within the following 30 hours,
- d. With one or more block valve (s) inoperable, within 1 hour:
(1) restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV and remo n power from its associated solenoid valve; and (2) apply the ACTION b. or c. above, as appropriate, for the isolated PORV(s),
- e. The provisions of Specification 3.0.4 are not applicable.
'\ SEABROOK - UNIT 1 3/4 4-11 i
REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:
- a. Performance of a CHANNEL CALIBRATION, and
- b. Operating the valve through one complete cycle of full travel.
4.4.4.2 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 ACTION b. or c. in Specification 3.4.4. O O SEABROOK - UNIT 1 3/4 4-12
REACTOR COOLANT SYSTEM f). 3/4.4.5 STEAM GENERATORS t.v! LIMITING 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 inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200"F. SURVEILLANCE REQUIREMENTS 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 Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the' minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum se ple size, inspection result classification, 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 3% of the l i total number of tubes in all steam' generators; the tubes selected for these l inspections shall be selected on a random basis except: I
- a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas; l b. The first sample of tubes selected for each inservice inspection l (subsequent to the preservice inspection) of each steam generator shall include:
i v SEABROOK - UNIT 1 3/4 4-13 _ - _ - _ _ _ _ _ . . _ . _ _ , _ .._- _ _ _ _ . . _ _ _ . _ _ - ~ _ _ - . _ _
REACTOR COOLANT SYSTEM STEAM GENER'iGORS SURVEILLANCE REQUIREMENTS 4.4.5.2b. (Continued)
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
- 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 (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
- 2) The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results C-1 Less than 5% 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 1% of the total tubes inspected, are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. SEABROOK - UNIT 1 3/4 4-14
l
. js
~
REACTOR COOLANT SYSTEM
-STEAM GENERATORS i
i ! SURVEILLANCE REQUIREMENTS ! 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection shall be performed after 6 Effective Full-Power' Months but within 24 calendar months of initial criticality..
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, not including the pre-l service inspection, result in all inspection results falling in Cate-
- gory C-1 or if two consecutive inspections demonstrate that previously l observed degradation has not continued and no additional degradation
- has occurred, the inspection interval may be extended to a maximum of l once per 40 months; j b. If the results of~the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a 6
s maximum of once per 40 months; and l c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection l specified in Table 4.4-2 during the shutdown subsequent to any of
- the following conditions
- 1) Primary-to-secondary tubes leak (not including leaks originating l from tube-to-tubesheet welds) in excess of the limits of Specification 3.4.6.2, or
- 2) A seismic occurrence greater than the Operating Basis Earthquake, or
- 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or l 4) A main steam line or feedwater line break.
L
\
SEABROOK - UNIT 1 3/4 4-15
REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS 4.4.5.4 Acceptance Criteria
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;
- 2) Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either the inside or outside of a tube;
- 3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
- 4) % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective;
- 6) Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness;
- 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-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 Specification 4.4.5.3c., above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot-leg side) completely around the U-bend to the top support of the cold leg; and
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy-current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
O SEABROOK - UNIT 1 3/4 4-16
p REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS 4.4.5.4 (Ccatinued)
- 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 Reports
- a. Within 15 days following-the completion of each inservice inspection of steam generator tubes, the n_ umber of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.8.2;
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.8.2 within 12 months following the completion of'the inspection. This Special Report shall include:
- 1) Number and exterit of tubes inspected,
- 2) Location and r;ercent of wall-thickness penetration for eac"h 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 in,a Special Report to the Commission pursuant-to Specification 6.8.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
I i l l SEABROOK - UNIT 1 3/4 4-17 i
TABLE 4.4-1 MIMIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION No. of Steam Generators per Unit Four Preservice Inspection Four First Inservice Inspection Two Sicond & Subsequent Inservice Inspections One (1) TABLE NOTATION (1) The third and fourth steam generators that were not inspected during the first inservice inspection shall be inspected during the second and third inspections, respectively. For the fourth and subsequent inspections, the inservice inspection may be limited to one steam generator on a rotating schedule encompassing 12% of the tubes if the results of the previous in-spections of the four steam generators indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions 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. O SEABROOK - UNIT 1 3/4 4-18
i TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 5 , SE IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION > Action Required Result Action Required h Sample Size Result Action Required Result s* A minimum of C-1 None N. A. N. A. N. A. N. A.
- S Tubes per S. G.
C .2 Plug defective tubes C-1 None N. A. N. A. and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G. Perform action for C-3 C-3 result of first , sample y Perform action for a C-3 C-3 result of first N. A. N. A. ' sample c C-3 Inspect all tubes in All other this S. G., plus de. S. G.s are None N. A. N. A. fective tubes and C-1 inspect 2S tubes in Some S. G.s Perform action for each other S. G. N. A. N. A. C-2 but no C-2 result of second additional ,,,,p g , I Notificatiert to NRC S. G. are l pursuant to $50.72 C-3 i (b)(2) of 10 CFR Additional Inspect all tubes in Part 50 S. G. is C-3 each S. G. and plug . defective tubes. Notification to NRC N. A. N. A. ; pursuant to 650.72 , (bl(2) of 10 CFR Part 50 j , S=3 % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected n during an inspection t
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SiSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- a. The Containment Atmosphere Particulate Radioactivity Monitoring System,
- b. The Containment Drainage Sump Level Moni.toring System, and
- c. Containment Radioactive Gas Monitor APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous and Particulate Monitoring Systems -
performance of CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
- b. Containment Drainage Semp Level Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.
O SEABROOK - UNIT 1 3/4 4-20
g3 REACTOR COOLANT SYSTEM
'w/ REACTOR COOLANT SYSTEM LEAKAGE OPERAlIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE B0UNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. 1 gpm total reactor-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,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 psig i 20 psig, and
- f. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in CT Table 3.4-1.
\'j-APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
/G
( ) v SEABROOK - UNIT 1 3/4 4-21
REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 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 particulate radioactivity monitor at least once per 12 hours;
- b. Monitoring the containment drainage sump inventory and discharge at least once per 12 hours;
- c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;
- d. Performance of a Reactor Coolant System water inventory balance within 12 hours after achieving steady-state operation
- and at least once per 72 hours thereafter during steady-state operation, except that not more than 96 hours shall elapse between any two successive inventory balances; and
- e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours.
- I,yg being changed by less than 5 F/ hour.
SEABROOK - UNIT 1 3/4 4-22
f. Ig) REACTOR COOLANT SYSTEM v REACTOR COOLANT SYSTEM LEAKAGE i OPERATIONAL LEAKAGE , SURVEILLANCE REQUIREMENTS l 4.4.6.2.2 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:
- a. At least once per 18 months,
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months,
- c. Prior to returning the valve to service follewing maintenance, repair, or replacement work on the valve, and
- d. Within 24 hours following valve actuation due to automatic or manual action or flow through the valve.
- e. As outlined in the ASME Code, Section XI, paragraph IW-3427(b).
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. SEABROOK - UNIT 1 3/4 4-23
TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE VALVE MAX. ALLOWABLE NUMBER SIZE FUNCTION LEAKAGE (GPM) SI-V144 1-1/2" SI to RCS Loop 1 Cold-Leg Injection 0.75 SI-V148 1-1/2" SI to RCS Loop 2 Cold-Leg Injection 0.75 SI-V152 1- 1/2" SI to RCS Loop 3 Cold-Leg Injection 0.75 SI-V156 1-1/2" SI to RCS Loop 4 Cold-Leg Injection 0.75 SI-V81 2" SI to RCS Loop 3 Hot-Leg Injection 1.0 SI-V86 2" SI to RCS Loop 2 Hot-Leg Injection 1.0 SI-V106 2" SI to RCS Loop 4 Hot-Leg Injection 1.0 SI-V110 2" SI to RCS Loop 1 Hot-Leg Injection 1.0 SI-V118 2" SI to RCS Loop 1 Cold-Leg Injection 1.0 SI-V122 2" SI to RCS Loop 2 Cold-Leg Injection 1.0 SI-V126 2" SI to RCS Loop 3 Cold-Leg Injection 1.0 SI-V130 2" SI to RCS Loop 4 Cold-Leg Injection 1.0 SI-V140 3" SI to RCS Cold-Leg Injection 1. 5 SI-V82 6" SI to RCS Loop 3 Hot-Leg Injection 3.0 SI-V87 6" SI.to RCS Loop 2 Hot-Leg Injection 3.0 RH-V15 6" RHR to SI Loop 1 Cold-Leg Injection 3.0 3H-V29 6" RHR to SI Loop 3 Cold-Leg Injection 3.0 RH-V30 6" RHR to SI Loop 4 Cold-Leg Injection 3.0 RH-V31 6" RHR to SI Loop 2 Cold-Leg Injection 3.0 RH-V52 6" SI to RCS Loop 1 Hot-Leg Injection 3.0 RH-V53 6" SI to RCS Loop 4 Hot-Leg Injection 3.0 RH-V50 8" RHR to RCS Loop 4 Hot-Leg Injection 4.0 RH-V51 8" RHR to RCS Loop 1 Hot-Leg Injection 4.0 SI-V5 10" SI to RCS Loop 1 Cold-Leg Injection 5.0 SI-V6 10" SI Tank 9A Discharge Isolation 5.0 SI-V20 10" SI to RCS Loop 2 Cold-Leg Injection 5.0 SI-V21 10" SI Tank 9B Discharge Isolation 5.0 SI-V35 10" SI to RCS Loop 3 Cold-Leg Injection 5.0 SI-V36 10" SI Tank 9C Discharge Isolation 5.0 SI-V50 10" SI to RCS Loop 4 Cold-Leg Injection 5.0 SI-VS1 10" SI Tank 90 Discharge Isolation 5.0 RC-V22 12" RHR Pump 8A Suction Isolation 5.0 RC-V23 12" RHR Pump 8A Suction Isolation 5.0 RC-V87 12" RHR Pump 88 Suction Isolation 5.0 RC-V88 12" RHR Pump 8B Suction Isolation 5.0 0 SEABROOK - UNIT 1 3/4 4-24
S l l I REACTOR COOLANT SYSTEM l V 3/4.4.7 CHEMISTRY l 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 all times. ACTION: MODES 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 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHulDOWN within the following 30 hours; 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 hours.
3 At All Other Times: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours 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 proceeding to MODE 4. SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters specified in Table 3.4-2 at least once per 72 hours.* b
\,
- Sample and analysis for dissolved oxygen is not required with T,yg 1 180 F.
SEABROOK - UNIT 1 3/4 4-25 t
- - - . - _ - _ - _ _ - - _ - - . _ - - - - - - - - - - . . ~ . - . - - - - _ - _ .
TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen * < 0.10 ppm 5 1.00 ppm Chloride < 0.15 ppm 5 1.50 ppm Fluoride 5 0.15 ppm 5 1.50 ppm
- Limit not applicable with T,yg less than or equal to 180 F.
O O SEABROOK - UNIT 1 3/4 4-26
REACTOR COOLANT SYSTEM k 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
- a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT 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. ACTION: MODES 1, 2, and 3*:
- a. With tne specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than avg 500 F within 6 hours; and q b. With_the specific activity of the reactor coolant greater than i
' 100/E microCuries per gram, be in at least HOT STANDBY with T avg less than 500 F within 6 hours. MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro-Curies per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-3 until the specific activity of the reactor coolant is restored to within its limits. SURVEILLANCE REQUIREMENTS 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 4.4-3.
*With T,yg greater than or equal to 500*F.
SEABROOK - UNIT 1 3/4 4-27 0
- 300 3
i
~
O t: \ .
~
E \ _J 250
\ i g > i s X u 200 $ \
E \ ""sme k150 8 13 B \ s b I
\
200 2 %ETgg
\'g - N b 50 a
2 o I w i i O O O 20 40 O 60 80 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131 O SEABROOK - UNIT 1 3/4 4-28
- . .. -- . . . - - . . - . ~ .. . _
i TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM )
=
TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED [
- 5 1. Gross Radioactivity At least once per 72 hours. 1,2,3,4
[ Determination
- 2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days. 1 l LENT I-131 Concentration
- 3. Radiochemical for 5 Determination
- 1 per 6 months ** 1
- 4. Isotopic Analysis for Iodine a) Once per 4 hours, whenever the 1#, 2#, 3#, 4#, 5#
- Including I-131, I-133, and I-135 specific activity exceeds 1 i pCi/ gram DOSE EQUIVALENT I-131
] or 100/E pCi/ gram of gross y radioactivity, and b) One sample between 2 and 6 hours 1, 2, 3 i following a THERMAL POWER change U$ exceeding 15% of the RATED THERMAL POWER within a 1-hour period.
*A radiochemical analysis for 5 shall consist of the quantitative, measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be used in the determination of E for the reactor coolant sample. Determination of the. contributors to E_shall be based upon those energy peaks identifiable with.a 95% confidence level.
** 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 hours or longer.
#Until the specific activity of the Reactor Coolant System is restored within its limits.
i
't,'4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS GENERAL LIMITING CONDITION FOR OPERATION 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 Fi 0ures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 100 F in any 1-hour period, t
- b. A maximum cooldown of 100 F in any 1-hour period, and s
- c. A maximum temperature change of less than or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION: With 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.sReactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS Tavg and pressure to less than 200 F and 1 500 psig, respectively, within the following 30 hours. 3 SURVEILLANCE REQUIREMENTS 4.4.9.1 The Reactor Coolant System temperature and pressure shall be deter-mined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. I
+
0 SEABROOK - UNIT 1 3/4 4-30 i
)
1
)
[ ') Controlling material: Base metal (,/ Copper content: Conservatively assumed to be 0.10 WT% (actual content = 0.06 WT%) RT initial: 40*F RT after 16 EFPY: 1/4T,110 F 3/4T,87 F Curve applicable for heatup rates up to 60 F/hr for the service period up to 16 EFPY and contains margins of 10 F and 60 psig for possible instrument errors 2800 2600 LEAK 1 CST. r j 2400 LIMIT _\ '
\ lf ] l 2200 l g
2000 ( l g 7 p 1800 Go . E5 1600 p) . %o ( l t w I / / U- e 1400-
@mw HEATUP i[ [
cnu- CURVE N/ / gg 1200 j j j y
'E ' / f },- . CRITICALITY LIMIT
$s Es 1000 / ,
/ BASED ON INSERVICE
-~~-
HYDROSTATIC TEST
~ j 800 TEMPERATURE (255 F) ---
I [I I' FOR THE SERVICE PERIDEL__ 600 / l l UP TO 16 EFPY
-- J<
400 200-i i g 100 200 300 35s 400 500 RCS ' TEMPERATURE (*F) (10 F PER DIVISION) FIGURE 3.4-2 b') REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY Q.// SEABROOK - UNIT 1 3/4 4-31
MATERIAL PROPERTY BASIS Controlling material: Base metal Copper content: Conservatively assumed to be 0.10 WT% (actual content = 0.06 WT%) RT initial: 40 F 1/4T,110 F RThafter16EFPY: 3/4T,87 F Curve applicable for cooldown rates up to 100 F/hr for the service period up to 16 EFPY and contains margins of 10 F and 60 psig for possible instrument errors 2800 2600 2400 2200 COOL DOWN LIMF15l s ! N / 2000 N
}
- 1800 !
sa > 1600 1400---- - mA ,/ wo 1200 Eui / 1000 j/
"E~ '
g' 800 A 600 on !ff 400
-M! $, 8
~r-u9 200 0
100 200 300 35e 400 RCS ' TEMPERATURE (*F) (10 F PER DIVISION) FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LOOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY SEABROOK - UNIT 1 3/4 4-32
l REACTOR COOLANT SYSTEM ( PRESSURE / TEMPERATURE LIMITS PRESSURIZER L'IMITING 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 str'uctural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least' HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig N within the following 30 hours. SURVEILLANCE REQUIREMENTS 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 . i limit at least once per 12 hours during auxiliary spray operation. O SEABROOK - UNIT 1 3/4 4-33
REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:
- a. Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig 3%, or
- b. Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 1.58 square inches.
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 329 F; MODE 5 and MODE 6 with the reactor vessel head on. ACTION:
- a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 1.58-square-inch vent within the next 8 hours.
- b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 1.58-square-inch vent within 8 hours.
- c. In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable.
O SEABROOK - UNIT 1 3/4 4-34
REACTOR COOLANT SYSTEM (-] i PRESSURE / TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering 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 isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.
4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection
. as follows:
O a. For RHR suction relief valve RC-V89
- 1) By verifying at least once per 31 days that RHR RCS Suction Isolation Valve RC-V88 is open with power to the valve operator removed, and
- 2) By verifying at least once per 12 hours that RC-V87 is open.
- b. For RHR suction relief valve RC-V24
- 1) By verifying at least once per 31 days that RC-V22 is open with power to the valve operator removed, and
- 2) By verifying at least once per 12 hours that RC-V23 is open.
- c. Testing pursuant to Specification 4.0.5.
4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overpressure protection.
.p *Except when the vent pathway is provided with a valve that is locked, sealed, e or otherwise secured in the open position, then verify these valves open at least once per 31 days. SEABROOK - UNIT 1 3/4 4-35
m g 2500 t I - . r
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' - VAUD FOR THE FIRST 16 ETPY. SETPOINT ,
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_ CONTAINS WARGIN OF 50 F FOR g U TRANSIENT EFFECTS. l f s 2000 i l i ! /j l I E i l l I k P=541 Psic. Ts160 F ! I P=412.7 + 9.597 e.016205T; T>t60 F ' l ) O 1500 ! I
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l i s f 500 , 6 200 50 100 150 200 25C 300 360 RCS TEMPERATURE ('F) l FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS 9 9 e-
. R O REACTOR COOLANT SYSTEM l \
d 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.'4.10 The structural -integrity of ASME (,ade Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:
- 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 4
_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 structural 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
,_ System temperature above 200 F. .( ) \g c. With the structural integrity of any ASME 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.
- d. The provisions of Specification 3.0.4 are not applicable.
4 SURVEILLANCE REQUIREMENTS 4.4.10 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. t SEABROOK - UNIT 1 3/4 4-37
REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolcnt System vent path consisting of one vent valve and one block valve 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:
- a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent peth is maintained closed with power removed from the valve ,
actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status
~
within 30 days, or, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With both Reactor Coolant System vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the vent valves and block valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.11.1 Each Reactor Coolant System vent path block valve not required to be closed by ACTION a. or b., above, shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel from the control room. 4.4.11.2 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,
- For an OPERABLE ~ vent path using a power-operated relief valve (PORV) as the vent path, the PORV block valve is not required to be closed.
SEABROOK - UNIT 1 3/4 4-38
_ - . _ . . . . . . . . - ._. ... . .._ -- .. -... - - ...- - -.-. . -_-.=. -.- - . ~ . -. -. i; i
.g REACTOR COOLANT SYSTEM l
,) REACTOR COOLANT SYSTEM VENTS i
1 SURVEILLANCE REQUIREMENTS t I 4.4.11.2 (Continued) , . b. Cycling each vent valve through at least one complete cycle of
. full travel from the control room, and I
- c. Verifying flow through the Reactor Coolant System vent paths during ;
! venting. 1 4 4 . t I E i I i i i i l' i o l i l ~
! SEABROOK - UNIT 1 3/4 4-39 F
[.
3/4.5 EMERGENCY CORE COOLING SYSTEMS j
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v
) 3/4.5.1 ACCUMULATORS HOT STANDBY, STARTUP, AND POWER OPERATION LIMITING-CONDITION FOR OPERATION 3.5.1.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:
- a. The isolation valve open and power removed,
- b. A contained borated water volume of between 6121 and 6596 gallons,
- c. A boron concentration of between 1900 and 2100 ppm, and
- d. A nitrogen cover pressure of between 585 and 664 psig.
APPLICABILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed isola-tion valve, restore the inoperable accumulator to OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours
,o and reduce pressurizer pressure to less than 1000 psig within the (m./ ) following 6 hours.
- 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 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 24 hours by:
- 1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and
- 2) Verifying that each accumulator isolation valve is open.
- b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulater solution; and
/^S 'V l
- Pressurizer pressure above 1000 psig.
SEABROOK - UNIT 1 3/4 5-1
EMERGENCY CORE COOLING SYSTEMS ACCUMULATORS HOT STANDBY, STARTUP,'AND POWER OPERATION SURVEILLANCE REQUIREMENTS
- 4. 5.1.1.1 (Continued)
- c. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected.
- d. At least once per 18 months by verifying that each accumulator isola-tion valve opens automatically under each of the following conditions:
- 1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, and
- 2) Upon receipt of a Safety Injection test signal.
4.5.1.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE:
- a. At least once per 31 days by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
O SEABROOK - UNIT 1 3/4 5-2
EMERGENCY CORE COOLING SYSTEMS Ch
/ ACCUMULATORS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.1.2 Each reactor coolant system accumulator isolation valve shall be shut with power removed from the valve operator.
APPLICABILITY: . MODES 4* and 5. - ACTION:
- a. With one or more accumulator isolation valve (s) open and/or power available to-the valve operator (s), immediately close the accumulator isolation valves and/or remove power from the valve operator (s).
- b. The provisions of Specification 3.0.4 are not applicable for entry into MODE 4 from MODE 3.
SURVEILLANCE REQUIREMENTS 4.5.1.2 Each accumulator isolation valve will be verified shut with power ('~')s (_ removed from the. valve operator at least once per 31 days. l l
*Within 12 hours prior to entry into MODE 3 from MODE 4 and if pressurizer pressure is greater than 1000 psig, each accumulator isolation valve shall be Os open as required by Specification 3.5.1.1.a.
SEABROOK - UNIT 1 3/4 5-3
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T avg GREATER THAN OR EQUAL TO 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE Safety Injection pump,
- c. One OPERABLE RHR heat exchanger,
- d. One OPERABLE RHR pump, and
- e. An OPERABLE flow path
- capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2, and 3**. ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in H0T SHUTDOWN within the following 6 hours,
- b. 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 6.8.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
*During MODE'3, the discharge paths of both Safety Injection pumps may be isolated by closing for a period of up to 2 hours to perform surveillance test.ing as required by Specification 4.4.6.2.2.
- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3.for the centrifugal charging pump and the Safety Injection pumps declared inoperable pursuant'to Specification 4.5.3.2 provided the centrifugal charging pump and the Safety Injection pumps are restored to OPERABLE status within at least 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first.
O SEABROOK - UNIT 1 3/4 5-4
EMERGENCY CORE COOLING SYSTEMS s ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350 F SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 24 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position SI-V-3 Accumulator Isolation Open* SI-V-17 Accumulator Isolation Open* SI-V-32 Accumulator Isolation Open* SI-V-47 Accumulator Isolation Open* SI-V-114 SI Pump to Cold-Leg Isolation Open RH-V-14 RHR Pump to Cold-Leg Isolation Open RH-V-26 RHR Pump to Cold-Leg Isolation Open RH-V-32 RHR to Hot-Leg Isolation Closed
, RH-V-70 RHR to Hot-Leg Isolation Closed d SI-V-77 SI-V-102 SI to Hot-Leg Isolation
~
SI to Hot-Leg Isolation Closed Closed
- 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 high points, and
- 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.
- c. 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:
- 1) For all accessible areas of the containment prior to establish-ing PRIMARY CONTAINMENT INTEGRITY, and
- 2) Of the areas affected within containment at the completion of ;
each containment entry when PRIMARY CONTAINMENT INTEGRITY is established.' l U
- Pressurizer pressure above 1000 psig.
SEABROOK - UNIT 1 3/4 5-5
EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350 F SURVEILLANCE REQUIREMENTS 4.5.2 (Continued)
- d. At least once per 18 months by:
- 1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that:
a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 365 psig, the interlocks prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 660 psig, the interlocks will cause the valves to automatically close.
- 2) A visual inspection of the containment sump and verifying that the subsystem 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.
- e. 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 (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and
- 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.
