ML20137G590
| ML20137G590 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/30/1985 |
| From: | Martin R Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1149, NUDOCS 8508270346 | |
| Download: ML20137G590 (500) | |
Text
{{#Wiki_filter:1 NUREG-1149 Technical Specifications , , Limerick Generating Station, , Unit No.1 Docket No. 50-352 Appendix "A" to License No. NPF-39 O Issued by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1985 pe "%,, O Pn r m =88sss. PDR
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NUREG-1149 i Technical Specifications Limerick Generating Station, Unit No.1 Docket No. 50-352 Appendix "A" to License No. NPF-39 Issued by the
! U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1985 p>"'%,
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(m ' DEFINITIONS SECTION.
- 1. 0 DEFINITIONS PAGE 1.1 ACTI0N....................................................... 1-1 1.2 AVERAGE PLANAR EXPOSURE...................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 1-1 1.4 CHANNEL" CALIBRATION.......................................... 1-1 1.5 CHANNEL CHECK................................................ 1-1 1.6 CHANNEL FUNCTIONAL TEST...................................... 1-1 1.7 CORE ALTERATION.............................................. 1-2 1.8 CRITICAL POWER RATI0......................................... 1-2
- 1. 9 DOSE EQUIVALENT I-131........................................ 1-2 g'- 1.10 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-2
~
1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME........... 1-2
- 1.12 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.... 1-2 1.13 FRACTION OF LIMITING POWER DENSITY........................... 1-3 1.14 FRACTION OF RATED THERMAL P0WER.............................. 1-3 1
1.15 FREQUENCY N0TATION........................................... 1-3
- 1.16 IDENTIFIED LEAKAGE........................................... 1-3
. 1.17 ISOLATION SYSTEM RESPONSE TIME............................... 1-3 r 1.18 LIMITING CONTROL R0D PATTERN................................. 1-3
-1.19 LINEAR HEAT GENERATION RATE.................................. 1-3
, 1.20 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-4 1.21 MAXIMUM FRACTION OF LIMITING POWER DENSITY................ .. 1-4 ( LIMERICK - UNIT 1 i 4
, , ~ . , - ,. ...w ., . nn,._ , , ,, ,, - _ - , - . . .-- -- - - -,- .. ,, _- __- _, . . _
DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.22 MEMBER (S) 0F THE PUBLIC....................... .............. 1-4 1.23 MINIMUM CRITICAL POWER RATIO..... ........................... 1-4 1.24 0FFSITE DOSE CALCULATION MANUAL.............. .............. 1-4 1.25 OPERABLE - OPERABILITY............................ .......... 1-4 1.26 OPERATIONAL CONDITION - CONDITION............ ............... 1-4 1.27 PHYSICS TESTS................... ...... ..................... 1-4 1.28 PRESSURE B0UNDARY LEAKAGE.................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITf................................ 1-5 1.30 PROCESS CONTROL PR0 GRAM...................................... 1-5 1.31 PURGE - PURGING................. ............................ 1-5 1.32 RATED THERMAL P0WER..................... .................... 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY............ 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME...................... 1-6 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY.............. 1-6 1.36 REPORTABLE EVENT..... ................. ........... ......... 1-7 1.37 R00 DENSITY......... .................. ........... .. ...... 1-7 1.38 SHUTDOWN MARGIN... ............................... .......... 1-7 1.39 SITE B0VNDARY................................................ 1-7 1.40 SOLIDIFICATION........ ...................................... 1-7 1.41 SOURCE CHECK............... .... ............ ............... 1-7 1.42 STAGGERED TEST BASIS......................................... 1-8 1.43 THERMAL POWER..... ............... ................ ........ 1-8 1.44 UNIDENTIFIED LEAKAGE.. ............................. ........ 1-8 LIMERICK - UNIT 1 ii
[ ? 4 i' ! INDEX ;
-DEFINITIONS I .~ SECTION F
DEFINITIONS (Continued) PAGE 1.45 UNRESTRICTED AREA............................................ 1-8 ! 1.46 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-8 11.47 VENTING...................................................... 1-8 r Table 1.1, Surveillance Frequency Notation........................ 1-9
. Table 1.2, Operational Conditions................................. 1-10 j i
i i r t I 3' ; l i I i t i t >
- e. .
f- ! i f
- i. .
4- i t i 4 1 9 1 t t i [ i, LIMERICK - UNIT 1- lii ,
! i i
i
~ - -- - .-- - --. - . - - - . . . - - - - - - ~ . _ _ . _ - _ . _ . , - - - -
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow. . . . . . . . ........... 2-1 THERMAL POWER, High Pressure and High Flow............. .... 2-1 Reactor Coolant System Pressure............................. 2-1 Reactor Vessel Water Level.................................. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protectio: System Instrumentation Setpoints......... 2-3 Table 2.2.1--1 Reactor Protection System Instrumentation Setpoints. ............. 2-4 BASES 2.1 SAFETY LIMITS O THERMAL POWER, Low Pressure or Low Flow................. ... B 2-1 THERMAL POWER, High Pressure and High Flow.................. B 2-2 Bases Table B 2.1.2-1 Uncertainties Used In The Determination Of The Fuel Cladding Safety Limit. . . . . . . . . . . . B 2-3 Bases Table B 2.1.2-2 Nominal Values Of Parameters Used In The Statistical Ana-lysis Of Fuel Cladding Integrity Safety Limit........ .. B 2-4 Reactor Coolant System Pressure............................. B 2-5 Reactor Vessel Water Level.................................. B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......... B 2-6 O LIMERICK - UNIT 1 iv
INDEX
,Im
\ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE
'3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS l
,3/4.1.1- SHUTD0WN' MARGIN.......................................... 3/4 1-1 3/4.1.2}REACTIVITYAN0MALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability.................................. 3/4 1-3 Control Rod Maximum-Scram Insertion Times................ 3/4 1 Control Rod Average Scram. Insertion Times................ 3/4 1-7 Four Control Rod Group Scram Insertion Times............. 3/4 1-8
, Cont ol Rod Scram Accumulators........................... 3/4 1-9 Control Rod Drive Coupling............................... 3/4 1-11 Control Rod Position Indication.......................... 3/4 1-13 Control Rod Drive Housing Support........................ 3/4 1-15 3'4.1.4
/ CONTROL R00 PROGRAM CONTROLS Rod Worth Minimizer...................................... 3/4~1-16
- j. Rod Sequence Control System.............................. 3/4 1-17
+
- i '
, Rod Block Monitor........................................ 3/4 1-18 3/4bl.5 STANDBY LIQUID CONTROL SYSTEM............................ 3/4 1-19 i
,,' Figure 3.1.5-1 Sodium Pentaborate Solution i '
Temperature / Concentration Requirements........................ 3/4 1-21 Figure 3.1.5-2 Sodium Pentaborate Solution c
,' Volume / Concentration Requirements... 3/4 1-22
+ 3'/4.2 POWER DISTRIBUTION LIMITS
- ( 3/4.2.1- ; AVERAGE PLANAR LINEAR HEAT GENERATION RATE...............
*i 3/4 2-1
[ : '(.; Figure 3.2.1-1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus t
^
Average Planar Exposure Initial Core Fuel Types P8CIB278............ 3/4 2-2 LIMERICK - UNIT 1 v g j
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) i Figure 3.2.1-2 Maximum Average Planar. Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB248........... 3/4 2-3 Figtec 3. 2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB263........... 3/4 2-4 Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB094........... 3/4 2-5 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB071........... 3/4 2-6 3/4 2.2 APRM SETP0INTS.......................................... 3/4 2-7 3/4 2.3 MINIMUM. CRITICAL POWER RATI0............................ 3/4 2-8 Figure 3.2.3-1 Minimum Critical Power Ratio (MCPR) Versus I at Rated Flow...... 3/4 2-10 Figure 3.2.3-2 K 7 Factor...................... .. 3/4 2-11 3/4.2.4 LINEAR WEAT GENERATION RATE............................. 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION....... ....... 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation....... ............. 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times...................... 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements...................... 3/4 3-7 O LIMERICK UNIT - 1 vi
INDEX n
- (v) LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . 3/4 3-9 Table 3.3.2-1 Isolation Actuation Instrumentation..................... 3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints........... 3/4 3-18 Table 3.3.2-3 Isolation System Instrumen-tation Response Time................ 3/4 3-23 Table 4.3.2.1-1 Isolation Actuation Instrumen-tit on Surveillance Requirements...................... 3/4 3-27 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION......................................... 3/4 3-32 Table 3.3.3-1 Emergency Core Cooling System Actuation Instrumentation........... 3/4 3-33
( Table 3.3.3-2 Emergency Core Cooling System
\s, Actuation Instrumentation Setpoints........................... 3/4 3-37 Table 3.3.3-3 Emergency Core Cooling System Response Times...................... 3/4 3-39 Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation j Surveillance Requirements......... 3/4 3-40 i
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION i ATWS Recirculation Pump Trip System Instrumentation..... 3/4 3-42 Table 3.3.4.1-1 ATWS Recirculation Pump Trip System Instrumentation............ 3/4 3-43 Table 3.3.4.1-2 ATWS Recirculation Pump Trip System Instrumentation Setpoints......................... 3/4 3-44 Table 4.3.4.1-1 ATWS Recirculation Pump Trip l Instrumentation Surveillance 3/4 3-45 Requirements...................... 3 End-of-Cycle Recirculation Pump Trip System N- Instrumentation......................................... 3/4 3-46 LIMERICK - UNIT 1 vii
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) Table 3.3.4.2-1 End-of-Cycle Recirculation Pump Trip System Instrumentation....... 3/4 3-48 Table 3.3.4.2-2 End-of-Cycle Recirculation Pump Trip Setpoints.................... 3/4 3-49 Table 3.3.4.2-3 End-0f-Cycle Recirculation Pump Trip System Response Time......... 3/4 3-50 Table 4.3.4.2.1-1 End-0f-Cycle Recirculation Pump Trip System Surveillance Requirements.................... 3/4 3-51 3/4.3.5 REACTOR CORE ISUutTION COOLING SYSTEM ACTUATION INSTRUMENTATION......................................... 3/4 3-52 Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instrumenta-tion................................ 3/4 3-53 Table 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation l Setpoints........................... 3/4 3-55 l Table 4.3.5.1-1 Reactor Core Isolation Cooling i System Actuation Instrumentation Surveillance Requirements.......... 3/4 3-56 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION....................... 3/4 3-57 Table 3.3.6-1 Control Rod Block Instrumenta-tion................................. 3/4 3-58 Table 3.3.6-2 Control Rod Block Instrumenta-tion Setpoints....................... 3/4 5-60 Table 4.3.6-1 Control Rod Block Instrumenta-tion Surveillance Requirements....... 3/4 3-61 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.................... 3/4 3-63 Table 3.3.7.1-1 Radiation Monitoring Instrumentation................... 3/4 3-64 LIMERICK - UNIT 1 viii
[] INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements...................... 3/4 3-66 Seismic Monitoring Instrumentation...................... 3/4 3-68 Table 3.3.7.2-1 Seismic Monitoring Instrumentation.................... 3/4 3-69 Table 4.3.7.2-1 Seismic Monitoring Instrumentation Surveillance Requirements....................... 3/4 3-71 Meteorological Monitoring Instrumentation............... 3/4 3-73 Table 3.3.7.3-1 Meteorological Monitoring Instrumentation................... 3/4 3-74 Table 4.3.7.3-1 Meteorological Monitoring (' Instrumentation Surveillance Requirements...................... 3/4 3-75 Remote Shutdown System Instrumentation and Controls. . . . . 3/4 3-76 Table 3.3.7.4-1 Remote Shutdown System Instrumentation and Controls...... 3/4 3-77 Table 4.3.7.4-1 Remote Shutdown System Instrumentation Surveillance Requirements...................... 3/4 3-83 i Accident Monitoring Instrumentation..................... 3/4 3-84 i Table 3.3.7.5-1 Accident Monitoring Instrumen-tation............................ 3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instrumenta-tion Surveillance Requirements.... 3/4 3-87
- Source Range Monitors................................... 3/4 3-88 Traversing In-Core Probe System......................... 3/4 3-89 O Chlorine Detection System............................... 3/4 3-90 Toxic Gas Detection System.............................. 3/4 3-91 Fire Detection Instrumentation.......................... 3/4 3-92 LIMERICK - UNIT 1 ix
}
i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) Table 3.3.7.9-1 Fire Detection Instrumentation.... 3/4 3-93 Loose-Part Detection System............................. 3/4 3-97 Radioactive Liquid Effluent Monitoring Instrumen-tation... .................................. ........... 3/4 3-98 Table 3.3.7.11-1 Radioactive Liquid Effluent Monitoring Instrumentation....... 3/4 3-99 Table 4.3.7.11-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements..... .. 3/4 3-101 Radioactive Gaseous Effluent Monitoring Instrumen-tation............................... ..... ............ 3/4 3-103 Table 3.3.7.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation....... 3/4 3-104 Table 4.3.7.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements........ 3/4 3-107 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM................ ..... 3/4 3-110 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............. ... . ..................... 3/4 3-112 Table 3.3.9-1 Feedwater/Mair. T ivine Trip System Actuation Instrumentation.... 3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpoints.............. ..... 3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrirmenta-tion Surveillance Require-ments................... .... . 3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.. .. ............ .. .. ........ 3/4 4-1 LIMERICK - UNIT 1 x
[] INDEX v LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REACTOR COOLANT SYSTEM (Continued) Figure 3.4.1.1-1 Thermal Power versus Core F10w............................. 3/4 4-3 Jet Pumps............................................... 3/4 4-4 Recirculation Pumps..................................... 3/4 4-5 Idle Recirculation Loop Startup......................... 3/4 4-6 3/4.4.2 SAFETY / RELIEF VALVES.................................... 3/4 4-7 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................... 3/4 4-8 Operational Leakage..................................... 3/4 4-9 p Table 3.4.3.2-1 Reactor Coolant System Pressure f Isolation Valves.................. 3/4 4-11 3/4.4.4 -CHEMISTRY............................................... 3/4 4-12 Table 3.4.4-1 Reactor Coolant System Chemistry Limits.............................. 3/4 4-14 3/4.4.5 SPECIFIC ACTIVITY....................................... 3/4 4-15 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program.......'.. 3/4 4-17 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................. 3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel Metal Temperature Vs. Reactor Vessel Pressure.................. 3/4 4-20 Table 4.4.6.1.3-1 Reactor Vessel Material Surveil-ance Program - Withdrawal Schedule........................ 3/4 4-21 Reactor Steam Dome...................................... 3/4 4-22 O) (
.v 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................ 3/4 4-23 3/4.4.8 STRUCTURAL INTEGRITY.................................... 3/4 4-24 LIMERICK - UNIT 1 xi
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown............................................ 3/4 4-25 Cold Shutdown....................... ................... 3/4 4-26 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - 0PERATING........................................ 3/4 5-1 3/4.5.2 ECCS - SHUTD0WN......................................... 3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER..................................... 3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity........................... 3/4 6-1 Primary Containment Leakage............................. 3/4 6-2 Primary Containment Air Lock............................ 3/4 6-5 MSIV Leakage Control System............ ................ 3/4 6-7 Primary Containment Structural Integrity................ 3/4 6-8 Drywell and Suppression Chamber Internal Pressure. . . . . . . 3/4 6-9 Drywell Average Air Temperature. . . ..................... 3/4 6-10 Drywell and Suppression Chamber Purge System............ 3/4 6-11 3/4.6.2 DEPRESSURIZATION SYSTEMS Suppression Chamber..................................... 3/4 6-12 Suppression Pool and Drywell Spray...................... 3/4 6-15 Suppression Pool Cooling................................ 3/4 6-16 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES........... ........ 3/4 6-17 Table 3.6.3-1 Primary Containment Isolation Valves 3/4 6-19 9 LIMERICK - UNIT 1 xii
. g'"Sg INDEX
't :
\' '
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE i CONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEF Suppression Chamber - Drywell Vacuum Breakers........... 3/4 6-44 3/4.6.5 SECONDARY CONTAINMENT Reactor Enclosure Secondary Containment Integrity....... 3/4 6-46 4 Refueling Area Secondary Containment Integrity.......... 3/4 6-47 Reactor Enclosure Secondary Containment Automatic Isolation Valves................... .................... 3/4 6-48 Table 3.6.5.2.1-1 Reactor Enclosure Secondary 4 Containment. Ventilation System Automatic Isolation Valves.......................... 3/4 6-49
/ Refueling Area Secondary Containment Automatic
(,_,<) Isolation Va1ves........................................ 3/4 6-50 Table 3.6.5.2.2-1 Refueling Area Secondary Contain-ment Ventilation System Automatic Isolation Valves................ 3/4 6-51 Standby Gas Treatment System............................ 3/4 6-52 Reactor Enclosure Recirculation System.................. 3/4 6-55 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Primary Containment Hydrogen Recombiner Systems. . . . . . . . . 3/4 6-57 - Drywell Hydrogen Mixing System.......................... 3/4 6-58 Drywell and Suppression Chamber Oxygen Concentration. . . . 3/4 6-59 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS Residual Heat Removal Service Water System.............. 3/4 7-1 Emergency Service Water System.......................... 3/4 7-3 r'^ (.N 3/4 7-5 i Ultimate Heat Sink...................................... LIMERICK - UNIT 1 xiii
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM.......... 3/4 7-6 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM.. ................ 3/4 7-9 3/4.7.4 SNUBB:RS...... ......................................... 3/4 7-11 Figure 4.7.4-1 Sample Plan 2) For Snubber Functional Test......... .......... 3/4 7-16 3/4.7.5 SEALED SOURCE CONTAMINATION.... ....... ..... .......... 3/4 7-17 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Suppression IIater System........... ............... 3/4 7-19 Spray and/or Sprinkl er Systems. . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-22 CO 2 Systems................... . ....................... 3/4 7-24 Halsn Systems.. ......... . ...................... ... 3/4 7-25 Fire Hose Stations.. ...... .... . ............. ....... 3/4 7-26 Table 3.7.E:5-1 Fire Hose Stations.. ... ....... 3/4 7-27 Yard Fire Hydrants and Hydrant Hose Houses.............. 3/4 7-29 Table 3.7.6.6-1 Yard Fire Hydrants and Associated Hydrant Hose Houses....... ....... 3/4 7-30 3/4.7.7 FIRE RATED ASSEMBLIES........ ........... .... . ....... 3/4 7-31 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. Sources - Operating... ....... . .... . ... .. . .. 3/4 8-1 Table 4.8.1.1.2-1 Diesel Generator Test Schedule..... . ...... ...... 3/4 8-8 A.C. Sources - Shutdown...... .. .... . ..... ..... 3/4 8-9 3/4.8.2 D.C. SOURCES D. C. Sources - Oper a ting. . . . . . . . . . .. . . ........ 3/4 8-10 LIMERICK - UNIT 1 xiv
._- . . _ . . - ~ - .. . - - . - _ . . . - .-
INDEX , .i
\
- i. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE l:
- ELECTRICAL POWER SYSTEMS (Continued) .