- f. By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:
- 1) Centrifugal charging pump, 1 2480 psid;
- 2) Safety Injection pump, 2 1445 psid; and
- 3) RHR pump, 1 176 psid.
O SEABROOK - UNIT 1 3/4 5-6
% EMERGENCY CORE COOLING SYSTEMS F
(d - ECCS SUBSYSTEMS --T,yg GREATER THAN OR EQUAL TO 350 F SURVElliANCE REQUIREMENTS 4.5.2 (Continued)
- g. By verifying the corre'ct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
- 1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and
- 2) At-least once per 18 months.
High Head SI System Intermediate Head SI System Valve Number Valve Number SI-V-143 SI-V-80 SI-V-147 SI-V-85 SI-V-151 SI-V-104 SI-V-155 SI-V-109 SI-V-117 [_ SI-V-121 i ( SI-V-125 SI-V-129
' h. By performing a' flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the l ' subsystem flow characteristics and verifying .that:
- 1) For centrifugal charging pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest ficw rate, is greater than or equal to 337 gpm, and b) The total pump flow rate is less than or equal to 550 gpm.
- 2) For Safety Injection. pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 445 gpm, and b) The total pump flow rate is less than or equal-to 660 gpm.
- 3) For RHR pump lines, with a single pump running, the sum of the
! \ injection line flow rates is greater than or equal to 2828 gpm. SEABROOK - UNIT 1 3/4 5-7
EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T avg LESS THAN 350 F LIMITING CONDITION FOR OPERATION 3.5.3.1 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE RHR heat exchanger,
- c. One OPERABLE RHR pump, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal cLarging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.
~
- b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T avg less than 350 F by use of alternate heat removal methods.
- c. 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 6.8.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
O SEABROOK - UNIT 1 3/4 5-8
f-~y EMERGENCY CORE COOLING SYSTEMS
/ i k_-)
s ECCS SUBSYSTEMS - T,y LESS THAN 350*F SURVEILLANCE REQUIREMENTS 4.5.3.1.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.'2. 4.5.3.1.2 All centrifugal charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps, shall be~ demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position withir) 4 hours-after entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325*F, whichever comes first, and at least once per 31 days thereafter, i
i !b'V 4
*An inoparable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed O' isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
SEABROOK - UNIT 1 3/4 5-9
EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T LESS THAN 350 F avg ECCS SUBSYSTEMS - T,y EQUAL TO OR LESS THAN 200 F LIMITING CONDITION FOR OPERATION 3.5.3.2 All Safety Injection pumps shall be inoperable. APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on. ACTION: With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours. SURVEILLANCE REQUIREMENTS 4.5.3.2 All Safety Injection pumps shall be demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position at least once per 31 days.
O
- An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
SEABROOK - UNIT 1 3/4 5-10
(~3 F BORON INJECTION SYSTEM
'- '3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:
- a. A minimum contained borated water volume of 477,000 gallons,
- b. A minimum boron concentration of 2000 ppm of boron,
- c. A minimum solution temperature of 50 F, and
- d. A maximum solution temperature of 98 F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT' STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. [ SURVEILLANCE REQUIREMENTS V) 4.5.4 The RWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
~1) Verifying the contained borated water volume in the tank, and
- 2) Verifying the boron concentration of the water.
- b. At.least once per 24 hours by verifying the RWST temperature.
n v SEABROOK - UNIT 1 3/4 5-11
~p) 3/4.6 CONTAINMENT SYSTEMS V 3/4.6.1 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 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: (o V)
- a. At least once per 31 days by verifying that all penetrations
- not capable of be'ing 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 except as provided in Specification 3.6.3;-
- 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 P,, 49.6 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,.
*Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured pI in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more l often than once per 92 days.
SEABROOK - UNIT 1 3/4 6-1
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT 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 a, 0.15% by weight of the containment air per 24 hours at P,, 49.6 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 Pa*
No individual penetration will be allowed to exceed 5% of the total allowed (0.05 L,); and
- c. A combined leakage rate of less than or equal to 0.60 L, for all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to Pa '
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With (a) the measured overall integrated containment leakage rate exceeding 0.75 L,, or (b) the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, r (c) the combined bypass leakage rate exceeding 0.60 L,, restore the overall integrated leakage rate to less than or equal to 0.75 L athe combined leakage rate for all penetrations and valves subject to Type B and C tests to less than 0.60 La, and the combined bypass leakage rate to less than 0.60a L prior to increasing the Reactor Coolant System temperature above 200 F. O SEABROOK - UNIT 1 3/4 6-2
,_. CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS 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) shall be conducted at 40 10 month intervals during shutdown at a pressure not less than Pa , 49.6 psig, during each 10 year service
, period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection;
- b. If any periodic Type A fails to meet 0.75 La , 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 La '
a Type A test shall be performed at least every 18 months.until two consecutive Typ'e A tests meet 0.75 L at which time the above test p a e 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 c, is in accordance with the following equation:
lLc - (L am + l o)l 5 0.25 L a where L is the measured Type A test leakage andgL is the am 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 suppl.inental test is between 0.75 L, and 1.25 La*
O SEABROOK - UNIT 1 3/4 6-3
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS 4.6.1.2 (Continued)
- d. Type B and C tests shall be conducted with gas at a pressure not less than Pa , 49.6 psig, at intervals no greater than 24 months except for tests involving:
- 1) Air locks, and
- 2) Purge supply and exhaust isolation valves with resilient material seals.
- e. The combined bypass leakage rate shall be determined to be less than or equal to 0.60 L aby applicable Type B and C tests at least once per 24 months.
- f. Purge supply and exhaust isolation valves with resilent material seals shall be tested and demonstrated OPERABLE by the requirements of Speci-fication 4.6.1.7.2 or 4.6.1.7.3, as applicable;
- g. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3; and
- h. The provisions of Specifications 4.0.2 are not applicable.
O SEABROOK - UNIT 1 3/4 6-4
l I
/
( TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS PENETRATION NO. FUNCTION RELEASE LOCATION X-16 Containment On-Line Purge Primary Auxiliary Building (Exhaust) X-17 Equipment Vent (RCDT) Waste Processing Building X-18 Containment On-Line Purge Primary Auxiliary Building (Supply) X-19 Post Accident Monitoring Primary Auxiliary Building Sample X-20 PCCW Loop A (Supply) Primary Auxiliary Building X-21 PCCW Loop A (Return) Primary Auxiliary Building X-22 PCCW Loop B (Return) Pri:::ary Auxiliary Building X-23 PCCW Loop B (Supply) Primary Auxiliary Building O X-32 Equipment and Floor Drainage (RCDT) Waste Processing Building X-34 Equipment and Floor Drainage Waste Processing Building (RC Sump) X-35A Safety Injection (Test Line) Waste Processing Building X-35B Reactor Coolant _(Pressurizer Primary Auxiliary Building Steam / Liquid Sample) X-35C Reactor Coolant Primary Auxiliary Building (RC Sample Loop I) X-35D Reactor Coolant Primary Auxiliary Building (RC Sample Loop III) X-36A Demineralized Water Demineralized Water Storage Tank (Outside) X-36B Nitrogen Gas (High Pressure) Primary Auxiliary Building X-36C Reactor Makeup Water Waste Processing Building (Tank Farm) X -37A Chemical and Volume Control Primary Auxiliary Building
,O (Letdown)
SEABROOK - UNIT 1 3/4 6-5
TABLE 3.6-1 (Continued) , SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS PENETRATION NO. FUNCTION -RELEASE LOCATION s X-37B Chemical and Volume Control ' Primary Auxiliary Building (Excess Letdown) X-38A/76A Fire Protection Fire Water Pumphouse/ Fire Water Tanks X-388/76B Combustible Gas Control Main Steam and Feedwater Pipe Chase-X-39 Spent Fuel Pool Cooling and Fuel Storage Building Cleanup X-40A Nitrogen Gas (Low Pressure) Primary Auxiliary Building X-40B PRT Sample Primary Auxiliary Builoing X-62 Fuel Transfer Tube Fuel Storage Building X-67 Service Air Main Steam and Feedwater Pipe Chase X-710/74D Leak Detection Main Steam and Feedwater Pipe Chase HVAC-1 Containment Air Purge Primary Auxiliary Building HVAC-2 Containment Air Purge Primary Auxiliary Building N.A. Equipment Hatch Outside N.A. Personnel Hatch Main St'eam and Feedwater Pipe Chase X-72/75 Combustible Gas Control Main Steam and Feedwater Pipe Chase O SEABROOK - UNIT 1 3/4 6-6
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT 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, 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 , 49.6 psig.
3 APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- 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 hours or lock the OPERABLE air lock door closed,
- 2. Operation may then continue until' performance of t.he next required overall air lock leakage test provided that the OPERABLE
~
air lock door is verified to be locked closed at least once per
'31 days,
- 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and
- 4. The provisions of Specification 3.0.4 are not applicable. ;
i
- 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 lock to GPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
*Except during entry to repair an inoperable inner door, for a cumulative time O not to exceed I hour per year.
SEABROOK - UNIT 1 3/4 6-7 l l 1
1 CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT AIR LOCKS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. Within 72 hours following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours, by s verifying that the seal leakage is less than 0.01 L as determined a
by precision flow measurements when measured for at least 30 seconds within the volume between the seal at a constant pressure of 49.6 psig;
- b. By conducting overall air lock leakage tests at not less than P3 ,
49.6 psig, and verifying the overall air lock leaftage rate is within
.; its limit:
- i i ,
- 1) At least once per 6 months,* and
- 2) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.**
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
\
L M i e i
*Thq provisions of Specification 4.0.2 are not applicable.
t i
**This represents an exemption to Appendix J, paragraph III.D.2.(b)(ii),
i of 10 CFR Part 50. 3 ',
, SEABROOK - UNIT 1 3/4 6-8
\
a
CONTAINMENT SYSTEMS V ~ PRIMARY CONTAINMENT
~ INTERNAL PRESSURE.
LIMITING CONDITION FOR OPERATION 3.6.1.4- Primary containment internal pressure shall be maintained between-
~
14.6 and 16.2 psia. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour or be in at least HOT STANDBY within the next 6 hours.and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be O within the limits at least once'per 12 hours.
~
m i. I SEABROOK - UNIT 1 3/4 6-9 i
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120 F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment average air temperature greater than 120 F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithme-tical average of the temperatures at the following locations and shall be determined at least once per 24 hours: Location
- a. Elevation 45 feet
- b. Elevation 71 feet
- c. Elevation 110 feet
- d. Elevation 130 feet O
SEABROOK - UNIT 1 3/4 6-10 L
- ~ . . - - _
CONTAINMENT SYSTEMS V PRIMARY CONTAINMENT CONTAINMENT VESSEL STRUCTURAL' INTEGRITY-LIMITING CONDITION FOR OPERATION i 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the structural integrity of the containment vessel not conforming to the
~
above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200 F. / SURVEILLANCE. REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined f -~g during the shutdown for each. Type A containment leakage rate test (reference ( ) Specification 4.6.1.2) by a visual inspection of the exposed accessible interior x- / and exterior surfaces of the vessel. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degrada-tion of the containment vessel detected during the above required inspections shall be reported to the Commission in a Special Report pursuant to Specifi-cation 6.8.2 within 15 days. ) 4 j \~~ SEABROOK - UNIT 1 3/4 6-11
--- --,,.m- . - -or, -- , - - - -,----,.m . _ . , - , .g. _,-, -, 1 -,w,. m., ~ , - .----,,-w,--,-,,r- ..,c,,7-.- -.-. m ym<.,,._ .,.
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:
- a. Each 36-inch containment shutdown purge supply and exhaust isolation valve shall be closed and locked closed, and
- b. The 8-inch containment purge supply and exhaust isolation valve (s) shall be sealed closed except when open for purge system operation for pressure control; for ALARA, respirable, and air quality consider-ations to facilitate personnel entry; and for surveillance tests that require the valve (s) to be open.
APPLICABILITY: MODES 1*, 2*, 3, and 4. ACTION:
- a. With a 36-inch containment purge supply or exhaust isolation valve open or not locked closed, close and lock close that valve or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one or more of the 8-inch containment purge supply or exhaust isolation valves open for reasons other than given in Specifica-tion 3.6.1.7.b above, close the open 8-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
- c. With one or more containment purge supply or exhaust isolation valves having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status or isolate the affected penetration (s) so that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3 within 24 hours and close the purge supply if the affected penetration is the exhaust penetration, otherwise be in at least' HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
- The 8-inch containment purge supply and exhaust isolation valves may not be opened while in MODE 1 or MODE 2 until installations of the narrow-range con-tainment pressure instrument channels and alarms are completed.
SEABROOK - UNIT 1 3/4 6-12
CONTAINMENT SYSTEMS D PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 36-inch containment purge supply and exhaust isolation valve shall be verified to be locked closed at least once per 31 days. 4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard isolation valves-with resilient material' seals in each sealed closed 36-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that'the measured leakage rate is less than or equal to 0.05 L, when pressurized to Pa* 4.6.1.7.3 'At least once per 92 days each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.01 Lawhen pressurized to P
- a 4.6.1.7.4 Each 8-inch containment purge supply and exhaust isolation valve.
shall be verified to be sealed closed or open in accordance with Specifi-cation 3.6.1.7.b at least once per 31 days. O SEABROOK - UNIT 1 3/4 6-13
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and transferring suction to the containment sump. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 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;
- b. By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 262 psi when tested pursuant to Specification 4.0.5;
- 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 Pressure-Hi-3 test signal, and
- 2) Verifying that each spray pump starts automatically on a Containment Pressure-Hi-3 test signal.
- 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.
O SEABROOK - UNIT 1 3/4 6-14
i CONTAINMENT SYSTEMS DEPRESSURIZATION AND COOLING SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:
- a. A spray additive tank containing a volume of between 9420 and 9650 gallons of between 19 and 21% by weight Na0H solution, and
- b. Two gravity. feed path's each capable of adding NaOH solution from the chemical additive tank to the Refueling' Water Storage Tank.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Spray Additive System inoperable,. restore the system to OPERABLE
-status within.72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive 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;
- b. At least once per 6 months by:
- 1) Verifying the contained solution volume in the tank, and
- 2) Verifying the concentration of the Na0H solution by chemical analysis,
- c. At least once per 18 months, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal.
L SEABROOK - UNIT 1 3/4 6-15
CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 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: With one or more of the isolation valve (s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or
- d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.3.1 Each isolation valve 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 perfor-mance of a cycling test and verification of isolation time. O SEABROOK - UNIT 1 3/4 6-16
O U ' CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE ~at least once per 18 months by:
- a. Verifying th'ta on a Phase "A" Isolation test signal, each Phase "A" Iso 1' a tion 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 Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.
4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. m O SEABROOK .- UNIT 1 3/4 6-17
CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDR 0 GEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION:
- 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 hours.
- b. With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours, 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:
- a. One volume percent hydrogen, balance nitrogen; and
- b. Four volume percent hydrogen, balance nitrogen.
O SEABROOK - UNIT 1 3/4 6-18
p CONTAINMENT SYSTEMS COMBUSTIBLE GAS CONTROL ELECTRIC HYDROGEN RECOMBINERS LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one Hydrogen Recombiner System inoperable, restore the inoperab'le system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:
- a. At least once per 6 months by verifying during a Hydrogen Recombiner i System functional test that the minimum heater sheath temperature
\ increases to greater than or equal 850 F within 90 minutes. l Upon reaching 850 F, increase the power setting to maximum power-for 2 minutes and verify that the power meter reads greater than or equal'to 65 kW; and
- b. At least once per 18 months by:
- 1) Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits,
- 2) Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and
- 3) Verifying the integrity of all~ heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any.
heater phase shall be greater than or equal to 10,000 ohms. L SEABROOK - UNIT 1 3/4 6-19
i s CONTAINMENT SYSTEMS COMBUSTIBLE GAS CONTROL HYDROGEN MIXING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.4.3 Two independent Containment Structure Recirculation Fan Systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one Containment Structure Recirculation Fan inoperable, restore the inoperable fan to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.6.4.3 Each Containment Structure Recirculation Fan System shall be demon-strated OPERABLE:
- a. At least once per 92 days on a STAGGERED TEST BASIS by starting each system from the control room and verifying that the system operates for at least 15 minutes, and
- b. At least once per 18 mcnths by verifying a system flow rate of at least 4000 cfm through the hydrogen mixing flow path.
O SEABROOK - UNIT 1 3/4 6-20 _ _ _ . _ _ _ . - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______________j
N O , CONTAINMENT SYSTEMS 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent Containment Enclosure Emergency Air Cleanup Systems shall be'0PERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one Containment Enclosure Emergency Air Cleanup System inoperable, re-store the inoperable system 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 hours. SURVEILLANCE REQUIREMENTS (o V) 4.6.5.1 Each Containment Enclosure Emergency Air Cleanup System 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 fil.ters and charcoal adsorbers and verifying that the system operates for at least 15 minutes;
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the system by:
- 1) Verifying that the cleanup system satisfies the in place pene-tration leakage testing acceptance criteria of less than 0.05%
and uses the test procedure guidance in Regulatory Posi-tions C.S.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2100 cfm i 10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*, meets the laboratory testing criteria i,
V)
- ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Rev. 2, March 1978.
SEABROOK - UNIT 1 3/4 6-21
CONTAINMENT SYSTEMS CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM SURVEILLANCE REQUIREMENTS 4.6.5.lb.2 (Continued) of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, by showing a methyl iodide penetration of less than 2.14% when tested at a temperature of 30 C and at a relative humidity of 95% in accordance with ASTM-D3803; and
- 3) Verifying a system flow rate of 2100 cfm i 10% during system operation when tested in accordance with ANSI N510-1980.
- c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 2.14% when tested at a tem-perature of 30 C and at a relative humidity of 95% in accordance with ASTM-03803.
- d. At least once per 18 months by:
- 1) 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 2100 cfm 10%,
- 2) Verifying that the system starts on a Safety Injection test signal,
- 3) Verifying that the filter cross connect valves can be manually opened,
- 4) Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch Water Gauge in the annulus within 4 minutes after a start signal, and
- e. After each complete or partial replacement of a high efficiency particulate air (HEPA) filter bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a dioctyl phthalate (00P) test aerosol while operating the system at a flow rate of 2100 cfm i 10%; and SEABROOK - UNIT 1 3/4 6-22
k CONTAINMENT SYSTEMS CONTAINMENT ENCLOSURE BUILDING 4 CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM i SURVEILLANCE REQUIREMENTS 4.6.5.1 (Continued) l
- f. After each complete or partial replacement.of a charcoal adsorber ,
bank, by verifying that the cleanup system satisfies the in place l penetration leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated. hydrocarbon refrigerant test gas while operating the system at a flow rate of . 2100 cfm t 10%. ) I .i i j l 1 ! SEABROOK - UNIT 1 3/4 6-23
CONTAINMENT SYSTEMS CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE BUILDING INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.2 CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3, and 4. < < ACTION: ' Without CONTAINMENT ENCLOSURE BUILDING INTEGRITY, restore CONTAINMENT ENCLOSURE BUILDING INTEGRITY within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.5.2 CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall be demonstrated at least once per 31 days by verifying that the door in each access opening is closed except when the access opening is being used for normal transit entry and exit. O SEABROOK - UNIT 1 3/4 6-24
bN.j' CONTAINMENT SYSTEMS CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.3 The structural integrity of the containment enclosure building shall be maintained at a' level consistent with the acceptance criteria in Specification 4.6.5.3. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the structural integrity of the containment enclosure building not con-forming to the above requirements, restore the structural integrity to within the limits within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS
-- 4.6.5.3 The structural integrity of the containment enclosure building shall
'\ be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed acces-
~s ible interior and exterior surfaces of the containment enclosure building and
~ verifying no apparent changes in appearance of the concrete surfaces or other .
I abnormal degradation. Any abnormal degradation of the containment enclosure building detected during the above required inspections shall be reported to
- the Commission in a Special Report pursuant to Specification 6.8.2 within 15 days.
i SEABROOK - UNIT 1 3/4 6-25
-3/4.7 PLANT SYSTEMS l, q l b) 3/4.7.1 TURBINE CYCLE l SAFETY 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:
- a. With four 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 hours either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-l'; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. The provisions of Specification 3.0.4 are not applicable.
2 g v)' SURVEILLANCE REQUIREMENTS
- 4. 7.1.1' No additional requirements other than those required by Speci--
fication 4.0.5. t I O SEABROOK - UNIT 1 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 MAXIMUM NUMBER OF IN0PERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 87 2 65 3 43 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER Loop 1 Loop 2 Loop 3 Loop 4 LIFT SETTING (i 1%)* ORIFICE SIZE V6 V22 V36 V50 1185 psig 16.0 sq. in. O V7 V23 V37 V51 1203 psig 16.0 sq. in. V8 V24 V38 V52 1220 psig 16.0 sq. in. V9 V25 V39 V53 1238 psig 16.0 sq. in. V10 V26 V40 V54 1255 psig 16.0 sq. in.
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating tenperature and pressure.
O SEABROOK - UNIT 1 3/4 7-2
m , PLANT SYSTEMS !I-v ) TURBINE CYCLE l l AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
- a. One motor-driven emergency feedwater pump, and one startup feedwater pump capable of being powered from an emergency bus and capable of being aligned to the dedicated water volume in the condensate storage tank, and
- b. One steam turbine-driven emergancy feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in H0T SHUTDOWN within the following 6 hours.
U
- b. With two emergency feedwater pumps inoperable, restore at least one emergency feedwater pump to OPERABLE status within 12 hours and restore both emergency feedwater pumps to OPERABLE status within 72 hours, or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With one emergency feedwater pump and the startup feedwater pump inoperable, restore both emergency feedwater pumps to OPERABLE status within 24 hours and all three pumps to OPERABLE status within 72 hours ,
or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- d. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1) Verifying that the motor-driven emergency feedwater pump
(" develops a discharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpm; SEABROOK - UNIT 1 3/4 7-3 1
PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.la. (Continued)
- 2) Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpm when the secondary steam supply pressure is greater than 500 psig. The provisions of Specification 4.0.4 are not applicable.for entry into MODE 3;
- 3) Verifying that the startup feedwater pump develops a discharge pressure of greater than or equal to 1375 psig at a flow of greater than or equal to 425 gpm.
- 4) 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; S) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER; and 6)' Verifying that valves FW-156 and FW-163 are operable for alignment of the startup feedwater pump to the emergency feedwater header.
- b. At least once per 18 months during shutdown by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal;
- 2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal;
- 3) Verifying that with all manual actions, including power source and valve alignment, the startup feedwater pump starts within the required elapsed time; and
- 4) Verifying that each emergency feedwater control valve closes on receipt of a high flow test signal.
O SEABROOK - UNIT 1 3/4 7-4
PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.2 Auxiliary feedwater flow paths to each steam generator shall-be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days, or after maintenance on an auxiliary feedwater pump that could have an effect upon pump performance, prior to entering MODE 2 by verifying normal flow to each steam generator from:
- a. Each emergency feedwater pump, and
- b. The startup feedwater pump via the main feedwater flow path and via the emergency feedwater header, i
t l SEABROOK - UNIT 1 3/4 7-5
- - - . . . -,_-,,,,--v_~~- , - , - . - - , , . . , , - - , ~ - - - - - + - - - , , - , - - . . . . . - - _ , _ . .,- - .-ro-.-- y..,m -.~ - , .-,
PLANT SYSTEMS TURBINE CYCLE CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) system shall be OPERABLE with
- a. A volume of 212,000 gallons of water contained in the condensate storage tank, and
- b. A concrete CST enclosure that is capable of retaining 212,000 gal-lons of water.
APPLICABILITY: MODES 1, 2, and 3. ACTION: With the CST or the CST enclosure inoperable, within 4 hours restore the CST and the CST enclosure to OPERABLE status or be in at least H0T STANDBY wi'5in the next 6 hours and in H0T SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.3 The CST and the CST enclosure shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume in the CST is within its limits and the CST enclosure integrity is maintained. O SEABROOK - UNIT 1 3/4 7-6
PLANT SYSTEMS TURBINE CYCLE 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 0.1 microcurie / gram DOSE EQUIVALENT I-131. AFPLICABILITY: -MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD. SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 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. ~ O j O SEABROOK'- UNIT 1 3/4 7-7
TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY
- 1. Gross Radioactivity At least once per 72 hours.
Determination
- 2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10% 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 10% of the allowable limit for radiciodines. O SEABROOK - UNIT 1 3/4 7-8
PLANT SYSTEMS TURBINE CYCLE MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: MODE 1: With one MSIV inoperable but ope'n, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY'within the next 6 hours and in HOT SHUT 00WN l within the following 6 hours. - l l MODES 2 and 3: , ,
~~
l With one MSIV inoperable, subsequent operation in MODE 2 ori 3 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT l STANDBY within the next 6 hours and in HOT SHUTDOWN within the following l . 6 hours. < SURVEILLANCE REQUIREMENTS ,' 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure ; within 5.0 seconds when tested pursuant to Specification 4.0:5. The i provisions of Specification 4.0.4 are not applicable for entry into MODE 3. I ) k iO SEABROOK - UNIT 1 3/4 7-9 L .. __ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
PLANT SYSTEMS TURBINE CYCLE ATMOSPHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION
- 3. 7.1. 6 At least four. atmospheric relief valves and associated manual controls including the safety-related gas supply systems shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.* ACTION:
- a. With one less than the required atmospheric relief valves OPERABLE, restore the required atmospheric relief valves to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 12 hours,
- b. With two less than the required atmospheric relief valves OPERABLE, restore at least three atmospheric relief valves to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.6 Each atmospheric relief valve and associated manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:
- a. At least once per 24 hours by verifying that the nitrogen accumulator tank is at a pressure greater than or equal to 500 psig.
- b. Prior to startup following any refueling shutdown or cold shutdown of 30 days or longer, verify that all valves will open and close fully by operation of manual controls.
*When steam generators are being used for decay heat removal.
'SEABROOK - UNIT 1 3/4 7-10
1
-i 5 s
k 11s \ PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION' c' LIMITTNG CONDITION FOR OPERATION 3.7.2 The temperatures of both the reactor and secondary coolants in the steam _ generators shall be greater than 70*F when the pressure of either l coolant in the steam generator is greater than 200 psig. APPLICABILITY: At all times. ' s ACTION: With the requirements'of the above specification not satisfied: 5 i }
- a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and j b. Perform an engineering evaluation to deter'mine the effect of
')
! the overpressurization on the structural ~ integrity of the
; steam generator. Determine that the steam generator ' remains acceptable for continued operation prior to increasing its ;
temperatures above 200 F. O SURVEILLANCE REQUIREMENTS 4.7.2 The pressure in each side of the stean, generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the reactor or secondary coolant is less than 70 F. S l s 1
\
t : v
~SEABROOK - UNIT 1 3/4 7-11
PLANT SYSTEMS 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent primary component cooling water loops shall be OPERABLE, including two OPERABLE pumps in each loop. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one primary component cooling water (PCCW) pump inoperable, restore the required primary component cooling water pumps to OPERABLE status within 7 days or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With two-primary component cooling water pumps inoperable, restore at least one of the inoperable primary component cooling water pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With two primary component cooling water pumps within one loop inoperable, restore at least one of the inoperable pumps to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.3 At least two primary component cooling water loops shall be demonstrated OPERABLE:
- a. 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 its associated Engineered Safety Fea-ture actuation signal, and
- 2) Each of the four primary Component Cooling Water System pump starts automatically upon loss of or failure to start of its redundant pump within the loop.
O SEABROOK - UNIT 1 3/4 7-12
PLANT SYSTEMS n$w) s 3/4.7.4 SERVICE' WATER SYSTEM LIMITING CONDITION FOR.0PERATION 3.7.4 At least two independent service water loops shall be OPERABLE, ia-cluding three OPERABLE pumps in each loop. - APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one service water pump inoperable, restore the' required service water pumps to OPERABLE status within 7 days or be in at least HOT l STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With two service water pumps inoperable, restore at least one of the inoperable service water pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With two service water pumps within one loop inoperable, restore at least one of the inoperable pumps to OPERABLE status within 24 hours 3' or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- d. With one cooling tower service water pump inoperable, restore the required cooling tower service water pump to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS I 4.7.4 At least two Station Service Water loops shall be demonstrated OPERABLE:
- a. 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 its associated Engineered Safety Fea-ture actuation test signal, and
- 2) Each of the four Station pumps aligned to the ocean ultimate heat sink (UHS) starts automatically upon loss of or failure m to start of the redundant pump within the loop and each of the l two pumps aligned to the cooling tower UHS starts on a cooling
~ d tower actuation (TA) signal.
SEABROOK - UNIT 1 3/4 7-13
PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with:
- a. A service water pumphouse water level at or above 5'-0", minus 36'-0" Mean Sea Level, USGS datum, and
- b. A mechanical draft cooling tower comprised of one cooling tower cell with one OPERABLE fan and a second cell with two OPERABLE fans, and a contained basin water level of equal to or greater than 42.15* feet at a bulk average water temperature of less than or equal to 67.3 F, and
- c. A portable tower makeup pump system stored to be OPERABLE for 30 days following a Safe Shutdown Earthquake.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With the service water pumphouse inoperable, restore the service water pumphouse to OPERABLE status within 72 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the mechanical draft cooling tower inoperable, restore the cooling tower to OPERABLE status within 72 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With the portable tower makeup pump system inoperable, continue operation and notify the NRC within 1 hour in accordance with the procedure of 10 CFR 50.72 of actions or contingencies to ensure an adequate supply of makeup water to the mechanical draft cooling tower for a minimum of 30 days.
- With the cooling tower in operation with valves aligned for tunnel heat treat-ment, the tower basin level shall be maintained at greater than or equal to 40.55 feet.
SEABROOK - UNIT 1 3/4 7-14
PLANT SYSTEMS l'LTIMATE HEAT SINK SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE:
- a. At least once per 24 hours by:
- 1) Verifying the water level in the service water pumphouse to be at or above 5'-0", minus 36'-0" Mean Sea Level, and
- 2) Verifying the water in the mechanical draft cooling tower basin to be greater than or equal to a level of 42.15 feet.
- b. At least once per week by verifying that the water in the mechanical draft cooling tower basin to be at a bulk average temperature of less than or equal to 67.3 F.
- c. At least once per 31 days by:
- 1) Starting from the control room each UHS cooling tower fan that is required to be OPERABLE and operating each of those fans for at least 15 minutes, and
- 2) Verifying that the portable tower makeup pump system is stored in its design operational readiness state.
- d. At least once per 18 months by verifying automatic actuation of each cooling tower fan on a Tower Actuation test signal.
I l l SEABROOK - UNIT 1 3/4 7-15
PLANT SYSTEMS 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two Control Room Area Ventilation Systems shall be OPERABLE. APPLICABILITY: All MODES. ACTION: MODES 1, 2, 3, and 4: With one Control Room Area Ventilation System inoperable, restore the inoperable system 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 hours. MODES 5 and 6:
- a. .With one Control Room Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Area Ventilation System in the recirculation mode.
- b. With both Control Room Area Ventilation Systems inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Area Ventilation System shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that the Control Room Area Ventilation System is maintaining the tempc a u e of equipment and instrumentation in the control room area . low its limiting equipment qualification temperature.
- b. At least once per 18 months or after any significant modificati7n to the Control Room Area Ventilation Systems by verifying a system flow rate of 25,700 cfm 10% through the air conditioner unit (3A and 3B) and a flow rate of at least 1200 cfm i 10% makeup from each intake to the emergency filtration unit with a discharge of 2000 cfm i 10% from the filtration unit.
O SEABROOK - UNIT 1 3/4 7-16
PLANT SYSTEMS CONTROL ROOM AREA VENTILATION SYSlEM SURVEILLANCE REQUIREMENTS 4.7.6 (Continued)
- c. At least once per 18 months by:
- 1) Verifying that on a high radiation signal from the control room makeup air intake, the subsystem automatically switches to the emergency recirculation mode of operation and the isolation dampers close within 5 seconds.
- 2) Verifying that on an S signal the emergency filtration fans start.
- 3) Verifying that the system maintains the control room area at a positive pressure of greater than or equal to a pressurization 1/8-inch Water Gauge relative to adjacent areas during system operation at a pressurization flow of 1200 cfm 10%.
PLANT SYSTEMS 3/4.7.7 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.7 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 installed 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: With one or more snubbers inoperable on any system, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation in accordance with the approved augmented inservice inspection pro-gram on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.7 Each snubber shall be demonstrated OPERABLE by performance of the re-quirements of an approved augmented inservice inspection program. O SEABROOK - UNIT 1 3/4 7-18
PLANT SYSTEMS h 3/4.7.8 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.8 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. 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 provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
[ \ v SURVEILLANCE REQUIREMENTS 4.7.8.1 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. 4.7.8.2 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.
SEABROOK --UNIT 1 3/4 7-19
PLANT SYSTEMS SEALED SOURCE CONTAMINATION SURVEILLANCE REQUIREMENTS 4.7.8.2 (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.
4.7.8.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination. O O SEABROOK - UNIT 1 3/4 7-20 1
~
1. 1 3-s- PLANT SYSTEMS
- 3/4.7.9 .'(This specification number is'not used')
4 1 ! 1 ,
- +
1^ . j-m 5 } i 1' b 1 l l ! i i e l i i 4 i 4-t i 1 O SEABROOK~- UNIT 1 3/4 7-21 i
PLANT SYSTEMS 3/4.7.10 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.10 The temperature of each area shown in Table-3.7-3 shall not be exceeded for more than 8 hours or by more than 30 F. APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. ACTION:
- a. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-3 for more than 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specifi-cations 3.0.3 and 3.0.4 are not applicable.
- b. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-3 by more than 30 F, prepare and submit a Special Report as required by ACTION a. above and within 4 hours either restore the area (s) to within the temperature limit (s) or declare the equip-ment in the affected area (s) inoperable.
SURVEILLANCE REQUIREMENTS 4.7.10 The temperature in each of the areas shown in Table 3.7-3 shall be determined to be within its limit at least once per 12 hours. O SEABROOK - UNIT 1 3/4 7-22
l I jr ss T'ABLE 3.7-3
\s_- AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT ( F) l 1.~ Control Room 75
- 2. Cable' Spreading Room 104
- 3. Switchgear Room - Train A 104
- 4. Switchgear Room'- Train B. ~ 104
- 5. . Battery Rooms - Trai.n A .97
.6. Battery Rooms - Train B 97
- 7. ECCS_ Equipment Vault - Train.A 104
- 8. ECCS Equipment Vault --Train B 104'
- 9. Centrifugal Charging Pump Room - Train A 104
- 10. Centrifugal Charging Pump Room - Train B 104
- 11. ECCS Equipment Vault Stairwell - Train A 104
- 12. ECCS Equipment Vault Stairwell - Train B 104
- 13. PCCW Pump Area 104
- 14. _ Cooling Tower Switchgear Room . Train A 104
- 15. Cooling Tower Switchgear Room - Train B 104
- 16. Cooling' Tower SW Pump Area .127
- 17. SW Pumphouse Electrical Room - Train A 104
- 18. SW Pumphouse Electrical Room - Train B 104
- 19. SW Pump Area 104
- 20. Diesel Generator Room - Train A 120
- 21. Diesel Generator Room - Train B l 0-22. 23.
EFW Pumphouse 120 104 Electrical Penetration Area - Train A 100
- 24. Electrical Penetration Area - Train B 85
- 25. Fuel Storage Building Spent Fuel Pool Cooling 104 Pump Area
- 26. Main Steam and Feedwater Pipe Chase - East 130 l
- 27. Main Steam.and Feedwater Pipe _ Chase - West 130 i l
I l SEABROOK - UNIT 1 3/4 7-23
_m 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION
- 3. 8.1.1 As a minimum, the following A.C. electrical power' sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System, and
- b. Two separate and independent diesel generators, each with:
- 1) A separate day fuel tank containing a minimum fuel volume fraction of 3/8 (600 gallons),
, 2) A separate Fuel Storage System containing a minimum volume of 60,000 gallons of fuel,
- 3) A separate fuel transfer. pump,
- 4) Lubricating oil storage containing a minimum total volume of
/~~% 275 gallons of lubricating oil, and b
- 5) Capability to transfer lubricating oil from storage to the diesel generator unit.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specification 4.8.1.1.la. within 1 hour'and at least once per 8 hours thereafter and~ Specification 4.8.1.1.2a.5) within 24 hours; restore at least two offsite circuits to OPERABLE status within 24 hours and two diesel generators to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-tion 4.8.1.1.la. within 1 hour and at least once per 8 hours there-after and Specification 4.8.1.1.2a.5) within 24 hours; restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least y two offsite circuits to OPERABLE status within 24 hours and two SEABROOK - UNIT 1 3/4 8-1 1
ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 (Continued) ACTION: diesel generators to OPERABLE status within 72 hours from the time of initial loss or be in at least H0T STANDBY within the~ next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With one diesel generator inoperable in addition to ACTION a. or
- b. above, verify that:
- 1. All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and
- 2. When in MODE 1, 2, or 3, the steam-driven emergency feedwater pump is OPERABLE.
If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing the requirements of Specification 4.8.1.1.2a.5) within 1 hour and at least once per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least H0T STANDBY within the next 6 hours. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 24 hours from time of initial loss or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours
- e. With two of the ab <e required diesel generators inoperable, demon-strate the OPERABILITY of two offsite A.C. circuits by performing the requirements of Specification 4.8.1.1.la. within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SEABROOK - UNIT 1 3/4 8-2
n
~
% ELECTRICAL POWER SYSTEMS A.C SOURCES
.0PER'ATING SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite
~
transmission. network and the Onsite Class 1E Distribution System.shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at.least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the normal circuit to the alternate circuit.
4.8.1.1.2 Each diesel generator shall b'e demonstrated OPERABLE:
- a. In accordance~with the frequency specified in Table 4.8-1-on a STAGGERED TEST BASIS by:
- 1) Verifying-the fuel level in the day fuel tank; 2)' Verifying the fuel level in the fuel storage tank;- '
- 3) Verifying.the fuel transfer. pump starts and transfers fuel from the. storage system to the day tank;
- 4) . Verifying the lubricating oil inventory in' storage;
- 5) Verifying the diesel starts'from ambient condition and acceler-ates to at least 514-rpm'in less than'or equal to 10. seconds.*
The generator voltage and frequency shall be. 4160 420 volts and 60 1.2 Hz within 10 seconds
- after the start signal. The diesel generator shall be started for this test by using one of the following signals:
a)~ Manual, or b) Simulated loss-of-offsite power by itself,'or
*All diesel generator starts for the purpose of this surveillance test may be
-preceded by an engine prelube period. Further, all' surveillance tests and all other engine starts for the purpose of-.this surveillance tests, with'the exception of once per 184 days, may-also be preceded by warmup. procedures (e.g.,
gradual acceleration and/or gradual loading greater than 60 seconds) as recommended by the manufacturer so that the mechanical stress and. wear on the diesel engine is minimized. O SEABROOK - UNIT 1 3/4 8-3
ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) c) Simulated loss-of-offsite power in conjunction with.an SI Actuation test signal, or d) An SI Actuation test signal by itself.
- 6) Verifying the generator is synchronized, loaded to greater than or equal to 6083 kW in less than or equal to 120 seconds *,
and operates with a load greater than or equal to 6083 kW for at least 60 minutes; and
- 7) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
- b. At-least once per 31 days and after each cperation of the diesel where the period of operation was greater than or equal to 1 hour by checking for and removing accumulated water from the day fuel tank;
- c. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks;
- d. By sampling new fuel oil in accordance with ASTM-04057-81 prior to addition to storage tanks and:
- 1) By verifying in accordance with the tests specified in ASTM-0975-81 prior to addition to the storage tanks that the sample has:
a) An API Gravity of within 0.3 degree at 60 F, or a specific gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60 F of greater than or equal to 0.81 but less than or equal to 0.89, or an API gravity of greater than or equal to 28 degrees but less than or equal to 42 degrees;
- All diesel generator starts for the purpose of this surveillance test may be preceeded by an engine prelube period. Further, all surveillance tests and all other engine starts for the purpose of this surveillance tests, with the exception of once per 184 days, may also be preceded by warmup procedures (e.g.,
gradual acceleration and/or gradual loading greater than 60 seconds) as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized. O SEABROOK - UNIT 1 3/4 8-4
,m ELECTRICAL POWER SYSTEMS V) A.C. SOURCES OPERATING
-SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) b) A kinematic viscosity at 40 C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification;.
c) A flash point greater than or equal to 125 F; and d)- A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82.
- 2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with~ ASTM-D975-81 except that the analysis for sulfur may be performed in a'cordance c with ASTM-D1552-79 or ASTM-D2622-82.
D T(g e. At least once every 31 days
- 1) By obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method A, and
- 2) By-visually inspecting the lagging in the area of the flanged joints on.the silencer outlet of the diesel exhaust system for leakage (also after an extended operation of greater than 24 hours).
- f. At least once per 18 months, during shutdown, by:
1
- 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service;
- 2) Verifying the generator capability to reject a load of-greater
~
than or equal to 671 kW while maintaining voltage at 4160 i 420 volts and frequency at 60 4.0 Hz; 3). Verifying the generator capability to reject a load of 6083 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the' load rejection;
) 4) Simulating a loss-of-offsite power by itself, and:
SEABROOK - UNIT 1 3/4 8-5
ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts on the loss of offsite power signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected shutdown loads through the emergency power sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 420 volts and 60 1.2 Hz during this test.
- 5) Verifying that on an SI actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 420 volts and 60 1.2 Hz within 10 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test;
- 6) Simulating a loss-of-offsite power in conjunction with an SI actuation test signal; and a) Verifying deenergization of the emergency busses and load shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the emergency power sequencer and operates for greater than or equal to 5 mi'nutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 420 volts and 60 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engine overspeed, low lube oil pressure, 4160-volt bus fault, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concur-rent with a Safety Injection actuation signal.
SEABROOK - UNIT 1 3/4 8-6
i I
\
' ELECTRICAL POWER SYSTEMS-
/ .
b A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.'1.1.2 (Continued)
- 7) Verifying the diesel generator operates .for at least 24 hours.
During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 6697.kW and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 6083 kW. The generator voltage and frequency shall be 4160 420 volts and 60 1.2 Hz within 10 seconds after the start signal; the steady-state generator-voltage ano frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2f.6)b);*
- 8) Verifying that the auto-connected loads to each diesel generator do not exceed the sho'rt time rating of 6697 kW;
- 9) Verifying the diesel generator's capability to:
p (' a) Synchronize'with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power,
.b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
- 10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;
- 11) Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the in-stalled cross-connection lines;
- 12) Verifying that the emergency power sequence timer is OPERABLE with the interval between each load block within i 10% of its design interval;
*If Specification 4.8.1.1.2f.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator
[d may be operated at 6083 'kW for 1 hour or until operating temperature has stabilized. SEABROOK - UNIT 1 3/4 8-7
ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) 13)- Verifying that the following diesel generator lockout features prevent diesel generator starting: a) Barring device engaged, or b) Differential lockout relay.
- 14) Simulating a Tower Actuation (TA) signal while the diesel generator is loaded with the permanently connected-loads and auto-connected emergency (accident) loads, and verifying that the service water pump automatically trips, and that the cool-ing tower pump and fan (s) automatically start. After energiza-tion the steady state voltage and frequency of the emergency buses shall be maintained at 4160 1 420 volts and 60 1.2 Hz; and
- 15) While diesel generator 1A is loaded with the permanently connected loads and auto-connected emergency (accident) loads, manually connect the 1500 hp startup feedwater pump to 4160-volt bus ES. After energization the steady-state voltage and frequency of the emergency bus shall be maintained at 4160 1 420 volts and 60 1 1.2 Hz.
- g. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 10 seconds; and
- h. At least once per 10 years by:
- 1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, or equivalent, and
- 2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.
O SEABROOK - UNIT 1 3/4 8-8
t
' ELECTRICAL POWER SYSTEMS O A.C. SOURCES 0PERATING SURVEILLANCE REQUIREMENTS i 4.8.1.1.3 Reports - All diesel generator failures, valid or nonvalid, shall be reported to.the Commission in a Special Report pursuant to Specification 6.8.2-within 30' days. Reports of diesel generator failures shall include the informa-tion recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revi-
- sion 1, August 1977. If the number of failures in the last 100 valid tests (on
- a per nuclear unit basis) is greater than or equal to 7,
- the report'shall be-
; supplemented to include the additional information recommended in Regulatory
- . . Position C.3.b of Regulatory Guide 1.108,. Revision 1, August 1977.'
i j a 1-( t f. i 1 4 l i. i i 4 F r SEABROOK - UNIT 1 3/4 8-9 I
._.y, -
~ ~ - -
TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN NUMBER OF FAILURES IN LAST 20 VALID TESTS
- LAST 100 VALID TESTS
- TEST FREQUENCY
<1 <5 At least once per 31 days
[2 [6 At least once per 7 days ** O
- Criteria for determining the number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August, 1977, but determined on a per diesel generator basis.
For the purpose of determining the required test frequency, the previous l test failure count may be reduced _ to zero if a complete diesel overhaul to l like-new condition is completed, provided that the overhaul, including l appropriate post-maintenance operation and testing, is specifica1ly approved ! by the manufacturer ard if acceptable reliability has been demonstrated. The l reliability criterion shall be the successful completion of 14 consecutive ! tests in a single series. Ten of these tests shall be in accordance with the routine Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 and four tests in accordance with the 184-day testing requirement of Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6. If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to transvalue the failure count to zero requires NRC approval. ! C*This test frequency shall be maintained until seven consecutive failure-free demands have been performed and the number of failures in the last 20 valid l demands has been reduced to one or less. SEABROOK - UNIT 1 3/4 8-10
ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a-minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit.between the offsite transmission network and the Onsite Class 1E Distrib~ution System, and
- b. One diesel generator with:
- 1) A day fuel tank containing a minimum fuel volume fraction of 3/8 (600 gallons of fuel),
- 2) A fuel storage system containing a minimum volume of 60,000 gallons of fuel,
- 3) A fuel transfer pump,
- 4) Lubricating oil storage containing a minimum ~ total volume of n 275 gallons of lubricating oil, and C h
- 5) Capability to transfer lubricating oil from storage to the diesel generator unit.
APPLICABILITY: MODES.5 and 6. ACTION: With less t'han the above minimum required A.C. electrical power sources-OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and within 8 hours, depressurize and vent the Reactor Coolant. System through a greater than or equal to 1.58-square-inch' vent. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor
-vessel flange,'immediately initiate correcti.ve action to restore the required sources to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.la', 4.8.1.1.2a [except for Specification 4.8.1.1.2a.6], and 4.8.1.1.3. /N (J SEABROOK - UNIT 1 3/4 8-11
ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE and energized:
- a. Train A
- 1) 125-volt Battery Banks 1A and IC,
- 2) One full-capacity battery charger on Bus #11A, and-
- 3) One full-capacity battery charger on Bus #11C.
- b. Train B
- 1) 125-volt Battery Banks 1B and 1D,
- 2) One full-capacity battery charger on Bus #11B, and
- 3) One full-capacity battery charger on Bus #110.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the required battery banks in one train inoperable, close the bus tie to connect the remaining operable battery bank to the'D.C.
bus supplied by the inoperable battery bank within 2 hours; restore the inoperable battery bank to OPERABLE status within 30 days
- or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one of the full-capacity chargers inoperable, restore the inoper-able charger to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The parameters in Table 4.8-2 meet the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 128 volts on float charge.
- b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below-110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:
- No more than one battery at a time may be taken out of service for more than 30 days.