- Table 4.8.2.1-1 Battery Surveillance Requirements...................... 3/4 8-13 D.C. Sources ; Shutdown................................. 3/4 8-14
[
- 1. 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS I
Distribution - Operating................................ 3/4 8-15 i ' Distribution - Shutdown................................. 3/4 8-18 { 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES. , Primary Containment Penetration Conductor Overcurrent Protective Devices.................................... 3/4 8-21 i Table 3.8.4.1-1 Primary Containment Penetration Conductor Overcurrent Protective
) Devices........................... 3/4 8-23 e /
i Motor-0perated Valves Thermal Overload Protection....... 3/4 8-27
- - Reactor Protection System Electric Power Monitoring..... 3/4 8-28 3/4.9 REFUELING OPERATIONS
-3/4.9.1 REACTOR MODE SWITCH..................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION......................................... 3/4 9-3 j 3/4.9.3 CONTROL R00 P0SITION.................................... 3/4 9-5 i
3/4.9.4 DECAY TIME.............................................. 3/4 9-6 1 3/4.9.5 -COMMUNICATIONS.......................................... 3/4 9-7 [ 3/4.9.6 REFUELING PLATF0RM...................................... 3/4 9-8 i 3/4.9.7 CRANE TRAVEL - SPENT-FUEL STORAGE P00L.................. 3/4 9-10 1 3/4.9.8 WATER LEVEL - REACTOR VESSEL............................ 3/4 9-11 F 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L................... 3/4 9-12 i, Ns . r/ I
- . LIMERICK - UNIT 1 xv i
. r r. - - , - - - -.-v 4 .--%--m._-,. - - . - , - - , - _ . , . . . - . . . . . . . . . - . . . . , _ ~ . . . . - _ - , _ . _ , - . . , . . . _ . . . , . , , . _ . . , - - . . - _ , - - . . -
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.10 CONTROL R0D REMOVAL Single Control Rod Removal........... .................. 3/4 9-13 Multiple Control Rod Remova1............................ 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level........................................ 3/4 9-17 Low Water Level......................................< . 3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY........................... 3/4 10-1 3/4.10.2 R0D SEQUENCE CONTROL SYSTEM........................... . 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.......................... 3/4 10-3 3/4.10.4 R EC I RC U LATI O N L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 0XYGEN CONCENTRATION.............. ..................... 3/4 10-5 3/4.10.6 TRAINING STARTUPS....................................... 3/4 10-6 3/4... RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration........................................... 3/4 11-1 Table 4.11.1.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program........................ 3/4 11-2 Dose.................................................... 3/4 11-5 Liquid Radwaste Treatment System... .................... 3/4 11-6 Liquid Holdup Tanks........ ........................... 3/4 11-7 3/4.11.2 GASE0US EFFLUENTS Dose Rate................. ..... ...... ...... . . ... 3/4 11-8 LIMERICK - UNIT 1 xvi
INDEX s_ - LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADIOACTIVE EFFLUENTS (Continued)- Table 4.11.2.1.2-1 Radioactive-Gaseous Waste Sampling and Analysis Program........................ 3/4 11-9 - Dose - Noble Gases...................................... 3/4 11-12 Do'se - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form..................... 3/4 11-13 P Ventilation Exhaust Treatment System.................... 3/4 11-14 i Explosive Gas Mixture................................... 3/4 11-15 Main Condenser.......................................... 3/4 11-16 Venting or Purging...................................... 3/4 11-17 3/4.11.3 SOLID RADWASTE TREATMENT................................ 3/4 11-18 - 3/4.11.4 TOTAL D0SE.............................................. 3/4 11-20 t 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM.......... ........................... 3/4 12-1 Table 3.12.1-1 Radiological Environmental Monitoring Program................. 3/4 12-3 Table 3.12.1-2 Reporting Levels For Radio-activity Concentrations In Environmental Samples.............. 3/4 12-9 Table 4.12.1-1 Detection Capabilities For Environmental Sample Analysis...... 3/4 12-10 3/4.12.2 LAND USE CENSUS......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... 3/4 12-14 ;
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LIMERICK - UNIT 1 xvii
INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY............ ............. ............ .... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTOOWN MARGIN......... ........ ....... ........ . B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALI ES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-1 3/4.1.3 CO NT RO L R0 D S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1- 2 3/4.1.4 CONTROL R0D PROGRAM CONTR0LS.................... ..... B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.. . ................. . B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE... .... ...... ...... .......................... B 3/4 2-1 Bases Table B 3/4 2.1-1 Significant Input Para-meters to the loss-Of-Cooling Accident Analysis.... ........... B 3/4 2-3 3/4.2.2 APRM SETPOINTS.. ................... ........ . ..... B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0.......................... B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...... .... . B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION..... ........... . B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION............ .......................... B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION..... B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION....................................... B 3/4 3-4 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION... ................. B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation............ .... B 3/4 3-4 LIMERICK - UNIT 1 xviii
t
/g INDEX
( BASES SECTION PAGE k INSTRUMENTATION (Continued) Seismic Monitoring Instrumentation...................... B 3/4 3-4 Meteorological Monitoring Instrumentation............... B 3/4 3-4 Remote Shutdown System Instrumentation and Controls..... B 3/4 3-5 Accident Monitoring Instrumentation..................... B 3/4 3-5 Source Range Monitors................................... B 3/4 3-5 Traversing In-Core Probe System......................... B 3/4 3-5 Chlorine and Toxic Gas Detection Systems................ B 3/4 3-6 l
- Fire Detection Instrumentation.......................... B 3/4 3-6 1
Loose-Part Detection System............................. B 3/4 3-6 l[ Radioactive Liquid Effluent F.onitoring Instrumentation......................................... B 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instrumentation......................................... B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM..................... B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION , INSTRUMENTATION......................................... B 3/4 3-7 4 Bases Figure B 3/4.3-1 Reactor Vessel Water ! Leve1...................... B 3/4 3-8 3 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM.................................... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................... B 3/4 4-3
- Operational Leakage..................................... B 3/4 4-3 3/4.4.4 CHEMISTRY............................................... B 3/4 4-3
( LIMERICK - UNIT 1 xix
INDEX e BASES SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY....................................... B 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS............................. B 3/4 4-4 Bases Table B 3/4.4.6-1 Reactor Vessel Toughness................. B 3/4 4-7 Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) At 1/4 T As A Function of Service Life...................... B 3/4 4-8 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................ B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY.................................... B 3/4 4-6 3/4.4.9 RE SI DUA L HE AT R EM0VA L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN............ B 3/4 5-1 3/4.5.3 SUPPRESSION CHAMBER................................ B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity...................... B 3/4 6-1 Primary Containment Leakage........................ B 3/4 6-1 Primary Contai nment Ai r Locks. . . . . . . . . . . . . . . . . . . . . . B 3/4 6-1 MSIV Leakage Control System........................ B 3/4 6-1 Primary Containment Structural Integrity. . . . . . . . . . . B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure......................................... B 3/4 6-2 Drywell Average Air Temperature. . . . . . . . . . . . . . . . . . . . B 3/4 6-2 Drywell and Suppression Chamber Purge System....... B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS........................... B 3/4 6-3 LIMERICK - UNIT 1 xx
fs INDEX
\s, / BASES SECTION PAGE CONTAINMENT SYSTEMS (Continued) 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES............... B 3/4 6-4 3/4.6.4 VACUUM RELIEF...................................... B 3/4 6-4 3/4.6.5 SEC0hDARY CONTAINMENT.............................. B 3/4 6-5 3/4.6.6' PRIMARY CONTAINMENT ATMOSPHERE CONTROL............. B 3/4 6-5 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS.............................. B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM..... B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM.............. B 3/4 7-1 3/4.7.4 SNUBBERS........................................... B 3/4 7-2
, 3/4.7.5 SEALED SOURCE CONTAMINATION........................ B 3/4 7-3 3/4.7.6 FIRE SUPPRESSION SYSTEMS........................ .. B 3/4 7-4 3/4.7.7 FIRE RATED ASSEMBLIES.............................. B 3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS............................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES............ B 3/4 8-3 i 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH................................ B 3/4 9-1
- 3/4.9.2 INSTRUMENTATION.................................... B 3/4 9-1 3/4.9.3 CONTROL R00 P0SITION............................... B 3/4 9-1 i
3/4.9.4 DECAY TIME......................................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS..................................... B 3/4 9-1 l g'e
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t LIMERICK - UNIT 1 xxi
INDEX BASES SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING PLATF0RM................................ B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L............ B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LE VEL - SPENT FUEL STORAGE P00L. . . . . . . . B 3/4 9-2 3/4.9.10 CONTROL R00 REM 0 VAL............................... B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION..... B 3/4 9-2 3/4.10 SPECIAL TEST EXCEFTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY..................... B 3/4 10-1 3/4.10.2 R0D SEQUENCE CONTROL SYSTEM....................... B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.................... B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS............................... B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION.................... ......... B 3/4 10-1 3/4.10.6 TRAINING STARTUPS................................. B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration..................................... B 3/4 11-1 Dose.............................................. B 3/4 11-1 Liquid Radwaste Treatment System................. B 3/4 11-2 Liquid Holdup Tanks........................... ... B 3/4 11-2 3/4.11.2 GASE0US EFFLUENTS Dose Rate........................... ......... ... B 3/4 11-2 Dose - Noble Gases................................ B 3/4 11-3 Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form............. ... B 3/4 11-3 Ventilation Exhaust Treatment System....... ...... B 3/4 11-4 LIMERICK - UNIT 1 xxii
- m. _ - . . . _. ._.. _._-. .-_. . _. - . _ -_ _ _ _ _ _ _ _ . _ . _ _ . _ . _
a INDEX BASES SECTION PAGE i RADI0 ACTIVE EFFLUENTS (Continued) Explosive Gas Mixture.............................. B 3/4 11-4 l Main Condenser..................................... B 3/4 11-5 i Venting or Purging................................. B 3/4 11-5 ; i l 3/4.11.3 SOLID RADWASTE TREATMENT........................... B 3/4 11-5 , 3/4.11.4 TOTAL 00SE......................................... B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l 3/4.12.1 MONITORING PR0 GRAM................................. B 3/4 12-1 i j 3/4.12.2 LAND USE CENSUS.................................... B 3/4 12-1 i ! 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM................. B 3/4 12-2 i I i i i i i t. l i t 4 l J l I l LIMERICK - UNIT 1- xxiii i w- ,,-ww.. -w,r --~m r,-m,, . m e. ,,,, e ww we m n .,.vm , m-mm,
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area............................................... 5-1 Figure 5.1.1-1 Exclusion Area.......................... 5-2 Low Population Zone..... ................................... 5-1 Figure 5.1.2-1 Low Population Zone..................... 5-3 Maps Defining UNRESTRICTED AREAS and SITE BOUNDARY for Radioactive Gaseous and Liquid Effluents............ 5-1 Figure 5.1.3-la Map Defining UNRESTRICTED AREAS for , Radioactive Gaseous and Liquid Effluents.............................. 5-4 Figure 5.1.3-lb Map Defining UHRESTRICTED AREAS for Radioactive Gaseous and Liquid Effluents.............................. 5-5 Meteorological Tower Location................................ 5-1 Figure 5.1.4-1 Meteorological Tower Location........... 5-6 5.2 CONTAINMENT Configuration................................................ 5-1 Design Temperature and Pressure.............................. 5-1 Secondary Containment........................................ 5-7 5.3 REACTOR CORE Fuel Assemblies.............................................. 5-7 Control Rod Assemblies....................................... 5-7 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature.............................. 5-7 Vo1ume....................................................... 5-8 5.5 FUEL STORAGE Criticality............................................... .. 5-8 LIMERICK - UNIT 1 xxiv
t i i i i j j INDEX i DESIGN FEATURES - i SECTION PAGE i 1 j FUEL STORAGE (Continued) ; i j' Drainage..................................................... 5-8 i ! Capacity..................................................... 5-8 1 5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT.......................... 5-8 ,! Table 5.6.1-1 Component Cyclic or Transient Limits..... 5-9 i .I . i k' . 4 i i i f t I i . l l 1 I . i i }. s i ! i
- LIMERICK - UNIT 1 xxv >
1 : t
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY........................ . ................. .. 6-1 6.2 ORGANIZATION............................ .................... 6-1 6.2.1 0ffsite................................................. 6-1 Figure 6.2.1-1 Offsite Organization...... ........ 6-3 6.2.2 Unit Staff.............................................. 6-1 Figure 6.2.2-1 Organization for Conduct of Plant Operations................... 6-4 Table 6.2.2-1 Minimum Shift Crew Composition......................... 6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP Function .............................................. 6-6 Composition.............................. ............. 6-6 Responsibilities........................ .............. 6-6 Records... ............................................ 6-6 6.2.4 SHIFT TECHNICAL ADVISOR............ ....... ........... 6-6 6.3 UNIT STAFF QUALIFICATIONS.. ................................ 6-6 6.4 TRAINING................... ............... ................ 6-7 6.5 REVIEW AND AUDIT 6.5.1 Plant Operations Review Committee (PORC) Function ............. ................. .. ....... . 6-7 Composition ............ ............................ . 6-7 Alternates.......... .. ............ .... .. ... .. .. 6-7 Meeting Frequency ............................... ..... 6-7 Quorum............... . . ....... ........... .. . .... 6-7 Responsibilities ....... .......... ....... ......... 6-8 Records.............. ....................... ......... 6-9 LIMERICK - UNIT 1 xxvi
,m INDEX t
i; ij ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 NUCLEAR REVIEW BOARD (NRB) Function .............................................. 6-9 Composition ........................ .................. 6-9 Alternates............................................. 6-10 Consultants............................................ 6-10 Meeting Frequency...................................... 6-10 Quorum................................................. 6-10 Review..................... ........................... 6-10 Audits................................................. 6-11 Records................................................ 6-12
,/N i 6.6 REPORTABLE EVENT ACTI0N..................................... 6-12 \,
l 6.7 SAFETY LIMIT VIOLATION...................................... 6-12 6.8 PROCEDURES AND PR0 GRAMS..................................... 6-13
- 6. 9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ....................................... 6-15 Startup Report......................................... 6-15 Annual Reports ........................................ 6-15 Monthly Operating Reports.............................. 6-16 Annual Radiological Environmental Operating Report..... 6-16 Semiannual Radioactive Effluent Release Report......... 6-17 4
6.9.2 SPECIAL REP 0RTS........................................ 6-18 6.10 RECORD RETENTION........................................... 6-19 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-20 O l 1 f 6.12 HIGH RADIATION AREA........................................ 6-20 V i LIMERICK - UNIT 1 xxvii
INDEX ADMINI Shn . . . e CONTROLS SECTION PAGE 6.13 PROCESS CONTROL PROGRAM (PCP).............................. 6-21 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)..................... 6-22 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS....... 6-22 O O LIMERICK - UNIT 1 xxviii
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4 1 0 I l l SECTION 1.0 < DEFINITIONS i 1 i i i i " 5 l , l > ! i i : ( i i i l 1 i 9
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1 i n 1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these j specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications. ACTION l
! 1.1 ACTION shall be that part of a Specification which prescribes remedial 1
measures required under designated conditions. AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the
; specified bundle at the specified height divided by the number of fuel rods in the fuel bundle. '
a AVERAGE PLANAR LINEAR HEAT GENERATION RATE i 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable } to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle. , CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or i total channel steps such that the entire channel is calibrated. i CHANNEL CHECK l
- 1. 5 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.
CHANNEL FUNCTIONAL TEST l 1.6 A CHANNEL FUNCTIONAL TEST shall be: l a. Analog channels - the injection of a simulated signal into the channel 4 as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.
- b. Bistable channels - the injection.of a simulated signal into the sensor j to verify OPERABILITY including alarm and/or trip functions.
The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested. i LIMERICK - UNIT 1 1-1 4
-.---_-.-r__.,__--.. - , - , , - _ - , , , , , . , - - - . , - - , - ~ - - - _ , , . - - - . - , _ - , - . - - . , . - - - -,m ---w,,,,---
DEFINITIONS CORE ALTERATION
- 1. 7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. DOSE EQUIVALENT I-131
- 1. 9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant. EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS 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. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.12 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:
- a. Turbine stop valves, and
- b. Turbine control valves.
LIMERICK - UNIT 1 1-2
1 DEFINITIONS k) This total system response time consists of two components, the instrumen-tation response time and the breaker arc suppression time. These times may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. , FRACTION OF LIMITING POWER DENSITY 1.13 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the specified LHGR limit for ,.c that bundle type. FRACTION OF RATED THERMAL POWER 1.14 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER. FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:
- a. Leakage into collection systems, such as pump seal or valve packing
( leaks, that is 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 interftre with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.
ISOLATION SYSTEM RESPONSE TIME 1.17 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. l Times shall include diesel generator starting and sequence loading delays
- where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
LIMITING CONTROL R00 PATTERN 1.18 A LIMITING CONTROL RCD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR. LINEAR HEAT GENERATION RATE 1.19 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit . \[ length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length. LIMERICK - UNIT 1 1-3
DEFINITIONS LOGIC SYSTEM FUNCTIONAL TEST 1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested. MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.21 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the highest value of the FLPD which exists in the core. MEMBER (S) 0F THE PUBLIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recrea-tional, occupational, or other purposes not associated with the plant. MINIMUM CRITICAL POWER RATIO 1.23 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core (for each class of fuel). OFFSITE DOSE CALCULATION MANUAL 1.24 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the current 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 environmental radiological monitoring program. OPERABLE - OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, cubsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL CONDITION - CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2. PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. LIMERICK - UNIT 1 1-4
i i i DEFINITIONS !g\ \vl PRESSURE B0UNDARY LEAKAGE 1.28 PRESSURE B0UNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall. PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:
- a. All primary containment penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.
- b. All primary containment equipment hatches are closed and sealed.
- c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.
- d. The primary. containment leakage rates are within the limits of Specification 3.6.1.2.
- e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
(g f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or 0 rings, is OPERABLE. PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the SOLIDIFICATION or dewatering and packaging of radioactive wastes results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for trans-portation to the disposal site and receipt at the disposal site. With SOLIDIFICATION, the PCP shall identify the process parameters influencing SOLIDIFICATION such as pH, oil content,2 H 0 content, solids content ratio of solidification agent to waste and/or necessary additives for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each waste type, based on laboratory scale and full scale testing or experience. With dewatering, the PCP shall include an identification of conditions that must be satisfied, based on full scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges.will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of the low-level radioactive waste disposal site. PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, h (j concentration or cther operating condition, in such a manner that replacement air or gas is required to purify the confinement. LIMERICK - UNIT 1 1-5
DEFINITIONS RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3293 MWt. REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:
- a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.5.2.1-1 of Specification 3.6.5.2.1.
- b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.
- c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
- d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.
- e. At least one door in each access to the reactor enclosure secondary containment is closed.
- f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
- g. The pressure withir the reactor enclosure secondary containment is less than or equal to the value required by Specific nion 4.6.5.1.la.
REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until do-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:
- a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.5.2.2-1 of Specification 3.6.5.2.2.
LIMERICK - UNIT 1 1-6 &
Q DEFINITIONS
'~
REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)
- b. All refueling floor secondary containment hatches and blowout panels are closed and sealed.
- c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
- d. At least one door in each access to the refueling floor secondary containment is closed.
- e. The sealing mechanism associated with each refueling floor secondary
, containment penetration, e.g. , welds, bellows, or 0-rings, is OPERABLE.
- f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.
! REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. ROD DENSITY 1.37 R0D DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% R0D DENSITY. V SHUTDOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown cor.11 tion; cold, i.e. 68 F; and xenon free. SITE BOUNDARY 1.39 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-1.a.
- SOLIDIFICATION 1.40 SOLIDIFICATION shall be the immobilization of wet radioactive wastes such as evaporator bottoms, spent resins, sludges, and reverse osmosis concen-trates as a result of a process of thoroughly mixing the waste type with a solidification agent (s) to form a free standing monolith with chemical and physical characteristics specified in the PROCESS CONTROL PROGRAM (PCP).
SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. A (Jl LIMERICK - UNIT 1 1-7
DEFINITIONS STAGGERED TEST BASIS 1.42 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.
- b. The testing of one system, subsystem, train, or other designated component at the beginaing of each subinterval.
THERMAL POWER 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reacto'r coolant. UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE. UNRESTRICTED AREA 1.45 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protec-tion of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. VENTILATION EXHAUST TREATMENT SYSTEM 1.46 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING 1.47 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. O\ LIMERICK - UNIT 1 1-8 l l
1 4 i i DEFINITIONS l TABLE 1.1 ) l SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY j 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. A At least once per 366 days. i R At least once per 18 months (550 days). S/U Prior to each reactor startup. P Prior to each radioactive release. N . I. . Not applicable. , i 1 4 b I \ LIMERICK - UNIT 1 1-9 w - -- w ---- - ----- - - --
DEFINITIONS TABLE 1.2 OPERATIONAL CONDITIONS NODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE
- 1. POWER OPERATION Run Any temperature
- 2. STARTUP Startup/ Hot Standby Any temperature
- 3. HOT SHUTDOWN Shutdown # *** > 200 F
- 4. COLD SHUTOOWN Shutdown # ## *** 1 200 F
- 5. REFUELING
- Shutdown or Refuel ** # $ 140 F O
#The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff. ##The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed. I
**See Special Test Exceptions 3.10.1 and 3.10.3.
- The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one-rod-out interlock is OPERABLE.
LIMERICK - UNIT 1 1-10 1
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i 4 4 l 4 ! SECTION 2.0 SAFETY LIMITS i AND I l LIMITING SAFETY SYSTEM SETTINGS i i l l L b t i (
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\s 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of' Specification 6.7.1. THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greate! than 10% of rated flow. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With MCPR less than 1.06 and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4. ACTION: With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours and comply with the requirements of Specification 6.7.1. b LIMERICK - UNIT 1 2-1
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued) REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel. APPLICABILITY: OPERATIONAL CONDITIONS 3, 4, and 5 ACTION: With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required. Comply with the requirements of Specification 6.7.1. O O LIMERICK - UNIT 1 2-2
__...__m- . _ _ _ SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
} l v
2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1. APPLICABILITY: As shown in Table 3.3.1-1. ACTION: With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply.the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value. [ ^ ( O G LIMERICK - UNIT 1 2-3
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS g , ALLOWABLE
;; FUNCTIONAL UNIT TRIP SETPOINT VALUES 7 1. Intermediate Range Monitor, Neutron Flux-High 5 120/125 divisions 5 122/125 divisions c of full scale of full scale z 2. Average Power Range Monitor:
[ a. Neutron Flux-Upscale, Setdown 5 15% of RATED THERMAL POWER $ 20% of RATED THERMAL POWER
- b. Neutron Flux-Upscale
- 1) Flow Biased -< 0.66 W+ 51%, with -< 0.66 W+ 54%, with a maximum of a maximum of
- 2) High Flow Clamped 5 116.5% of RATED 5 118.5% of RATED THERMAL POWER THERMAL POWER
- c. Inoperative N.A. N.A.
- d. Downscale -> 4% of RATED -> 3% of RATED y THERMAL POWER THERMAL POWER S 3. Reactor Vessel Steam Dome Pressure - High 5 1037 psig 5 1057 psig l 4. Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above instrument 2 11.0 inches above
- zero* instrument zero
( 5. Main Steam Line Isolation Valve - Closure 5 8% closed 5 12% closed
- 6. Main Steam Line Radiation - High 5 3.0 x full power background 5 3.6 x full power background
- 7. Drywell Pressure - High 5 1.68 psig 5 1.88 psig I 8. Scram Discharge Volume Water Level - High l a. Level Transmitter -< 260' 9 5/8" elevation ** < 261' 5 5/8" elevation l
- b. Float Switch 5 260' 9 5/8" elevation ** 5 261' 5 5/8" elevation l
- 9. Turbine Stop Valve - Closure 5 5% closed 5 7% closed
- 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 2 500 psig 2 465 psig
- 11. Reactor Mode Switch Shutdown Position N.A. N.A.
- 12. Manual Scram N.A. N.A.
*See Bases Figure B 3/4.3-1.
** Equivalent to 25.45 gallons / scram discharge volume.
O O O
1 1 i 1 i t ; q "c k i I 1 i i 1 i BASES i l FOR t t l SECTION 2.0 1 1 ! SAFETY LIMITS i AND
- LIMITING SAFETY SYSTEM SETTINGS
- l
} 4 i i i e i I I ,i i l 1 l
l 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 . =
f7 2.1 SAFETY LIMITS i \
' V BASES
2.0 INTRODUCTION
The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.06. MCPR greater than 1.06 represents a con-servative margin relative to the conditions required to maintain fuel cladding i ntegri ty. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While m fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations [V T signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signi-ficant departure from the condition intended by design for planned operation. 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10s lb/h, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lb/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. l p) l t
'O LIMERICK - UNIT 1 B 2-1 i I
1
1 I SAFETY LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The Safety Limit MCPR is determined using the General Electric Thermal a Analysis Basis, GETAB , which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L), (GEXL), correlation. The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation. The required input to the statistical model are the uncertainties listed in Bases Table 82.1.2-1 and the nominal values of the core parameters listed in Bases Table B2.1.2-2. The bases for the uncertainties in the core parameters are given in b NED0-20340 and the basis for the uncertainty in the GEXL correlation is given a in NED0-10958-A . The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.
- a. " General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application," NED0-10958-A.
- b. General Electric " Process Computer Performance Evaluation Accuracy" NED0-20340 and Amendment 1, NE00-20340-1 dated June 1974 and December 1974, respectively.
LIMERICK - UNIT 1 B 2-2
t BASES TABLE B2.1.2-1 , UNCERTAINTIES USED IN THE DETERMINATION OF'THE FUEL CLADDING SAFETY LIMIT
- STANDARD DEVIATION
' QUANTITY (% of Point)
Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow 2.5 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings 6.3 R Factor 1.5 Critical Power 3.6 i
- The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of_ quadrant power symmetry for the reactor core.
i i LIMERICK - UNIT 1 B 2-3
BASES TABLE B2.1.2-2 NOMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT THERMAL POWER 3323 N'4 Core Flow 108.5 M1b/h Dome Pressure 1010.4 psig Channel Flow Area 0.1089 ft2 R-Factor High enrichment - 1.043 Medium enrichment - 1.039 Low enrichment - 1.030 O l l l 9 l LIMERICK - UNIT 1 B 2-4 l l l
SAFETY. LIMITS v BASES 2.1.3 REACTOR COOLANT SYSTEM PRESSURE . The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code 1968 Edition, including Addenda through Summer 1969,-which permits a maximum pres-sure transient of 110%, 1375 psig, of design pressure 1250 psig. The Safety ' Limit of 1325 psig, as measured by the reactor vessel steam dome pressure , indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor and Pressure Vessel Code, 77tg lant system Edition, is designed including to the ASME Addenda through Boiler Summer 1977 for the reactor recirculation piping, which permits a maximum pressure transient
.of 110%, 1375 psig of design pressure, 1250 psig for suction piping and 1500 psig for discharge piping. The pressure Safety Limit is selected to be the '
lowest transient overpressure allowed by the ASME Boiler and Pressure Vessel Code Section_III, Class I. 2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor.is shutdown, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the-top of the active irradiated fuel during this period, the ability to remove decay heat is reduced.- This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action. O ! LIMERICK - UNIT 1 B 2-5
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
- 1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.
The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15.4 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal /gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal nf control rods in sequence and provides backup protection for the APRM.
- 2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform con-trol rod withdrawal is the most probable cause of significant power increase.
LIMERICK - UNIT 1 B 2-6
' -g [^N q LIMITING SAFETY SYSTEM SETTINGS v
BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0INTS (Continued) Average Power Range Monitor (Continued) Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to cnange power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in the Run position. The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-Upscale flow bias setpoint; i.e, for a pcwer increase, the THERMAL PCWER of the fuel will be less a than that indicatr.d by the neutron flux due to the time constants of the heat transfer associated with the fusl. The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown. The flow referenced trip setpoint must be adjusted by the specifisd formula in Specification 3.2.2 in order to maintain these margins when MFLPD is greater than or equal to FRTP.
- 3. Reactor Vessel Steam Dome Pressure-High
^
High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve
, and control fast closure trips are bypassed. For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
1 i v l LIMERICK - UNIT 1 8 2-7
}
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
- 4. Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
S. Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature, and low steam line pressure. The MSIVs closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
- 6. Main Steam Line Radiation-High The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected, a trip is initiated to reduce the continued failure of fuel cladding. At the same time the main steam line isolatinn valves are closed to limit the release of fission products. The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel cladding.
- 7. Drywell Pressare-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment. The trip setting was selected as low as possible without causing spurious trips.
O LIMERICK - UNIT 1 B 2-8
LIMITING SAFETY SYSTEM SETTING (m. (d/ BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
- 8. Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped. The trip setpoint for each scram discharge volume is equivalent to a contained volume of 25.45 gallons of water.
- 9. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst design basis transient.
p (Lj 10. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident failure of the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the Reactor Protection System. This trip setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve. Relevant transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis Report.
- 11. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.
- 12. Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
s V) LIMERICK - UNIT 1 B 2-9
k J f a l l i SECTIONS 3.0 and 4.0 l
- LIMITING CONDITIONS FOR OPERATION AND l.
- SURVEILLANCE REQUIREMENTS i
4 d 1 i l 1 r I i l l ) i 1 , 4 l
jQ 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliar:ce with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS 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 cet, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the Specification does not apply by placing it, as applicable, in:
- a. At least STARTUP 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 (m requirements, the ACTION may be taken in accordance with the specified time V) limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Speci-1 fications.
This Specification is not applicable in OPERATIONAL CONDITION 4 or 5. 3.0.4 Entry into an OPERATIONAL CONDITION 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 CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications. O). t
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LIMERICK - UNIT 1 3/4 0-1
APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be ~+. during the OPERATIONAL CONDITIONS or other conditions specified tc. " ividual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirements. 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 3 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 Specificatons. Surveillance requirements do not have to be per-formed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condi-tion shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, & 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 Coamission pursuant to 10 CFR Part 50, Section 50.55a(g) (6) (i).
- 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 Code and applicable Addenda for 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 every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days LIMERICK - UNIT 1 3/4 0-2
', r 2 L i [ APPLICABILITY t 6 F' . SURVEILLANCE REQUIREMENTS (Continued)' .
- c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing '
- i. activities.
!' d. Performance of the above inservice inspection and testing activities i shall be in addition to other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed j to supersede the requirements of any Technical Specification.
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i 1 1 6 6 { 4 ) ( l e s } i ' [ lC e 1 j LIMERICK - UNIT 1 3/4 0-3
, 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:
- a. 0.38% ak/k with the highest worth rod analytically determined, or l
- b. 0.28% Ak/k with the highest worth rod determined by test. '
1 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5. ACTION: With the SHUTDOWN MARGIN less than specified:
- a. In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours or be in at least HOT SHUTDOWN within the next 12 hours. -
- b. In OPERATIONAL CONDITION 3 or 4. immediately verify all insertable l N I
control rods to be, inserted and suspend all activities that could reduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4, establish SECONDARY CONTAINMENT INTEGRITY within 8 hours.
- c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS and other..
activities that could reauce the SHUTDOWN MARGIN and insert all
' insertable control rods within 1 hour. Establish SECONDARY CONTAIN-MENT INTEGRITY within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:
- a. By measurement, prior to or during the first startup after each refueling.
t b.' By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.
- c. Within 12 hours after detection of a withdrawn control rod that is immovable, as a result of excessive friction or mechanical inter .
ference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod. O LIMERICK - UNIT 1 3/4 1-1
REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES O', I LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity equivalence of the difference between the actual R0D DENSITY and the predicted R0D DENSITY shall not exceed 1% Ak/k. APPLICABILITY: OPERATIONAL CONDITION 1 and 2. ACTION: With the reactivity equivalence difference exceeding 1% Ak/k:
- a. Within 12 hours perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.
- b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENTS 4.1.2 The reactivity equivalence of the difference between the actual R0D DENSITY and the predicted R0D DENSITY shall be verified to be less than or equal to 1% Ak/k:
- a. During the first startup following CORE ALTERATIONS, and
- b. At least once per 31 effective full power days during POWER OPERATION.
O LIMERICK - UNIT 1 3/4 1-2
l 1 l f3 REACTIVITY CONTROL SYSTEMS
\ ) 3/4.1.3 CONTROL RODS CONTROL ROD OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
- a. With one control rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to be untrippable:
- 1. Within 1 hour:
a) Verify that the inoperable control rod, if withdrawn, is separated from all other inoperaole control rods by at least two control cells in all directions. b) Disarm the associated directional control valves ** either:
- 1) Electrically, or
- 2) Hydraulically by closing the dfive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours. p) ('~ 2. Restore the inoperable control rod to OPERABLE status within 48 hours or be in at least HOT SHUTOOWN within the next 12 hours,
- b. With one or more control rods trippable but inoperable for causes other than addressed in ACTION a, above:
- 1. If the inoperable control rod (s) is withdrawn, within 1 hour:
a) Verify that the inoperable withdrawn control rod (s) is separated from all other inoperable withdrawn control rods by at least two control cells in all directions, and b) Demonstrate the insertion capability of the inoperable with-drawn control rod (s) by inserting tha control rod (s) at least one notch by drive water pressure within the normal operating range *.
-Otherwise, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves ** either:
a) Electrically, oc b) Hydraulically by closing the drive water and exhaust water isolation valves.
*The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperable.
t **May be rearmed intermittently, under administrative control, to permit < V testing associated with restoring the control rod to OPERABLE status. LIMERICK - UNIT 1 3/4 1-3
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- 2. If the inoperable control rod (s) is inserted, within 1 hour disarm the associated directional control valves ** either:
a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- 3. The provisions of Specification 3.0.4 are not applicable.
- c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWN within 12 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
- a. At least once per 31 days verifying each valve to be open,* and
- b. At least once per 92 days cycling each valve through at least one complete cycle of full travel.
4.1.3.1.2 When above the preset power level of the RWM and RSCS, all withdrawn control rods not required to have their directional control valves disarmed electrica'.ly or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
- a. At least once per 7 days, and
- b. At least once per 24 hours when any control rod is immovable as a result of excessive friction or mechanical interference.
4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.6, and 4.1.3.7.
*These valves may be closed intermittently for testing under administrative controls.
- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
l LIMERICK - UNIT 1 3/4 1-4 o
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS'(Continued) 4.1.3.1.4 .The scram discharge volume shall be determined OPERABLE by demonstrating:
. a. The scram discharge volume drain and vent valves OPERABLE, when control rods are scram tested from a normal control rod configura- ~i tion of less than or equal to 50% R0D DENSITY at least once per 18 months, by verifying that the drain and vent valves:
- 1. Close within 30 seconds after receipt of a signal for control rods to scram, and
- 2. Open when the scram signal is reset.
- b. Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation at least once per 31 days.
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1 I 3 4 i 1 O LIMERICK - UNIT 1 3/4 1-5
- - . + + , - - - ~ - ,,,,,-w - v. m
l l REACTIVITY CONTROL SYSTEMS CONTROL R0D MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
- a. With the maximum scram insertion time of one or more control rods exceeding 7 seconds:
- 1. Declare the control rod (s) with the slow insertion time inoperable, and
- 2. Perform the Surveillance Requirements of Specification 4.1.3.2c.
at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds. Otherwise, be in at least HOT SHUTDOWN within 12 hours,
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:
- a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERAT. TONS or after a reactor shutdown that is greater than 120 days.
- b. For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and
- c. For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.
O LIMERICK - UNIT 1 3/4 1-6
REACTIVITY CONTROL SYSTEMS , s CONTROL R00 AVERAGE SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.3 The average scram insertion time of all OPERABLE control rods from the fully withdrawn position, based on deenergization of=the scram pilot valve solenoids as time zero, shall no+ exceed any of the following: Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With the average scram insertion time exceeding any of the above limits, be in at least HOT SHUTDOWN within 12 hours. SURVEILLANCE REQUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2. 1 1-Ns _. - , LIMERICK - UNIT 1 3/4 1-7 ; l l
REACTIVITY CONTROL SYSTEMS FOUR CONTROL R00 GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve sole-noids as time zero, shall not exceed any of the following: Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.45 39 0.92 25 2.05 5 3.70 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
- a. With the average scram insertion times of control rods exceeding the above limits:
- 1. Declare the control rods with the slower than average scram insertion times inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and
- 2. Perform the Surveillance Requirements of Specification 4.1.3.2c.
at least once per 60 days when operation is continued with an average scram insertion time (s) in excess of the average scram insertion time limit. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2. O LIMERICK - UNIT 1 3/4 1-8
- - ~
REACTIVITY CONTROL SYSTEMS ~ V CONTROL R0D SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*. ACTION:
- a. In OPERATIONAL CONDITION 1 or 2:
- 1. With r,ne control rod scram accumulator inoperable, within 8 hours:
a) Restore the inoperable accumulator to OPERABLE status, or b) Declare the control rod associated with the inoperable accumulator inoperable. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- 2. With more than one control rod scram accumulator inoperable, declare the associated control rods inoperable and:
a) If the control rod associated with any inoperable scram accumulator is withdrawn, immediately verify that at least one control rod drive pump is operating by inserting at least one withdrawn control rod at least one notch or place the reactor mode switch in the Shutdown position. [ } b) Insert the inoperable control rods and disarm the associated C/ control valves ' ther:
- 1) Electrical:y, or
- 2) Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within 12 hours.
- b. In OPERATIONAL CONDITION 5*:
- 1. With one withdrawn control rod with its associated scram accumulator inoperable, insert the affected control rod and disarm the associated directional control valves within one hour, either:
a) Electrically, or b) Hydraulically by closihy the t/Tip water and exhaust W4f hr isolation valves.
- 2. With more than one withdrawn cohl(9) rod with the qssocjated scram accumulator inoperable or 'fo/ Li!Di t o) rod di f /8 DWP oror-ating, immediately place the reactor mode PdIWl 14 Om $hutag position.
Wo pi wit tre f m ni f h at inn 1. O i .n e not and te +1e.
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dpplitablf i. t z i 1- > jM por g6 11itation3.9.10.fdr3,d,ld2. l LIMIklU - UNIT 1 3/4 1-9 L_
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:
- a. At least c~ e per 7 days by verifying that the indicated pressure is greater than or equal to 955 psig unless the control rod is inserted and disarmed or scrammed.
- b. At least once per 18 months by:
- 1. Performance of a:
a) CHANNEL FUNCTIONAL TEST of the leak detectors, and b) CHANNEL CALIBRATION of the pressure detectors, and verifying an alarm setpoint of 970 + 15, psig on decreasing pressure.
- 2. Measuring and recording the time for up to 10 minutes that each individual accumulator check valve maintains the associated accumulator pressure above the alarm set point with no control rod drive pump operating.
O 1 t 9 LIMERICK - UNIT 1 'c 1-10
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y REACTIVITY CONTROL SYSTEMS 1
\s_ -). CONTROL R0D DRIVE COUPLING LIMITING CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*. ACTION:
- a. In OPERATIONAL CONDITIONS 1 and 2 with one control rod not coupled to its associated drive mechanism, within 2 hours:
- 1. If permitted by the RWM and RSCS, insert the control red drive mechanism to accomplish recoupling and verify recoupling by with-drawing the control rod, and:
a) Observing any ir:d!cated respor a of the nuclear instrumenta-tion, and . I b) Demonstrating ita: ~"e control red wil? not go to the over- t travel positio: Otherwise, be in at l ea r. ' SHUTDC -ittin the next 12 hours.
- 2. If recoupling is not accon Tished en ne first attempt or, if not permitted by the RWM c r RSCS, then until permitte - by the RWM and RSC5, declare the contro' od inoperable, : sert the O
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control rcd and disarm t? < asso
+ h e r--
9d directional control valves ** U E3ectrically, L Hydraulically t closing ~2e di s c fer and er ust water isolation vale , Ot wrwise, be i- ieast HOT SHiTuCWN within the n' ' 12 hours. In OPERis i a NAL C0h01 ~ '4 5* witr hdrawn control rod n- coupled to its associated ati, ~ochanist thin 2 hours either:
- 1. Insert the contro: to accm; 'ish recoupling and verify recoup-ling by withdrawirm control oI and demonstrating that the l control rad will not to the overtrayel position, or
, If recoupi ng is amplished, inself (fx
- rod and disarm the associutt. :rectional contr61 ,alte
- eithert a) Electrically, or b) Hydraulically by cle og the W water and exilaust water isolation valves.
- c. The provisions of Spec 'ication 3.0.4 are not applicable.
-~s 'r t least e ich withdrawn con ' ul rec. hct applicable to control rods removed per Specif cation 3.9.10.1 ar 3.9.12.2.
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' '/ **May he rearmed intermittently, under administrative control, to permit test ig associated with restoring the control rod to OPERABLE status.
LirLRICK - UNIT 1 3/4 1-11 w .
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS O 4.1.3.6 Each affected control rod shall be demonstrated to be coupled to its drive mechanism by observing any indicated response of the nuclear instrumen-tation while withdrawing the control rod to the fully withdrawn position and then verifying that the control rod drive does not go to the overtravel position:
- a. Prior to reactor criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity,
- b. Anytime the control rod is withdrawn to the " Full out" position in subsequent operation, and
- c. Following maintenance on or modification to the control rod or control rod drive system which could have affected the control rod drive coupling integrity.
O O LIMERICK - UNIT 1 3/4 1-12
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REACTIVITY CONTROL SYSTEMS v CONTROL R0D POSITION INDICATION LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod position indication system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*. ACTION:
- a. In OPERATIONAL CONDITION 1 or 2 with one or more control rod position indicators inoperable, within 1 hour:
- 1. Determine the position of the control rod by utilizing the RSCS substitute position display (within preset power level), or:
a) Moving the control rod, by single notch movement, to a position with an OPERABLE position indicator, b) Returning the control rod, by single notch movement, to its original position, and c) Verifying no control rod drift alarm at least once per 12 hours, or D N 2. Move the control rod to a position with an OPERABLE position
) indicator, or
- 3. When THERMAL POWER is:
a) Within the preset power level of the RSCS, declare the control rod inoperable. b) Greater than the preset power level of the RSCS, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves ** either:
- 1) Electrically, or
- 2) Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTOOWN within the next 12 hours.
- b. In OPERATIONAL CONDITION 5* with a withdrawn control rod position indicator inoperable, move the control rod to a position with an OPERABLE position indicator or insert the control rod,
- c. The provisions of Specification 3.0.4 are not applicable.
*At least each withdrawn control rod. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
Q **May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status. (v ) LIMERICK - UNIT 1 3/4 1-13
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying:
- a. At least once per 24 hours that the position of each contro' rod is indicated,
- b. That the indicated control rod position changes during the movement of the control rod drive when performing Surveillance Requirement 4.1.3.1.2, and
- c. That the control rod position indicator corresponds to the control rod position indicated by the " Full out" position indicator when performing Surveillance Requirement 4.1.3.6b.
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9 LIMERICK - UNIT 1 3/4 1-14 l
m ,. ._ REACTIVITY CONTROL SYSTEMS f- . (s CONTROL ROD DRIVE HOUSING SUPPORT s LIMITING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place. APPLICABILITYi OPERATIONAL CONDITIONS 1, 2 and 3. ACTION: With the control rod drive housing support not in place, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.1.3.8 The control rod drive housing support shall be verified to be in place by a visual inspection prior to startup any time it has been disassembled or when maintenance has been pet.* formed in the control rod drive housing support area.
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N LIMERICK - UNIT 1 3/4 1-15 L_
REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM) snuil he OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*' **, when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER, the minimum allowable preset power level. ACTION:
- a. With the RWM inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console. Otherwise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.4.1 The RWM shall be demonstrated OPERABLE:
- a. In OPERATIONAL CONDITION 2 within 8 hours prior to withdrawal of control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 1 hour after RM automatic initia-tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.
- b. In OPERATIONAL CONDITION 2 within 8 hours prior to withdrawal of control rods for the purpose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.
- c. In OPERATIONAL CONDITION 1 within 1 hour after RWM automatic initiation when reducing THERMAL POWER, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.
- d. By verifying that the control rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer.
*See Snecial Test Exception 3.10.2.
R* Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality. LIMERICK - UNIT 1 3/4 1-16
/~ REACTIVITY CONTROL SYSTEMS .( y R0D SEQUENCE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.4.2 The rod sequence control system (RSCS) shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*' **, when THERMAL POWER is less than or equal to 20% RATED THERMAL POWER, the minimum allowable preset power level. 1 ACTION:
- a. With the RSCS inoperable, control rod movement shall not be permitted, j except by a scram.
- b. With an inoperable control rod (s), OPERABLE control rod movement may continue by bypassing the inoperable control rod (s) in the RSCS provided that:
- 1. The position and bypassing of inoperable control rods is verified by a second licensed operator or other technically qualified member of the unit technical staff, and
) 2. There are not more than three inoperable control rods in any V RSCS group.
1 SURVEILLANCE REQUIREMENTS 4.1.4.2 The RSCS shall be demonstrated OPERABLE by:
- a. Performance of a system diagnostic function:
- 1. Within 8 hours prior to each reactor startup, and
- 2. Prior to movement of a control rod after rod inhibit mode i automatic initiation when reducing THERMAL POWER.
i b. Attempting to select and move an inhibited control rod:
- 1. After withdrawal of the first insequence control rod for each reactor startup, and
- 2. Within one hour after rod inhibit mode automatic initiation (preset power level) when reducing THERMAL POWER.
*See Special Test Exception 3.10.2 Oj t
** Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RSCS
\ prior to withdrawal of control rods for the purpose of bringing the reactor
- .to criticality.
LIMERICK - UNIT 1 3/4 1-17
REACTIVITY CONTROL SYSTEMS R00 BLOCK MONITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER. ACTION:
- a. With one RBM channel inoperable:
- 1. Verify that the reactor is not operating on a LIMITING CONTROL R0D PATTERN, and
- 2. Restore the inoperable RBM channel to OPERABLE status within 24 hours.
Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour.
- b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within 1 hour. '
SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a:
- a. CHANNEL FUNCTIONAL TEST and Cf'ANNEL CALIBRATION at the frequencies and for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1.
- b. CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CONTROL R0D PATTERN.
O LIMERICK - UNIT 1 3/4 1-18
REACTIVITY CONTROL SYSTEMS G 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE. I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*
. ACTION:
- a. In OPERATIONAL CONDITION 1 or 2:
- 1. With only one pump and corresponding explosive valve OPERABLE, i restore one inoperable pump and correspondng explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
- 2. With the standby liquid control system otherwise inoperable, restore the system to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours.
- b. Ir. OPERATIONAL CONDITION 5*:
- 1. With only one pump and corresponding explosive valve OPERABLE, 7sg restore one inoperable pump and corresponding explosive valve t
i to OPERABLE status within 30 days or insert all insertable
\s_ / control rods within the next hour.
- 2. With the standby liquid control system otherwise inoperable, insert all insertable control rods within 1 hour.
SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE: s a. At least once per 24 hours by verifying that; i
- 1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.
. 2. The available volume of sodium pentaborate solution is within i the limits of Figure 3.1.5-2.
- 3. The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping to be greater than or equal to 70 F.
i
*With any control rod withdrawn. Not applicable to control rods removed per.
7s s Specification 3.9.10.1 or 3.9.10.2. [v} LIMERICK - UNIT 1 3/4 1-19
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 9
- b. At least once per 31 days by:
- 1. Verifying the continuity of the explosive charge.
- 2. Determining that the availabl.e weight of sodium pentaborate is greater than or equal to 5500 lbs and the concentration of boron in solution is within the limits of Figure 3.1.5-2 by chemical analysis.*
- 3. 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. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a p. essure of greater than or equal to 1190 psig is met.
- d. At least once per 18 months during shutdown by:
- 1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch success-fully fired. All injection loops shall be tested in 36 months.
- 2. ** Demonstrating that all heat traced piping is unblocked by pumping from the storage tank to the test tank and then draining and flushing the piping with demineralized water.
- 3. Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized.
*This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below 70 F.
- This test shall also be performed whenever all three heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, overlapping or total flow path steps such that the entire flow path is included.
O LIMERICK - UNIT 1 3/4 1-20
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\ \s LIMERICK - UNIT 1 3/4 1-21
C E x g 35-- x 4 Z
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REGION OF APP'10VED g 14 -- 13.8.4820 VOLUME -CONCENTRATION 0
- i LOW LEVEL .
, /
/ MARGIN /, OVERFLOW
,9 VOLUME 13.4
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12.9,4860 A 5110 N MINIMUM REQUIRED 9 CONCENTRATION LINE S 8 12 I I I I l i l l l 1 I I I I I I 4 00 4500 4600 4700 4800 4900 5000 5100 5200 V- NET VOLUME (GALLONS) S0DIUM PENTABORATE SOLUTION VOLUME / CONCENTRATION REQUIREMENTS FIGURE 3.1.5-2 O __ O O_-_----
- 3/4.2 POWER DISTRIBUTION LIMITS
\
~J 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERA. ION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and 3.2.1-5.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: With an APLHGR exceeding the limits of Figure 3. 2.1- 1, 3. 2.1-2, 3. 2.1-3, 3. 2.1-4, or 3.2.1-5 initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. e SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and 3.2.1-5:
~
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL R0D PATTERN for APLHGR.
- d. The provisions of Specification 4.0.4 are not applicable.
(v) LIMERICK - UNIT 1 3/4 2-1
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3 u._ a - i- ! i , i i, ! i ., 0 5000 10,000 15,000 20,000 25.000 30,000 35,000 40,000 45,000 AVERAGE PLANAR EXPOSURE (MWD /T) MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE INITIAL CORE FUEL TYPES P8CIB071 FIGURE 3.2.1-5 O O O
POWER DISTRIBUTION LIMITS Q' \ 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-upscale scram trip setpoint (S) and and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships: TRIP SETPOINT ALLOWABLE VALUE l S < (0.66W + 51%)T S < (0.66W + 54%)T S RB 5 (0.66W + 42%)T S RB $ (0.66W + 45%)T where: S and S are in percent of RATED THERMAL POWER, RB W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY. T is applied only if less than or equal to 1.0. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: With the APRM flow biased neutron fiux-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or h. ( S as above determined, initiate corrective action within 15 minutes ab,adjustSand/ ors to be consistent with the Trip Setpoint values
- within6hoursorred0$eTHERMALPOWERtolessthan25%ofRATEDTHERMAL POWER within the next 4 hours. .
SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased neutron flux-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating with MFLPD greater than or equal to FRTP.
- d. The provisions of Specification 4.0.4 are not applicable.
*With MFLPD greater than the FRTP during power ascension up to 90% of RATED 4
THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be i adjusted such that the APRM readings are greater than or equal to 100% times p MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel. 1
%)
LIMERICK - UNIT 1 3/4 2-7 t
POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING-CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figure 3.2.a . +imes the Kf shown in Figure 3.2:3-2, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2, with:
; (Iave IB)
T -I A B where: TA = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, N i 1 T 31 (0.052), B = 0.688 + 1.65[ " N. I 1 i=1 I ave = i=1 "i i I n I N. I i=1 n = number of surveillance tests performed to date in cycle, N.I = number of active control rods measured in the i th surveillance test, 1 9 = average scram time to notch 39 of all rods measured in the i th surveillance test, and N = total number of active rods measured in Specification 1 4.1.3.2.a. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. O LIMERICK - UNIT 1 3/4 2-8
. -. .. - - =_.-
POWER DISTRIBUTION LIMITS
~
V LIMITING CONDITION FOR OPERATION (Continued) ACTION ,
- a. With the end-of-cycle recirculation pump trip system inoperable per
- Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within 1 hour, MCPR is determined to be greater than or equal to the MCPR limit as a function of the average scram time shown in Figure 3.2.3-1, E0C-RPT inoperable curve, times the K shown 7 in Figure 3.2.3-2.
- b. With MCPR less than the applicable MCPR limit shown in Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required. limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED. THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS , 4.2.3 MCPR, with:
- a. T = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or
- b. I as defined in Specification 3.2.3 used to determine the limit within 72 hours of the conclusion of each ssram time surveillance test required by Specification 4.1.3.2, shall-be determined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of I
at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is
- operating with a LIMITING CONTROL R0D PATTERN for MCPR.
I i- d. The provisions of Specification 4.0.4 are not applicable. LIMERICK - UNIT 1 3/4 2-9
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' ' ' "1"~' ' ' ' ' ' ' ' l' 1.20 1.20 O. O.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 7
MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS T AT RATED FLOW FIGURE 3.2.3-1 l i LIMERICK - UNIT 1 3/4 2-10 l
O O 'O 1.4 i C
~
1.3 x
- AUTOMATIC FLOW CONTROL w 1.1 1 MANUAL FLOW CONTROL e 4
SCOOP TUBE SETPOINT CAtlBRATION POSITIONED SO THAT FLOW MAXIMUM - 102.6% gj __. FLOW MAXIMUM - 107A% / v FLOW MAXilVIUM = 112.0% "/ FLOW MAXIMUM ,117.0% 0 30 40 60 60 70 80 90 100 Core Flow, % Of Rated Core Flow K FACTOR f FIGURE 3.2.3-2
POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kW/ft. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING CONTROL R0D PATTERN for LHGR.
3
- d. The provisions of Specification 4.0.4 are not applicable.
O LIMERICK - UNIT 1 3/4 2-12
. , ~~3 3/4.3 INSTRUMENTATION .I \ \s l 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION
~ LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2 APPLICABILITY: As shown in Table 3.3.1-1. ACTION:
- a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Syst em requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition
- within 1 hour. The provisions of Specification 3.0.4 are not applicable.
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within 1 hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS ('- Each reactor protection system instrumentation channel shall be
' 4.3.1.1 demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per trip system 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 system.
*An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.
**The trip system need not be placed in the ' tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both
('~') systems have the same number of inoperable channels, place either trip system
- \ ) in the tripped condition.
4 LIMERICK - UNIT 1 3/4 3-1
TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 5 n
' APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS E FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION w
w 1. Intermediate Range Monitors (b).
- a. Neutron Flux - High 2 3 1 3, 4 3 2 5(c) 3(d) 1
- b. Inoperative 2 3 1 ~
3, 4 3 2 5 3(d) 3 Average Power Range Monitorf *): { 2. Y a. Neutron Flux - Upscale, Setdown 2 2 1 3, 4 2 2 5(c) 2(d) 3 l b. Neutron Flux - Upscale i
- 1) Flow Biased 1 2 4
- 2) High Flow Clamped 1 2 4 l c. Inoperative 1, 2 2 1 3, 4 2 2 5(c) 2(d) 3
- d. Downscale 1(g) 2 4
- 3. Reactor Vessel Steam Dome Pressure - High 1, 2(f) 2 1
- 4. Reactor Vessel Water Level - Low, Level 3 1, 2 2 1
- 5. Main Steam Line Isolation Valve -
Closure 1(g) 1/ valve 4 O O O
L i O t j ._ TABLE 3.3.1-1 (Continued) 1 r-j REACTOR PROTECTION SYSTEM INSTRUMENTATION i 5
' /
I 1 APPLICABLE -MINIMUM i j OPERATIONAL OPERABLE CHANNELS , l E FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION
! H t - 6. Main Steam Line Rad tion - '
i High \ 1, 2(f) 2 5 i 1 7. Drywell Pressure - High 1, 2(h) 2 1 I
- 8. Scram Discharge Volume Water ,
i Level - High I i a. Level Transmitter 1, 2 2 1 i y 5(i) 2 3 Y
- b. Float Switch 1, 2 2 1 !
! 5 (i) 2 3 i l 9. Turbine Stop Valve - Closure 1(j) 4(k) 6
- i i 10. Turbine Control Valve Fast Closure, l
Trip Oil Pressure - Low 1(j) 2(k) 6 ,
- 11. Reactor Mode Switch Shutdown 1 Position 1, 2 2 1 s
3, 4 2 7 l 5 2 3 l 1 12. Manual Scram 1, 2 2 1 3, 4 2 8-
- 5 2 9 i
' TABLE 3. 3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 - Be in at least HOT SHUTOOWN within 12 hours. ACTION 2 - Verify all " ertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour. ACTION 3 - Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour. ACTION 4 - Be in at least STARTUP within 6 hours. ACTION 5 - Be in STARTUP with the main steam line isolation valves closed within 6 hours or in at least HOT SHUTDOWN within 12 hours. ACTION 6 - Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure until the function is automatically bypassed, within 2 hours. ACTION 7 - Verify all insertable control rods to be inserted within 1 hour. ACTION 8 - Lock the reactor mode switch in the Shutdown position within 1 hour. ACTION 9 - Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour. O LIMERICK - UNIT 1 3/4 3-4
TABLE 3.3.1-1 (Continued) V REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the trip system in the tripped ~ condition provided at least one OPERABLE channgl in the same trip system is monitoring that parameter. (b) This function shall be automatically bypassed when the reactor mode switch is in.the Run position and the associated APRM is not downscale. (c) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn
- and shutdown margin
-demonstrations performed per Specification 3.10.3.
(d) The noncoincident NMS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 IRMs and 2 SRMs. (e) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel. (f) This function is not required to be OPERABLE when the reactor pressure d vessel head is removed per Specification 3.10.1. (g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position. (h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required. (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. l (j) This function shall be automatically bypassed when turbine first stage i pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL l POWER. (k) Also actuates the E0C-RPT system. t
- *Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
<f C LIMERICK - UNIT 1 3/4 3-5
TABLE 3.3.1-2 h REACTOR PROTECTION SYSTEM RESPONSE TIMES 9_;, R RESPONSE TIME i FUNCTIONAL UNIT (Seconds) E 1. Intermediate Range Monitors:
- a. Neutron Flux - High N.A.
w
- b. Inoperative N.A.
l
- 2. Average Power Range Monitor *:
! a. Neutron Flux - Upscale, Setdown N.A.
- b. Neutron Flux - Upscale l 1) Flow Biased 50.09 l 2) High Flow Clamped 50.09 t' c. Inoperative N.A.
+
l y d. Downscale N.A.
- 3. Reactor Vessel Steam Dome Pressure - High 5 0.55
- 4. Reactor Vessel Water Level - Low, Level 3 $ 1.05 1
- 5. Main Steam Line Isolation Valve - Closure 5 0.06
- 6. Main Steam Line Radiation - High N.A.
- 7. Drywell Pressure - High N.A.
- 8. Scram Discharge Volume Water Level - High
- a. Level Transmitter N.A.
l b. Float Switch N.A.
- 9. Turbine Stop Valve - Closure 5 0.06
- 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low $ 0.08**
- 11. Reactor Mode Switch Shutdown Position N.A.
- 12. Manual Scram N.A.
*Heutron detectors are exempt from response time testing. Response time shall be measured from the detector output or from the input of the first electronic component in the channel.
** Measured from start of turbine control valve fast closure.
O O O
j TABLE 4.3.1.1-1 i C
$ REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
5 Q CHANNEL OPERATIONAL
, CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH l FUNCTIONAL UNIT CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED 4
U 1. Intermediate Range Monitors: . ]
- a. Neutron Flux - High S/U S(b) S/U(c), W R 2 i S W(j) R 3, Ir , 5 J t I b. Inoperative N.A. W(j) N.A. 2,3,4,5
- 2. Average Power Range Monitor (I):
! a. Neutron Flux - S/U,5(b) S/U(c),.W SA 2 ; Upscale, Setdown S W(j) SA 3, 5 I
m b. Neutron Flux - Upscale i } 1) Flow Biased S,D(g) S/U(c), W W(d)(e),SA, 1
$ 2) High Flow Clamped S S/U(c), W W(d)(e), SA 1
- c. Inoperative N.A. W(j) N.A. 1, 2, 3, 5
- d. Downscale S W SA 1
- 3. Reactor Vessel Steam Dome Pressure - High -S M R 1, 2(h) ,
- 4. Reactor Vessel Water Level -
) Low, Level 3 S M R 1, 2 l l 5. Main Steam Line Isolation Valve - Closure N.A. M R 1 1
- 6. Main Steam Line Radiation -
- High S M R 1, 2(h)
- 7. Drywell Pressure - High S M R 1, 2
TABLE 4.3.1.1-1 (Centinued) REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS h CHANNEL OPERATIONAL g CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH p; FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 7 8. Scram Discharge Volume Water e Level - High 5 a. Level Transmitter S M R 1, 2, 5(g)
- b. Float Switch N.A. M R 1, 2, 5(5)
]
.9. Turbine Stop Valve - Closure N.A. M R 1
- 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N. A. M R 1
- 11. Reactor Mode Switch Shutdown Position N.A. R 3. . A . 1,2,3,4,5
$ 12. Manual Scram N.A. M N.A. 1,2,3,4,5 [ (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. (b) The IRM and SRM channels shall be determined to overlap for at least decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least decades during each controlled shutdown, if not performed within the previous 7 days. (c) Within 24 hours prior to startup, if not performed within the previous 7 days. (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater tiian 2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference. (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal. (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system. (g) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1. (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. (j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position. 9 O O
Q INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in-Table 3.3.2-3. APPLICABILITY: As shown in Table 3.3.2-1. ACTION:
- a. With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition
- within 1 hour. The provisions of Specification 3.0.4 are not applicable.
v
- c. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within 1 hour and take the ACTION required by Table 3.3.2-1.
*An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken. **The trip system need not be placed in the' tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place'either trip system in the tripped condition.
( LIMERICK - UNIT 1 3/4 3-9
INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1. 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per trip system 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 isolation trip system. O O LIMERICK - UNIT 1 3/4 3-10
. [
4 TABLE 3.3.2-1 j cx , ISOLATION ACTUATION INSTRUMENTATION E MINIMUM APPLICABLE l [n ISOLATION OPERABLE CHANNELS OPERATIONAL 1 i TRIP FUNCTION SIGNAL-(a) PER TRIP SYSTEM (b) CONDITION ACTION i l 1. MAIN STEAM LINE ISOLATION I #
- a. Reactor Vessel Water Level
, 1) Low, Low-Level 2 B 2 1,2,3 21 , ! 2) Low, Low, Low-Level 1 C 2 1,2,3 21
- b. Main Steam Line -
4 Radiation - High D 2 1,2,3 21 ^! i 1
- c. Main Steam Line i l Pressure - Low P 2 1 22 w
A d. Main Steam Line ' l w Flow - High E 2/line 1, 2, 3 20 j
- e. Condenser Vacuum - Low Q 2 1, 2**, 3** 21 ,
1 , t j f. Main Steam Line Tunnel : j Temperature - High F(#) 6 1,2,3 21 l g. Turbine Enclosure - Main Steam i Line Tunnel Temperature - High F(f) 8 1,2,3 21 j ! h. Manual Initiation NA 2 1, 2, 3 24
- 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION !
i a. Reactor Vessel Water Level ; 1 Low - Level 3 A 2 1,2,3 23 ' i b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High V 2 1,2,3 23 1 c. Manual Initiation NA 1 1, 2, 3 24 ! i i ,- - . ____
TABLE 3.3.2-1 (C:ntinued) g ISOLATION ACTUATION INSTRUMENTATION 1 E MINIMUM APPLICABLE y ISOLATION OPERABLE CHANNELS OPERATIONAL i TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. RWCS A Flow - High J 1 1,2,3 23
- b. RWCS Area Temperature - High J 6 1,2,3 23
- c. RWCS Area Ventilation A Temperature - High J 6 1,2,3 23
- d. SLCS Initiation Y(d) NA 1,2,3 23 w e. Reactor Vessel Water Level -
A Low, Low - Level 2 B 2 1,2,3 23
- f. Manual Initiation NA 1 1,2,3 24
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
- a. HPCI Steam Line A Pressure - High L 1 1,2,3 23
- b. HPCI Steam Supply Pressure - Low LA 2 1,2,3 23
- c. HPCI Turbine Exhaust Diaphragm Pressure - High L 2 1,2,3 23
- d. HPCI Equipment Room Temperature - High L 1 1,2,3 23
- e. HPCI Equipment Room A Temperature - High L 1 1,2,3 23 O O O
,m ,- ,.
I \,_)
\J J TABLE 3.3.2-1 (Continued)
C ISOLATION ACTUATION INSTRUMENTATION in 5 MINIMUM APPLICABLE ISOLATION
- OPERABLE CHANNELS OPERATIONAL 7 TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION E 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION (Continued)
- f. HPCI Pipe Routing Area Temperature - High L 4 1,2,3 23
- g. Manual Initiation NA(e) 1/ system 1, 2, 3 24
- h. HPCI Steam Line a Press Timer NA 1 1,2,3 23
- 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
- a. Reactor Steam Line
{ A Pressure - High K 1 1,2,3 23
- b. RCIC Steam Supply Pressure - Low KA 2 1,2,3 23
- c. RCIC- Turbine Exhaust Diaphragm Pressure - High K 2 1,2,3 23
- d. RCIC Equipment Room Temperature - High K 1 1,2,3 23
- e. RCIC Equipment Room a Temperature - High K 1 1,2,3 23
- f. RCIC Pipe Routing Area Temperature - High K 5 1,2,3 23
- g. Manual Initiation NA(e) 1/ system 1,2,3 24
- h. RCIC Steam L'ine A Pressure Timer NA 1 1,2,3 23
C M TABLE 3.3.2-1 (Continued) 5 ISOLATION ACTUATION INSTRUMENTATION 9
, MINIMUM APPLICABLE c
= ISOLATIg) OPERABLECHANNEg) OPERATIONAL TRIP FUNCTION SIGNAL PER TRIP SYSTEM CONDITION ACTION ~
- 6. PRIMARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Low - Level 2 B 2 1,2,3 20
- 2) Low, Low, Low - Level 1 C 2 1,2,3 20
- b. Drywell Pressure - High H 2 1,2,3 20
- c. North Stack Effluent Radiation - High (9) W 1 1,2,3 23 R d. Refueling Area Ventilation
- Exhaust Duct-Radiation - High R 2 1,2,3 23
- e. Reactor Enclosure Ventilation Exhaust Duct-Radiation - High S 2 1,2,3 U
- f. Outside Atmosphere to Reactor Enclosure a Pressure - Low U 1 1,2,3 23
- g. Outside Atmosphere To Refueling Area a Pressure - Low T 1 1,2,3 23
- h. Drywell Pressure - High/
Reactor Pressure - Low G 2/2 1,2,3 26
- i. Primary Containment Instrument M 1 1,2,3 26 Gas Line to Drywell a Pressure-Low
- j. Manual Initiation NA 1 1,2,3 24 O O O
~s 7-N. .s C
pi TABLE 3.3.2-1.(Continued) 51 . ISOLATION ACTUATION INSTRUMENTATION 9 APPLICABLE MINIMUM j c- ISOLATIgt) OPERABLE CHANNE OPERATIONAL E TRIP FUNCTION SIGNAL PERTRIPSYSTEM(g) CONDITION ACTION e
- 7. SECONDARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level Low, Low - Level 2 B 2 1,2,3 25
- b. Drywell Pressure - High H 2 1,2,3 25
- c. Refueling Area Ventilation Exhaust Duct Radiation - High R 2 25
- d. Reactor Enclosure Ventilation Exhaust u, Duct Radiation - High S 2 1,2,3 25 '
1 u, e. Outside Atmosphere To Reactor
- 1. Enclosure A Pressure - Low U 1 1,2,3 25 m
- f. Outside Atmosphere To Refueling Area A Pressure - Low T 1 25
- g. Manual Initiation NA 1 1, 2, 3, and
- 24
c i TABLE 3.3.2-1 (Continued) ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 - Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. ACTION 21- Be in at least STARTUP with the associated isolation valves closed within 6 hours or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. ACTION 22 - Be in at least STARTUP within 6 hours. ACTION 23 - In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within 1 hour and declare the affected system inoperable. In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours. ACTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour. ACTION 26 - Close the affected system isolation valves within 1 hour. TABLE NOTATIONS When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel. May be bypassed under administrative control, with all turbine stop valves closed. (a) See Specification 3.6.3, Table 3.6.3-1 for valves which are actuated by these isolation signals. (b) A channel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the channel or trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter. In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours for required surveillance without placing the channel or trip system in the tripped condition. (c) Actuates valves shown in Table 3.6.5.2.1-1 and/or 3.6.5.2.2-1 and signals B, H, S, and U also start the standby gas treatment system. (d) RWCU system inlet outboard isolation valve closes on SLCS "B" initiation. RWCU system inlet inbcard isolation valve closes on SLCS "A" or SLCS "C" initiation. LIMERICK - UNIT 1 3/4 3-16
i TABLE 3.3.2-1 (Continued) TABLE NOTATIONS . (e) Manual initiation isolates the steam supply line outboard isolation
- valve and only following manual or automatic initiation of the system.
(f) In the event of a loss of ventilation in.the Main Steam Line Tunnel Area, the main steam line tunnel temperature - high setpoint may be raised by
- 50 F for a period not to exceed 30 minutes to permit restoration of the ventilation flow without a spurious trip. During the 30 minute period, an operator, or other qualified member of the technical staff, shall observe the temperature indications continuously, so that, in the event of rapid increases in temperature, the main steam lines shall be manually 1 isolated.
(g) Wide range accident monitor per Specification 3.3.7.5. .i I I i i i } i l 4 s , r I i LIMERICK - UNIT 1 3/4 3-17 f
, - , - , - , ---2--- y,ew.-m.----- - - - - - , , , , -_ --,,-,,r,,-,,-m..,-yc-..c-n,,,v-
TABLE 3.3.2-2 h ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 9 R
' ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE E
Z 1. MAIN STEAM LINE ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Low - Level 2 > - 38 inches *
- 2) Low, Low, Low - Level 1
_ 1 - 45 inches 2 - 129 inches * > - 136 inches
- b. Main Steam Line < 3.0 x Full Power < 3.6 x Full Power Radiation - High Background Background
- c. Main Steam Line Pressure - Low 1 756 psig 1 736 psig
- d. Main Steam Line Y Flow - High 5 108.7 psid 5 111.7 psid E
- e. Condenser Vacuum - Low 10.5 psia 1 10.1 psia /5 10.9 psia
- f. Main Steam Line Tunnel Temperature - High 5 192 F 5 200 F
- g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High 5 165 F 5 175 F
- h. Manual Initiation N.A. N.A.
- 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
- a. Reactor Vessel Water Level Low - Level 3 2 12.5 inches
- 1 11.0 inches
- b. Reactor Vessel (RHR Cut-in Permissive) Pressure - High 5 75 psig 5 95 psig
- c. Manual Initiation N.A. N.A.
I O G G
g- s g-
'D N TABLE 3.3.2-2 (Continued) h ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M
ALLOWABLE
' TRIP SETPOINT VALUE TRIP FUNCTION E
Z 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. RWCS a Flow - High 5 54.9 gpm 5 65.2 gpm
- b. RWCS Area Temperature - High 5 135 F or 122 F** 1 145 F or 130 F**
- c. RWCS Area Ventilation a Temperature - High 5 32*F 5 40 F
- d. SLCS Initiation N.A. N.A.
- e. Reactor Vessel Water Level -
Low, Low, - Level 2 1 -38 inches
- 1 -45 inches i f. Manual Initiation N.A. N.A.
T 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION G
- a. HPCI Steam Line a Pressure - High 5 343" H 2O 5 358" H 2O
- b. HPCI Steam Supply Pressure - Low 1 100 psig 2 90 psig
- c. HPIC Turbine Exhaust Diaphragm Pressure - High 5 10 psig 5 20 psig
- d. HPCI Equipment Room Temperature - High 175 F 1 165 F, 1 200 F
- e. HPCI Equipment Room a Temperature - High 5 80 F 5 88 F
- f. HPCI Pipe Routing Area Temperature - High 175 F 1 165*F, 1 200*F
- g. Manual Initiation N.A. N.A.
- h. HPCI Steam Line a Pressure - Timer 3 $ r 5 12.5 seconds 2.5 5 1 5 13 seconds
TABLE 3.3.2-2 (Continued) C pj ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 5 9
, ALLOWABLE c: TRIP FUNCTION TRIP SETPOINT VALUE 5
-d
- 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
- a. Reactor Steam Line a Pressure - High 1 213" H 2O 5 223" H 2O
- b. RCIC Steam Supply Pressure - Low 1 64.5 psig 3 56.5 psig
- c. RCIC Turbine Fxhaust Diaphragm Pressure - High 5 10.0 psig 5 20.0 psig
,, d. RCIC Equipment Room ); Temperature - High 175 F 2 165 F, < 200 F )' c)
- e. RCIC Equipment Room A Temperature - High 5 80 F 5 88 F
- f. RCIC Pipe Routing Area Temperature - High 175 F 2 165 F, 5 200 F
- g. Manual Initiation N.A. N.A.