SEABROOK - UNIT 1 3/4 8-12
x ELECTRICAL POWER SYSTEMS (, ) (/ D.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.2.lb (Continued)
- 1) The parameters in Table 4.8-2 meet the Category B limits,
- 2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 6 ohm,* and
- 3) The average electrolyte temperature of 16 connected cells (4 cells per row) is above 65 F.
- c. At least once per 18 months by verifying that:
- 1) The cells, cell plates, and battery racks show no visual.
indication of physical damage or abnormal deterioration,
- 2) The cell-to-cell and terminal connections are clean', tight, and
("'T coated with anticorrosion material, \
) \# 3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 6 ohm,* and
- 4) Each battery charger will supply at least 150 amperes at a minimum of 132 volts for at least 8 hours.
- d. At least once per 18 months, c*uring shutdown, by verifying that the battery capacity is adequate to supply and maintain'in OPERABLE status all of the actual or simulated emergency loads for the design duty' cycle when the battery is subjected to a battery service test;
- e. At'least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60-month interval this' performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and
- f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has ~ reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
p (j *0btained by subtracting the . normal resistance of: (1) the cross room rack connector (210 x 10 6 ohm, typical) and (2) the bi-level rack connector (35 x 10 6 ohm, typical) from the measured cell-to-cell conection resistance. SEABROOK - UNIT 1 3/4 8-13
TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2) LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOT- LIMITS FOR EACH VALUE FOR EACH PARAMETER CELL CONNECTED CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mark, ' indication mark, plates, and < " above and C h" above and not maximum level maximuin leve.1 overflowing indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (6) > 2.07 volts Not more than 0.020 below the
- " 9' *' '
Specific > 1.200(5) > 1.195 connected cells Gravity (4) s Average of all Average of all connected cells connected cells
> 1.205 > 1.195(5)
TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Cate-gory B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within their allowable values and provided the Category B parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery. (4) Corrected for electrolyte temperature and level. (5) Or battery charging current is less than 2 amps when on float charge. (6) Corrected for average electrolyte temperature. O SEABROOK - UNIT 1 3/4 8-14
i s 6 ELECTRICAL POWER SYSTEMS
~
w/ - D.C. SOURCES , ESHUTDOWN
., LIMITING CONDITION FOP JPERATION-3.8.2.2 As a minimum, two~ 125-volt battery banks in one D.C. Train and the associated full-capacity chargers shall be.0PERABLE.
APPLICABILITY: MODES.5 and 6. ACTION:
- a. With one of'the required b'att'ery banks inoperable, immediately.close the-bus tie to the alternate.0PERABLE battery.
- b. With both required battery banks and/or full-capacity chargers inoperable, ,
immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of-irradiated fuel;' initiate corrective-
. action-to restore the required battery banks and full-capacity chargers to-OPERABLE status as soon as possible,.and within 8-hours, depressurize and vent'the Reactor. Coolant System through a 1.58-square-inch vent.
-SURVEILLANCE REQUIREMENTS 4.8.2.2 The above required 125-volt battery b'anks and full-capacity chargers.
shall be demonstrated OPERABLE.in accordance with= Specification 4.8.2.1. O SEABROOK - UNIT 1 3/4 8-15
i ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION s OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified manner:
- a. Train A, A.C. Emergency Busses consisting of:
- 1) 4160-volt Emergency Bus #E5,
- 2) 480-volt Emergency Bus #E51,** and
- 3) 480-volt Emergency Bus #E52.**
- b. Train B, A.C. Emergency Busses ccnsisting of:
- 1) 4160-volt Emergency Bus #E6,
- 2) 480-volt Emergency Bus #E61,**
- 3) 480-volt Emergency Bus #E62,** and
- 4) 480-volt Emergency Bus #E64.
- c. 120-volt A.C. Vital Panel #1A energized from its associated inverter connected to D.C. Bus #11A,*
- d. 120-volt A.C. Vital Panel #1B. energized from its associated inverter connected to D.C. Bus #11B,*
- e. 120-volt A.C. Vital Panel #1C energized from its associated inverter connected to D.C. Bus #11C,*
- f. 120-volt A.C. Vital Panel.#10 energized from its associated inverter connected to D.C. Bus #110,*'
- g. 120-volt A.C. Vital Panel #1E energized from its associated inverter connected to D.C. Bus #11A,*
- h. 120-volt,d.C.'(italPanel#1Fenergizedfromitsassociatedinverter connected to D.C. Bus #11B,*
*Two ir;Jerters may be disconnected from their D.C. bus for up to 24 hours as necessary, for the purpose of performing an equalizing charge on their associated battery bank provided: (1) their vital busses are energized, and (2) the vital busses associated with the other battery bank are energized from their associated inverters and connected to their associated D.C. bus.
**These busses can be considered OPERABLE if the 480 volt bus ties are closed.
These bus ties will be under administrative control to ensure loading is within transformer rating. SEABROOK - UNIT 1 3/4 8-16
. - ~. .. - . . ~ .. ~ , , _. . . . -. . _ .
. ELECTRICAL POWER-SYSTEMS
- 0NSITE POWER' DISTRIBUTION !
OPERATING MMITINGCONDITIONFOROPERATION j 4 3.8.3.1 (Continued).
- 1. Train.A,125-volt D.C. Busses consisting of:
- 1) 125-volt D.C. Bus #11A energized from Battery Bank 1A* or 1C*, and 7 2) 125-volt D.C. Bus #11C energized from Battery Bank 1C* or 1A*.
C j. Train B,125 volt D.C. Busses consisting of:
- 1) .125-volt D.C. Bus #11B energized from Battery Bank 1B* or 1D*, and
- 2) 125-volt D.C. Bus #11D energized from Battery Bank 1D* or 1B*.
APPLICABILITY: MODES 1, 2, 3,.and 4. ACTION:
- a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the train within 8 hours or be in at least HOT STANDBY within'the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
.L b. With one A.C. vital panel either not energized from its assoc _iated ; f inverter, or with the inverter not' connected to its associated D.C. I bus: (1) reenergize the A.C. Vital panel within'2 hours or be in at L least H0T STANDBY within the next 6 hours and'in COLD SHUTDOWN within the following 30 hours;.;and (2) reenergize the A.C. vital .4 panel from its associated inverter connected to its associated D.C. 4-i bus within 24 hours or be in at least HOT STANDBY within the next
- 6. hours and in COLD SHUTDOWN within the following 30 hours.
- c. With one.D.C. bus not energized from its associated battery bank, reenergize the D.C.-bus from its associated battery bank or close the bus tie to the alternate OPERABLE battery of the same train 4
within 2 hours
- or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.3.1 The'specified busses and panels shall be determined energizdd in the , required manner.at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. i *No more.than one Battery Bank (1A,18,1C, or 10) at a time may be taken out of service.for more than 30 days. SEABROOK - UNIT 1 3/4 8-17
~ _ _ _ .__. _ _ _ _ _ _ _ _ _ _ _ _ .. _ _ _ . . _ , _ _ _ .
ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:
- a. One train of A.C. emergency busses consisting of the 4160-volt and the 480-volt A.C. emergency busses listed in 3.8.3.la. and b.
(excluding 480-volt Emergency Bus #E64);
- b. Two of the four 120-volt A.C. vital Panels 1A, 1B, 1C, and ID energized from their associated inverters connected to their respective D.C. busses;
- c. One of the two 120-volt A.C. Vital Panels 1E or IF energized from its associated inverter connected to the respective D.C. bus; and
- d. Two 125-volt D.C. busses (in the same train) energized from their associated battery banks.
APPLICABILITY MODES 5 and 6. ACTION: With any of the above required electrical busses and panels not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses and panels in the specified manner as soon as possible, and within 8 hours, depressurize and vent the RCS through at least a 1.58-square-inch vent. SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses and panels shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. O SEABROOK - UNIT 1 3/4 8-18
. . _ ._ _ _ . . ~ . . _ _ .. . _
2 s, ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION ^ TRIP CIRCUIT FOR INVERTER'I-2A LIMITING CONDITION FOR OPERATION 3'8.3.3 The safety-related trip circuit t' hat trips the D.C. feed from D.C. i Bus #11C to inverter #I-2A after 15 minutes of discharge from the battery.
'shall be OPERABLE. Note that this LIMITING CONDITION FOR OPERATION is appli-cable only when D.C. Bus #11C is required to be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, 4, 5, and 6.' ACTION: With this safety-related trip circuit inoperable, restore the trip circuit to OPERABLE status within 7 days or de energize the D.C. feed to inverter #I-2A by tripping the D.C. circuit breaker in D.C. Bus #11C. Verify that~this breaker is open once per 7 days thereafter.
-SURVEILLANCE REQUIREMENTS
(
\
4.8.3.3 The safety-related trip circuit'shall be demonstrated operable at least once per 18 months. I 1 4 a r SEABROOK - UNIT 1 3/4 8-19
ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. CIRCUITS INSIDE PRIMARY CONTAINMENT LIMITING CONDITION FOR OPERATION 3.8.4.1 The ci,cuit breakers feeding the following loads inside primary con-tainment shall be padlocked in the open position: Loads Circuit Panel l Refueling Canal Skimmer Pump 1-SF-P-272 1-ED-MCC-111 - Polar Gantry Crane 1-MM-CR-3 1-ED-US-11 Distribution Panel 1-ED-PP-7A 1-ED-US-11 Distribution Panel 1-ED-PP-7B 1-ED-US-23 Rod Control Cluster Change Fixture 1-FH-RE 12 1-ED-MCC-111 APPLICABILITY: MODES 1, 2, and 3. ACTION: With any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel (s) within 1 hour. EXCEPTION: If any of the above-mentioned loads are required for brief durations (not to exceed 72 hours) during plant operation, the pertinent circuit breaker can be unlocked and closed for the required duration provided this change in breaker position becomes part of the applicable operating procedure used for the work inside containment. SURVEILLANCE REQUIREMENTS 4.8.4.1 Verify at least once per 31 days that the circuit breakers listed above are locked in the open position. O SEABROOK - UNIT 1 3/4 8-20 l
ELECTRICAL POWER SYSTEMS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES AND PROTECTIVE DEVICES FOR CLASS IE POWER SOURCES CONNECTED TO NON-CLASS IE CIRCUITS LIMITING CONDITION FOR OPERATION 3.8.4.2 Each containment penetration conductor overcurrent protective device and each protective device for Class IE power sources connected to non-class IE circuits shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, 4, 5,* and 6.* ACTION:
- a. With one or more of the containment penetration conductor overcurrent pro-tective device (s) inoperable:
- 1) Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated circuit breaker or racking out or removing'the inoperable protective device within 72 hours, declare the affected system or component ino'perable, and verify the circuit breaker to be tripped or the inoperable protective device to be racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, or gj their inoperable protective devices racked out, or removed; or
- 2) Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one or more of the Class IE power source protective device (s) inoper-able, restore the protective device (s) to-OPERABLE status or deenergize the circuit (s) by tripping the circuit breaker or racking out or removing the inoperable protective device within 72 hours, deleare the affected com-ponent inoperable, and verify the circuit breaker to be tripped or the inoperable protective device to be racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their circuit breakers tripped, or their inoperable protective devices racked out, or removed.
SURVEILLANCE REQUIREMENTS 4.8.4.2 Each containment penetration conductor overcurrent and Class IE power source protective device shall be demonstrated OPERABLE:
- a. At least once per 18 months:
- 1) By verifying that the medium voltage 13.8-kV and 4.16-kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least one of the circuit breakers, and performing the following:
( ') (/ *0nly for Class IE power source protective devices. SEABROOK - UNIT 1 3/4 8-21
ELECTRICAL POWER SYSTEMS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES . CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES AND PROTECTIVE DEVICES FOR CLASS IE POWER SOURCES CONNECTED TO NONCLASS IE CIRCUITS SURVEILLANCE REQUIREMENTS 4.8.4.2.a.1) (Continued) a) A CHANNEL CALIBRATION of the associated protective relays (because of the large currents involved, it is impractical to inject primary side signals to current transformers; therefore, the channel calibration will be performed by injecting a signal on the secondary side of those trans- = formers at their test plug), b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and c) For each circuit breaker found inoperable during these func- m tional tests, one additional circuit breaker of the inoper-able type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
- 2) By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers and overload devices. Circuit breakers and overload devices selected for functional testing shall be selected on a rotating basis.
Testing of air circuit breakers shall consist of injecting a cur-rent with a value equal to 300% of the pickup of the long-time delay trip element and 150% of the pickup of the short-time delay trip element. The instantaneous element shall be tested by inject-ing a current equal to 20% of the pickup value of the element. Testing of thermal magnetic molded-case circuit breakers shall consist of injecting a current with a value equal to 300% of the circuit breaker trip rating and -25% to +40% of the circuit breaker instantaneous trip range or setpoint. Testing of combination starters (a magnetic only molded-case circuit breaker in series with a motor starter and integral overload device) shall consist of injecting a curr.ent with a value equal.to -25% to +40% of the circuit breaker instantaneous trip setpoint, and 200% and 300% of the thermal overload device trip rating to the respective devices. Circuit breakers and/or overload devices found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker and or overload de-vices found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers and or overload devices of the inoperable type shall also be func-tionally tested until no more failures are found or all circuit breakers and or overload devices of that type have been func-tionally tested. SEABROOK - UNIT 1 3/4 8-22
' ELECTRICAL POWER SYSTEMS O
ELECTRICAL ~ EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SURVEILLANCE REQUIREMENTS 4.8.4.2.a (Continued)
- 3) Corrective actions for any generic degradation of overcurrent protective devices, such as setpoint drift, manufacturing deficiencies, material defects, etc., shall-be applicable to all (Class 1E and non-Class 1E) protective devices of identical design.
- b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.
\ SEABROOK - UNIT 1 3/4 8-23
ELECTRICAL POWER SYSTEMS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION LIMITING CONDITION FOR OPERATION 3.8.4.3 Each thermal overload protection for safety related motor-operated valves shall be OPERABLE. APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE. ACTION: With the thermal overload protection for one or more of the above-required valves inoperable, bypass the inoperable thermal overload within 8 hours, restore the inoperable thermal overload to OPERABLE status within 30 days, or declare the affected valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected system (s). SURVEILLANCE REQUIR'EMENTS 4.8.4.3 The thermal overload protection for the above required valves shall be demonstrated OPERABLE at least once per 18 months and following maintenance on the motor starter by selection of a representative sample of at least 25% of all thermal overloads for the above-required valves and replacing them with precalibrated devices that have been subjected to a CHANNEL CALIBRATION. O SEABROOK - UNIT 1 3/4 8-24
j- . 3/4.9 REFUELING OPERATIONS e i bl 3/4.9.1- BORON CONCENTRATION LIMITING' CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled ^ portions of the Reactor Coolant System and the refuel.ing canal shall be maintained uniform and sufficient to ensure that the more' restrictive of the following reactivity conditions is met; either:
- a. A k,ff of 0.95 or -less, or
- b. A boron concentration of greater than or equal to 2000 ppm.
APPLICABILITY: MODE 6.* ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity-changes and initiate and continue boration_at greater than or equal-to 30 gpm
~of a solution containing greater than or equal to 7000 ppm boron or its equivalent until k,77 is reduced to less than or equal to 0.95 or the boron-concentration is restored to greater than or equal to 2000 ppm, whichever is the more restrictive.
v SURVEILLANCE REQUIREMENTS
- 4. 9.1.1 The more restrictive of the above two reactiv'ity conditions shall be determined prior to:
- a. Removing or unbolting the' reactor vessel head, and
- b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal.shall be determined by chemical analysis at least once per 72 hours.
*The reactor shall be maintained in MODE 6 whenever fuel is in the reactor ) vessel with the vessel head closure bolts less than fully tensioned or with v the head removed.
SEABROOK - UNIT 1 3/4 9-1
-REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment and control room.
APPLICABILITY: MODE 6. ACTION:
- a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
- b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours.
SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE
-by performance of:
- a. A CHANNEL CHECK at least once per 12 hours,
- b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
O SEABROOK - UNIT 1 3/4 9-2
. . . _ . _________-_____.-____________________m______
REFUELING OPERATIONS ~ ss. - 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION
~3.~9.3 --The reactor shall be subtritical for at least 100 hours.
APPLICABILITY: During movement of irradiated fuel-in the reactor vessel. ACTION: With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. SURVEILLANCE REQUIREMENTS
'4 9.3 The reactor shall be determined to have been-subcritical for at least 100 hours by verification of the'date and. time of subcriticality prior to movement of irradiated fuel in the reactor. vessel.
\ ( SEABROOK --UNIT 1 3/4 9-3 ,
REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door closed and held in place by a minimum of four bolts,
- b. A minimum of one door in each airlock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1) Closed by an isolation valve, blind flange, or manual valve; or
- 2) Be capable of being closed by an OPERABLE automatic containment "
purge and exhaust isolation valve. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. SURVEILLANCE REQUIREMENTS
- 4. 9.' 4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:
- a. Verifying the penetrations are in their closed / isolated condition, or
- b. Testing the containment purge and exhaust isolation valves per the applicable portions of Specification 4.6.3.2.
4 O SEABROOK - UNIT 1 3/4 9-4
REFUELING OPERATIONS O \ V 3/4.9.5 COMMUNICATIONS l LIMITING CONDITION FOR OPERATION
-3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.
APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct communications'between the control room and personnel at the refueling station cannot be' maintained, suspend all CORE ALTERATIONS. SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room'and personnel at the refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS. O h SEABROOK - UNIT 1 3/4 9-5
REFUELING GPERATIONS 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine and auxiliary hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:
- a. The refueling machine used for movement of fuel assemblies having:
- 1) A minimum capacity of 4000 pounds, and
- 2) An overload cutoff limit less than or equal to 3900 pounds. -
- b. The auxiliary hoist used for latching and unlatching drive rods having:
- 1) A minimum capacity of 2100 pounds, and
- 2) A load indicator which shall be used to prevent lifting loads in excess of 1000 pounds. {
APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel. ACTION: With the' requirements for refueling machine and/or hoist OPERABILITY not satis-fled, suspend use of any inoperable refueling machine and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 4000 pounds and demonstrating an automatic load cutoff when the refueling machine load exceeds 3900 pounds. 4.9.6.2 The auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 2100 pounds. O SEABROOK - UNIT 1 3/4 9-6
p REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS LIMITING' CONDITION FOR OPERATION 3.9.7 Loads in excess of 2100 pounds shall be prohibited from travel over fuel assemblies in the storage pool. APPLICABILITY: With fuel assemblies in the storage pool. i ACTION:
- a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are'not applicable.
SURVEILLANCE REQUIREMENTS 4.9.7'-Crane interlocks that prevent crane travel with loads in excess of 2100 pounds over fuel assemblies sha11 be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. l l l
-1 l
O SEABROOK - UNIT 1 3/4 9-7 I
REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2750 gpm at least once per 12 hours.
- The RHP loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
SEABROOK - UNIT 1 3/4 9-8
1 p REFUELING OPERATIONS RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION LOW WATER LEVEL LIMITING CON 9ITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in. operation.* APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet. ACTI0'N:
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal,to 23 feet of water above the reactor vessel flange, 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. Close all containment penetrations providing A direct access from the containment atmosphere to the outside atmosphere within 4 hours.
SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2750.gpm at least once per 12 hours. 1
,o
- Prior to initial criticality, the RHR loop may be removed from operation for (v) up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
SEABROOK - UNIT 1 3/4 9-9
)
REFUELING OPERATIONS 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust Isolation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION:
- a. With the Containment Purge and Exhaust Isolation System inoperable, close each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Purge and Exhaust Isolation System shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment purge and exhaust isolation occurs on manual initiation and on a High Radiation test signal from each of the manipulator crane area radiation monitoring instrumentation channels. O SEABROOK - UNIT 1 3/4 9-10
, REFUELING OPERATIONS /
k_,)\ s 3/4.9.10 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange. APPLICABILITY: During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6. ACTION: With the requirements of the above specification not satisfied,' suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies or control rods. o) l
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( ) 'N_ l l l SEABROOK - UNIT 1 3/4 9-11 l 1 l
REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. ACTION:
- a. With the requirements of the above specification not satisfied, suspend all movement of. fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. O SEABROOK - UNIT 1 3/4 9-12
,q REFUELING OPERATIONS L 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 Two independent trains of the Fuel Storage Building Emergency Air Cleaning System shall.be OPERABLE whenever irradiated fuel is in the storage pool and shall be OPERABLE with one train operating during fuel movement.
APPLICABILITY: Whenever irradiated fuel is in the storage pool.
~
ACTION:
- a. With one train of the Fuel Storage Building Emergency Air Cleaning System inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the OPERABLE train of the Fuel Storage Building Emergency Air System is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.
i
- b. With no trains of the Fuel Storage Building Emergency Air Cleaning System OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the O storage pool until at least one train of the Fuel Storage Building
(") Emergency Air Cleaning System is restored to OPERABLE status and is in operation.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS 4.9.12 The above required trains of the Fuel Storage Building Emergency Air Cleaning System 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 adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;
- b. At least once per 18 months or (1) after any structural maintenance l on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
O-SEABROOK - UNIT 1 3/4 9-13
REFUELING OPERATIONS FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM SURVEILLANCE REQUIREMENTS 4.9.12b (Continued)
- 1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,* and the system flow rate is 16,450 cfm 10%;
- 2) Verifying, within 31 days after removal, that a laboratory.
analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 1.0% when tested at a temperature of 30 C and at a relative hu-midity of 95% in accordance with ASTM-D-3803; and
- 3) Verifying a system flow rate of 16,450 cfm i 10% during system operation when tested in accordance with ANSI N510-1980.
- c. After every 720 hours of charcoal adsorber operation by 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,*
meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,* by showing a methyl iodide penetration of less than 1.0% when tested at a tem-perature of 30 C and at a relative humidity of 95% in accordance with ASTM-D-3803.
- d. At least once per 18 months by:
- 1) 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 16,450 cfm i 10%,
- 2) Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere during system operation,
- ANSI N510-1980 shall be used in place of ANSI N510-1975 as referenced in Regulatory Guide 1.52, Rev. 2, March 1978.
SEABROOK - UNIT 1 3/4 9-14
REFUELING OPERATIONS FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM SURVEILLANCE REQUIREMENTS I 4.9.12d (Continued)
- 3) Verifying that the filter cross connect valve can be manually opened, and
- 4) Verifying that the heaters dissipate 95 i 11 kW when tested in accordance with ANSI N510-1980.
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a 00P test aerosol while operating the system at a flow rate of 16,450 cfm.1 10%.
- f. After each complete or partial replacement of-a charcoal adsorber
- bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05%
in accordance with ANSI N510-1980 for a halogenated hydrocarbon re-frigerant test gas while operating the system at a flow rate of 16,450 cfm i 10%. k SEABROOK - UNIT 1 3/4 9-15
3/4.10 SPECIAL TEST EXCEPTIONS ( ) (j 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2. ACTION:
- a. With any full-length control rod not fully inserted and with less than the_above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immedi-ately initiate and continue boration at greater than or equal to 30'gpm of a solution containing greater than or equal to 7000 ppm
( boron or its equivalent until the SHUTDOWN MARGIN required by L Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS 4.10.1.1 .The position of each full-length control .od either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. N , SEABROOK - UNIT 1 3/4 10-1
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SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
- b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.
APPLICABILITY: MODE 1. ACTION: With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
- b. Be in HOT STAND 8Y within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.2 The requirements of Specifications 4.2.2.2, 4.2.2.3, and 4.2.3.2 shall be performed at least once per 12 hours during PHYSICS TESTS: O SEABROOK - UNIT 1 3/4 10-2 L_.