- h. RCIC Steam Line a Pressure Timer 3 5 t 5 12.5 seconds 2.5 5 1 1 13 seconds O O O
s. i i TABLE 3.3.2-2 (Continued) kh ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 9 4 74 i
ALLOWABLE 1
TRIP FUNCTION TRIP SETPOINT VALUE I E
< Z 6. PRIMARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level
- 1. Low, Low - Level 2 > -38 inches * > -45 inches Low, Low, Low, Level 1 [-136 inches
- 2. [-129 inches *
- b. Drywell Pressure - High 5 1.68 psig 5 1.88 psig i c. North Stack Effluent '
4 Radiation - High 5 2.1 pCi/cc 5 4.0 pCi/cc i , t' d. Refueling Area Ventilation Exhaust ' ** Duct - Radiation - High -< 2.2 mR/h
-< 2.0 mR/h t w i d' e. Reactor Enclosure Ventilation Exhaust j Duct - Radiation - High 5 1.35 mR/h 5 1.5 mR/h 1
j f. Outside Atmosphere To Reactor Enclosure i a Pressure - Low -> 0.1 inch -> 0.0 inch 4 i g. Outside Atmosphere To Refueling Area l A Presssure - Low 1 0.1 inch 1 0.0 inch j h. Drywell Pressure - High/ 5 1.68 psig/ $ 1.88 psig/ ! Reactor Pressure - Low 1 455 psig (decreasing) 1 435 psig (decreasing) i ! i. Primary Containment Instrument > 2.0 psig 1 1.9 psig Gas to Drywell a Pressure-Low i j
- j. Manual Initiation N.A. N.A.
1 ] 1 i
TABLE 3.3.2-2 (Continued) h ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 9 R 7 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE E q 7. SECONDARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level -
Low, Low - Level 2 2 -38 inches
- 1 -45 inches
- b. Drywell Pressure - High 5 1.68 psig 5 1.88 psig
- c. Refueling Area Ventilation Exhaust Duct Radiation - High 5 2.0 mR/h 1 2.2 mR/h
- d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High 5 1.35 mR/h 5 1.5 mR/h y e. Outside Atmosphere To Reactor Enclosure y A Pressure - Low 1 0.1 inch 1 0.0 inch
- f. Outside Atmosphere To Refueling Area A Pressure - Low 1 0.1 inch 2 0.0 inch
- g. Manual Initiatien N.A. N.A.
*See Bases Figure B 3/4 3-1. **The low setpoints are for the RWCU Heat Exchanger Rooms; the high setpoints are for the pump rooms.
O O O
N TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME . TRIP FUNCTION RESPONSE TIME (Seconds)#
- 1. MAIN STEAM LINE ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Low - Level 2 < 13(a),,
- 2) Low, Low, Low - Level 1 31.0*
b. Main Steam Radiation Line (b)
- Higb i 1.0*/5 13(a),,
. c. Main Steam Line Pressure - Low 1 1.0*/1 13(a),,
- d. Main Steam Line Flow - High 1 0.5*/1 13(a),a
- e. Condenser Vacuum - Low N.A.
i f. Main Steam Line Tunnel l Temperature - High N.A.
- g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
- h. Manual Initiation N.A.
i 2. RHR SYSTEM SHUTOOWN COOLING MODE ISOLATION
\ a. Reactor Vessel Water Level Low - Level 3 1 13(a)
- b. Reactor Vessel (RHR Cut-In 2
Permissive) Pressure - High N.A.
- c. Manual Initiation N.A.
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. RWCS A' Flow - High ##
1 13 .
- b. RWCS Area Temperature - High N.A.
- c. RWCS Area Ventilation A Temperature - High N.A.
- d. -SLCS Initiation N.A.
- e. Reactor Vessel Water Level -
Low, Low - Level 2 1 13(,)
- f. Manual Initiation N.A.
. a' + LIMERICK - UNIT 1 3/4 3-23
TABLE 3.3.2-3 (Continued) ISOLATION SYSTEM INSTRUMENTATIuN RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
- a. HPCI Steam Line a Pressure - High 5 13(a)
- b. HPCI Steam Supply Pressure - Low a) 5 13
- c. HPCI Turbine Exhaust Diaphragm Pressure - High N.A.
- d. HPCI Equipment Room Temperature - High N.A.
- e. HPCI Equipment Room a Temperature - High N.A.
- f. HPCI Pipe Routing Area Temperature - High N.A.
- g. Manual Initiation N.A.
- 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
- a. Reactor Steam Line a Pressure - High 5 13(a)
- b. RCIC Steam Supply Pressure - Low 5 13(a)
- c. RCIC Turbine Exhaust Diaphragm Pressure - High N.A.
- d. RCIC Equipment Room Temperature - High N.A.
- e. RCIC Equipment Room A Temperature - High N.A.
- f. RCIC Pipe Routing Area Temperature - High N.A.
- g. Manual Initiation N.A.
O LIMERICK - UNIT 1 3/4 3-24
TABLE 3.3.2-3 (Continued) 7 kj ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME i TRIP FUNCTION RESPONSE TIME (Seconds)#
- 6. PRIMARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Low - Level 2 i 13(a)
- 2) Low, Low, Low - Level 1 5 13(a)
- b. Drywell Pressure - High 1 13(a)
- c. North Stack Ef fluent Radiation - High N.A.
- d. Refueling Area Ventilation Exhaust Duct - Radiation - High N.A.
- e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High N.A.
- f. Outside Atmosphere To Re.sctor Enclosure A Pressure - Low N.A.
- g. Outside Atmosphere To Refueling Area A Pressure - Low N.A.
- h. Drywell Pressure - High/
Reactor Pressure - Low N.A. l i. Primary Containment Instrument Gas to N.A. Drywell A Pressure-Low
- j. Manual Initiation N.A.
I 7. SECONDARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level Low, Low - Level 2 N.A.
- b. Drywell Pressure - High N.A.
- c. Refueling Area Ventilation Exhaust .
Duct Radiation - High N.A.
- d. Reactor Enclosure Ventilation Exhaust
- Duct Radiation - High N.A.
- e. Outside Atmosphere to Reactor
! Enclosure A Pressure - Low N.A. A LIMERICK - UNIT 1 3/4 3-25
TABLE 3.3.2-3 (Continued) ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
- f. Outside Atmosphere To Refueling Area A Pressure - Low N.A.
- g. Manual Initiation N.A.
TABLE NOTATIONS (a) Isolation system instrumentation response time specified includes 10 seconds diesel generator starting and 3 seconds for sequence loading delays. (b) Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in the channel.
- Isolation system instrumentation response time for MSIV only. No diesel generator delays assumed for MSIVs.
** Isolation system instrumentation response time for associated valves except MSIVs.
# Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
##With 45 second time delay.
O LIMERICK - UNIT 1 3/4 3-26
p f V- (./ O TABLE 4.3.2.1-1 ISOLATION' ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS B n CHANNEL OPERATIONAL , CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED w H 1. MAIN STEAM LINE ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Low, Level 2 S M R 1,2,3
- 2) Low, Low, Low - Level 1 S M R 1,2,3
- b. Main Steam Line Radiation - High S M R 1,2,3
- c. Main Steam Line
{ Pressure - Low S M R 1 T d. Main Steam Line D Flow - High S M R 1,2,3
- e. Condenser Vacuum - Low S M R 1, 2**, 3**
- f. Main Steam Line Tunnel Temperature - High S M R 1,2,3
- g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High S M R 1,2,3
- h. Manual Initiation N.A. R N.A. 1,2,3
- 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
- a. Reactor Vessel Water Level Low - Level 3 S M R 1,2,3
- b. Reactor Vessel (RHR Cut-In S M R 1,2,3 Permissive) Pressure - High
- c. Manual Initiation N.A. R N.A. 1,2,3
TABLE 4.3.2.1-1 (Continu::d) C g ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 i p CHANNEL OPERATIONAL
, CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH c TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 5
- 3.- REACTOR WATER CLEANUP SYSTEM ISOLATION i
- a. RWCS A Flow - High S M R 1,2,3 l
l b. RWCS Area Temperature - High S M R 1,2,3 I c. RWCS Area Ventilation A Temperature - High S M R 1,2,3
- d. SLCS Initiation N.A. R N.A. 1,2,3
- e. Reactor Vessel Water Level -
w Low, Low, - Level 2 S M R 1,2,3 A f. Manual Initiation N.A. R N.A. 1,2,3
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
- a. HPCI Steam Line A Pressure - High S M R 1,2,3
- b. HPCI Steam Supply Pressure - Low S M R 1,2,3
- c. HPCI Turbine Exhaust Diaphragm Pressure - High S M R 1,2,3
- d. HPCI Equipment Room Temperature - High S M R 1,2,3
- e. HPCI Equipment Room A Temperature - High S M R 1,2,3
- f. HPCI Pipe Routing Area Temperature - High S M R 1,2,3
- g. Manual Initiation N.A. R N.A. 1,2,3
- h. HPCI Steam Line A Pressure Timer N.A. M R 1,2,3 O O O
,- /~ ,e~
N.] 0 N-] y TABLE 4.3.2.1-1 (Continued) 3! g ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS M Pc CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH g TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED C y 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
- a. Reactor Steam Line a Pressure - High S M R 1,2,3
- b. RCIC Steam Supply Pressure - Low S M R 1,2,3
- c. RCIC Turbine Exhaust Jiaphragm m Pressure - High S M R 1,2,3 1
m d. RCIC Equipment Room 4 Temperature - High S H R 1,2,3 e
- e. RCIC. Equipment Room A Temperature - High S M R 1,2,3
- f. RCIC Pipe Routing Area Temperature - High S M R 1,2,3
- g. Manual Initiation N.A. R N.A. 1,2,3
- h. RCIC Steam Line a Pressure Timer N.A. M R 1,2,3
D HE CR L ARR NO II HU WQ E O OFE 33 3 3 3 3 3 3 3 3 3 I C TSN ,, , , , , , , , , , ANA 22 2 2 2 2 2 2 2 2 2 ROL EIL ,, , , , , , , , , , PTI 11 1 1 1 1 1 1 1 1 1 OIE DV NR OU CS N S O T LI . N ET E NA A. M NR RR R R R R Q Q R Q N E AB R HI I CL U A Q C E R E C N A ) L d L e I u E L n V LA i R EN t U NOT n S NIS MM M Q M M M M M M R o ATE C N HCT ( O CN O I U 1 T F
- A 1 T . N 2 E . M 3 U . R 4 T S L . . . .
E N EK A L I NC A. A. A. B NE SS S S S S N N S N. N A N AH T O HC I C h h g w T g g n t o A i ni r i nL U 1 H oH o l e T i t e m-C A l e l e n-o t - a cw ao f u u re N v v i n l n eL e / tr N O e2e h t o i o R R h su O I L L g ai ti - gw ns I T A re-l i H l t nt ea o ow i o I s e T t i a Te To HL A L ev n ti Vi r L t r L O t ew - e nd d eu e - - nP O S aL o u ea ea rs r- e n S I I W - L e r l h f g VR rR u es h e e h e ee rr mA n i o T l , u fi a- s - pr pr uu il t N E eww soo s s EH e rt o l t sP o su os ss ss al t e i a M N sLL e e r k - c Ac u cc nu t mA ms t e ee rr nw oy t i I V ,, P an gD ED Ae Ar PP C r n A T roo ww l t o Si n it rt eu r e P l r y D I N oLL l e t l s eu os d s dA l o ro l O t h a t u io i et at a N C c a w y ti rd ua fh ca ah sl t c sa t e wc ya is m n u O Y e)) r oa ex ex un ur re ra a I R R12 D NR RE RE OE OA DR PG M T A C M N I U R . . . . . . . . . . a c e O F P b d f g h i j P I R . T 6 Cg5m7 c5H s { ,I$
.p . f%
O O O i TABLE 4.3.2.1-1 (Continued) ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS e o CHANNEL OPERATIONAL 7 TRIP FUNCTION CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH c CHECK TEST' CALIBRATION SURVEILLANCE REQUIRED z Z 7. SECONDARY CONTAINMENT ISOLATION w
- a. Reactor Vessel Water Level Low, Low - Level 2 S M R 1,2,3
- b. Drywell Pressure - High S M R 1, 2, 3
- c. Refueling Area Ventilation Exhaust Duct Radiation - High S M R
- l l d. Reactor Enclosure Ventilation
- $ Exhaust Duct Radiation - High S M R 1,2,3 Y e. Outside Atmosphere To Reactor j d Enclosure A Pressure - Low N.A. M Q 1,2,3
- f. Outside Atmosphere To Refueling Area a Pressure - Low N.A. M Q
- g. Manual Initiation N.A. R N.A. 1, 2, 3, and
- l l
*When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
**When not administrative 1y bypassed and/or when any turbine stop valve is open.
INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3. APPLICABILITY: As shown in Table 3.3.3-1. ACTION:
- a. With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.
- c. With either ADS trip system subsystem inoperable, restore the inoperable trip system to OPERABLE status within:
- 1. 7 days, provided that the HPCI and RCIC systems are OPERABLE.
- 2. 72 hours.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1-1. 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per trip system 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 ECCS trip system. LIMERICK - UNIT 1 3/4 3-32
-- - - . . . - - . _ -- - _ - _- . . . . . . _ . ~ . . . . . .. -- . - - - - - - - - _ . ~ .
s O
- s. (m[ ,
TABLE 3.3.3-1 C E EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION 5 Q MINIMUM OPERABLE
, CHANNELS PER APPLICABLE c- TRIP OPERATIONAL g TRIP FUNCTION FUNCTION (a) CONDITIONS ACTION w
- 1. CORE SPRAY SYSTEM ***
- a. Reactor Vessel Water Level - Low Low Low, Level 1 2/ pump 1, 2, 3, 4* , 5* 30
- b. Drywell Pressure - High 2/ pump 1,2,3 30
- c. Reactor Vessel Pressure - Low (Permissive) 6(b) 1, 2, 3 31 4*, 5* 32 '
- d. Manual Initiation 2(*) 1, 2, 3, 4*, 5* 33
( 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM *** a , y a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1, 2, 3, 4*, 5* 30 I
$ b. Drywell Pressure - High 2 1,2,3 30
- c. Reactor Vessel Pressure - Low (Permissive) 2 1,2,3 31
- d. Injection Valve Differential Pressure-Low 1/ valve 1, 2, 3, 4* , 5* 31
- e. Manual Initiation 1 1, 2, 3, 4*, 5* 33
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM U
j a. Reactor Vessel Water Level - Low Low Level 2 4 1,2,3 34
- b. Drywell Pressure - High 1, 2, 3 34
- c. Condensate Storage Tank Level - Low 4(c) 2 1,2,3 35
- d. Suppression Pool Water Level - High 2 1, 2, 3 35
- e. Reactor Vessel Water Level - High, Level 8 4(d) 1,2,3 31
- f. Manual Initiation 1/ system 1,2,3 33 i
l
TABLE 3.3.3-1 (Continued) C 3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION Q MINIMUM OPERABLE
, CHANNELS PER APPLICABLE c TRIP OPERATIONAL y TRIP FUNCTION FUNCTION (3) CONDITIONS ACTION H 4. AUTOMATIC DEPRESSURIZATION SYSTEM #***
- a. Reactor Vessel Water Level - Low Low Low, Level 1 2 1,2,3 30
- b. Drywell Pressure - High 2 1,2,3 30
- c. ADS Timer 1 1,2,3 31
- d. Core Spray Pump Discharge Pressure - High (Permissive) 2 1,2,3 31
- e. RHR LPCI Mode Pump Discharge Pressure High (Permissive) 4 1,2,3 31
- f. Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 31 g g. Manual Initiation 2 1,2,3 33 g h. ADS Drywell Pressure Bypass Timer 2 1,2,3 31 Y
g MINIMUM APPLICABLE e TOTAL NO. CHANNELS CHANNELS OPERATIONAL OF CHANNELS (f) TO TRIP OPERABLE CONDITIONS ACTION
- 5. LOSS OF POWER
- 1. 4.16 kV Emergency Bus Under-voltage (Loss of Voltage) 1/ bus 1/ bus 1/ bus 1, 2, 3, 4**, 5** 36
- 2. 4.16 kV Emergency Bus Under-voltage (Degraded Voltage) 1/ source / 1/ source / 1/ source / 1, 2, 3, 4**, 5** 37 bus bus bus
***The Minimum OPERABLE Channels Per Trip Function is per subsystem. -
O O O
- ('N TABLE 3.3.3-1 (Continued)
\
V) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE NOTATIONS (a) A charniel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. (b) Also provides input to actuation logic for the associated emergency diesel generators. j (c) One trip system. Provides signal to HPCI pump suction valves only. (d) On 1 out of 2 taken twice logic, provides a signal to trip the HPCI pump turbine only. (e) The manual initiation push buttons start the respective core spray pump and diesel generator. The "A" and "B" logic manual push buttons also actuate an initiation permissive in the injection valve opening logic. (f) A channel as use.d here is defined as the 127 bus relay for Item 1 and the 127,127Y, and 127Z feeder relays with their associated time delay relays i taken together for Item 2. p
- When the system is required to be OPERABLE per Specification 3.5.2.
# Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
** Required when ESF equipment is required to be OPERABLE.
-## Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
b U LIMERICK - UNIT 1 3/4 3-35
TABLE 3.3.3-1 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 30 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. With one channel inoperable, place the inoperable channel in the tripped condition within 1 hour
- or declare the associated system inoperable.
- b. With more than one channel inoperable, declare the associated system inoperable.
ACTION 31 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ECCS inoperable. ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour. ACTION 33 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours or declare the associated ECCS inoperable. ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. For one channel inoperable, place the inoperable channel in the tripped condition within 1 hour
- or declare the HPCI system inoperable.
- b. With more than one channel inoperable, declare the HPCI system inoperable.
ACTION 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per frip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour
- or declare the HPCI system inoperable.
ACTION 36 With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate. ACTION 37 - With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the tripped condition within 1 hour;* operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST.
- The provisions of Specification 3.0.4 are not applicable.
LIMERICK - UNIT 1 3/4 3- 36
g 3 'N-Y TABLE 3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS E ALLOWABLE-7 TRIP FUNCTION TRIP SETPOINT VALUE E 1. CORE SPRAY SYSTEM Z s a. Reactor Vessel Water Level - Low Low Low, Level 1 > -129 inches * > -136 inches
- b. Drywell Pressure - High 51.68psig 31.88psig
- c. Reactor Vessel Pressure - Low > 455 psig,(decreasing) > 435 psig, (decreasing)
- d. Manual Initiation N.A. N.A.
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM i
- a. Reactor Vessel Water Level - Low Low Low, Level 1 > -129 inches
- 2 -136 inches
- b. Drywell Pressure - High 5 1.68 psig 5 1.88 psig
, c. Reactor Vessel Pressure - Low > 455 psig,(decreasing) > 435 psig, (decreasing) g d. Injection Valve Differential Pressure - Low 2 78 psid, (decreasing) > 68 psid and 5 88 psid i
- e. Manual Initiation N.A. N.A. '
d 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
- a. Reactor Vessel Water Level - (Low Low, Level 2) > -38 inches * > -45 inches
- b. Drywell Pressure - High 31.68psig 31.88psig
- c. Condensate Storage Tank Level - Low > 167.8 inches ** > 164.3 inches
- d. Suppression Pool Water Level - High 324 feet 1.5 inches 524 feet 3 inches
- e. Reactor Vessel Water Level - High, Level 8 < 54 inches < 60 inches
- f. Manual Initiation R.A. H.A.
- 4. AUTOMATIC DEPRESSURIZATION SYSTEM
- a. Reactor Water Level - Low Low Low, Level 1 2 -129 inches * > -136 inches
- b. Drywell Pressure - High 5 1.68 psig 5 1.88 psig
- c. ADS Timer < 105 seconds < 117 seconds
- d. Core Spray Pump Discharge Pressure - High [145psig,(increasing) [125psig,(increasing),
- e. RHR LPCI Mode Pump Discharge Pressure-High > 125 psig,(increasing) > 115 psig, (increasing)
- f. leactor Vessel Water Level-Low, Level 3 > 12.5 inches > 11.0 inches
- g. Manual' Initiation H.A. H.A.
- h. ADS Drywell Pressure Bypass Timer 5 420 seconds 1 450 seconds
*See Bases Figure B 3/4.3-1.
** Corresponds to 2.25 feet indicated.
TABLE 3.3.3-2 (Continued) EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS C z ALLOWABLE @ TRIP FUNCTION TRIP SETPOINT VALUE R 5. LOSS OF POWER RELAY [ a. 4.16 kV Emergency Bus Undervoltage NA NA z (Loss of Voltage) 127-11X w s b. 4.16 kV Emergency Bus Undervoltage RELAY (Degraded Voltage) 127-11X0X a. 4.16 kV Basis and 2905 i 115 volts 2905 i 145 volts 102-11X0X b. 120 V Basis 83 3 volts 83 1 4 volts
- c. 5 1 second time $ 1.5 second time delay delay 127Y-11X0X** a. 4.16 kV Basis and 3640 1 91 volts 3640 1 182 volts R
- 127Y-1-11X0X b. 120 V Basis 104 3 volts 104 5.2 volts T c. < 52 second time < 60 second time S 3e1ay 3e1ay 127Z-11X0X a. 4.16 kV Basis and 3745 i 94 volts 3745 1 187 volts 162Y-11X0X b. 120 V Basis 107 3 volts 107 1 5.4 volts
- c. < 10 second time < 11 second time 3elay Helay 127Z-11X0X a. 4.16 kV Basis and 3745 94 volts 3745 i 187 volts 162Z-11X0X b. 120 V Basis 107 1 3 volts 107 5.4 volts
- c. < 61 second time < 64 second time delay 3elay
**This is an inverse time delay voltage relay. The voltages shown are the maximum that will not result in a trip. Some voltage conditions will result in decreased trip times.
O O O
l l t s, 1 TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES 3 ECCS RESPONSE TIME (Seconds)
- 1. CORE SPRAY SYSTEM $ 27
) 2. LOW PRESSURE COOLANT INJECTION MODE 0F RHR SYSTEM i 5 40
- 3. AUTOMATIC DEPRESSURIZATION SYSTEM N.A.