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SPECIAL TEST' EXCEPTIONS , i . V '3/4.10.3' PHYSICS ~ TESTS '
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. LIMITING CONDITION FOR OPERATION -
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-3.-10.3 'The'1 imitations of Specifications 3.1.1.3, 321;ir4, 3.1.3.1, 3.1.3.5, and'3.1.3.6 may be suspended during the performance of PHYSICS TESTS'provided:
- a. TheTHERMALPOWERdoesnotexceed5[nfRATEDTHERMALPOWER,-
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- b. . The Reactor Trip Setpoliits on the OPERABLE' Intermediate and Power Range channels are set at less than o,tr THERMAL POWER, and. " equal to 25% of RATED
+- .
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- c. The Reactor Coolant System best op$ rating loop' temperature (Tavg is greater than or equal to S41*F.
y APPLICABILITY: MODE'2.' '[ ACTION: o+
- a. With the' THERMAL POWER greater than 5% d ' RATED' THERMAL POWER, immediately open the Reactor trip breakers..
~
b. With a Reactor Coolant System operatin0 loopetempeiature (T"V9) less than 541*F, restore T,yg to within its, limit >within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. J .
~ -
SURVEILLANCE REQUIREMENTS . 4.10.3.1 The THERMAL POWER shall be' determined to be less than c[ edual' to 5% of RATED THERMAL POWER 'at least once per' hour during PHYSICS,1ESTS..
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4.10.3.2 Each Intermediate and Power Range channel shall..be subjected to an ANALOGCHANNELOPERATIONALTEST41 thin 12hoursUEior'i.8InitiatingPHYSICS TESTS. ' C 4.10.3.3 The Reactor Coolant System temperature (T,. ) shall be determined to be greater than or equal to 541*F at least once per 30 minutes'during PHYSICS TESTS.~ . . J
.SEABROOK - UNIT 1 3/4 10-3 b_
SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of STARTUP and PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, cnd
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
APPLICABILITY: During operation below the P-7 Interlock Setpoint. ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint during the performance of STARTUP and PHYSICS TESTS, immediately open the Reactor trip breakers. SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during STARTUP and PHYSICS TESTS. 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating STARTUP and PHYSICS TESTS. O SEABROOK - UNIT 1 3/4 10-4
i i I [
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SPECIAL TEST EXCEPTIONS- i V 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN ' s LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 m.ay be suspended during the _ performance of individual full-length shutdown and control rod drop time j measurements provided; ; I { , a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time', and
- b. The rcd position indicator is OPERABLE during the withdrawal of-the i
rods.* ' l APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements. ACTION: I With the Position Indication Systems inoperable or with more than one bank o' rods withdrawn, immediately open the Reactor trip breakers. ' s SURVEILLANCE REQUIREMENTS 4.10.5 The above r'equir'ed Position Indication Systems shall be' determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop ti.ne measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps when the rods are stationary. (
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i
*This requirement is not applicable during the initial calibration of the s Digital Rod Position Indication System provided: (1) k is maintained eff O' less'than or equal to 0.95, and (2) only one shutdown or control rod bank is withdrawn from the fully inserted position at one time.
e , SEABROOK - UNIT 1 3/4 10-5 h _ _ ------------
'3/4.11 RADI0 ACTIVE EFFLUENTS '/q \
( ,)_ 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION i 3.11.1.1 The concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser (see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or' entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcurie /ml total activity. APPLICABILITY: At all times. ACTION: With the concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser exceeding the above limits, restore the concentration to within the above limits within 15 minutes. SURVEILLANCE REQUIREMENTS 3 4.11.1.1.1 . Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Part A of the ODCH. 4 4.11.1-1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. T SEABROOK - UNIT 1 3/4 11-1
. -- - - - . - - - -......-.-..-.=. . - ..... - - - ,
RADI0 ACTIVE EFFLUENTS LIQUID EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and
- b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. O SEABROOK - UNIT 1 3/4 11-2
-RADI0 ACTIVE EFFLUENTS IyV) LIQUID EFFLUENTS LIQUID RADWASTE TREATMENT ~ SYSTEM LIMITING CONDITION FOR OPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate-portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid _ effluent, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem-to any organ in a 31-day period.
APPLICABILITY: At all times. ACTION:
- a. With racicactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System which could reduce the radioactive liquid waste dis-charged not in operation, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that includes the following information:
('N 1. Explanation of why liquid radwaste was being discharged without
' N]
l treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The. provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized. 4.11.1.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2. v SEABROOK - UNIT 1 3/4 11-3 l
RADI0 ACTIVE EFFLUENTS LIQUID EFFLUENTS LIQUID HOLDUP TANKS
- LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each temporary unprotected outdoor tank shall be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases:
APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any temporary unprotected outdoor tank exceeding the above limit, immediately suspend all addi-tions of radioactive material to the tank, within 48 hours reduce the tank contents to within the limit, and describe the events leading to this condition-in the next Semiannual Radioactive Effluent Release Re-port, pursuant to Specification 6.8.1.4.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS O 4.11.1.4 The quantity of radioactive material contained in each temporary unprotected outdoor tank shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.
- Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows.and surrounding area drains connected to the Liquid Radwaste Treatment System.
SEABROOK - UNIT 1 3/4 11-4
,, ._ ~ - . . - - _ _ - . _ , _ _ . . - - - .
t t - . 1 RADI0 ACTIVE EFFLUENTS 4-4
- 3/4.11.2 GASEOUS EFFLUENTS l-DOSE RATE
~ LIMITING CONDITION FOR OPERATION ,
3.11.2.1 The dose rate due-to radioactive materials released in gaseous.
- effluents'from the site to areas at and beyond the SITE B0UNDARY-(see Figure 5.1-1) shall be limited to the following: 1 4
- a. For noble gases: _. Less than or equal'to 500 mrems/yr to the whole n
body.and less than or equal to 3000 mrems/yr to the skin, and !' b. For Iodine-131, fo~r Iodine-133, for~ tritium, and for all radio-nuclides,in particulate form with half-lives greater than 8 days: e Less than or equal to 1500 mrems/yr to any organ. APPLICABILITY: At all times. 2 ACTION: With the dose rate (s) exceeding the above limits, decrease the release rate within 15 minutes'to within the above limit (s). SURVEILLANCE REQUIREMENTS. F 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents'shall be ~ determined to be within the~above limits in accordance with the methodology
- and parameters in the ODCM.
i 4.11.2.1.2 The dose rate due to. Iodine-131, Iodine-133, tritium, and all ^ - radionuclides in' particulate form with half-lives greater than'8 days in gaseous effluents shall-be determined to be within the'above limits in
~
accordance with the methodology and parameters.in the ODCM by obtaining' l l representative samples and performing analyses in accordance with the sampling i j and analysis program specified in Part A of the ODCM. ' i j j-i SEABROOK -' UNIT 1 3/4 11-5 i
-~ s v. ,m ....,._,m._._.~..,..-_..,-.-_.,,.._._..--.._.___..._-._-..---,.,~..,..,__,..._,c.-
RADIOACTIVE EFFLUENTS l GASE0US EFFLUENTS DOSE - NOBLE GASES l LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE B0UNDARY (see Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation-and less than or equal to 10 mrads for beta radiation, and
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times. ACTION
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. O SEABROOK - UNIT 1 3/4 11-6
(9
/
/
RADIOACTIVE EFFLUENTS
' GASEOUS EFFLUENTS DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL-IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides _in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE B0UNDARY (see Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
- b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and. radionuclides in particulate form with half-lives (Vn) greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit the the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be.in compliance with the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131, Iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCH at least once per 31 days. jh 4 d SEABROOK - UNIT 1 3/4 11-7
RADI0 ACTIVE EFFLUENTS GASEOUS EFFLUENTS GASE0US RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and the GASE0US RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and be-yond the SITE B0UNDARY (see Figure 5.1-1) would exceed:
- a. 0.2 mrad to air from gamma radiation, or
- b. 0.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
APPLICABILITY: At all times. ACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized. 4.11.2.4.2 The installed VEliTILATION EXHAUST TREATMEllT SYSTEM and GASE0US RADWASTE TREATMENT SYSTEM shall be considered OPERABLE by meeting Specifica-tions 3.11.2.1 and 3.11.2.2 or 3.11.2.3. O SEABROOK - UNIT 1 3/4 11-8
-l RADI0 ACTIVE EFFLUENTS D GASEOUS EFFLUENTS -
EXPLOSIVE GAS MIXTURE - SYSTEM i LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the GASEOUS RADWASTE TREATMENT SYSTEM shall be limited to less than or equal to 2%'by volume. APPLICABILITY: At all times. ACTION: [
- a. With the' concentration of oxygen. in the GASEOUS RAN;A'STE TREATMENT ' i SYSTEM greater than 2% by volume,-reduce the oxygen concentration to the above limit within 48 hours unless the hydrogen concentration is verified to be less,than 4% by volume.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS x . . 4.11.2.5 The concentration of hydrogen or oxygen in the GASEOUS RADWASTE. TREAT-MENT SYSTEM shall be determined to.be within the above limit by continuously monitoring the waste gases in the GASEOUS RADWASTE TREATMENT SYSTEM with the. hydrogen or oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.10. b i SEABROOK - UNIT 1 3/4 11-9 e - wam,+' = -- . .---
RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTES LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and' transportation requirements during transit, and disposal site requirements when received at the disposal site. APPLICABILITY: At all times. ACTION:
- a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures, and/or the Solid Waste System as necessary to prevent recurrence,
- b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.3 For cement SOLIDIFICATION of at least one representative test speci-men from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:
- a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM;
- b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.12, to assure SOLIDIFICATION of subsequent batches of waste; and O SEABROOK - UNIT 1 3/4 11-10
s I RADI0 ACTIVE EFFLUENTS SOLID RADIOACTIVE WASTES l l b SURVEILLANCE REQUIREMENTS 4.11.3 (Continued) 4
- c. With the installed equipment incapable'of meeting Specification 3.11.3 or declared inoperable, restore the equipment to 0PERABLE status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.
I i ! 4 5 4 , i i
.r 4
1
- J 4
. SE'ABROOK - UNIT 1 3/4 11-11
RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel-cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. APPLICABILITY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or 3.11.2.3b., calculations shall be made including direct radiation contributions from the units and from outside storage tanks to deter-mine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.8.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.
This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) ccvered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition result-ing in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM. 4.11.4.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in ACTION a. of Specification 3.11.4. SEABROOK - UNIT 1 3/4 11-12
m 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING
/ T
\
/ 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION i
3.12.1 The Radiological Environmental Monitoring Program (REMP) shall be I conducted as specified in the ODCM. APPLICABILITY: At all times. ACTION:
- a. With the Radiological Environmental Monitoring Program not being conducted as specified, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.8.1.3, a description _of the reasons for not conducting
'he program as required and the plans for preventing a recurrence.
- b. nith the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of the REMP when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from receipt of the laboratory analyses, pursuant to Specification 6.8.'2, a Special Report that identifies the cause(s) for exceeding the
~
_A limit (s) and defines the corrective actions *o be taken to reduce radioactive effluents so that the potential annual. dose
- to a MEMBER (V) 0F THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than one of the radio-nuclides in the REMP are detected in the sampling medium, this report shall be submitted if:
concentration (1) . concentration (2) . ***3 1.0 reporting level (1) + reporting level (2) - When radionuclides other than those listed in the REMP are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and describeo in the Annual Radiological Environmental Operating Report required by Specification 6.8.1.3.
O} G
*The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
SEABROOK - UNIT 1 3/4 12-1
RADIOLOGICAL ENVIRONMENTAL MONITORING MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 (Continued) ACTION:
- c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by the Radiological Environ-mental Monitoring Program (REMP), identify specific locations for obtaining replacement samples and add them within 30 days to the REMP given in the ODCM. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Specification 6.13, submit in the next Semiannual Radioactive Ef-fluent Release Report documentation for a~ change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new loca , -
tion (s) with supporting information identifying the cause of the un-availability of samples and justifying the selection of the new loca-tion (s) for obtaining samples.
- d. The provisions of Specifications 3.0.3 and 3.'0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to the REMP from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of and the detection capabilities required by the REMP. O SEABROOK - UNIT 1 3/4 12-2
[-s RADIOLOGICAL ENVIRONMENTAL MONITORING V) 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden
- of' greater than 50 m2 (500 ft 2)' producing broad leaf vegetation.
APPLICABILITY: At all' times. ACTION:
- a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification ~4.11.2.3, pursuant to Specifi_ca-tion 6.8.1.4, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report.
- b. With a Land Use Census identifying a location (s) that yields a calcu-lated dose or dose commitment (via the same exposure pathway) 20%
greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new location (s) within 30 days to the Radiological Environmenta1 Moni-(mj toring Program given in the ODCM, if permission.from the owner to V _ collect samples can be obtained and sufficient sample volume is available. The sampling location (s), excluding the control station location, having the lowest calculated dose or. dose commitment (s), via the same exposure pathway, may~be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Specification 6.13, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s).for the ODCM reflecting the new location (s) with information supporting the change in sampling locations.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
4
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed'at the SITE B0UNDARY_in each of two different direction sec-tors with the highest predicted relative deposition values (D/Qs) in lieu of (Vp) the garden census. Specifications for broad leaf vegetation sampling in the REMP, shall be followed, including analysis of control samples.
SEABROOK - UNIT 1 3/4 12-3
RADIOLOGICAL ENVIRONMENTAL MONITORING LAND USE CENSUS SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months using a method such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities, as described in the ODCM. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.8.1.'3. O O SEABROOK - UNIT 1 3/4 12-4
. . . .- . ~ .
_f s- RADIOLOGICAL ENVIRONMENTAL MONITORING V) I 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 -Analyses shall be performed on all radioactive materials, supplied as part of an.Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by the REMP. APPLICABILITY: At all times. ACTION:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission
~in the Annual Radiological Environmental Operating Report pursuant to Specification 6.8.1.3.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS. e s- 4.12.3 The Interlaboratory Comparison Program shall be identified in the-ODCM.
~
A summary of the results obtained as part of the above required Interlaboratory (\d) Comparison Program shall be included-in the Annual Radiological Environmental Operating Report pursuant to Specification 6.8.1.3. SEABROOK - UNIT 1 3/4 12-5
4 _ a .a. --a m -_J % mm. 2_-ii -. _ . .. E.1.u u. A.e-m -_ __ l 1 l BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I i
~
l l 1 l i O l NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. O 1 I l { l l l l l O
, 3/4.0 APPLICABILITY (m I
\.j -
BASES The specifications of this section provide the general requirements appli-cable to each of the Limiting Conditions for Operation and Surveillance Re-quirements within Sections 3.0 and 4.0. In the event.of a disagreement between the requirements stated in these Technical Specifications and those stated in an applicable Federal Regulation or Act, the requirements stated in the appli-cable Federal Regulation or Act shall take precedence and shall be met. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL' MODES or other specified conditions and is pro-- vided to delineate specifically when each specification is applicable. l 3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and
~
associated ACTION requirement. 3.0.3 The specification delineates the measures to be taken for those circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specifi-cation 3.5.2~ requires two independent ECCS subsystems to be OPERABLE and pro-vides explicit ACTION requirements if one ECCS subsystem is inoperable. Under the requirements of Specification 3.0.3, if both the required ECCS subsystems (N are inoperable, within 1 hour measures must be initiated to place the unit in (*) at least HOT STANDBY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 nours. As a.further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit t ACTION requirements if one Spray System is inoperable. Under the requirements I of Specification 3.0.3, if both the required Containment Spray Systems are in-operable, within 1 hour measures must be initiated to place the unit'in at least HOT STANDBY within the next 6 hours, in at least HOT SHUTDOWN withir, the follow-l . i nD 6 hours, and in COLD SHUTDOWN within the subsequent 24 hours. It is ac-ceptable.to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused por-tion of this allowable out-of-service time to that provided for operation in subsequent lower OPERATIONAL MODE (S). Stated allowable out of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability is discovered, but the times provided for achieving a-mode reduction are not l applicable if the inoperability is discovered in a mode lower than the appli-I cable mode. For example if the Containment Spray System was discovered to be l inoperable while in STARTUP, the ACTION Statement would allow up to 156 hours to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours rather than the allowed 78 hours, 140 hours would still be available before the plant would be required to be in COLD SHUTDOWN However, if this system was discovered to be inoperable while in HOT STANDBY, the 6 hours provided to achieve HOT STANDBY would not be additive to the time available to' achieve COLD SHUT 00WN so that the total allowable time is reduced from 156 hours to 150 hours. 3.0.4 This specification provides that entry into an OPERATIONAL MODE or O other specified applicability condition must be made with: (1) the' full com-plement of required systems, equipment, or components OPERABLE and (2) all SEABROOK - UNIT 1 B 3/4 0-1
APPLICABILITY BASES O other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out-of-service provisions con-tained in the ACTION statements. The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other speci-fied limits being exceeded. Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications. 4.0.1 This specification provides that surveillance activities necessary to ensure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements.
~
Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification. 4.0.2 This specification provides allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance in-terval. These tolerances are necessary to provide operational flexibility be-cause of scheduling and performance considerations. The phrase "at least" as-sociated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities. The tolerance values, taken either individually or consecutively over three test intervals, are suf ficiently restrictive to ensure that the reli-ability associated with the surveillance activity is not significantly degraded beyond that obtained'from the nominal specified interval. 4.0.3 This specification provides the criteria for determination of com-pliance with the OPERABILITY requirements of the Limiting Conditions for Opera- . tion. Under these criteria, equipment, systems, or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. No, thing in this provision is to be construed as defining equipment, systems, or components OPERABLE when such items are found or known to be inoperable although still meeting the Surveil-lance Requirements. Items may be determined inoperable during use, during surveillance tests, or in accordance with this specification. Therefore, ACTION statements are entered when the Surveillance Requirements should have been performed rather than at the time it is discovered that the tests were not performed. 4.0.4 This specification ensures that the surveillance activities asso- ) ciated with a Limiting Condition for Operation have been performed within the SEABROOK - UNIT 1 B 3/4 0-2
)
. _ _ - . . _ . _ - - . . - _ = . - . . . . . - .. .
k APPLICABILITY O_ BASES
=specified time interval prior to entry into an OPERATIONAL MODE or other ap-
-plicable condition. The intent of this provision is to ensure that surveil-lance activities have been. satisfactorily demonstrated on a current basis as required to. meet the OPERABILITY requirements of the Limiting Condition for Operation.
.Under the terms of this specification,- for example,'during initial plant ~
STARTUP or following extended plant outages,.the applicable surveillan'ce acti-
, vities must be' performed within the. stated surveillance interval prior to placing or returning the system or equipment into OPERABLE' status. ,
4.0.5 This specification ensures that inservice inspection of ASMEl Code Class 1, 2, and.3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be' performed in accordance with-a periodically updated 1 version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10.CFR 50.55a. Relief from any of the above requirements has 4 - been provided in writing by the Commission and is not a part of these Technical
' Specifications.
This specification includes a~ clarification of the frequencies _for per-forming the inservice inspection and testing activities required by.Section'XI i % of the ASME Boiler and Pressure Vessel Code and applicable' Addenda. This-
- clarification is provided to ensure consistency in surveillance intervals i
throughout these Technical Specifications and to remove any ambiguities re-L lative-to the frequencies for performing the required inservice inspection and ! testing activities. Under the terms of this specification, the more restrictive requirements-I of the. Technical Specifications take precedence over the ASME Boiler and Pres-sure Vessel Code'and applicable Addenda. For example, the requi_rements of Specification 4.0.4 to perform surveillance activities prior to entry into an
~
- - OPERATIONAL MODE or other specified applicability condition takes precedence i over the ASME Boiler and Pressure Vessel Code provision which allows pumps to
, beltested up to 1 week after return to normal operation. And for example, f the Technical. Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function
!- is dec.lared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code. provision which allows a valve to be incapable of performing its i specified function for up to 24 hours before being declared inoperable. i l J $ ( l i ! SEABROOK - UNIT 1 B 3/4 0-3
, - , . . - ~ . - - . - , ,.,m -,-r_,,, , , , . . - ... _.,..%..g . -....c,, ,...___w.e, , , , , . ---, , . , __._.e+_m,,..gr vm.
I m 3/4.1. - REACTIVITY CONTROL SYSTEMS
/ T YM BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg. The most restrictive condition occurs at E0L, with T at no-load operating temperature, and is avg associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% Ak/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200 F, the reactivity transients resulting from a postulated steam-line break cooldown are minimal. A 1.2% Ak/k SHUTDOWN MARGIN and a boron [ concentration of greater than 2000 ppm are required to permit sufficient time (* ]/ for the operator to terminate an inadvertent boron dilution event with T less than 200 F. avg 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate conparison. The most negative MTC,'value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.2 x 10 4 Ak/k/.F. The MTC value
'of -3.3 x 10 4 Ak/k/ F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron-concentration and is obtained by making these corrections to the limiting MTC T value of -4.2 x 10 4 ak/k/ F.
d SEABROOK - UNIT 1 B 3/4-1-1
REACTIVITY CONTROL SYSTEMS BASES O B0 RATION CONTROL 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued) The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS baron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551 F. This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT NDT temperature. 3/4.1.2 B0 RATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators. With the RCS in MODES 1, 2, or 3, a minimum of two boron injection flow paths are required to ensure single functioaal capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 22,000 gallons of 7000 ppm borated water from the boric acid storage tanks or a minimum contained volume of 477,000 gallons of 2000 ppm borated water from the refueling water storage tank (RWST). The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the. Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can be relieved by operation of a single PORV or an RHR suction relief valve. As a result of this, only one boron injection system is available. This is acceptable on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable. SEABROOK - UNIT 1 B 3/4 1-2
s REACTIVITY CONTROL SYSTEMS I b
\ _.)
BASES 3/4.1.2 B0 RATION SYSTEMS (Continued) The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1.2% Ak/k after xenon decay and cooldown from 200 F to 140 F. This condition requires a minimu'm contained volume of 6500 gallons of 7000 ppm borated water from the boric acid storage tanks or a minimum contained volume of 24,500 gallons of 2000 ppm borated water from the RWST. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume and boron concentration of the RWST. also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6. The limitations on OPERABILITY of isolation provisions for the Boron ,O Thermal Regeneration System and the Reactor Water Makeup System in Modes 3, 4, (,'d) 5, and 6 ensure that the boron dilution flow rates cannot exceed the value assumed in the transient analysis. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Canks and 18, 210, and 228 steps with-drawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions pro-vide assurance of fuel rod integrity during continued operation; In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm f} that the results remain valid during future operation. V l SEABROOK - UNIT 1 B 3/4 1-3 l
s d '! REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued) The maximum rod drop time restriction'is consistent with the assumed rod drop time used in the safety analyses. Measusement with T avg greater than or equal to 551 F and with all reactor coolant pumps operating ensores that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions. Control rod positions a.nd OPERABILITY of the-rod position indicators are required to be verified on a nominal basis of once per 12 hours with more fre-quent verifications required if an automatic monitoring' channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied. For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod (s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rate (WestinghOase) to C. O. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the e.ontrol rod stepping mecaanism. In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this al)ows approximately 4 hours for this verification. O SEABROOK - UNIT 1 B 3/4 1-4
3/4.2 POWER DISTRIBUTION LIMITS
\ ) BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and'II (Incidents of Moderate Frequency) events by: (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation.and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the.ECCS acceptance criteria limit of 2200 F is' not exceeded.
The definitions of certain hot channel and peaking factors as used in
.these specifications are as follows:
Fq (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the, surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F g Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated. power to the average rod power; and. p) V- Fxy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z. 3/4.2.1 AXIAL FLUX DIFFERENCE
-The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fn(Z) upper' bound envelope of 2.32 times the normalized axial peaking factor is not-exceeded dur-ing either normal operation or in the event of xenon redistribution following power changes.
Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion. limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional _ THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations'. Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This a deviation will not affect the xenon redistribution sufficiently to change the k envelope of peaking factors which may be reached on a subsequent return to SEABROOK - UNIT 1 B 3/4 2-1
POWER DISTRIBUTION LIMITS BASES 3/4.2.1 A)'I AL FLUX DIFFERENCE (Continued) RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within established limits while at THERMAL POWER levels between\50% and 90% of RATED THERMAL POWER.