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM 5 30 i
j 5. LOSS OF POWER N.A. i l U a i 4 I 1 1
'\
t ! LIMERICK - UNIT 1 3/4 3-39
TABLE 4.3.3.1-1 h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS "5 E CHANNEL OPERATIONAL
^ CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 1
Z 1.. CORE SPRAY SYSTEM s
- a. Reactor Vessel Water Level -
! Low Low Low, Level 1 S M R 1, 2, 3, 4*, 5* l b. Drywell Pressure - High S M R 1,2,3 l c. Reactor Vessel Pressure - Low S M R 1, 2, 3, 4* , 5*
- d. Manual Initiation N.A. R N.A. 1, 2, 3, 4*, 5*
l 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM
- a. Reactor Vessel Water Level -
R Low Low Low, level 1 S M R 1, 2, 3, 4*, 5*
- 1,2,3
- b. Drywell Pressure - High S M R Y c. Reactor Vessel Pressure - Low S M R 1,2,3
$ d. Injection Valve Differential Pressure - Low (Permissive) S M R 1, 2, 3, 4*, 5*
- e. Manual Initiation N.A. R N.A. 1, 2, 3, 4*, 5*
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM ***
, a. Reactor Vessel Water Level - Low Low, Level 2 S M R 1,2,3
- b. Drywell Pressure - High S M R 1,2,3 j c. Condensate Storage Tank Level -
Low S M R 1,2,3
- d. Suppression Pool Water Level -
High S M R 1,2,3
- e. Reactor Vessel Water Level -
High, Level 8 S M R 1,2,3
- f. Manual Initiation N.A. R N.A. 1,2,3 O O O
O O O TABLE 4.3.3.1-1 (Continued) , h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9 p; CHANNEL OPERATIONAL
^ CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED h 4. AUTOMATIC DEPRESSURIZATION SYSTEM j a. Reactor Vessel Water Level -
Low Low Low, Level 1 5 M R 1,2,3
- b. Drywell Pressure - High S M R 1,_2, 3
- c. ADS Timer N.A. M Q 1,2,3
- d. Core Spray Pump Discharge Pressure - High S M R 1,2,3
- e. RHR LPCI Mode Pump Discharge Pressure - High S M R 1,2,3
- f. Reactor Vessel Water Level - Low, R Level 3 S M R 1,2,3
- g. Manual Initiation N. A. R N.A. 1,2,3
- h. ADS Drywell Pressure Bypass Timer N.A. M Q 1,2,3
- 5. LOSS OF POWER
- a. 4.16 kV Emergency Bus Underp, voltage (Loss of Voltage) N.A. R N.A. 1, 2, 3, 4**, 5**
- b. 4.16 kV Emergency Bus Under-voltage (Degraded Voltage) S M R 1, 2, 3, 4**, 5**
j When the system is required to be OPERABLE per Specification 3.5.2. Required OPERABLE when ESF equipment is required to be OPERABLE.
*** Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.
# Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.
## Loss of Voltage Relay 127-11X is not field setable.
~_
INSTRUMENTATION 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2. APPLICABILITY: OPERATIONAL CONDITION 1. ACTION:
- a. With an ATWS recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, piace the inoperable channel (s) in the tripped condition within 1 hour.
- c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:
- 1. If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure channel, place both inoperable channels in the tripped condition within 1 hour, or, if this action will initiate a pump trip, declare the trip system inoperable.
- 2. If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip system inoperable.
- d. With one trip system inoperable, restore the inoperable trip s3 stem to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.
- e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour or be in at least STARTUP within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.3.4.1.1. Each ATWS recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.1-1. 4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. LIMERICK - UNIT 1 3/4 3-42
4 I. i e 3 . TABLE 3.3.4.1-1 i 4 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION i ! MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP SYSTEM * ! 1. Reactor Vessel Water Level -
- Low Low, level 2 2
!i 2. Reactor Vessel Pressure - High 2 l 1. 1 } i 1-i 4 4 4 ) t l 1 i ! is I-1 l-F i. l 5 i i ) ) *0ne channel may be placed in an inoperable status for up to 2 hours for
- h. required surveillance provided the other channel.is OPERABLE. .
i I i }~ !
- LIMERICK - UNIT 1. 3/4 3-43 r
. ~. . _ _ _ _ _ . _ _ . . . . _ . . . _ . . . . , . _ _ _ _ , . _ _ _ _ . _ _ _ _ . _ . . _ . _ _ . . . _ . - _ _ _ . . . - . . . _ . . _ _ _ . _ _ . . . _ _ _ . _ . . _ _ . _ _ _ _ , . , . . - . _ , , _ _ _ _ _ _ _ , .
TABLE 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE
- 1. Reactor Vessel, Water Level -
Low Low, Level 2 1 -38 inches
- 2 -45 inches
- 2. Reactor Vessel Pressure - High 5 1093 psig 5 1108 psig O
- See Bases Figure 83/4 3-1.
O LIMERICK - UNIT 1 3/4 3-44
i l
~
L- TABLE 4.3.4.1-1 i 1 lJ ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION -
. SURVEILLANCE REQUIREMENTS i
i l-CHANNEL CHANNEL FUNCTIONAL CHANNEL. !- TRIP FUNCTION CHECK _ TEST CALIBRATION l } 1. Reactor Vessel Water Level - 1 Low-Low, Level 2 S M R -)
- i. t
- 2. . Reactor' Vessel Pressure - High S M R l-i i' '
i' i i t 1 J 1 ! t i i i i t 1 i I-I i ' t I l i ? - l-L. , t t ,' i r i LIMERICK - UNIT 1 3/4 3-45 i i
~ - __,,,, , _ _ _ _ - _ . - - _ _ _ - .
INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TPIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pu...p trip (E0C-RPT) system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER. ACTION:
- a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within 1 hour.
- c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:
- 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 1 hour.
- 2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.
- d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours or take the ACTION required by Specification 3.2.3.
- e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or take the ACTION required by Specification 3.2.3.
O LIMERICK - UNIT 1 3/4 3-46
s 7.s INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.4.2.1 .Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.2.1-1. 4.3.4.2.2. LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. 4.3.4.2.3 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested
. at least once per 36 months. The .neasured time shall be added to the most i
recent breaker arc suppression tine and the resulting END-0F-CYCLE-RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be verified to be within its limit. 4.3.4.2.4 The time interval necessary for breaker arc suppression from energi-zation of the recirculation pump circuit breaker trip coil shall be measured at least once per 60 months. b V 4
- LIMERICK - UNIT 1 3/4 3-47 l
TABLE 3.3.4.2-1 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS TRIP FUNCTION PER TRIP SYSTEM *
- 1. Turbine Stop Valve - Closure 2**
- 2. Turbine Control Valve-Fast Closure 2**
*A trip system may be placed in an inoperable status for up to 2 hours for required surveillance provided that the other trip system is OPERABLE.
- This function shall be automatically bypassed when turbine first stage ,
pressure is equivalent to THERMAL POWER less than 30% of RATED THERMAL POWER. I O LIMERICK - UNIT 1 3/4 3-48 , 1
- ..---.-.-...-.-.--. _ . - . - - . - . _ . . . . . - . - - . . . _ - - - - - . . - - - - . _~
i~ !~
/h TABLE 3.3.4.2-2
, END-0F-CYCLE- RECIRCULATION PUMP TRIP SETPOINTS ! . ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
- 1. Turbine Stop Valve-Closure 1 5% closed i 7% closed l- 2. Turbine Control Valve-Fast Closure > 500 p,sig
> 465 psig 4
4 i 4
- u i
l f 3 i-l' I' i i I I I I l l . i i i ! f i l l l i LIMERICK - UNIT 1 3/4 3-49 i - - , . -
l l TABLE 3.3.4.2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Milliseconds)
- 1. Turbine Stop Valve-Closure 5 175
- 2. Turbine Control Valve-Fast Closure 5 175 O
O : 1 LIMERICK - UNIT 1 3/4 3-50
TABLE'4.3.4.2.1-1 l
'END-0F-CYCLE RECIRCULATON PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS i
CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION TEST CALIBRATION ' i
- 1. Turbine Stop. Valve-Closure. M* R 2.' Turbine Control Valve-Fast Closure M* R
- Including trip system logic testing.
Y S I f
- LIMERICK - UNIT 1 3/4 3-51
INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. ACTION:
- a. With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With one or more RCIC system actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.5-1.
SURVEILLANCE REQUIREMENTS 4.3.5.1 Each RCIC system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.5.1-1. 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. O LIMERICK - UNIT 1 3/4 3-52
'"'g TABLE 3.3.5-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE CHANNELS FUNCTIONAL UNITS PER TRIP FUNCTION
- ACTION
- a. Reactor Vessel Water Level -
Low Low, Level 2 4# 50
- b. Reactor Vessel Water Level -
High, Level 8 4# 51
- c. Condensate Storate Tank Water Level - Low 2** 52
- d. Manual Initiation 1/ system *** 53 2
{ \s_ / *A channel may be placed in an inoperable status for-up to 2 hours for J required surveillance without placing the trip system in the tripped con-dition provided all other channels monitoring that parameter are OPERABLE.
**0ne trip system with one-out-of-two logic.
***0ne trip system with one channel.
- #0ne trip system with one-out-of-two twice logic.
1 1 l I L) '
. LIMERICK - UNIT 1 3/4 3-53
?
r TABLE 3.3.5-1 (Continued) REACTOR CORE ISOLATION COOLING SYSTEM ACTION STATEMENTS ACTION 50 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
- a. With one channel inoperable, place the inoperable channel in the tripped condition within 1 hour or declare the RCIC system inoperable.
- b. With more than one channel inoperable, declare the RCIC system inoperable.
ACTION 51 - With the number of OPERABLE channels less than required by the minimum OPERABLE channels per Trip System requirement, declare the RCIC system inoperable. ACTION 52 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, place at least one inoperable channel in the tripped condition within 1 hour or declare the RCIC system inoperable. ACTION 53 - With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE status within 8 hours or declare the RCIC system inoperable. O LIMERICK - UNIT 1 3/4 3-54
p TABLE 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT VALUE
- a. ' Reactor Vessel Water Level -
Low Low, Level 2 1-38 inches
- 1-45 inches
- b. Reactor Vessel Water Level -
High, Level 8 5 54 inches 5 60 inches
- c. Condensate Storage Tank Level -
Low 2 135.8** inches 1 132.3 inches
- d. Manual Initiation N.A. N.A.
*See Bases Figure B 3/4.3-1.
** Corresponds to 2.25. feet indicated.
J l i LIMERICK - UNIT 1 3/4 3-55
TABLE 4.3.5.1-1 REACTOR CORE ISOLATION SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL FUNCTIONAL UNITS CHECK TEST CALIBRATION
- a. Reactor Vessel Water Level -
Low Low, Level 2 S M R
- b. Reactor Vessel Water Level -
High, Level 8 5 M R
- c. Condensate Storage Tank Level - Low S M R
- d. Manual Initiation N.A. R N.A.
I O O LIMERICK - UNIT 1 3/4 3-56
INSTRUMENTATION O 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.. APPLICABILITY: As shown in Table 3.3.6-1. ACTION:
- a. With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of
- Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.
SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1. v} LIMERICK - UNIT 1 3/4 3-57
TABLE 3.3.6-1 r- CONTROL RG3 BLOCK INSTRUMENTATION 3: m MINIMUM APPLICABLE 3 ' OPERABLE CHANNELS OPERATIONAL E TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION [ 1. ROD BLOCK MONITOR (a)
- a. Upscale 2 1* 60 5 1* 60
-i b. Inoperative 2 M' c. Downscale 2 1* 60
- 2. APRM
- a. Flow Biased Neutron Flux -
Upscale 4 1 61
- b. Inoperative 4 1, 2, 5 61
- c. Downscale 4 1 61
- d. Neutron Flux - Upscale, Startup 4 2, 5 61
- 3. SOURCE RANGE MONITORS
- a. Detector not full in(b) 3 2 61
$ 2 5 61 3 2 61 m b. Upscale (c) 2 5 61 m
- c. Inoperative (C) 2 3 2 6
- d. Downscale(d)
- 4. INTERMEDIATE RANGE MONITORS Detector not full in 6 2, 5 61 a.
Upscale 6 2, 5 61 b.
- c. 6 2, 5 61 Inoperatijg) 6 2, 5 61
- d. Downscale
- 5. SCRAM DISCHARGE VOLUME
- a. Water Level-High 2 1, 2, 5** 62
- 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW 2 1 62
- a. Upscale Inoperative 2 1 62 b.
Comparator 2 1 62 c. 2 3, 4 63
- 7. REACTOR MODE SWITCH SHUTDOWN POSITION 9 9 e
l l TABLE 3.3.6-1 (Continued) CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 - Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3. ACTION 61 - With the number of OPERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour. ACTION 62 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour. ACTION 63 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block. NOTES
- With THERMAL POWER > 30% of RATED THERMAL POWER.
U ** With more than one control rod withdrawn. Not applicable removed per Specification 3.9.10.1 or 3.9.10.2. to control rods t (a) The RBM shall be automatically bypassed when a peripheral control rod is , selected or the reference APRM channel indicates less than 30% of RATED THERMAL POWER. (b) This function shall be automatically bypassed if detector count rate is
> 100 cps or the IRM channels are on range 3 or higher.
(c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher. - (d) This function is automatically bypassed when the IRM channels are on range 3 or higher. (e) This function is automatically bypassed when the IRM channels are on range 1. J LIMERICK - UNIT 1 3/4 3-59 ,
TABLE 3.3.6-2 CONTROL R0D BLOCK INSTRUMENTATION SETP0INTS TRIP SETPOINT ALLOWABLE VALUE TRIP FUNCTION
!a N 1. R0D BLOCK MONITOR
- a. Upscale 5 0.66 W + 40% 5 0.66 W + 43%
7 Inoperative N.A. N.A. c b. z c. Downscale 1 5% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER
~ 2 .' APRM
- a. Flow Biased Neutron Flux - Upscale 5 0.66 W + 42%* 5 0.66 W + 45%*
- b. Inoperative N.A. N.A.
- c. Downscale > 4% of RATED THERMAL POWER > 3% of RATED THERMAL POWER
- d. Neutron Flux - Upscale, Startup 312%ofRATEDTHERMALPOWER {14%ofRATEDTHERMALPOWER
- 3. SOURCE RANGE MONITORS
- a. Detector not full in N.A. N.A.
- b. Upscale 5 1 x 105 cps 5 1.6 x 105 cps
- c. Inoperative N.A. N.A.
- d. Downscale 2 3 cps ** 1 1.8 cps **
- 4. INTERMEDIATE RANGE MONITORS T a. Detector not full in N.A. N.A.
di isions of
$ b. Upscale 5 108/125 divisions of 5 110.2 .2 full s. ale full scale
- c. Inoperative N.A. N.A.
- d. Downscale 1 5/125 divisions of full scale 1 3/125 jivisions of full scale
- 5. SCRAM DISCHARGE VOLUME
- a. Water Level-High 5 257' 5 9/16" elevation *** 5 257' 7 9/16" elevation
- a. Float Switch
- 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW
- a. Upscale 5 108% of rated flow $ 111% of rated flow
- b. Inoperative N.A. N.A.
- c. Comparator 5 10% flow deviation i 11% flow deviation N.A. N.A.
- 7. REACTOR MODE SWITCH SHUTDOWN POSITION
*The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow
! (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.
**May be reduced to 0.7 cps provided the signal-to-noise ratio is 2 2.
*** Equivalent to 13 gallons / scram discharge volume.
O O O
. . _ . _ ..._ .__ _ _ - - - --- - ~ -__
j r TABLE 4.3.6-1 , ( CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS l $ CHANNEL OPERATIONAL i ' CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH )i g TRIP FUNCTION CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED ! Z 1. ROD BLOCK MONITOR i ,-. l a. Upscale N.A. S/U(b)(c) (c) SA 1* I b. Inoperative N.A. S/U(b)(c) (c) N.A. 1* l' c. Downscale N.A. S/U(b)(c) (c) , SA 1* l 2. APRM
- a. Flow Biased Neutron Flux -
, Upscale N.A. S/U(b) M, SA 1 4
- b. Inoperative N.A. S/U ,M N.A. 1, 2, 5 l c. Downscale N.A. S/U SA 1
- d. Neutron Flux - Upscale, Startup N.A. S/U(b),M,M SA 2, 5 l {
! <a 3. SOURCE RANGE MONITORS
- a. Detector not full in N.A. S/U ,W N.A. 2, 5 i b. Upscale N.A. S/U(b),W SA 2, 5
- c. Inoperative N.A. S/U W N.A. 2, 5
- d. Downscale N.A. S/U(b),W, SA 2, 5
- 4. INTERMEDIATE RANGE MONITORS
- a. Detector not' full in N.A. S/U(b) W N.A. 5, 5
- b. Upscale N.A. S/U(b),W, SA 2, 5 l c. Inoperative N.A. S/U ,W N.A. 2, 5 j d. Downscale N.A. S/U ,W SA 2, 5 l 5. SCRAM DISCHARGE VOLUME
- a. Water Level-High N.A. M R' 1, 2, 5**
! 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW
- a. Upscale N.A. -S/U ,M SA 1 j b. Inoperative N.A. S/U(b),M N.A. 1
- c. Comparator N.A. S/U ,M SA 1 i
l 7. REACTOR MODE SWITCH SHUTDOWN l POSITION N.A. R N.A. 3, 4 l
TABLE 4.3.6-1 (Continued) O CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. (b) Within 24 hours prior to startup, if not performed within the previous 7 days. (c) Includes reactor manual control multiplexing system input.
- With THERMAL POWER > 30% of RATED THERMAL POWER.
- With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
O O LIMERICK - UNIT 1 3/4 3-62
~s INSTRUMENTATION s- 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE with their alarm / trip setpoints within the specified limits.
APPLICABILITY: As shown in Table 3.3.7.1-1. ACTION:
- a. With a radiation monitoring instrumentation channel alarm / trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours or declare the channel j inoperable.
. b. With one or more radiation monitoring channels inoperable, take the > ACTION required by Table 3.3.7.1-1.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
( SURVEILLANCE REQUIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the conditions and at the frequencies shown in Table 4.3.7.1-1.
'\
LIMERICK - UNIT 1 3/4 3-63 i _ .m
TABLE 3.3.7.1-1 h RADIATION MONITORING INSTRUMENTATION 9 M Pc MINIMUM CHANNELS APPLICABLE ALARM / TRIP
' INSTRUMENTATION OPERABLE CONDITIONS SETPOINT ACTION E
q 1. Main Control Room Normal 4 1,2,3,5 1 x 10 5 pCi/cc 70 - Fresh Air Supply Radiation and
- Monitor
- 2. Area Monitors
- a. Criticality Monitors
- 1) Spent Fuel 2 (a) > 5 mR/h and $20mR/h(b)71 Storage Pool
- b. Control Room Direct 1 At All Times N.A.(b) 73 w Radiation Monitor k
w 3. Reactor Enclosure Cooling a Water Radiation Monitor 1 At All Times 5 3 x Background (b) 72 + 0 0 0
, TABLE 3.3.7.1-1 (Continued)
( (/ RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS
*When irradiated fuel is being handled in the secondary containment.
(a) With fuel in the spent fuel storage pool. (b) Alarm only. ACTION STATEMENTS ACTION 70 - With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within the next 6 hours, initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation. With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control room emergency filtration system in the radiation mode of operation. ACTION 71 - With one of the required monitor inoperable, assure a portable continuous monitor with the same alarm setpoint is OPERABLE in the vicinity of the installed monitor during any fuel movement.
- If no fuel movement is being made, perform area surveys of the
( j monitored area with portable monitoring instrumentation at least
\ / once per 24 hours.
ACTION 72 - With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours. ACTION 73 - With the required monitor inoperable, assure a portable alarming monitor is OPERABLE in the vicinity of the installed monitor or perform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours. , ,O
\d LIMERICK - UNIT 1 3/4 3-65
TABLE 4.3.7.1-1 h RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9 Es OPERATIONAL
- CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE E INSTRUMENTATION CHECK TEST CALIBRATION IS REQUIRED Z
~ 1. Main Control Room Normal Fresh Air Supply Radiation Monitor S M R 1, 2, 3, 5 and *
- 2. Area Monitors
- a. Criticality Monitors
- 1) Spent Fuel Storage S M R (a)
Pool M
** b. Control Room Direct S M R At All Times Y Radiation Monitor
- 3. Reactor Enclosure Cooling Water Radiation Monitor S M R(b) At All Times l
l 9 O O
5 k TABLE 4.3.7.1-1 (Continued) RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS l < *When irradiated fuel is being handled in the secondary containment. , (a) With fuel in the spent fuel storage pool. l (b) The initial CHANNEL CALIBRATION shall be performed using one or more j 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.
[ t 1 1 4 I 4 i i I \~ 3 j ci 1 LIMERICK - UNIT 1 3/4 3-67
INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.2 The seismic monitoring instrumentation shown in Table 3.3.7.2-1* shall be OPERABLE. 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 Report to the Commission pursuant to Specification 6.9.2 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 4.3.7.2.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNC-TIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.2-1. 4.3.7.2.2 Each of the above required seismic monitoring instruments actuated during a seismic event greater than or equal to 0.01g shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 5 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 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect upon unit features important to safety.
- Shared with Unit 2.
O LIMERICK - UNIT 1 3/4 3-68
TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1.- Triaxial Time-History Accelerographs (T/A's)
- a. Sensors
- 1) XE-VA-102 Primary Containment 0 to 1 g 1 Foundation (Loc. 109-R15-177)
- 2) XE-VA-103 Containment Structure 0.to 1 g 1 (Diaphragm Slab)
- 3) XE-VA-104 Reactor Enclosure 0 to 1 g 1 Foundation (Loc. 111-R11-177)
- 4) XE-VA-105 Reactor Piping Support 0 to 1 g 1 (Mn. Stm. Line 'D', El 313',
in containment)
- 5) XE-VA-106 Outside Containment 0 to 1-g 1 on Seismic Category I O Equipment (RHR Heat Exchanger, Loc. 102-R15-177)
- 6) XRSH-VA-1078 Foundation of an 0 to 1 g 1 Independent Seismic Category I Structure (Spray Pond Pump House, El 237')
- b. Recorders (Panel 00C693)
- 1) XR-VA-102 for XE-VA-302 N.A. 1
- 2) XR-VA-103 for XE-VA-103 N.A. 1 3)- XR-VA-104 for XE-VA-104 N.A. 1
- 4) XR-VA-105 for XE-VA-105 N.A. 1
- 5) XR-VA-106 for XE-VA-106 N.A. 1
- Includes sensor, trigger, recorder, and backup power supply, s
LIMERICK - UNIT 1 3/4 3-69
TABLE 3.3.7.2-1 (Continued) SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- c. Triaxial Seismic Trigger (S/T)
- 1) XSH-VA-001 (Activates Items 0 to 1 g 1 1.b.1) thru 5) above)
- 2. Triaxial Peak Recording Accelerograph (P/A's)
- a. XR-VA-151 Reactor Equipment 0-2g 1 (Top of reactor vessel head)
- b. XR-VA-152 Reactor Piping 0-2g 1 (Mn. Stm. Line 'D,' El 313',
in containment)
- c. XR-VA-153 Reactor Equipment Outside 0 -2g 1 Containment (RHR Heat Exchangar, Loc. 203-R15-201)
- 3. Triaxial Seismic Switches
- a. XSHH-VA-001 Primary Containment 0 - 0.15 g Horiz. 1*
Foundation (Loc. 118-R16-117) 0 - 0.10 g Vert.
- 4. Triaxial Response Spectrum Analyzer 1-33.5 Hz 1*, **
(RSA)
*With reactor control room indication and annunciation.
- Receives signal from playback unit fed with data from the Triaxial Accelerographs, Item 1.a above.