^
For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target , band are less significant. The penalty of 2 hours' actual time reflects this reduced significance.
' Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE "
excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels - between 50% and 90V and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm rwsage when the penaltyideviation accumulates beyond the linits of 1 hour and 2 hours, respectively. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density- i and minimum DNBR are not exceeded and (2) in the event of a LOCA, the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion differing by more than t 12 steps, indicated, from the group demand position;
- b. Control rod groups are sequenced'with overlapping groups as described in Specification 3.1.3.6;
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERE.NCE, is maintained within the limits.
O SEABROOK - UNIT 1 B 3/4 2-2
n POWER DISTRIBUTION LIMITS I v
)
BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Lontinued) F g will be maintained within its limits provided Conditions a. through
- d. above'are maintained. The relaxation of F H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
Fuel rod bowing reduces the value of DNBR. Credit is available to offset this reduction in the generic margin. The generic margins, totaling 9.1% DNBR completely offset any rod bow penalties. This margin includes the following:
- a. Design limit DNBR of 1.30 vs. 1.28,
- b. Grid spacing (K s
) f 0.046 vs. 0.059,
- c. Thermal diffusion coefficient of 0.038 vs. 0.059,
- d. DNBR multiplier of 0.86 vs. 0.88, and
- e. Pitch reduction.
7 (v) The applicable values of rod bow penalties are referenced in the FSAR. When an qF measurement is taken, an allowance for both experimental error
-and manufacturing tolerance must be made. An allowance of 5% is appropriate for a. full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
The Radial Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit. The RT F limit for RATED THERMAL POWER (F ,P)qas provided in the Radial Peaking Factor Limit Report per Specification 6.8.1.6 was determined from expected power control maneuvers over the full range of burnup' conditions in the core. When RCS F H is measured, no additional allowances are necessary prior to comparison with the established limit of a measurement error of.4% for F has H been allowed for in determination of the design DNBR value. 3/4.2.4 QUADRANT POWER TILT RATIO The purpose of this specification is to detect gross changes in core power , distribution between monthly incore flux maps. During normal operation the ! QUADRANT POWER ~ TILT RATIO is set equal to zero once acceptability of core l peaking factors has been established by review of incore maps. The limit of (n\ j) 1.02 is established as an indication that the power distribution has changed enough to warrant further investigation. SEABROOK - UNIT 1 B 3/4 2-3
POWER DISTRIBUTION LIMITS BASES O 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3 F for T avg and 2205 psig for pressurizer are not exceeded. The measurement error of 2.1% for RCS total flow rate is based upon per-forming a precision heat balance and using the result to normalize the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a noncon-servative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is applied. Any fouling which might bias the RCS flow rate measureme.nt greater than 0.1% can be detected by monitoring and trending vari-ous plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified limit. O SEABROOK - UNIT 1 8 3/4 2-4
,m 3/4.3 INSTRUMENTATION i
QJ BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to provide'the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.
~
The Surveillance Requirements speci-fied for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveil-lance tests performed at the minimum frequencies are sufficient to demonstrate this capability. g y ) Specified surveillance intervals and surveillance and maintenance outage times ~have been determined in accordance with WCAP-10271, " Evaluation of Sur-veillance Frequencies and Out of Service Times for the Reactor Protection In-strumentation System," and supplements to that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety
~ Features instrumentation. (Implementation of ouarterly testing of RTS is being postponed until after approval of a similar testing interval for ESFAS.) The NRC Safety Evaluation Report for WCAP-10271 was provided in a letter dated February 21, 1985, from C. O. Thomas (NRC) to J. J. Sheppard (WOG-CP&L).
The Engineered Safety Features Actuation System Instrumentation Trip Set-points specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Opera-tion with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the
~
O "as measured". deviation from the specified calibration point for rack and (j sensor components in conjunction with a statistical combination of the other SEABROOK - UNIT 1 B 3/4 3-1 1
INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + R S < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the. statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span; R or Rack Error is the "as measured" deviation, in the percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor ~and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. Na credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time. The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety SEABROOK - UNIT 1 B 3/4 3-2
INSTRUMENTATION
'tj BASES 3/4.3.1 and~3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)-
Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position,. (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) emergency feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, and (11) automatic service water valves position. The Engineered Safety Features Actuation System interlocks perform the following functions: P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T avg below Setpoint, prevents _the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped. Reactor'not tripped prevents manual block of Safety Injection. P-11 On increasing pressurizer pressure', P-11 automatically reinstates Q Safety Injection actuation on low pressurizer pressure. On decreasing pressure, P-11 allows the manual block of Safety Injection actuation on low pressurizer pressure, and the manual block of SI and steamline isolation on steamline low pressure. On the manual block of steamline low pressure, manual block of steamline low pressure automatically initiates steamline isolation on steam generator pressure negative rate - high. P-14 On increasing steam generator water leval, P-14 automatically trips the turbine and all feedwater isolation valves; inhibits feedwater control valve modulation; and blocks the start of the startup feed- , water pump. 3/4.3.3 MONITORING INSTRUMENTATION 2 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant - operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance. The radiation monitors for plant operations sense radiation levels in selected plant systems and locations and determine whether or not (Vn) predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents SEABROOK - UNIT 1 B 3/4 3-3
. _ _ _ - _ ~_ __. _ __ _. . __ ._ _
INSTRUMENTATION BASES O MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS (Continued) and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems. 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. For the purpose of measuring9F (Z) or F H, a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earth-quakes," April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. 3/4.3.3.5 REMOTE SHUTDOWN SYSTEM The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of Appendix A to 10 CFR Part 50. SEABROOK - UNIT 1 B 3/4 3-4
4 INSTRUMENTATION BASES MONITORING INSTRUMENTATION 3/4.3.3.5 REMOTE SHUTDOWN SYSTEM (Continued) The OPERABILITY of the Remote Shutdown System ensures that a-fire will not preclude achieving safe _ shutdown. The Remote Shutdown System instrumentation, control, and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is consistent with General Design Criterion.3 and Appendix R to 10 CFR Part 50. 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consis-tent with the recommendations of Regulatory Guide 1.97, Revision 3, "Instrumen-tation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. p\ U
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3/4.3.3.7 (This specification number is not used.) 3/4.3.3.8 (This specification number is not used.) 3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING-INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as. applicable, the releases of radioactive materials in liquid efflu-ents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3/4.3.3.10 RADI0 ACTIVE ~ GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumenta-tion also includes provisions for monitoring (and controlling) the concentra-O tions of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements SEABROOK - UNIT 1 8 3/4 3-5
INSTRUMENTATION BASES MONITORING INSTRUMENTATION 3/4.3.3.10 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Continued) of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10 6 pCi/cc are measurable. 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment, or structures. O O aEABR00K - UNIT 1 8 3/4 3-6
3/4.4 REACTOR COOLANT SYSTEM iV
\J BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in opera-tion and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 6 hours.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor c_oolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all. times. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides p; D sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting an RCP in MODES 4 and 5 are provided to pre-vent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the~ secondary water temperature of each steam generator is less than 50 F above each of the RCS cold-leg temperatures. 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The
- relief capacity of a single safety valve is adequate to relieve any overpres-sure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides over-pressure relief capability and will prevent RCS overpressurization. In addi-O 'V tion, the Overpressure Protection System provides a diverse means of. protection against RCS overpressurization at low temperatures.
SEABROOK - UNIT 1 B 3/4 4-1
REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES (Continued) During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete. loss of load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reache'd (i.e., no credit is taken for a direct Reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to re-lieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capabil-ity should a relief valve become inoperable. The PORVs and their associated block valves are powered from Class 1E power supply busses. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation, so that corrective measures can be taken. SEABROOK - UNIT 1 B 3/4 4-2
n REACTOR COOLANT SYSTEM / \ v !
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BASES 3/4.4.5 STEAM GENERATORS (Continued) The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result.in stress corrosion cracking. The extent of cracking during plant' operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leak-age in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. 77 Plugging will be required for all tubes with imperfections exceeding the plug-ging limit of 40% of the tube nominal wall thickness. Steam generator tube ('v) inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20%-of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commis-sion in a Special Report pursuant to Specification 6.8.2 within 30 days and prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revi-sion of the Technical Specifications, if necessary. 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are pro-vided to monitor and detect leakage from the reactor coolant pressure boundary. These. Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, O) (d the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. SEABROOK - UNIT 1 B 3/4 4-3
e REACTOR COOLANT SYSTEM BASES REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. The total steam generator tube leakage limit of 1 gpm for all steam generators r.ot isolated from the RCS ensures that the dosage contribution from the i Ae leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analys's of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. The CONTROLLED LEAKAGE limitation restricts operation when the total flow ' supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the. event of a LOCA, the safety injection flow will not be less than assumed in the safety analyses. The specified allowed leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. O SEABROOK - UNIT 1 B 3/4 4-4
- REACTOR COOLANT SYSTEM i
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s__ BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reac-tor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are. time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in ex-cess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting con-tinued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits. The Surveillance Requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take cor-rective action. m 3/4.4.8 SPECIFIC ACTIVITY l i \j - The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE B0UNDARY will not exceed an appro-priately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Seabrook site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates a possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 800 hours per year (approximately 10% of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture. The reporting of cumulative operating time over 500 hours in any 6-month consecutive period with greater than 1 microcurie / gram DOSE EQUIVALENT I-131 will allow sufficient. time for Commission evaluation of the circumstances prior to reaching the 800-hour limit. v SEABROOK - UNIT 1 B 3/4 4-5
REACTOR COOLANT SYSTEM BASES O 3/4.4.8 SPECIFIC ACTIVITY (Continued) The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radioiodine level is to be determined every 4 hours. If the gross specific activity level and radioiodine level in the reactor coolant we're at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radio-iodine contribution would probably be about 20%. The exclusion of radio-nuclides with half-lives less than 10 minutes from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes' decay time. The choice of 10 minutes for the half-life cutoff was'made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radio-nuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE B0UNDARY under any accident condition. Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours between sample taking and completing the initial analysis is based upon a typical time necessary to per-form the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour, about 2 hours, about 1 day, about 1 week, and about 1 month. Reducing T avg to less than 500'F prevents the release of activity should a steam generator tube rupture, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. O SEABROOK - UNIT 1 B 3/4 4-6
REACTOR COOLANT SYSTEM ig) BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,-Section III, Appendix G:
- 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
- a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines ~for cooldown rates between those pre-sented may be obtained by interpolation; and
- b. Figures 3.4-2 and 3.4-3 define limits to' assure prevention of' non-ductile failure only. For normal' operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates ~that can be achieved over certain pressure-temperature ranges.
- 2. These limit lines shall be calculated periodically using methods pro-
.(nV) vided below,
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70 F,
- 4. The pressurizer heatup and cooldown rates shall not exceed 100 F/h and 200 F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 F, and
- 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the ' calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975. Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 16 effec-h
\J tive full power years (EFPY) of service life. The 16 EFPY service life period SEABROOK - UNIT 1 B 3/4 4-7
REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) is chosen such that the limiting RT NDT at the 1/4T location in the core region is greater than the RT NDT f the limiting unirradiated material. The selection of such a limiting RT NDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor opera-tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART NDT computed by either Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjust-ments for this shif t in RT NDT at the end of 16 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments. Values of ART NDT determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the cap-sule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the t.RT NDT determined from the surveillance capsule exceeds the calculated ART NDT f r the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of tim ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mecha,ics (LEFM) technology. In the calculation procedures, a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T O SEABROOK - UNIT 1 B 3/4 4-8
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CU% 0.20 BASE 0.15 WELD---- 3- --- ---1 -
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I I l 10 10** 2 3 4 5 5 7 8 91d' 2 3 4 5 6 7 8 91d* 2 FAST NEUTRON FLUENCE (N/CM , E > 1 MeV) FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RTNDT FOR REACTOR VESSELS EXPOSED TO 550 F SEABROOK - UNIT 1 B 3/4 4-10
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Q \ j . TABLE B 3/4.4-1 REACTOR VESSEL TOUCHNESS 8 o Avg. Shelf Energy T RT Material Cu P NDT NDT NMWO* MWD *^ E Component Code No. Spec. No. (%) (%) ( F) ( F) (it-lb) (ft-Lb)
~ Closure Head Dome R1809-1 A533B,CL.1 0.15 0.011 -40 10 80.5 -
Closure Head Torus R1810-1 A533B,CL.1 0.08 0.012 -50 0 104 - Closure Head Flange R1802-1 A508,CL.2 - 0.013 10 10 105.5 - Vessel Flange R1801-1 A508,CL.2 - 0.012 20 30 91 - Inlet Nozzle R1804-1 A508,CL.2 0.10 0.011 0 0 125 - Inlet Nozzle R1804-2 A508,CL.2 0.09 0.010 -20 -20 125 - Inlet Nozzle R1804-3 A508,CL.2 0.08 0.010 -20 -20 131 - I Inlet Notzle R1804-4 A508,CL.2 0.10 0.013 -20 -20 128 - Outlet Nozzle R1805-1 A508,CL.2 - 0.003 -20 -10 115 - Outlet Nozzle R1805-2 A508,CL.2 - 0.004 -20 -20 132 - t' Outlet Nozzle R1805-3 A508,CL.2 - 0.009 -10 _10 128 - Outlet Nozzle R1805-4 A508,CL.2 - 0.005 -10 -10 117 - i Nozzle Shell R1807-1 A533B,CL.1 0.08 0.011 -30 30 66 - p Nozzle Shell R1807-2 A533B,CL.1 0.09 0.012 -40 30 66.5 - Nozzle Shell R1807-3 A5338,CL.1 0.06 0.010 -20 10 107 - Inter. Shell R1806-1 A533B,CL.1 0.04 0.012 -30 40 82 139.5 Inter. Shell R1806-2 A5338,CL.1 0.05 0.007 -30 0 102 143.5 Inter. Shell R1806-3 A533B,CL.1 0.07 0.007 -40 10 115 138 Lower Shell R1808-1 A533B,CL.1 0.05 0.005 -30 40 78 120.5 Lower Shell R1808-2 A533B,CL.1 0.05 0.007 -20 10 77 127 Lower Shell R1808-3 A533B,CL.1 0.06 0.007 -20 40 78 130.5 Bottom Head Torus R1811-1 A533B,CL.1 0.15 0.010 -50 0 94.5 - Bottom Head Dome R1812-1 A533B,CL.1 0.09 0.009 -30 0 97.5 - Inter. & Lower Shell Long Weld Seams G1.72 Sub Arc Weld 0.07 0.008 -50 -50 200 - Inter. & Lower Shell Girth Weld Seam G1.72 Sub Arc Weld 0.07 0.008 -50 -50 200 -
*NMWD - Normal to Major Working Direction
** MWD - Major Working Direction i
REACTOR COOLANT SYSTEM BASES O 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time. K IR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: K = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) IR Where: K is the reference stress intensity factor as a function of the metal IR temperature T and the metal nil-ductility reference temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: (2) CKIM + kit <KIR Where: K IM = the stress intensity factor caused by membrane (pressure) stress, K = the stretc intensity factor caused by the thermal gradients, It KIR = c nstant provided by the Code as a function of temperature relative to the RT NDT f the material, C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations. At any time duriN the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value SEABROOK - UNIT 1 B 3/4 4-12
,m REACTOR COOLANT SYSTEM i
d BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, kit, f r the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, frcm these, the allowable pressures are calculated. C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the (n (_/ i 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situa-tion. It follows that at any given reactor coolant temperature, the AT devel-oped during cooldown results in a higher value of K IR at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if condi-tions exist such that the increase in K IR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period. HEATUP l Three separate calculations are required to determine the limit curves i for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions I as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup t produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K IR f r the 1/4T crack during heatup is lower than the K for the 1/4T crack during steady-state IR > v l SEABROOK - UNIT 1 B 3/4 4-13
REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) HEATUP (Continued) conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different KIRs for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. O SEABROOK - UNIT 1 B 3/4 4-14
f7 REACTOR COOLANT SYSTEM
! i l Qj BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued)
COLD OVERPRESSURE PROTECTION
-The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) a maximum pres-sure overshoot beyond the PORV Setpoint which can occur as a result of time de-lays in signal processing and valve opening; (2) a 50 F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; (3) instrument uncertain-ties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lock-out of both Safety Injection pumps and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow st' art of an RCP if secondary coolant temperature is more than 50 F above reac-tor coolant temperature. Exceptions to these requirements are acceptable as described below. ,m l j Operation above 350 F but less than 375 F with only centrifugal charging Lj pump OPERABLE and no Safety Injection pumps OPERABLE is allowed for up to 4 hours. As shown by analysis, LOCAs occurring at low temperature, low pres-sure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time duration and the condition of having only one centrifugal charging pump OPERABLE and the probability of a LOCA occurring dur-ing this time, the failure of the single centrifugal charging pump is not assumed.
Operation below 350 F but greater than 325 F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours. During low pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure - High are blocked. In normal conditions, a single failure of the ESF actuation circuitry will result in the starting of at most one train of Safety Injection (one centrifugal charging pump, and one Safety Injection pump). For temperatures above 325 F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORVs without exceeding Appendix G limit. Given the short time duration that this condition is allowed and the low probability of a single failure of a PORV is not assumed. Initiation of both trains of Safety Injection during this 4-hour time frame due to operator error or a single fail-ure occurring during testing of a redundant channel are not considered to be credible accidents. p)
;V OperationwithallcentrifugalchargingpumpsandbothSafetyInjection pumps OPERABLE is acceptable when RCS temperature is greater than 350 F, a single PORV has sufficient capacity to relieve the combined flow rate of all SEABROOK - UNIT 1 B 3/4 4-15 I
REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) COLO OVERPRESSURE PROTECTION (Continued) pumps. Above 350 F two RCPs and all pressure safety valves are required to be OPERABLE. Operation of an RCP eliminates the possibility of a 50 F difference existing between indicated and actual RCS temperature as a result of heat trans-port effects. Considering instrument uncertainties only, an indicated RCS tem-perature of 350 F is sufficiently high to allow full RCS pressurization in ac-cordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pressurizer safety valves provide acceptable and redun-dant overpressure protection. The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System will be revised on the basis of the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H. 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure. Vessel Code, 1983 Edition and Addenda through Summer 1983. 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible' gases and/or steam from the Reactor Coolant System that could inhibit natural circu-lation core cooling. The OPERABILITY of least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures that the capability exists to perform this function. The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation wnile ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plant Requirements," November 1980. SEABROOK - UNIT 1 8 3/4 4-16
- p. 3/4.5 EMERGENCY CORE COOLING SYSTEMS O
BASES
'3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.
The limits en accumulator volume, boron concentration, and pressure ensure' that the assumptions used for accumulator injection in the safety analysis are met. The accumulator power-operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive-conditions are not met. In addition, as these accumulator isolation valves fail to meet single-failure criteria, removal of power to the valves is required.
The limits for operation with an accumulator inoperable for any reason
.except an isolation valve closed minimizes the time exposure of the plant to a O LOCA event occurring concurrent with failure of an additional accumulator which th may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of'two independent ECCS subsystems ensures that sufficient emergency' core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any' single-failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying suf.ficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double-ended break of the largest RCS cold-leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limitegi core cooling requirements. The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumps except the required OPERABLE charging pump to be in-operable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on pro-vides assurance that a mass addition pressure transient can be relieved by the (Vn) operation of a single PORV or RHR suction relief valve. SEABROOK - UNIT 1 B 3/4 5-1
EMERGENCY CORE COOLING SYSTEMS - BASES 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS (Continued) The Surveillance Requirements provided to ensure OPERABILITY of each com-ponent ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOC \ Maintenance of proper flow resistance and pressure drop in the piping system to each injec-tion point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) pro-vide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the. ECCS-LOCA analyses. 3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems.and components. O SEABROOK - UNIT 1 8 3/4 5-2
("N 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1' PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE B0UNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations.on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa . As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L a during performance of the periodic tests to account for pos-sible degradation of the containment leakage barriers between leakage tests. (O The surveillance testing for measuring leakage rates is consistent with V the requirements of Appendix J of 10 CFR Part 50.
~
3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that-the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment. internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure' differential with respect to the annulus atmosphere of 3.5 psi and (2) the containment peak pressure does not exceed the design pressure of 52 psig during LOCA conditions.
.The maximum peak pressure expected to be obtained from a LOCA event is
-49.6 psig. The limit of 16.2 psia for initial positive containment pressure will limit the total pressure to 49.6 psia which is less than the design pres-sure and is consistent with the safety analyses.
b
'EAB: 1 - UNIT 1 S B 3/4 6-1
CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT (Continued) 3/4.6.1.5 AIR TEMPERATURE The limitation in containment average air temperature ensures that the containment average air temperature does not exceed the initial temperature condition assumed in the overall safety analysis for a steam line break acci-dent. _ Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 52 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability. 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 36-inch containment purge supply and exhaust isolation valves are required to be sealed closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining the'se valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Standard Review Plan Section 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator. The use of the containment purge lines is restricted to the 8-inch purge supply and exhaust isolation valves since, unlike the 36-inch valves, the 8-inch valves are capable of closing during a LOCA or steam line break accident. There-fore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not bc - exceeded in the event of an accident during contain' ment PURGING operation. The total time the containment purge (vent) system isolation valves may be open dur-ing MODES 1, 2, 3, and 4 in a calendar year is determined by the actual need for opening the valves for safety-related reasons; e.g., containment pressure control or the reduction of airborne radioactivity to facilitate personnel access for surveillance and maintenance activities. Leakage integrity tests with a maximum allowable leakage rate for contain-ment purge supply and exhaust supply valves will provide eariy indication of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 La le kage limit of Specification 3.6.1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. SEABROOK - UNIT 1 B 3/4 6-2
J CONTAINMENT SYSTEMS ~
! )
BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses. The two independent Containment Spray Systems provide post accident cool-ing of the containment atmosphere. The Containment Spray Systems also provide a mechanism for removing iodine from.the containment atmosphere, and, therefore, the time requirements for restoring an inoperable. Spray System to OPERABLE status have been maintained consistent with those assigned other inoperable ESF equipment. 3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient Na0H is added to the' containment spray in the event of a LOCA. The limits on Na0H volume and concentration ensure a pH value of between 8.5 and 11.0 for the 7 solution recirculated within containment after a LOCA. This pH band' minimizes the evolution of iodine and minimizes the effect of chloride and_ caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not' usable because of tank discharge line location or other physical characteristics. These assumptions are con-sistent with the iodine removal efficiency assumed in the safety analyses. 3/4.6.3 CONTAINMENT ISOLATICN VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of con-p trolling the expected hydrogen generation associated with: (1) zirconium-water SEABROOK - UNIT 1 B 3/4 6-3
- - _ _ ~ _ . _ .