LIMERICK - UNIT 1 3/4 3-70
o' , l.f TABLE 4.3.7.2-1 M SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHECK TEST CALIBRATION
- 1. Triaxial Time-History Accelerographs (T/A's)
- a. Sensors
- 1) XE-VA-102 Primary Contain- N.A. SA R ment Foundation <
-(Loc. 109-R15-177)
- 2) XE-VA-103 Containment N.A. SA R Structure (Diaphragm Slab)
- 3) XE-VA-104 Reactor Enclosure N.A. SA R Foundation (Loc. 111-R11-117)
- 4) XE-VA-105 Reactor Piping N.A. SA R Support (Mn Stm. Line 'D,'
El 313', in containment)
- 5) XE-VA-106 Outside Contain- N.A. SA R ment on Seismic Category I Equipment, (RHR Heat Exchanger, Loc. 102-R15-177)
- 6) XRSH-VA-107* Foundation of N.A. SA R an Independent Seismic Category I Structure (Spray Pond Pump House, El 237')
- b. Recorders (Panel 00C693)
- 1) XR-VA-102 for XE-VA-102 N.A. SA R
- 2) XR-VA-103 for XE-VA-103 N.A. SA R
- 3) XR-VA-104 for XE-VA-104 N.A. SA R
- 4) XR-VA-105 for XE-VA-105 N.A. SA R t
- 5) XR-VA-106 for XE-VA-106 N.A. SA R
- Includes sensor, trigger, recorder, and backup power supply.
k a LIMERICK - UNIT 1 3/4 3-71
TABLE 4.3.7.2-1 (Continued) SEI5MIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHECK TEST CALIBRATION
- c. Triaxial Seismic Trigger (S/T)
- 1) XSH-VA-001 (Activities N.A. SA R Itecs 1.b.1) thru 5) above)
- 2. Triaxial Peak Recording Accelerograph (P/A's)
- a. XR-VA-151 Reactor Equipment N.A. N.A. R (Top of reactor vessel head)
- b. XR-VA-152 Reactor Piping N.A. N.A. R (Mn. Stm. Line 'D,' El 313',
in containment)
- c. XR-VA-153 Reactor Equipment N.A. N.A. R Outside Containment (RHR Heat Exchanger, Loc. 203-R15-201)
- 3. Triaxial Seismic Switches
- a. XSHH-VA-001 Primary Containment N.A. SA R Foundation (Loc. 118-R16-177)
- 4. Triaxial Response Spectrum Analyzer N.A. SA R (RSA)
O LIMERICK - UNIT 1 3/4 3-72
1 l l INSTRUMENTATION METEOROLOGICAL MONITORING INSTRUMENTATION LIMITING CON 0ITION FOR OPERATION 3.3.7.3 The meteorological monitoring instrumentation channels shown in Table 3.3.7.3-1 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more meteorological monitoring instrumentation channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
Os SURVEILLANCE REQUIREMENTS 4.3.7.3 Each of the above required meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.3-1. LIMERICK - UNIT 1 3/4 3-73
TABLE 3.3.7.3-1 METEOROLOGICAL MONITORING INSlRUMENTATION MINIMUM Tower 1 Tower 2 INSTRUMENTS INSTRUMENT (Primary) (Backup) OPERABLE
- 1. Wind Speed
- a. Elevation 1 30 feet or 159 feet 1
- b. Elevation 2 175 feet or 304 feet 1
- 2. Wind Direction
- a. Elevation 1 30 feet or 159 feet 1
- b. Elevation 2 175 feet or 304 feet 1
- 3. Air Temperature Difference
- a. Elevations 266 feet- 300 feet-26 feet or 26 feet 1 O
I O LIMERICK - UNIT 1 . 3/4 3-74 1
l' I. ! TABLE 4.3.7.3-1 ! i i METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS f CHANNEL CHANNEL l j-. INSTRUMENT CHECK CALIBRATION ! I. } 1. Wind Speed r 5 l-
- a. Elevation 1 (Tower 1 and Tower 2) D SA ,
j b. Elevation 2 (Tower 1 and Tower 2)- D SA l 2. Wind Direction j a. Elevation 1 (Tower 1 and Tower 2) D SA , i ! b. Elevation 2 (Tower 1 and Tower 2) D- ~ SA I 3. Air Temperature Difference ! l ,
- a. Elevations 266 - 26 ft (Tower 1) D SA
- b. Elevations 300 - 26 ft (Tower 2) D- SA i-l ,
1 i , f r i + i i 4 5 LIMERICK - UNIT 1 3/4 3-75 l 1
-,, -v,,,--,-,.--,. 4.--.- --. . - - - -
INSTRUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS !.IMITING CONDITION FOR OPERATION 3.3.7.4 The remote shutdown system instrumentation and controls shown in Table 3.3.7.4-1 shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
- a. With the number of OPERABLE remote shutdown system instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTOOWN within the next 12 hours.
- b. With the number of OPERABLE remote shutdown system controls less than required in Table 3.3.7.4-1, restore the inoperable control (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
O SURVEILLANCE REQUIREMENTS 4.3.7.4.1 Each of the above required remote shutdown monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.4-1. 4.3.7.4.2 Each of the above remote shutdown control switch (es) and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended function (s) at least once per 18 months. 1 l Oll l LIMERICK - UNIT 1 3/4 3-76
't O O I 1 ! TABLE 3.3.7.4-1 h REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS i 'E i j- g MINIMUM- ,
^ INSTRUMENTS' ', ' INSTRUMENT- OPERABLE
- E' q 1. Reactor Vessel Pressure 1 j - 2. Reactor Vessel Water Level 1 f 3. Safety / Relief Valve Position, 3 valves 1/ valve
) 4. Suppression Chamber Water Level 1 5.. Suppression Chamber Water Temperature (Actually RHR Pump "A" Suction Temperature) 1
- 6. Drywell Pressure 1 i 7. Drywell Temperature 1
, 8. RHR System Flow 1 s
- 9. RHR Service Water Pump Discharge Pressure 1
,_ {"
- 10. RHR Heat Exchanger Service Water Outlet Pressure 1
- 11. RCIC Syst'en Flow .1
, 12. RCIC Turbine Speed- 1 ;
- 13. Emergency Service Water Pump Discharge Pressure 1 l 14. Condensate Storage Tank Level 1 t
I j 15. RHR Heat Exchanger Bypass Valve (HV51-1F048A) Position Indication (0 - 100%) 1 j 16. RCIC Turbine Tripped Indication. 1 ) 17. RCIC Turbine Bearing Oil Pressure Low Indication 1 4 j 18. RCIC LP Bearing 011 Temperature High Indication 1 t j 19. RHR Heat Exchanger Discharge Line High Radiation Indication 1 I 1 i i i
) ! 'w.,
4
TABLE 3. 3. 7.1-1 REMOTE SHUTDOWN SYSTEM CONTROLS RCIC SYSTEM HSS-49-191 Control-Transfer Switch HSS-49-192 Control-Transfer Switch HSS-49-193 Control-Transfer Switch HSS-49-195 Control-Transfer Switch HSS-49-196 Control-Transfer Switch HV-49-1F076 Control-Steam Line warmup bypass valve HV-49-1F060 Control-RCIC turb exhaust to suppression pool isolation HV-50-112 Control-Turb trip throttle valve HV-50-1F045 Control-Turbine steam supply valve HV-49-1F008 Control-Turbine steam line outboard isolation valve HV-49-1F007 Control-Turbine steam line inboard isolation valve HV-49-1F031 Control-RCIC pump suction from suppression pool HV-49-1F029 Control-RCIC pump suction from suppression pool HV-49-1F010 Control-RCIC pump suction from condensate storage tank HV-49-1F019 Control-Minimum flow bypass to suppression pool HV-49-1F022 Control-Test return to condensate storage tank HV-50-1F046 Control-RCIC turbine cooling water valve HV-49-1F012 Control-RCIC pump disch valve HV-49-1F013 Control-RCIC pump disch valve 10P220 Control-Vacuum tank condensate pump 10P219 Control-Barometric condenser vacuum puma HV-49-1F002 Control-Barometric condenser vacuum pump disch 9 LIMERICK - UNIT 1 3/4 3-78
RCIC SYSTEM (Continued) HV-49-1F080 . Control-Vacuum breaker outboard isolation valve HV-49-1F084 Control-Vacuum breaker inboard isolation valve FIC-49-1R001 Controller-RCIC discharge flow control E51-545 RCIC Turbine Trip Bypass NUCLEAR BOILER SYSTEM HSS-41-191 Control-Transfer switch PSV-41-1F013A Control-Main steam line safety / relief valve f PSV-41-1F013C Control-Main steam line safety / relief valve
~PSV-41-1F013N Control-Main steam line safety / relief valve RHR SYSTEM HSS-51-192 Control-Transfer switch O- HSS-51-193 Control-Transfer switch HSS-51-194 Control-Transfer switch HSS-51-195 Control-Transfer switch HSS-51-196 Control-Transfer switch HSS-51-197 Control-Transfer switch '
HSS-51-198 Control-Transfer switch HV-51-1F009 Control-RHR pump shutdown cooling suction inboard j isolation HV-51-1F008 Control-RHR shutdown cooling suction outboard isolation HV-51-1F006A Control-1A RHR loop shutdown cooling suction
.gV-51-1F0068 Control-1B RHR loop shutdown cooling suction HV-51-1F004A Control-IA RHR pump suction 1AP202 Control-1A RHR pump 'u LIMERICK - UNIT 1 3/4 3-79
Table 3.3.7.4-1 (Continued) RHR SYSTEM (Continued) HV-43-1F023A Control-Recirculation pump A suction valve HSS-43-191 Control-Transfer switch HV-51-1F007A Control-1A RHR pump minimum flow bypass valve HV-51-1F048A Control-1A heat exchanger shell side bypass HV-51-1F015A Control-1A shutdown cooling injection valve HV-51-1F022 Control-RHR head spray inboard isolation valve HV-51-1F023 Control-RHR head spray outboard isolation HV-51-1F016A Control-Reactor containment spray HV-51-1F011A Control-1A heat exchanger flow to suppression pool HV-51-1F017A Control-1A RHR loop injection valve HV-51-1F024A Control-1A RHR loop test return HV-51-1F027A Control-Suppression pool sparger isolation HV-51-1F047A Control-1A Heat exchanger shell side inlet HV-51-1F003A Control-1A Heat exchanger shell side outlet HV-51-1F026A Control-1A Heat exchanger flow to RCIC HV-51-1F049 Control-RHR Discharge to radwaste outboard isolation HV-51-125A Control-1A/1C test return line to suppression pool HV-51-]F052A Control-HPCI steam to RHR heat exchanger HV-51-153A Control-HPCI steam to RHR heat exchanger warm-up bypass RHR SERVICE WATER SYSTEM HSS-12-015A-2 Control-Spray' pond / cooling tower select HSS-12-015C-2 Control-Spray pond / cooling tower select HSS-12-016A-2 Control-Spray / bypass select HSS-12-016C-2 Control-Spray / bypass select l l l LIMERICK - UNIT 1 3/4 3-80 l
/>
Table 3.3.7.4-1 (Continued) RHR SERVICE WATER SYSTEM (Continued) HSS-12-094 Control-Transfer switch HSS-12-093 Control-Transfer switch HSS-51-1F014A Control-1A RHR heat exchanger tube side inlet 0AP506 Control-RHR Service Water pump HV-51-1F068A Control-1A RHR Heat exchanger tube side outlet EMERGENCY SERVICE WATER SYSTEM OAP548 Control-1A emergency service water pump HS-11-011A-2 Control-1A emergency service water disch to RHR service water HSS-11-091 Control-Transfer switch HSS-11-092 Control-Transfer switch ]v HSS-11-093 Control-Transfer switch The following valves of the'ESW and RHRSW systems are ac.tuated by signals from the transfer switches: HV-12-005 ESW and RHRSW pumps wetwell intertie gate HV-11-015A ESW loop A discharge to RHRSW loop B HV-12-017A ESW and RHRSW cooling tower return cross-tie STANDBY AC POWER SUPPLY 152-11509/CSR 101-D11 Safeguard SWGR feeder bkr. 152-11609/CSR 101-D12 Safeguard SWGR feeder bkr. 152-11709/CSR 101-013 Safeguard SWGR feeder bkr. 152-11502/CSR 201-011 Safeguard SWGR feeder bkr. 152-11602/CSR 201-D12 Safeguard SWGR feeder bkr. n 152-11702/CSR 201-D13 Safeguard SWGR feeder bkr. 152-11505/CSR D114 Safeguard LC XFMR breaker LIMERICK - UNIT 1 3/4 3-81
Table 3.3.7.4-1 (Continued) STANDBY AC POWER SUPPLY (Continued) 152-11605/CSR D124 Safeguard LC XFMR breaker 152-11705/CSR D134 Safeguard LC XFMR breaker 143-115/CS Transfer switch 143-116/CS Transfer switch 143-117/CS Transfer switch O l O l l LIMERICK - UNIT 1 3/4 3-82 1 1 __ _____- __________ _____-_- - ____l
TABLE 4.3.7.4-1 r_ REMOTE' SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
-Q CHANNEL CHANNEL INSTRUMENT CHECK
- CALIBRATION
' 1. Reactor Vessel Pressure M R-k 2. Reactor' Vessel Water Level M R
[ 3. ' Safety / Relief Valve Position, 3 valves M NA
- 4. Suppression Chamber Water Level M R
- 5. Suppression Chamber Water Temperature M R
- 6. Drywell Pressure M R
- 7. Drywell Temperature M R
- 8. RHR System Flow M R
- 9. RHR Service Water Pump Discharge Pressure M R
$ 10. RHR Heat Exchanger Service Water Outlet Pressure M R Y 11. RCIC System Flow M R O 12. RCIC Turbine Speed M R
- 13. Emergency Service Water Pump Discharge Pressure M R
- 14. Condensate Storage Tank Level M R
- 15. RHR Heat Exchanger Bypass Valve (HV51-1F048A) Position Indication (0 - 100%) M R
- 16. RCIC Turbine Tripped Indication M R
- 17. RCIC Turbine Bearing Oil Pressure Low Indication M R
- 18. RCIC LP Bearing Oil Temperature High Indication M R
- 19. RHR Heat Exchanger Discharge Line High Radiation Indication M R
" Control is not required to be transferred to perform this CHANNEL CHECK.
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1 shall be OPERABLE. APPLICABILITY: As shown in Table 3.3.7.5-1. ACTION: With one or more accident monitoring instrumentation channels inoperable, take the ACTION required by Table 3.3.7.5-1. SURVEILLANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.5-1. O O LIMERICK - UNIT 1 3/4 3-84
h '] i TABLE 3.3.7.5-1 h m ACCIDENT MONITORING INSTRUMENTATION m E MINIMUM APPLICABLE !
- REQUIRED NUMBER CHANNELS OPERATIONAL
' OF CHANNELS OPERABLE CONDITIONS ACTION INSTRUMENT E
Z 1. Reactor Vessel Pressure 2 1 1,2 80
" 2. Reactor Vessel Water Level 2 1 1,2 80
- 3. Suppression Chamber Water Level 2 1 1,2 80
- 4. Suppression Chamber Water Temperature 8, 6 locations 6, 1,2 80 1 location
- 5. Suppression Chamber Air Temperature 1 1 1,2 80
- 6. Drywell Pressure 2 1 1,2 80
- 7. Drywell Air Temperature 1 1 1,2 80 w -
g i 8. Drywell Oxygen Concentration Analyzer 2 1 1,2 80 T 9. Drywell Hydrogen Concentration Analyzer 2 1 1,2 80 co j
- 10. Safety / Relief Valve Position Indicators 1/ valve 1/ valve 1,2 80
# # 81
- 11. Primary Containment Post-LOCA Radiation Monitors 4 2 1,2 ,3
# # 81
- 12. North Stack Wide Range Accident Monitor ** 3* 3* 1,2 ,3
- 13. Neutron Flux 2 1 1,2 80
. l l 1
#Not required to be OPERABLE until initial criticality.
I
Table 3.3.7.5-1 (Continued) ACCIDENT MONITORING INSTRUMENTATION TABLE NOTATIONS
*Three noble gas detectors with overlapping ranges (10 7 to 10 1, 10 4 to 102, 10 2 to 105 pCi/cc).
- High range noble gas monitor.
ACTION STATEMENTS ACTION 80 -
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours,
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
ACTION 81 - With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitor-ing the appropriate parameters within 72 hours, and
- a. Either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or
- b. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
O LIMERICK - UNIT 1 3/4 3-86
g
/ s) '
l m/ sj TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E CHANNEL CHANNEL INSTRUMENT CALIBRATION 7 CHECK g 1. Reactor Vessel Pressure M R Z. 2. Reactor Vessel Water Level M R
- 3. Suppression Chamber Water Level M R
- 4. Suppression Chamber Water Temperature M R
- 5. Suppression Chamber Air Temperature M R
- 6. Primary Containment Pressure M R
- 7. Drywell Air Temperature M R
- 8. Drywell Oxygen Concentration Analyzer M Q q 9. .Drywell Hydrogen Concentration Analyzer M Q*
[ 10. Safety / Relief Valve Position Indicators M R g 11. Primary Containment Post LOCA Radiation Monitors. M R**
- 12. North Stack Wide Range Accident Monitor *** M R
- 13. Neutron Flux M R t
*Using calibration gas containing:
- a. Zero volume percent hydrogen, balance nitrogen.
i
- b. Five volume percent hydrogen, balance nitrogen.
** CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.
***High range noble gas monitors.
#Using calibration gas containing:
- a. Zero volume percent oxygen, balance nitrogen.
- b. Five volume percent oxygen, balance nitrogen.
INSTRUMENTATION SOURCE RANGE MONITORS LIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:
- a. In OPERATIONAL CONDlTION 2*, three.
- b. In OPERATIONAL CONDITION 3 and 4, two.
APPLICABILITY: OPERATIONAL CONDITIONS 2*, 3, and 4. ACTION:
- a. In OPERATIONAL CONDITION 2* with one of the above required source range monitor channels inoperable, restore at least three source range monitor channels to OPERABLE status within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours.
- b. In OPERATIONAL CONDITION 3 or 4 with one or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour.
SURVEILLANCE REQUIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERABLE by:
- a. Performance of a:
- 1. CHANNEL CHECK at least once per:
a) 12 hours in CONDITION 2*, and b) 24 hours in CONDITION 3 or 4.
- 2. CHANNEL CALIBRATION ** at least once per 18 months.
- b. Performance of a CHANNEL FUNCTIONAL TEST:
- 1. Within 24 hours prior to moving the reactor mode switch from the Shutdown position, if not performed within the previous 7 days, and
- 2. At least once per 31 days.
- c. Verifying, prior to withdrawal of control rods, that the SRM count rate is at least 3.0 cps *** with the detector fully inserted.
*With IRM's on range 2 or below.
l
** Neutron detectors may be excluded from CHANNEL CALIBRATION.
j ***May be reduced to 0.7 cps provided the signal-to-noise ratio is > 2. LIMERICK - UNIT 1 3/4 3-88
i INSTRUMENTATION
'y/ TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.7 The traversing in-core probe system shall be OPERABLE with:
- a. Five movable detectors, drives and readout equipment to map the core, and
- b. Indexing equipment to allow all five detectors to be calibrated in a common location.
APPLICABILITY: When the traversing in-core probe is used for: I
- a. Recalibration of the LPRM detectors, and b.* Monitoring the APLHGR, LHGR, MCPR, or MFLPD.
ACTION: With the traversing in-core probe system inoperable, suspend use of the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
\
SURVEILLANCE REQUIREMENTS 4.3.7.7 The traversing in-core probe system shall be demonstrated OPERABLE by normalizing each of the above required detector outputs within 72 hours prior to use for the LPRM calibration function.
*0nly the detector (s) in the required measurement location (s) are required to be OPERABLE.
.Q LIMERICK - UNIT 1 3/4 3-89
INSTRUMENTATION CHLORINc ou. 1UN SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8.1 Two independent chlorine detection system subsystems shall be OPERABLE with their alarm and trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 0.5 ppm. APPLICABILITY: All OPERATIONAL CONDITIONS. ACTION:
- a. With one chlorine detection subsystem inoperable, restore the inoperable detection system to OPERABLE status within 7 days or, within the next 6 hours, initiate and maintain operation of at least one control room emergency filtration system subsystem in the chlorine isolation mode of operation.
- b. With both chlorine detection subsystems ineparable, within 1 hour initiate and maintain operation of at least one control room emer-gency filtration system subsystem in the chlorine isolation mode of operation.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.8.1 Each of the above required chlorine detection system subsystems shall be demonstrated OPERABLE by performance of a:
- a. CHANNEL CHECK at least once per 12 hours, .
- b. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- c. CHANNEL CALIBRATION at least once per 18 months.
O LIMERICK - UNIT 1 3/4 3-90
s INSTRUMENTATION T0XIC GAS DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8.2 Two independent toxic gas detection system subsystems shall be OPERABLE with their alarm setpoints adjusted to actdate at a toxic gas concen-tration of less than or equal to: MONITOR SET POINT CHEMICAL (ppm) Ammonia 25 Ethylene Oxide 50 Formaldehyde 5
-Vinyl Chloride 10 Phosgene 0.4 APPLICABILITY: All OPERATIONAL CONDITIONS.
ACTION: [^ a. With one toxic gas detection subsystem inoperable, restore the ; i- inoperable detection system to OPERABLE status within 7 days or, within the next 6 hours, initiate and maintain operation of at least one control room emergency filtration system subsystem in the T chlorine isolation mode of operation.
- b. With both toxic gas detection subsystems inoperable, within 1 hour initiate and maintain operation of at least one control room emer-gency filtration system subsystem in the chlorine isolation mode of operation,
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.8.2 Each of the above required toxic gas detection system subsystems shall.be demonstrated OPERABLE by performance of a:
- a. CHANNEL CHECK at least once per 12 hours, l b. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- c. CHANNEL CALIBRATION at least once per.18 months.
LIMERICK - UNIT 1 3/4 3-91 L v . - - .
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.9 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3.7.9-1 chall be OPERABLE. APPLICABILITY: Whenever equipment protected oy the fire detection instrument is required to be OPERABLE. ACTION:
- a. With the number of OPERABLE fire detection instruments in one or more zones:
- 1. Less than, but more than one-half of, the Total Number of Instruments shown in Table 3.3.7.9-1 for Function A, restore the inoperable Function A instrument (s) to OPERABLE status within 14 days or within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside an inaccessible zone, then inspect the area surrounding the inaccessible zone at least once per hour.
- 2. One less than the Total Number of Instruments shown in Table 3.3.7.9-1 for Function D, or one-half or less of the Total Number of Instruments shown in Table 3.3.7.9-1 for Function A, or with any two or more adjacent instruments inoperable, within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located insiae an inaccessible zone, then inspect the area surrounding the inaccessible zone at least once per hour,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.9.1 Each of the above required fire detection instruments which are accessible during unit operation shall be demonstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST. Fire detectors which are not accessible during unit operation shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months. 4.3.7.9.2 The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. O LIMERICK - UNIT 1 3/4 3-92
7; x TABLE 3.3.7.9-1
-t i
\__ / FIRE DETECTION INSTRUMENTATION INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS
- FIRE.