CONTAINMENT SYSTEMS BASES 3/4.6.4 COMBUSTIBLE GAS CONTROL (Continued) reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentra-tions in Containment Following a LOCA," March 1971. The Hydrogen Mixing Systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit. 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING 3/4.6.5.1 CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM The OPERABILITY of the Containment Enclosure Emergencv Air Cleanup System ensures that during LOCA conditions containment vessel leakage into the annulus, and radioactive materials leaking from engineered safety features equipment, from the electrical penetration areas, and from the mechanical penetration tunnel, will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. 3/4.6.5.2 CONTAINMENT ENCLOSURE BUILDING INTEGRITY CONTAINMENT ENCLOSURE BUILDING INTEGRITY ensures that the release of radio-active materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with operation of the Containment Enclosure Emergency Air Cleanup System, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions. 3/4.6.5.3 CONTAINMENT ENCLOSURE BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment enclosure building will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to provide: (1) protection for the steel vessel from external missi.les, (2) radiation shielding in the event of a LOCA, and (3) an annulus surrounding the primary containment that can be maintained at a negative pressure during accident con-ditions. A visual inspection is sufficient to demonstrate this capability. SEABROOK - UNIT -1 B 3/4 6-4
3/4.7 PLANT SYSTEMS
,m
{Vb BASES 3/4.7.1 ' TURBINE CYCLE 3/4.7.1.1 SAFETYVALy,FJ The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1320 psia) of
'its design pressure of 1200 psia during the most severe anticipated systen operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed 1_oss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve ~ lift settings and relieving capacities are in accor-dance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 1.839 x 107 lbs/hr which is 121% of the total secondary steam flow of 1.514 x 107 lbs/hr at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2. STARTUP and/or POWER OPERATION is allowable with safety valves ~ inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the - ") s reduced Reactor trip settings of the Power Range Neutron Flux channels. The
. Reactor Trip Setpoint reductions are derived on the following bases:
For four loop'op'eration: SP = (X).X- (Y)(V) x 109 Where: SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, X = Total relieving capacity of all safety valves per steam line in lbs/hr, and Y = Maximum relieving capacity of any one safety valve in 1bs/hr CN V SEABROOK - UNIT 1 B 3/4 7-1
PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE (Continued) 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss-of-offsite power. The electric motor-driven emergency feedwater pump is capable of deliver-ing a total feedwater flow of 650 gpm at a pressure of 1221 psig to the en-trance of the steam generators. The steam-driven emergency feedwater pump is capable of delivering a total feedwater flow of 650 gpm at a pressure of 1221 psig to the entrance of the steam generators. The startup feedwater pump serves as the third auxiliary feedwater pump and can be manually aligned to be powered from an emergency bus (Bus 5). The startup feedwater pump is capable of taking suction on the dedicated emergency feedwater volume of water in the condensate storage tank and delivering a total feedwater flow of in excess of 650 gpm at a pressure of 1221 psig to the entrance of the steam generator via either the main feedwater header or with manual alignment to the emergency feedwater flow path. This capacity is sufficient to ensure that adequate feed-water flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the Residual Heat Removal System may be placed into operation. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water vol-ume ensures that sufficient water is available to cool the RCS to a temperature of 350*F. The OPERABILITY of the concrete enclosure ensures this availability of water following rupture of the condensate storage tank by a tornado generated missile. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in.the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one' steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses. SEABROOK - UNIT 1 B 3/4 7-2
PLANT SYSTEMS V BASES l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION l The liinitation on steam generator pressure and temperature ensures that i the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator RT of 60 F and are sufficient to NDT prevent brittle fracture 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Primary Component Cooling Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses. 3/4.7.4 SERVICE WATER SYSTEM The Service Water System consists of two independent loops, each of which can operate with either a service water pump train or a cooling tower pump train. The OPERABILITY of the Service Water System ensures that sufficient CNi cooling capacity is available for continued operation of safety-related equip - \, ment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses, which also~ assumes loss of either'the cooling tower or ocean cooling. 3/4.7.5 ULTIMATE HEAT SINK
- The limitations on service water pumphouse level, and the OPERABILITY requirements for the mechanical draft cooling tower and the portable tower makeup pump system, ensure that sufficient cooling capacity is available to either: (1) provide. normal cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits. This cooling capabil-ity is provided by the Atlantic Ocean except during loss of ocean tunnel water flow, when the cooling capability is provided by the mechanical draft cooling tower with tower makeup using portable pumps.
The limitations on minimum water level and the requirements for mechanical draft cooling tower OPERABILITY are based on providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis tempera-ture and is consistent with the recommendations of Regulatory Guide 1.27,
" Ultimate Heat Sink for Nuclear Plants," March 1974.
3/4.7.6 CONTROL-ROOM AREA VENTILATION SYSTEM The OPERABILITY of the Control Room Area Ventilation System ensures that: (1) the allowable temperature for continuous-duty rating.for the equipment and v SEABROOK - UNIT 1 B 3/4 7-3
PLANT SYSTEMS BASES 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM (Continued) instrumentation cooled by this system is not exceeded; and (2) the control room will remain habitable for operations personnel during and following credible accident conditions. The OPERABILITY of this system in conjunction with con-trol room design provisions is based on limiting the radiation exposure to per-sont.el occupying the control room to 5 rems or less whole body, or its equiva-lent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50. 3/4.7.7 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads. Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer. A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Station Operation Review Committee (50RC). The determination shall be based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50. Surveillance to demonstrate OPERABILITY is by performance of the require-ments of an approved inservice inspection program. Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions. The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal SEABROOK - UNIT 1 B 3/4 7-4
,, PLANT SYSTEMS N j BASES 3/4.7.7 SNUBBERS (Continued) replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These. records will provide statistical bases for future consideration of snubber service life.
3/4.7.8 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based-on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with. Surveillance Req 0irements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored - ["'s and need not be tested unless they are removed from the shielded mechanism. ('# 3/4.7.9 (This specification number is not used.) 3/4.7.10 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive. temperatures may degrade equipment and can cause a loss of its OPERABILITY. The' temperature limits include an allowance for instrument error of i 4.5 F. py L.] SEABROOK - UNIT 1 B 3/4 7-5
'3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1,-3/4.8.2, and 3/4.8.3' A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION The OPERABILITY of the A.C. and D.C power sources and' associated d'istribu-tion systems during operation ensures that sufficient power will.be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C.~ and D.'C.
power sources and di.stribution systems satisfy'the requirements of General Design Criterion 17 of-Appendix A to 10 CFR Part 50. The ACTION requirements specified.for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources is consis-tent with the initial condition assumptions ~of the safety analyses and is based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distr.ibution systems OPERABLE during accident conditions
-coincident with an assumed loss-of-offsite power and single failure of.the other onsite A.C. source. The.A.C. and D.C. source-allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE,-and that the steam-driven emergency-feedwater pump is OPERABLE. This requirement.is intended to provide assurance that a loss-of-offsite power event will not result in a com-plete' loss.of safety function of critical. systems during the period one of the diesel generators is inoperable. The term, " verify," as used in this context means to administratively check by examining logs or other information to deter-mine;if~certain components are out of service for maintenance or other reasons. 'It does not mean-to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.
The OPERABILITY of the niinimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that:
-(1) the facility can be maintained in the shutdown or refueling condition for ,
extended time periods and (2) sufficient' instrumentation and control capability is available for monitoring and maintaining the unit status. The Surveillance Requirements for demonstrating the OPERABILITY of the , diesel generators'are in accordance with the recommendations of Regulatory Guides 1.9, " Selection of Diesel ~ Generator Set Capacity for Standby Power Supplies," March 10, 1971; 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric ~ Power Systems at Nuclear Power Plants," Revision 1, August 1977; and 1.137, " Fuel-0il Systems for Standby Diesel Generators," Revision 1, October 1979. i O SEABROOK - UNIT 1 B 3/4 8-1 i
ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued) The Surveillance Requirement for demonstrating the OPERABILITY of the sta-tion batteries are based on the recommendations of Regulatory Guide 1.129,
" Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std. 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity. Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's-full charge specific gravity, ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function. O SEABROOK - UNIT 1 B 3/4 8-2
ELECTRICAL POWER SYSTEMS k BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES -
. Containment electrical' penetrations and penetration conductors are pro-tected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protec-tion circuit breakers during periodic surveillance.
The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker. Each manufacturer's air circuit breakers, molded case circuit breakers, and overload devices are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufactur-er.'s brand of circuit breakers, it is necessary to divide that manufacturer's breakers into groups and. treat each group as a separate type of breaker for surveillance purposes. The OPERABILITY of the ' motor-operated valves thermal overload protection ensures that the thermal overload protection will not prevent safety-related valves from pe~rforming their function. The Surveillance Requirements for-demonstrating the OPERABILITY of the thermal overload protection are in accor-
'N dance with Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1, March 1977.
o V SEABROOK - UNIT 1 B 3/4 8-3
, , - - , . - - . - - , - . . - , . - - , _ . - - - . - - - - - - . ,,,7,-..,,
3/4.9 REFUELING OPERATIONS
. /, \
V BASES 3/4.9.1 BORON CONCENTRATION
.The limitations on reactivity conditions during REFUELING ensure that:
(1) the reactor will remain subcritical during CORE ALTERATIONS and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The value of 0.95 or less for k eff includes a 1% Ak/k conservative allowance for uncertainties. Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm baron. 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of-the core. 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of
/O irradiated fuel assemblies in the reactor vessel ensures that sufficient (j time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration. closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a
- fuel element rupture based upon the lack of containment pressurization.
potential while in the REFUELING MODE. 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. 3/4.9.6 REFUELING MACHINE The OPERABit.1TY requirements for the refueling machine ensure that: (1) refueling muchine will be used for movement of drive rods and fuel assem-blies, (2) each hoist has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during Q lifting operations. SEABROOK - UNIT 1 B 3/4 9-1
REFUELING OPERATIONS. BASES O 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyse.s. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140 F as required during the REFUELING MODE. and (2) sufficient coolant circulation is maintained through the core to mini n ' S effect of a boron dilution incident and prevent baron stratification. The requirement to have twc RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency proc:2dures to cool the core. 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. 3/4.9.10 an-d 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The. restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is cc sis-tent with the assumptions of the safety analysis. 3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM The. limitations on the Fuel Storage Building Emergency Air Cleaning System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to dis-charge to the atmosphere. Operation of the system with the heaters operating SEABROOK - UNIT 1 B 3/4 9-2
REFUELING OPERATIONS-BASES 3/4.9.12' FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM (Continued). for at. least 10 continuous hours in a 31-day period'.-is sufficient' to reduce -the buildup.of moisture on the adsorbers and HEPA filters. zThe OPERABILITY of this system and the resulting iodine removal capacity are consistent with.the assumptions of the safety analyses. ANSI N510-1980 will be used as a proce-dural guide for surveillance testing.
.SEABROOK - UNIT 1 B 3/4 9-3
l l l
- f. 3/4.10 SPECIAL TEST EXCEPTIONS V) i BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed
, for control rod worth measurement. This.special test exception is required to l permit the periodic verification of the actual versus predicted core reactivity I condition occurring as a result of fuel burnup or fuel cycling operations.
-3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be posi-tioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to: (1) measure control rod worth and (2) determine the reactor stability index and damping factor under xenon oscillation conditions. I 3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the RCS T slightly lower avg fm than normally allowed so that the fundamental nuclear. characteristics of the core and related instrumentation can be verified. In order for various charac-(U )
teristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at B0L, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 and the RCS T,yg may be below the minimum temperature of Specification 3.1.1.4 during the measurement. l 3/4.10.4 REACTOR COOLANT LOOPS l This special test exception permits reactor criticality under no flow conditions and is required to perform certain STARTUP and PHYSICS TESTS while i at low THERMAL POWER levels. 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to. determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position. Indication Systems remain OPERABLE. O SEABROOK - UNIT 1 B 3/4 10-1
3/4.11 RADIOACTIVE EFFLUENTS
<g > %.)
BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents at the point of discharge from the multipart diffuser will be less than the concentration levels speci-fied in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation pro-vides additional assurance that the levels of radioactive materials in bedies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Sec-tion II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE. PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in' air (sub-mersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sec-77 tions II.A, III.A, and IV.A of Appendix I to 10 CFR Part 50. The Limiting Con-( v
) dition for Operation implements the guides set forth in Section II.A of Appen-
-dix I. The ACTION statements provide the required operating flexibility and at
-the same time implement the guides. set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRE-STRICTED AREAS wil1 be kept as' low as is reasonably achievable. The dose cal-culation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with th'e guides of Appendix I be shown by calculational procedures b.ased on models and data, such that the actual exposure of a MEMBER-0F THE PUBLIC through appropriate pathways is un-likely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual-release rates of radioactive mate-rials in liquid ef.fluents are consistent with the methodology provided in Regu-latory Guide 1.109, " Calculation of Annual Dases to Man from Routine Releases of Reactor Effluents for the Pur ose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113,
" Estimating Aquatic Dispersien of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.
This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioa'ctivity concentrations, or,'if not practicable, the treated effluent O- releases may be allocated equally to each of the radioactive waste producing
' units sharing the Radwaste Treatment System. For determining conformance to
-SEABROOK - UNIT 1- B 3/4 11-1
RADI0 ACTIVE EFFLUENTS BASES LIQUID EFFLUENTS 3/4.11.1.2 DOSE (Continued) LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require tr'eatment prior to release to the environment. The requirement that the appropriate por-tions of this system be used when specified provides assurance that the releases of radi7 active materials in liquid effluents will be kept as low as is reason-ably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design ob-jective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treat-ment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR Part 50 for liquid effluents. 3/4.11.1.4 LIQUID HOLDUP TANKS The temporary tanks include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System. Restricting the quantity of radioactive material contained in the speci-fied tanks provides assurance that in the' event of an uncontrolled release of the tank's contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column I. These limits provide reasonable as-surance that radioactive material discharged in gaseous effluents will not re-sult in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations ex-ceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be SEABROOK - UNIT 1 B 3/4 11-2
RADI0 ACTIVE EFFLUENTS ! ] BASES GASE0US EFFLUENTS 3/4.11.2.1 DOSE RATE (Continued) sufficiently low to compensate- for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE B0UNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. 3/4.11.2.2 DOSE - N0BLE GASES This specification is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I to 10 CFR Part 50. The Limiting Condition for Operati.on implements the guides set forth in Section I.B of Appendix I. The ACTION statements provide the required operating flexibility and at the fN same time implement the guides set forth in Section IV.A of Appendix I at the ( V) SITE B0UNDARY that the' releases of radioactive material.in gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The Surveillcnce Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calcula-tional procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The dose calculation methodology and parameters estab-lished in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the method-ology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Com-pliance with 10 CFR Part 50, Appendix I," Revision I, October 1977 and Regu-latory Guide 1.111, " Methods for Estimating Atmospheric Transport and Disper-sion of Gaseous Effluents in Routine Releases from Light-Water Ccaled Reactors," Revision.1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE B0UNDARY are based upon the historical average atmospheric conditions. 3/4.11.2.3 OOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I to 10 CFR Part 50. The Limiting Conditicns I for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same p time. implement the guides set forth in Section IV.A of Appendix I to assure b SEABROOK - UNIT 1 B 3/4 11-3
RADI0 ACTIVE EFFLUENTS BASES GASEOUS EFFLUENTS 3/4.11.2.3 DOSE - IODINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM (Continued) that the releases of radioactive materials in gaseous effluents at the SITE B0UNDARY will be kept as low as is reasonably achievable. The ODCM calcula-tional methods specified in the Surveillance Requirements implement the require-ments in Section III.A of Appendix I that conformance with the guides of Appen-dix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodol - ogy and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regula-tory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR.Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equa-tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifica_tions for Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were exam-ined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vege-tation with subsequent consumption by man, (3) deposition of radionuclides onto grassy areas where milk animals and meat producing animals graze followed by human consumption of that milk and meat, and (4) deposition of radionuclides on the ground followed by subsequent human exposure. 3/4.11.2.4 GASE0US RADWASTE TREATMENT SYSTEM The OPERABILITY of the GASE0US RADWASTE TREATMENT SYSTEM and the VENTILA-TION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in. Sections II.B and II.C of Appendix I to 10 CFR Part 50, for gaseous effluents. O SEABROOK - UNIT 1 B 3/4 11-4
l l 6 RADIOACTIVE EFFLUENTS U BASES GASE0US EFFLUENTS 3/4.11.2.5 EXPLOSIVE GAS MIXTURE FOR-THE GASE0US RADWASTE SYSTEM T.his specification is provided to ensure that the concentration of poten-- tially. explosive gas mixtures contained in the GASE0US RADWASTE SYSTEM is main-tained below the flammability limits of hydrogen and oxygen. Maintaining the
-concentration of hydrogen and oxygen below their flammability limits'provides assurance that the releases of radioactive materials will be controlled in conformance.with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
3/4.11.3 SOLID RADI0 ACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and General. Design Criterion 60 of Appendix A to 10 CFR Part 50. -The process parameters included in establishing the_ PROCESS CONTROL' PROGRAM.may include, but are not limited to, waste type, wa'ste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content,' waste principal chemical constituents, and. mixing-and curing times. [ '\ 3/4.11.4 TOTAL-DOSE This specification is provided to meet the dose limitations.of 40 CFR Part 190.that have been incorporated into 10 CFR Part 20 by 46 FR_18525. The specification requires the preparation and submittal of a Special Report when-ever the calculated doses due to releases of radioactivity and to radiation
~
from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the.resul-tant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the. individual reactors remain within twice the dose design objec-tives of Appendix I,-and if direct radiation doses from the units (including outside storage tanks ~, etc.) are kept small. The Special Report will describe a course of action that should result in the limitation of.the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other~ uranium fuel cycle. sources is negligible, with the excep-tion that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any. MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release condi-tions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c,-is con-sidered to be a-timely request.and fulfills the requirements of 40 CFR Part 190
.until NRC staff action is completed. The variance only relates.to the limits of.40 CFR Part 190, and does .not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and
\ 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she-is engaged in carrying out any operation that is part of the nuclear fuel cycle.
,SEABROOK - UNIT 1 B 3/4 11-5
RADI0 ACTIVE EFFLUENTS BASES O 3/4.11.4 TOTAL DOSE (Continued) THE PUBLIC from other uranium fuel cycle sources is negligible, with the excep-tion that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release condi-tions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is con-sidered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. O O SEABROOK - UNIT 1 8 3/4 11-6
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING
/- ,\
t ) V BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this specifi-cation provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring,
~
Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.
~
The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LL'us). The LLDs required by the ODCM are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as p an a priori (before the fact) limit representing the capability of a measure-g
\
ment system and not as an a posteriori (after the fact) limit for a particular measurement. Detailed discussion of the LLD and other detection limits can be found in Currie, L.A., " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually). 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE B0UNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census. Information from methods such as the door-to-door survey, from aerial survey, or from consulting with local agricul-tural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I~to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored, since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of , leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. l To determine this minimum garden size, the following assumptions were made: ) (1) 20% of the garden was used for growing broad-leaf vegetation (i.e., similar to lettuce and cabbage) and (2) there was a vegetation yield of 2 kg/m2, { j / l G \ l SEABROOK - UNIT 1 B 3/4 12-1
RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material.in environmental sample matrices are performed as part of the quality assurance program for environ-mental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. O i SEABROOK - UNIT 1 B 3/4 12-2
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(N 5.0 DESIGN FEATURES l 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5.1-2. MAPS DEFINING UNRESTRICTED AREAS AND SITE B0UNDARY FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall-be shown in Figures 5.1-1 and 5.1-3, respectively. The definition of UNRESTRICTED AREA used in implementing these Technical Speci-fications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but'the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE s
.B0UNDARY, is utilized in the Limiting Conditions-for Operation to keep levels T of radioactive materials in liquid and gaseous effluents as low as is reasonably Q achievable, pursuant to 10 CFP. 50.36a.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
- a. Nomin 1 inside diameter = 140 feet.'
- b. Nominal inside height = 219 feet,
- c. Minimum thickness of concrete walls = 4 feet 6 inches.
- d. Minimum thickness of concrete dome = 3 feet 6 inches.
- e. Minimum thickness of concrete floor pad = 10 feet.
- f. Nominal thickness of steel liner = 1/4, 3/8, and 1/2 inch for the floor, wall, and dome, respectively,
- g. Net free volume = 2.704 X 108 cubic feet.
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DESIGN FEATURES ( ( ,/ DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 52.0 psig and a temperature of 296 F. 5.3 REACTOR CORE FUEL ASSEMBLIES l 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly con-taining 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar in phy-sical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235. CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80% silver, 15% in-dium, and 5%' cadmium. All control rods shall be clad with stainless steel tubing. 5.4 REACTOR COOLANT SYSTEM '
/ 9 DESIGN PRESSURE AND TEMPERATURE V 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 ,
of the FSAR, with allowance for normal degradation pursuant to the~ l applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650 F, except for the pressurizer which is 680 F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,265' cubic feet at a nominal T f 588.5 F. avg 5.5 METEOROLOGICAL TOWER LOCATION
- 5. 5. -1 The meteorological tower shall'be located as shown on Figure 5.1-1.
l SEABROOK - UNIT 1 5-9 _ - - _ - _ - - - - - - _ - - - - - -- I
DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. Ak eff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 1.5% Ak/k for uncertainties as described in Section 4.3 of the FSAR, and
- b. A nominal 10.35 inch center-to-center distance between fuel assemblies placed in the storage racks.
5.6.1.2 The k eff f r new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed. DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 14 feet 6 inches. CAPACITY. 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1236 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. O SEABROOK - UNIT 1 5-10
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o : li . TABLE 5.7-1 m , i 9 COMPONENT CYCLIC OR TRANSIENT LIMITS i E i 8 i CYCLIC OR DESIGN CYCLE l '. COMPONENT TRANSIENT LIMIT- OR TRANSIENT i I E . j Q Reactor Coolant System 200 heatup cycles'at < 10n F/h Heatup cycle'- T. avg from 1 200 F j ~ and 200 cooldown cycles at to > 550 F. j
-< 100*F/h.
Cooldown cycle - Tavg from _> 550 F to _< 200 F i 200 pressurizer cooldown cycles Pressurizer'cooldown cycle l at 5 200 F/h. temperatures from'> 650*F to . j $ 200 F. f 80 loss of load cycles, without > 15% of RATED THERMAL POWER to i immediate Turbine or Reactor trip. D% of RATED THERMAL POWER.
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- 1 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical !
* . A.C. electrical power. ESF Electrical System.
l 80 cycles.of loss of flow in one Loss of only one reactor i reactor coolant loop. . coolant pump. .j I 400 Reactor trip cycles. 100% to 0% of RATED THERMAL' POWER. , i ! '10 auxiliary. spray actuation Spray water temperature' differential ' cycles. > 320 F. 1 i 200 leak tests. Pressurized to > 2250 psig. 10 hydrostatic pressure tests. Pressurized to > 3106'psig. ) . Secondary Coolant System 1 steam line break. Break in a > 6-inch steam line. l 10 hydrostatic pressure tests. Pressurized to-> 1481 psig. i t : P _ .c - - -.
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1 6.0 ADMINISTRATIVE CONTROLS 6.1 -RESPONSIBILITY 6.1.1 The Station Manager'shall be responsible for overall station opera-
' tion and shall delegate in writing the succession to this responsibility during his absence.
6.1.2 The Shift Superintendent (or during his absence from the control' room, a designated individual) shall be responsible for the control room command function. .A management directive to this effect, signed by the Vice President, i Nuclear Production shall be reissued to all station personnel on an annual basis. 6.2- ORGANIZATION
- 0FFSITE 4
6'.2.1' The offsite organization for station management and technical support shall be as shown in Figure 6.2-li STATION STAFF 6.2.2 The station organization shall be as shown in Figure 6.2-2 and: ! a. Each on-duty shift shall be composed of at least the minimum shift r crew composition shown in Table 6.2-1;
- b. - At least one licensed Operator shall be in the control room when i . fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;
- c. A Health Physics Technician
- shall be on site when fuel is in the j reactor;
- d. All CORE ALTERATIONS shall'be observed and directly supervised by
- either a licensed Senior Operator or licensed Senior Operator Limited l- to Fuel Handling who has no other. concurrent responsibilities during
- this operation; and ,
a 4- e. Administrative procedures shall be developed and implemented to limit j the working hours of station staff who perform _ safety-related func-i tions, e.g. , licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel. The l amount of overtime worked by station ~ staff members performing safety-related functions shall be limited in accordance with the NRC Policy i Statement on working hours (Generic Letter No. 82-12). } 4 i *The Health Physics Technician may be less than the minimum re quirements for a period of time not to exceed 2 hours, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions. SEABROOK - UNIT 1 6-1 s
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O PSNH PRESIDENT
& CHIEF EXECUTIVE OFFICER tai YAP 4(EE PRESIDENT & CEO SENIOR VICE PRESIDENT I I VICE PRESIDENT VICE PRESIDENT MEM APO DIRECTOR OF PRODUCTION 00ALITY PROGRAMS STATION MANAGER NUCLEAR OUALITY MANAGER DIRECTOR OF DIRECTOR OF DIRECTOR OF CORPORATE MGMT. CONTROL ENGDEERING SERVICES FIGURE 6.2-1 0FFSITE ORGAN!ZATION SEABROOK - UNIT 1 6-2
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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION (1) POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6 33(2,4) y 1 SR0(4) 1 None R0 2 1 A0 1 STA 2(3) 1 None { SS - Shift Superintendent with a Senior Operator license on Unit 1 SRO - Individual with a Senior Operator license on Unit 1 R0 - Individual with an Operator license on Unit 1 A0 - Auxiliary Operator STA - Shift Technical Advisor < TABLE NOTATIONS (1) The shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within
~
the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. (2) During any absence of the Shift Superintendent from the control room while the unit is in MODE l', 2, 3, or 4, an individual with a valid Senior Operator license shall be designated to assume the control ~ room command function. During any absence of the Shift Superintendent from the control room while the unit is in MODE 5 or 6, an individual with a valid 1,enior Operator license or Operator license shall be designated to assume the control room command function. (3) The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Superintendent or the individual with a Senior Operator license meets the qualifications for the STA as required by the NRC. (4) While the ulit is in MODE 1, 2, 3 or 4, a licensed senior operator, either the SS or SRO, shall be on shift having had at least 6 months of hot operat-ing experience. O SEABROOK - UNIT 1 6-4
ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) FUNCTION 6.2.3.1 The ISEG shall function to examine station operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of station design and operating experience information, including units of similar design, which may indicate areas for improving station safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving station safety to the Executive Assistant to the Senior Vice President. COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field. RESPONSIBILITIES 6.2.3.3 .The ISEG shall be responsible for maintaining surveillance of station activities to provide independent verification
- that these activities are performed correctly and that human errors are reduced as much as practical.
V RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to the Executive Assistant to the Senior Vice President. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Superintendent in the areas of thermal hydraulics, reactor engi-neering,~and plant analysis with regard to the safe operation of the station. 6.3 TRAINING 6.3.1 A retraining and replacemen_t licensed training program for the station staff shall be maintained under the. direction of the Training Center Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-197 and Appendix A of 10 CFR Part 55 and the supplemental requirements'specified in Sections A and C of Enclosure 1 of the NRC letter dated March 28, 1980 to all licensees, and shall include familiarization with relevant industry operational experience. O *Not responsible for sign-off function. SEABROOK - UNIT 1 6-5
ADMINISTRATIVE CONTROLS 6.4 REVIEW AND AUDIT 6.4.1 STf. TION OPERATION REVIEW COMMITTEE (50RC) FUNCTION 6.4.1.1 The 50RC shall function to advise the Station Manager on all matters related to nuclear safety. COMPOSITION 6.4.1.2 The 50RC shall be composed of the: Chairman: Station Manager Member: Assistant Station Manager Member: Operations Manager Member: Technical Services Manager Member: Maintenance Department Supervisor Member: Instrumentation and Control Department Supervisor Member: Reactor Engineering Department Supervisor Member: Health Physics Department Supervisor Member: Technical Support Department Supervisor Member: _ Chemistry Department Supervisor ALTERNATES 6.4.1.3 All alternate members shall be appointed in writing by the 50RC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in 50RC activities at any one time. MEETING FREQUENCY 6.4.1.4 The 50RC shall meet at least once per calendar month and as convened by the 50RC Chairman or his designated alternate. QUORUM 6.4.1.5 The quorum of the SORC necessary for the performance of the 50RC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES 6.4.1.6 The 50RC shall be responsible for:
- a. Review of: (1) all proposed procedures required by Specification 6.7 anj changes thereto, (2) all proposed programs required by Specifi-cation 6.7 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety;
- b. Review of all proposed tests and experiments that affect nuclear safety; SEABROOK - UNIT 1 6-6
1 ADMINISTRATIVE CONTROLS Q' RESPONSIBILITIES 6.4.1.6 (Continued)
- c. Review of all proposed changes to Appendix "A" Technical Specifications;
- d. Review of all proposed changes or modifications to station systems or equipment that affect nuclear safety;
- e. Investigation of all. violations of the Technical Specifications, 1 including the preparation and forwarding of' reports covering evalua-tion and recommendations to prevent recurrence, to the Vice President-Nuclear Production and to .the Nuclear Safety Audit Review Committee (NSARC);
- f. Review of all REPORTABLE EVENTS;
- g. Review of station operations to detect potential hazards to nuclear safety;
- h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Station Manager or the NSARC;
- i. Review of the Security Plan and implementing procedures and submittal l of recommended changes to the NSARC;
- j. Review of the Emergency Plan and implementing procedures and submittal l of recommended changes to the NSARC;-
- k. Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President-Nuclear Production and to the NSARC;
- 1. Review of changes to the PROCESS CONTROL PROGRAM,'0FFSITE DOSE CALCULATION MANUAL, and the Radwaste Treatment System; and
- m. Review of the Fire Protection Program and implementing instructions and submittal of recommended. changes to the NSARC.
6.4.1.7 The SORC shall:
- a. Recommend in writing to the Station Manager approval or disapproval of items considered under Specification 6.4.1.6a. through d;
- b. Render determinations in writing with regard to whether or not each item considered under Specification 6.4.1.6a. through e. constitutes an unreviewed safety question; and
- c. Provide written notification within 24 hours to the Vice President-Nuclear Production and the NSARC of disagreement between the 50RC SEABROOK - UNIT 1 6-7
ADMINISTRATIVE CONTROLS RESPONSIBILITIES 6.4.1.7 (Continued) and the Station Manager however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1. RECORDS 6.4.1.8 The 50RC shall maintain written minutes of each SORC meeting that, -at a minimum, document the results of all SORC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President-Nuclear Production and the NSARC. 6.4.2 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC) FUNCTION 6.4.2.1 The NSARC shall function to provide independent review and audit of designated activities in the areas of:
- a. Nuclear power plant operations,
- b. Nuclear engineering,
- c. Chemistry and radiochemistry,
- d. Metallurgy,
- e. Instrumentation and control,
- f. Radiological safety,
- g. Mechanical and electrical engineering, and
- h. Quality assurance practices.
The NSARC shall report to and advise the Senior Vice President on those areas of responsibility specified in Specifications 6.4.2.7 and 6.4.2.8. COMPOSITION 6.4.2.2 The NSARC shall be composed of at least five (5) individuals. The Chairman, Vice Chairman and members, including designated alternates, shall be appointed in writing by the Senior Vice President. Collectively, the individuals appointed to the NSARC should be competent to conduct reviews identified by Specification 6.4.2.1. Each member shall meet the qualifica-tions of ANSI 3.1-1978, Section 4.7. ALTERNATES 6.4.2.3 All alternate members shall be appointed in writing by the Senior Vice President to serve on a temporary basis; however, no more than a minority shall participate as voting members in NSARC activities at any one time. CONSULTANTS 6.4.2.4 Consultants shall be utilized as determined by the NSARC to provide expert advice to the NSARC. SEABROOK - UNIT 1 6-8
ADMINISTRATIVE CONTROLS n MEETING FREQUENCY 6.4.2.'5 The NSARC shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and thereafter at least once per 6 months 1 6 weeks. QUORUM 6.4.2.6 -The quorum of the NSARC necessary for the performance of the NSARC review and audit functions of these Technical Specifications shall consist of the Chairman or Vice-Chairman and at least four NSARC members including alter-nates. No more than a minority of the quorum shall have line responsibility for operation of the unit. The Vice Chairman, or his designated alternate, can
. participate as an NSARC member when the Chairman is in attendance.
REVIEW 6.4.2.7 The NSARC shall be responsible for the review of:
- a. The safety evaluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question;
- b. Proposed changes to procedures, equipment, or systems that involve an unreviewed safety question as defined in 10 CFR 50.59; V c. Proposed tests or experiments that involve an unreviewed safety ques-tion as defined in 10 CFR 50.59;
- d. Proposed changes to Technical Specifications or this Operating License;
- e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;
- f. Significant operating abnormalities or deviations from normal and expected performance of station equipment that affect nuclear safety;
- g. All REPORTABLE EVENTS;
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and
- i. Reports and meeting minutes of the SORC.
AUDITS 6.4.2.8 Audits of station activities shall be performed under the cognizance of i (~ the NSARC. The audits shall be performed within the specified time interval ( with a maximum allowable extension not to exceed 25% of the specified interval SEABROOK - UNIT 1 6-9
ADMINISTRATIVE CONTROLS AUDITS 6.4.2.8 (Continued) provided the combined time interval for any three consecutive intervals shall not exceed 3.25 times the specified interval. These audits shall encompass:
- a. The conformance of station operatien to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months;
- b. The performance, training, and qualifications of the entire station staff at least once per 12 months;
- c. The results of actions taken to correct deficiencies occurring in station equipment, structures, systems, or method of operation that affect nuclear safety, at least once per 6 months;
- d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months;
- e. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA personnel;
- f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee
~
fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year;
- g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;
- h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;
- i. The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months;
- j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months;
- k. The Emergency Plan and implementing procedures at least once per 12 months;
- 1. The Security Plan and implementing procedures at least once per 12 months; and
- m. Any other area of station operation considered appropriate by the NSARC or the Senior Vice President.
SEABROOK - UNIT 1 6-10
ADMINISTRATIVE CONTROLS fm i) - ( ,/ RECORDS 6.4.2.9 Records of NSARC activities shall be prepared and distributed as
~
indicated below:
- a. -Minutes of each NSARC meeting'shall be prepared and forwarded to the
-Senior Vice President within 14 days following each meeting;.
- b. Reports of reviews encompassed by Specification 6.4.2.7 shall be included in the minutes where applicabl_e or forwarded under sepa-rate cover to the Senior Vice President within 14 days following completion of the review; and
- c. Audit reports encompassed by Specification 6.4.2.8 shall be forwarded to the Senior Vice President and to the management positions respons-ible for the areas audited within 30 days'after completion of the audit by the auditing organization.
6.5 REPORTABLE EVENT ACTION The following actions shall be.taken for REPORTABLE EVENTS:
- a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and
- b. Each REPORTABLE EVENT shall be reviewed by the 50RC and the
(,/. results of this review shall be submitted to the NSARC and the Vice President-Nuclear Production. 6.6 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety Limit is violated:
- a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour. The Vice President-Nuclear Production and the NSARC shall be notified within 24 hours;
- b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the 50RC. This report shall describe: (1) applicable circumstances preceding the violation,' (2) effects of the vio.lation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;
- c. The Safety Limit Violation Report shall be submitted to the Commission, the NSARC, and the Vice President-Nuclear Production within 14 days of the violation; and
- d. Operation of the station shall not be resumed until authorized by the Commission.
O SEABROOK - UNIT 1 6-11
ADMINISTRATIVE CLNTROLS 6.7 PROCEDURES AND PROGRAMS 6.7.1 Written procedures shall te established, implemented, and maintained covering the activities referenced below:
- a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;
- b. The emergency operating procedures required to' implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33;
- c. Security Plan implementation;
- d. Emergency Plan implementation;
- e. PROCESS CONTROL PROGRAM implementation;
- f. OFFSITE DOSE CALCULATION MANUAL implementation;
- g. Quality Assurance Program for effluent and environmental monitoring;
- h. Fire Protection Program implementation; and
- i. Technical Specification Improvement Program implementation.
6.7.2 Each procedure of Specification 6.7.1, and changes thereto, shall be reviewed by the 50RC and shall be approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures. 6.7.3 Temporary changes to procedures of Specification 6.7.1 may be made pro-vided:
- a. The intent of the original procedure is not altered;
- b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and
- c. The change is documented, reviewed by the SORC, and approved by the Station Manager within 14 days of implementation.
6.7.4 The following programs shall be established, implemented, and maintained:
- a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outride containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the RHR and containment spray, Safety Injection, chemical and volume control. The program shall include the following:
SEABROOK - UNIT 1 6-12
ADMINISTRATIVE CONTROLS 7 ~5) ( ,) PROCEDURES AND PROGRAMS 6.7.4a. (Continued)
- 1) Preventive maintenance and periodic visual inspection require-ments, and
- 2) Integrated leak test requirements for each system at refueling cycle intervals or less.
- b. In-Plant Radiation Monitoring A program that will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
- 1) Training of personnel,
- 2) Procedures for monitoring, and
- 3) Provisions for maintenance of sampling and analysis equipment.
- c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit /\ steam generator tube degradation. This program shall include:
k 1) Identification of a sampling schedule for the critical variables and control points for these variables,
- 2) Identification of the procedures used to measure the values of
- the critical variables,
- 3) Identification of' process sampling points, which shall include ;
monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,
- 4) Procedures for the recording and management of data,
- 5) Procedures defining corrective actions for all off-control point chemistry conditions, and i
- 6) A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
- d. Backup Method for Determining Subcooling Margin A program that will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include t
, -s the following:
' SEABROOK - UNIT 1 6-13
-..a-m. , , . , , - . _ . e ~m-. , , . _ , . , , , ,,.,,,_.-,.,--,r. . , , . - - - , - , - - , _ - , - . . - , _ - . - _ - - - , . , . ~ . . .
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.4d. (Continued)
- 1) Training of personnel, and y, , ,,2
- 2) Procedures for monitoring.
- e. Post-Accident Sampling A program that will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:
- 1) Training of personnel,
- 2) Procedures for sampling and analysis, and
- 3) Provisions for maintenance of sampling and analysis equipment.
6.8 REPORTING REQUIREMENTS ROUTINE REPORTS 6.8.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted. STARTUP REPORT
- 6. 8.1.1 A summary report of station startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear,
. thermal, or hydraulic performance of the station. The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license condi-tions based on other commitments shall be included in this report. Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of. commercial operation), supplementary reports shall be suomitted at least every 3 months until all three events have been completed. SEABROOK - UNIT 1 6-14
i ADMINISTRATIVE CONTROLS
- M t < 'd ANNUAL REPORTS *
- 6. 8.1. 2 Annual Reports covering the activities of the station as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.
Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number.of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions ** (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [ describe maintenance), waste processing, and refueling).
.The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling.less.than 20% of the individual total. dose need not be accounted for. In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions;
- b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded
[V ) (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one
- analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (pCi/gm) and one other radio-iodine isotope concentration (pCi/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
- c. Documentation of all challenges to the pressurizer power-operated relief valves (PORVs) and safety valves.
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT *** 6.8.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the station during the previous calendar year shall be submitted 4
*A single submittal may be made for a multiple unit station. The submittal i should combine those sections that are common to all units at the station.
**This tabulation supplements the requirements of S20.407 of 10 CFR Part 20.
Q ***A single submittal may be made for a multiple unit station. SEABROOK - UNIT 1 6-15
l ADMINISTRATIVE' CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT O 6.8.1.3 (Continued) prior to May 1 of each year. The initial report shal; be submitted prior to May 1 of the year following initial. criticality and shall include copies of the preoperational Radiological Environmental Pro (ram of the unit for at least 2 years prior to criticality. The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, an1 an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls, as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operatipn on the environment. The reports shall also include the results of the Land Use Census required by Specification 3.12.2. The Annual Radiologicai Environmental Operating Reports shall include the results of analysis of all radiological environmental-samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some indi-vidual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be sebmitted as soon as possible in.a supplementary report. , The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legit'ie maps
- covering all sampling locations keyed to a tatale giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Specification 3.12.3; reason for not conducting the Radiological Environmental Monitoring Program as required by specification 3.12.1, and discussio6 of all deviations fro:n the sampling schedule; discussion of environmental sample measurements that exceed the reporting levels but are not the result of plant effluents, pursuant to ACTION b. of Specification 3.12.1; and discussion of all Analyses in which the LLD required was not achievable.
t
*0ne map shall cover locations near the SITE B0UNDARY; the more distant locations shall be covered by one or more additional maps.
**A single submittal mai be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
SEABROOK - UNIT 1 6-16
- _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . - _ _ . _ _ . . _ _ _ _ . _ ___--______-___-_---____a
g ADMINISTRATIVE CONTROLS' SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT ** 6.8.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the station during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality. The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the station as outlined in Regulatory Guide 1.21,
" Measuring, Evaluating, and Reporting Radioactivity.in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.
For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement). The Semiannual Radioactive Effkent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year **. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured),
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1 or in the form of joint. frequency distributions of wind speed, wind direction, Q and atmospheric stability.* This same report shall include an Lssessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station' during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (0DCM). The Semiannual Radioactive Effluent Release Report to be. submitted within 60 days after January 1 of.each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year "In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC p ' upon request.
**The dose calculations may be reported in a' supplement submitted 30 days later.
SEABROOK - UNIT 1 6-17
ADMINISTRATIVE CONTROLS SEHIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Continued) to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977. The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS.of radioactive materials in gaseous and liquid effluents made during the reporting period. The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, pursuant to Specifications 6.12 and 6.13, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.14. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2. The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively. MONTHLY OPERATING REPORTS 6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report. RADIAL PEAKING FACTOR LIMIT REPORT 6.8.1.6 The F xy limits for RATED THERMAL POWER (F P) shall be provided to the NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, Attention: Chief, Reactor Systems Branch, DPL-A, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555, for all core planes containing Bank "D" control rods and all unrodded core planes and the plot of predicted (F PRel) vs Axial Core Height with the limit envelope at least 60 days prior to each cycle initial criticalit.v unless otherwise approved by the Commission by letter. In addition, in the event that the limit should change requiring a new substantial or an amended submittal to the Radial Peaking Factor Limit Report, it will be submitted 60 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. Any informa-tion needed to support F xRTP will De by request from the NRC and need not be included in this report. SEABROOK - UNIT 1 6-18
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ADMINISTRATIVE CONTROLS
! h V SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
6.9 RECORD RETENTION 6.9.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following' records shall be retained for at least the minimum period indicated. 6.9.2 The following records shall be retained for at least 5 years:
- a. Records and logs of station operation covering time interval at each power level;
- b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;
- e. Records of changes made to the procedures required by Specifi-v cation 6.7.1;
- f. Records of radioactive shipments;
- g. Records of sealed source and fission detector leak tests and results; and
- h. Records of annual physical inventory of all sealed source material of record.
6.9.3--The following' records shall be retained for the duration of the station Operating License:
- a. Records and drawing changes reflecting station design modifications made to systems and equipment described in the Final Safety Analysis-Report;
- b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories;
- c. Records of radiation exposure for all individuals entering radiation control areas;
- d. Records of gaseous and liquid radioactive material released to the environs; k e. Records of transient or operational cycles for those station components identified in Table 5.7-1;
'SEABROOK - UNIT 1. 6-19
ADMINISTRATIVE CONTROLS RECORD RETENTION 6.9.3 (Continued)
- f. Records of reactor tests and experiments;
- g. Records of training and qualification for current members of the station staff;
- h. Records of inservice inspections performed pursuant to these Technical Specifications;
- i. Records of quality assurance activities required by the Operational Quality Assurance Manual;
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59;
- k. Records of meetings of the SORC and the NSARC;
- 1. Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.7 including the date at which the service life commences and associated installation and maintenance records;
- m. Records of secondary water sampling and water quality; and
- n. Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.
6.10 RADIATION PROTECTION PROGRAM 6.10.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. 6.11 HIGH RADIATION AREA 6.11.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the " control device" or " alarm signal" required by paragraph 20.203(c), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface that the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radia-tion areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high SEABROOK - UNIT 1 6-20
ADMINISTRATIVE CONTROLS
/ \ \. v } HIGH RADIATION AREA' 6.-11.1 (Continued) radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device.that continuously indicates the radiation dose rate in the area; or
- b. A radiation aonitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or
- c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified ire the Radiation Work Permit.
6.11.2 In addition to the requirements of Specification 6.11.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface that the radiation penetrates I, m shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Superintendent on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where - no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device. 6.12 PROCESS CONTROL PROGRAM (PCP) 6.12.1 The PCP shall be approved by the Commission prior to implementation.
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6.12.2 Licensee-initiated changes to the PCP:
- a. Shall be submitted to the Commission in the Semiannual Radioactive p Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
SEABROOK - UNIT 1- 6-21
ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) 6.12.2 (Continued)
- 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
- 2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- 3) Documentation of the fact that the cnange has been reviewed and found acceptable by the 50RC.
- b. Shall become effective upon review and acceptance by the 50RC.
6.13 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.13.1 The ODCM shall be approved by the Commission prior to implementation. 6.13.2 Licensee-initiated changes to the ODCM:
- a. Changes to Part A shall be submitted to and approved by the NRC staff prior to implementation,
- b. Changes to Part B shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:
- 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s);
- 2) A determination that the change will not' reduce the accuracy or reliability of dose calculations or Setpoint determinations; and
- 3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC.
- c. Changes to Part B shall become effective upon review and acceptance by the 50RC.
1 Ol SEABROOK - UNIT 1 6-22
o 1 ADMINISTRATIVE CONTROLS 6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS
- 6.14.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):
- a. Shall be reported to the Comtnission in' the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the SORC. The discussion of each change shall contain:
- 1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;
- 2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- 3) A detailed description of the equipment, components, and processes involved and the interfaces with other p] ant systems;
- 4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;
,A 5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto; 6)' A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the change is to be made;
- 7) An estimate of the exposure to plant operating personnel as a result of the change; and
- 8) Documentation of the fact that the change was reviewed and found acceptable by the 50RC.
- b. Shall become effective upon review and acceptance by the 50RC.
O\ j
- Licensees may choose to submit the information called for in this Specification
'u as part of the annual FSAR update.
SEABROOK - UNIT 1 6-23
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BIBLIOGRAPHIC DATA SHEET NUREG-1207 2 Leeve b eca E AND $v5 T 81 L E 4 HECsesENT $ ACCE SSION Nov.E R echnical Specifications for Seabrook Station, Unit 1 b D A T E R E POH T COMP L E T E D Appendix "A" to License No. NPF-56 * " l" OCTOBER 1986
& AUTHOMES, 7 D ATE REPORT ISSUED MONTM YEAR OCTOBER 1986 9 PHOJECT, T Asn 'WORet UNIT NUMSE R g
PERFOHviNG ORG ANil ATiON NAME AND MAILING ADDHE S5 trace ude 2 o Codel Division of PWR-Licensing-A Office of Nuclear Reactor Regulation io i~ Nuu.E R U. S. Nuclear Regulatory Commission Washington, D. C. 20555 11 $PONSORING ORG ANil ATION N AME AND MAILsNG ADDHE SS flac'ude I,a Codri 12a T YPE OF REPORT Same as 8. above Technical 120 PE RIOD COvf RED unctusage deres) October 1986 -
- 1) $UPPLEVE NT ARY NOTES Docket Nos. 50-443 & 50-444 14 A85TR ACT I200 woras or Arsst
{ ) The Seabrook Station, Units 1 and 2 Technical Specifications were prepared by the U. S. Nuclear Regulatory Commission to Set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set f~rth o in Section 50.36 of 10 CFR Part 50 for the protectbn of the health and safety of the public. 15e K E Y WORDS AND DOCUutNT ANALV$iS 150 DE SCR IP T OR $ Seabrook License Technical Specifications
+ vail ARittTY ST ATE Yt NT 17 SE CURIT Y C L AS$tt eC A T ION is Nsv6EH DhPAGES UNCLASSIFIED UNLIMITED i sECuH.T v u Assi iCA TION 20 P ICE UNCLASSIFIED s
UNITED STATES ,,,c,,t ,ovare. ctass nar ' NUCLEAR REGULATORY COMMISSION Postaon ertisraio WASHINGTON, D.C. 20555 wfs'7"se. FfRMIT hio G47 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 9 9
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