ZONE STRUCTURE ELEV. AREA HEAT SM0KE FLAME (x/y) (x/y) (x/y) IL Control 200' Control Structure Chillers and NA 3/0 NA Chilled Water Pump Area 258 1M Control 200' Control Structure Chillers and NA 3/0 NA Chilled Water Pump Area 263 2 Control 217' 13-kV Switchgear Area 336 NA' 34/0 NA 3 Control 217' Battery Room 323 (10) 1/0 1/0 NA 4 Control 217' Battery Room 324 (1C) 1/0 1/0 NA r 7 Control 239' Corridor 437 NA' 5/0 NA 8 Control 239' Battery Room 425 (181/182) 1/0 2/0 NA 9 Control 239' Battery Room 436 (1A1/1A2) 1/0 2/0 NA 12 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 434 (D13) 7-g 13 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 435 (011) g } 14 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 432 (D14) 15 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 433 (012) 20 Control 254' Static Inverter Room Unit 1, NA 4/0 NA Area 452 22 Control 254' Cable Spreading Room Unit 1, NA 14/0 NA Area 449 24A Control 269' Control Room 533 NA 23(a)/0 NA 11(b)/0 24B Control 269' Control Room Utility Room 529 NA 1/0 NA 24C Control 269' Control Room Office 531 NA 1/0 NA 24D Control 269' Control Room Shift Supt. 532 NA 1/0 NA 24E Control 269' Control Room Shop 534 NA 1/0 NA (Photo-Elect) 24F Control 269' Control Room Instrument NA 1/0 NA Lab 535 (Photo-Elect) [\s_-)/ 24G Control 269' Control Room Shift Supt. 532A NA 1/0 NA l LIMERICK - UNIT 1 3/4 3-93
TABLE 3.3.7.9-1 (Continued) FIRE DETECTION INSTRUMENTATION INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS
- FIRE ZONE STRUCTURE ELEV. AREA HEAT SM0KE FLAME (x/y) (x/y) (x/y) 25 Control 289' Auxiliary Equipment Room 542 0/112 57/0 NA (PGCC (Ceiling)
Floor) 56/0 (PGCC Floor) 0/15 14/0 (twn- (Non-PGCC PGCC Floor) Floor) 32/0 (Terminal Cabinets) 26 Control 289' Remote Shutdown Panel Area 540 0/4 3/0 NA (Non- (Ceiling PGCC Level) Floor) 2/0 (Non-PGCC Floor) 27 Control 304' Control Structure 0/23 10/0 NA Fan Room 619 4/0 (inside plenum) 28A Control 332' SGTS Access Area 625 (SGTS 4/0 NA NA Room Ventilation Exhaust) (inside plenum) 28B Control 332' SGTS Filter Compartment 624 4/0 NA NA (inside plenum) 28C Control 332' Control Room Fresh Air NA 3/0 NA Intake Plenum 31 Unit 1 177' RHR Heat Exchanger & NA 6/0 NA Reactor Pump Room 103 (B&D) 32 Unit 1 177' RHR Heat Exchanger & NA 5/0 NA Reactor Pump Room 102 (A&C) 33 Unit 1 177' RCIC Pump Room 108 0/3 2/0 NA Reactor 34 Unit 1 177' HPCI Pump Room 109 0/4 3/0 NA Reactor 35 Unit 1 177' 'A' Core Spray Pump NA 2/0 NA l Reactor Room 110 i LIMERICh - UNIT 1 3/4 3-94 1
i l l )
- [']
TABLE 3.3.7.9-1 (Continued) C FIRE DETECTION INSTRUMENTATION INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS
- FIRE ZONE STRUCT!!RE ELEV. AREA HEAT SM0KE FLAME (x/y) (x/y) (x/y) 36 Unit 1 177' 'C' Core Spray Pump NA 2/0 NA Reactor Room 113 37 Unit 1 177' 'D' Core Spray Pump NA 2/0 NA Reactor Room 114 38 Unit 1 177' 'B' Core Spray Pump NA 2/0 NA Reactor Room 117 39 Unit 1 177' Sump Room 115; NA 4/0 NA Reactor Passageway 118 40 Unit 1 177' Corridor III NA 2/0 NA Reactor 41 Unit 1 201' RECW Equipment Area 207 0/10 3/0 NA Reactor 42A Unit 1 201' Safeguard System Access 0/12 3/0 NA Reactor Area 200
- [N 43 Unit 1 217' Safeguard System Isolation NA 8/0 NA
( 44 Reactor Unit 1 217' Valve Area 309 Safeguard System Access 0/8 27/0 NA Reactor Area 304 (Southwest) 0/14 (Northeast) 45A Unit 1 253' CRD Hydraulic Equipment 0/16 20/0 NA Reactor Area 402 45B Unit 1 253' Neutron Monitoring 0/2 2/0 NA Reactor System Area 406 45C Unit 1 253' CRD Repair Room 403 NA 1/0 NA Reactor , 47A Unit 1 283' Corridor 506; General 0/18 21/0 NA Reactor Equipment Area 500 47B Unit 1 295' Isolation Valve NA 2/0 NA Reactor Compartment 523 47C Unit 1 283' Fuel Pool Cooling Water NA 2/0 NA Reactor Pump and Heat Exchanger Area 511 470 Unit 1 283' Isolation Valve NA 1/0 NA Reactor Compartment 510/522
/^\
(")
#0nly 13 of these heat detectors are required to be OPERABLE until prior to exceeding 5% of RATED THERMAL POWER.
##Not required to be OPERABLE until prior to exceeding 5% of RATED THERMAL POWER.
LIMERICK - UNIT 1 3/4 3-95 i
TABLE 3.3.7.9-1 (Continued) FIPE DETECTION INSTRUMENTATION INSTRbMENT LOCATION TOTAL NUMBER OF INSTRUMENTS
- FIRE ZONE STRUCTURE ELEV. AREA HEAT SM0KE FLAME (x/y) (x/y) (x/y) 48A Unit 1 313' Laydown Areas 601 and 602; NA 8/0 NA Reactor Corridor and RERS Fan Area 605 51A Unit 1 331' RERS Filter 2/0 NA NA Reactor Compartment 618 (inside plenum) 51B Unit 1 331' RERS Filter 2/0 NA NA Reactor Compartment 612 (inside plenum) 79 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1 80 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1 81 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1 82 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1 122A Spray 268' ESW and RHRSW Pump Area NA 4/0 NA Pond Pump Structure 122E Spray 251' RHRSW Valve Compartment NA 2/0 NA Pond Pump Structure 123A Spray 268' ESW and RHRSW Pump Area NA 4/0 NA Pond Pump Structure 123E Spray 251' RHRSW Valve Compartment NA 2/0 NA Pond Pump Structure 124A Diesel- 217' Diesel-Generator Access NA 4/0 NA Generator Corridor 313 126A Common 412' North Stack Instrument NA 2/0 NA Reactor Room 712 0 (x/y): X is the number of Function A (Early Warning Fire Detection and Notification Only) Instruments.
Y is the number of Function B (Activation of Fire Suppression System and Early Warning Notification) Instruments. (a) These smoke detectors are located below the suspended ceiling in the Ccntrol' Room. (b) These smoke detectors are located above the suspended ceiling in t'e n Control Room. LIMERICK - UNIT 1 3/4 3-96
d r~s INSTRUMENTATION
/ i 3 \ e)
~- LOOSE-PART DETECTION SYSTEM 1
LIMITING CONDITION FOR OPERATION 1 t 3.3.7.10 The loose part detection system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
, a. With one or more loose part detection system channels inoperable for
!, more than 30. days, prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 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 Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
~
(h
\~~-
4.3.7.10 Each channel of the loose part detection system shall be demonstrated OPERABLE by performance of a:
- - .a. CHANNEL CHECK at least once per 24 hours,
- b. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- c. CHANNEL CALIBRATION at least once per 18 months.
i o 1 l [ l LIMERICK - UNIT 1 3/4 3-97
INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.11 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their alarm / trip setpoints set to e.1sure 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). APPLICABILIITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitort.d by the affected channel, or declare the channel inoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.11-1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.11 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.11-1.
- Excluding the flow rate measuring devices which are not determined and adjusted in accordance with the ODCM.
LIMERICK - UNIT 1 3/4 3-98
4 i TABLE 3.3.7.11-1 + It RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION I
. MINIMUM 1 CHANNELS j- INSTRUMENT- OPERABLE ACTION 1
- 1. GROSS RADI0 ACTIVITY MONITORS PROVIDING ,
a [ AUTOMATIC TERMINATION OF RELEASE
- a. Liquid Radwaste Effluent Line 1 100
- b. RHR Service Water System Effluent Line 1/ loop 101
- 2. GROSS RADI0 ACTIVITY MONITORS NOT I PROVIDING AUTOMATIC TERMINATION OF RELEASE
{
- a. Service Water System Effluent Line 1 101 I
! 3. FLOW. RATE MEASUREMENT DEVICES
- a. Liquid Radwaste Effluent Line 1 102 1.
l b. Discharge Line 1 102 i i t l- l 1 e ? I' i I i i-LIMERICK - UNIT 1 3/4 3-99
TABLE 3.3.7.11-1 (Continued) ACTION STATEMENTS ACTION 100 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 14 aays provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Specification 4.11 1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 101 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 8 hours, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10 7 microcurie /mL. ACTION 102 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump curves generated in situ may be used to estimate flow. O LIMERICK - UNIT 1 3/4 3-100
rw s O J O. l TABLE 4.3.7.11-1 C '
-M RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS S
- S
, CHANNEL :
c CHANNEL SOURCE CHANNEL FUNCTIONAL ! 5 INSTRUMENT CHECK CHECK CALIBRATION TEST i
;
- 1. GROSS RADI0 ACTIVITY MONITORS PROVIDING AUTOMATIC TERMINATION OF RELEASE i
- a. Liquid Radwaste Effluent Line P P R(3) Q(1)
- b. RHR Service Water System Effluent Line D M R(3) Q(1)
- 2. GROSS RADI0 ACTIVITY MONITORS NOT PROVIDING i ^
..UTOMATIC TERMINATION OF RELEASE
- a. Service Water System Effluent Line D M R(3) Q(2) l' w !
4 3. FLOW RATE MEASUREMENT DEVICES S -
- a. Liquid Radwaste Effluent Line D(4) N.A. R Q l
- b. Discharge Line D(4) N.A. R Q
TABLE 4.3.7.11-1 (Continued) TABLE NOTATIONS (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that 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. (4) CHANNEL 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. O LIMERICK - UNIT 1 3/4 3-102
s - INSTRUMENTATION l
-(\s_/ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.12 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.7.12-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints* of the applicable channels chall be determined in accordance with the methodology and parameters in the ODCM.
APPLICABILITY: As shown in Table 3.3.7.12-1 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the i
above Specification, immediately suspend the release of radioactive gaseous affluents monitored by the affected channel or declare the channel inoperable. , b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown i [} ( ,,f in Table 3.3.7.12-1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain why this inoperability was not corrected in a timely manner in the next Semiannual Radioactive Effluent Release Report. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
~
c. SURVEILLANCE REQUIREMENTS i i 4.3.7.12 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.12-1.
- *The alarm / trip setpoints for the Main Condenser Offgas Treatment System l' Explosive Gas Monitoring System and the Main Condenser Offgas Pretreatment g' Radiation Monitor are set in accordance with Specification 3.11.2.5 and j
g 3.11.2.6, respectively. t LIMERICK - UNIT 1 3/4 3-103
h TABLE 3.3.7.12-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION c5 MINIMUM CHANNELS $ INSTRUMENT OPERABLE APPLICABILITY ACTION ~
- 1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
- a. Hydrogen Monitor 1 110
- 2. SOUTH STACK EFFLUENT MONITORING SYSTEM
- a. Noble Gas Activity Monitor 1 111 M *
= b. Iodine Sampler 1 112
- c. Particulate Sampler 1 112
- d. Effluent System Flow Rate Monitor 1 113
- e. Sampler Flow Rate Monitor 1 113
- 3. NORTH STACK EFFLUENT MONITORING SYSTEM
- a. Noble Gas Activity Monitor 1 114
- b. Iodine Sampler 1 112
- c. Particulate Sampler 1 112
- d. Effluent System Flow Rate Monitor 1 113
- e. Sampler Flow Rate Monitor 1 113 O O O
.___..m... 4.-... . . . _.
1 TABLE 3.3.7.12-1 (Continued) c-y RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 2 n . ,
' MINIMUM CHANNELS i INSTRUMENT OPERABLE APPLICABILITY . ACTION i E i Z 4. MAIN CONDENSER OFFGAS PRE-TREATMENT l - RADI0 ACTIVITY MONITOR
- a. Noble Gas Activity Monitor' 1. ** 115 ;
- 5. HOT MAINTENANCE SHOP VENTILATION EXHAUST RADIATION MONITOR.
1 a. Iodine Sampler 1 *** 112 , b. Particulate' Sampler 1 *** 112 i Y i w
- c. Effluent System Flow Rate Monitor 1 *** 113 j . t i
U d. Sampler Flow Rate Monitor 1 *** 113
- l w -
t l ! i l i l 2 I I I l 4 ! o i ! l l 1 } l 1
TABLE 3.3.7.12-1 (Continued) TABLE NOTATIONS
*At all times. **During operation of the main condenser steam jet air ejector and offgas treatment system.
C**During operation of the hot maintenance shop ventilation exhaust system. ACTION STATEMENTS ACTION 110 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser offgas treatment system may continue for up to 30 days provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. ACTION 111 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours. ACTION 112 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided samples are con-tinuously collected with auxiliary sampling equipment as required in Table 4.11.2.1.2-1. ACTION 113 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. ACTION 114 - With the number of char.nels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours and these samples are analyzed for gross activity within 24 hours and provided the mechanical vacuum pumps are not operated. ACTION 115 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, releases to the environment may continue for up to 72 hours provided that the North Stack Effluent Noble Gas Activity Monitor is OPERABLE; otherwise, be in at least HOT SHUTDOWN within 12 hours. O LIMERICK - UNIT 1 3/4 3-106
rN [N -{N
\,
TABLE 4.3.7.12-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS e: o CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL . SURVEILLANCE : lE . INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED w
- 1. MAIN CONDENSER OFFGAS TREATMENT,.
SYSTEM EXPLOSIVE GAS MONITORING SYSTEM
- a. Hydrogen Monitor N.A. M **
D Q(3)
- 2. SOUTH STACK EFFLUENT MONITORING SYSTEM
- a. Noble Gas Activity Monitor. D M R(2) Q(1) s
[ b. Iodine Sampler W (4) N.A. N.A. N.A.
- i t Ej c. Particulate Sampler W (4) N.A. N.A. N.A.
- d. Effluent System Flow Rate Monitor D N.A. R Q
- e. Sampler Flow Rate Monitor D N.A. R Q
- 3. NORTH STACK EFFLUENT !
MONITORING SYSTEM
- a. Noble Gas Activity Monitor D M R(2) Q(1)
- b. Iodine Sampler W (4) N.A. N.A. N.A.
- c. Particulate Sampler W (4) N.A. N.A. N.A.
- d. Effluent System Flow Rate Monitor D N.A. R Q
- e. Sampler Flow Rate Monitor D N.A. R Q w
TABLE 4.3.7.12-1 (Continued) 3! RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 n
' CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE E INSTRUMENT CHECK CHECK CALIBRATION TEST ^ IS REQUIRED Z
s 4. MAIN CONDENSER OFFGAS PRE-TREATMENT l RADI0 ACTIVITY MONITOR (STEAM JET AIR EJECTOR) i
- a. Noble Gas Activity Monitor 0 M R(2) Q(1) **
- 5. HOT MAINTENANCE SHOP VENTILATION EXHAUST RADIATION MONITOR
- a. Iodine Sampler W(4) N.A. N.A. N.A. ***
- b. Particulate Sampler W(4) N.A. N.A. N.A. ***
l
$ c. Effluent System Flow Rate j
f Monitor D N.A. R Q j d. Sampler Flow Rate Monitor D N.A. R Q i { l 9 - O O
+
1 4 (y-L sm ,/
)
TABLE 4.3.7.12-1 (Continued) TABLE NOTATIONS l
- At all. times.
** During operation of the main condenser steam jet air ejector and offgas 4
treatment system.
*** During operation of'the hot maintenance shop ventilation exhaust system. ,
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that' control room alarm annunciation occurs if any of the following conditions. exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
'(2) The initial-CHANNEL CALIBRATION shall be performed using one or more of 4
the reference standards certified by the National Bureau of Standards (NBS) e or using standards that have been obtained from suppliers that participat'e in measurement assurance activities with NBS. These standards shall permit i f ~se calibrating the system over its intended range of energy and measurement -( j\s_-) range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (3) The CHANNEL CALIBRATION shall include the use of standard gas samples . containing a nominal: l 1. 0.0 volume percent hydrogen, balance nitrogen, and i
- 2. 4 volume percent hydrogen, balance nitrogen.
' (4) The iodine cartridges and particulate filters will be changed at least once per 7 days. 4 r N l .' i . LIMERICK - UNIT 1 3/4 3-109
l INSTRUMENTATION 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.8 At least one turbine overspeed protection system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
- a. With one turbine control valve and/or one turbine stop valve per high pressure turbine steam lead inoperable and/or with one turbine combined intermediate valve per low pressure turbine steam lead inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours or close at least one valve in the affected steam lead (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 4.3.8.1 The provisions of Specification 4.0.4 are not applicable. 4.3.8.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Cycling each of the following valves through at least one complete cycle from the running position:
a) For the overspeed protection control system;
- 1) Six low pressure turbine intercept valves b) For the electrical overspeed trip system and the mechanical overspeed trip system;
- 1) Four high pressure turbine stop valves, and
- 2) Six low pressure turbine intermediate stop valves.
O LIMERICK - UNIT 1 3/4 3-110
~
i INSTRUMENTATION ( (s_, SURVEILLANCE REQUIREMENTS (Continued) . b. At least once per 31 days by:
- 1. Cycling each of the following valves through at least one complete cycle from the running position:
$ a) For the overspeed protection control system;
- 1) Four high pressure turbine control valves b) For the electrical overspeed trip system and the mechanical overspeed trip system; 4
3 1) Four high pressure turbine control valve
- - c. At least once per 18 months by performance of a CHANNEL CALIBRATION of the turbine overspeed protection instrumentation.
- 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-all valve seats, disks and stems and verifying no unacceptable flaws i or excessive corrosion. If unacceptable flaws or excessive corrosion are found, all other valves of that type shall be inspected.
~ {'"'
4 V L 4 [ t 1 4 l\ ? LIMERICK - UNIT 1 3/4 3-111 1 1
INSTRUMENTATION 3/4.3.9 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The feedwater/ main turbine trip system actuation instrumentation channels shown in Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2. 1 APPLICABILITY: As shown in Table 3.3.9-1. ACTION:
- a. With a feedwater/ main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, declare the channel inoper-able and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip set-point adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable.
- b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours.
- c. With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status within 72 hours or be in at least STARTUP within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.3.9.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.9.1-1. 4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. O LIMERICK - UNIT 1 3/4 3-112
i i 4 . l; t l TABLE 3.3.9-1 l'
- j. FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION i
i MINIMUM OPERABLE APPLICABLE ':j CHANNELS PER OPERATIONAL -
; TRIP FUNCTION TRIP SYSTEM CONDITIONS
- 1. Reactor Vessel Water l1 Level-High, Level 8 4 1 1
i F i i I
?
I t LIMERICK - UNIT 1 3/4 3-113
l
~ TABLE 3.3.9-2 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE
- 1. Reactor Vessel Water Level-High, Level 8 5 54 inches * $ 55.5 inches l
l l
- See Bases Figure B 3/4.3-1 O
O LIMERICK - UNIT 1 3/4 3-114
l
< TABLE 4.3.9.1-1 x, / FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION t SURVEILLANCE REQUIREMENTS OPERATIONAL CONDITIONS CHANNEL FOR WHICH
-, CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE , TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED i 1. Reactor Vessel Water D M R 1 4 Level-High, level 8 l 1 i 1
, U 4
4 e i i r
\
LIMERICK - UNIT 1 3/4 3-115
.----.-.m - - - - - - - - - - - - - - , - - y-,- ,- ,-,y -
,Q 3/4.4 REACTOR COOLANT SYSTEM b] 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:
- a. Total core flow greater than or equal to 45% of rated core flow, or
- b. THERMAL POWER less than or equal to the limit specified in Figure 3.4.1.1-1.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*. ACTION:
- a. With one reactor coolant system recirculation loop not in operation, immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours and initiate measures to place the unit in at least HOT SHUTDOWN within 12 hours.
- b. With no reactor coolant system recirculation loops in operation, d immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours and initiate measures to place the unit in at least STARTUP within 6 hours and in HOT SHUTDOWN within the next 6 hours.
- c. With two reactor coolant system recirculation loops in operation and total core flow less than 45% of rated core flow and THERMAL POWER greater than the limit specified in Figure 3.4.1.1-1:
- 1. Determine the APRM and LPRM** noise levels (Surv'eillance 4.4.1.1.3):
a) At least once per 8 hours, and b) Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.
- 2. With the APRM or LPRM** neutron flux noise levels greater than three times their established baseline noise levels, immediately initiate corrective action to restore the noise levels to within the required limits within 2 hours by increasing core flow to greater than 45% of rated core flow or by reducing THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1.
*See Special Test Exception 3.10.4.
[] (,) ** Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored. LIMERICK - UNIT 1 3/4 4-1
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup* prior to THERMAL POWER exceeding 25% of RATED THERriAL POWER. 4.4.1.1.2 Each pump MG set sccop tube mechanical and electrical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 105% and 102.5%, respectively, of rated core flow, at least once per 18 months. 4.4.1.1.3 Establish a baseline /PRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage. O
*If not performed within the previous 31 days.
- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
LIMERICK - UNIT 1 3/4 4-2
.i i i f
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f \ v (031YW %) U3 mod 1VWH3H13HOO LIMERICK - UNIT 1 3/4 4-3
REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours. SURVEILLANCE REQUIREMENTS 4.4.1.2 Each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once per 24 hours
- by dete'rmining recirculation loop flow, total core flow and dif f user-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumps are operating at the same speed.
- a. The indicated recirculation loop flow differs by more than 10% from the established pump speed-loop flow characteristics.
- b. The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.
- c. The indicated diffuser-to-lower plenum differential pressure of any individual jet punip dif fers from the established patterns by more than 10%.
- During the startup test program, data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships.
Comparisons of the actual data in accordance with the criteria listed shall commence upon the conclusion of the startup test program. O LIMERICK - UNIT 1 3/4 4-4
_ _ _ _ . = ._ _ _ _ _ _ _ - . _ - _ . . . m _ . _ _ . - _ _ _ ._.. . . _ J
! REACTOR COOLANT SYSTEM f
\ RECIRCULATION PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation pump speed shall be maintained within:
- a. 5% of each other with core flow greater than or equal to 70% of rated core flow,
- b. 1(%E of each other with core flow less than 70% of rated core flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2^. ACTION: With the recirculation pump speeds different by more than the specified limits, either: l
- a. Restore the recirculation pump speeds to within the specified limit
- j. within 2 hours, or
! b. Declare the recirculation loop of the pump with the slower speed not I
(g in operation and take the ACTION required by Specification 3.4.1.1.
\ /
SURVEILLANCE REQUIREMENTS
- 4.4.1.3 Recirculation pump speed shall be verified to be within the limits
! at least once per 24 hours. i j *See Special Test Exception 3.10.4. f h i t lO v 1 LIMERICK - UNIT 1 3/4 4-5
- - . . . ... _...,.,.-.,,._,--,-,.,.,,,----.-,--.,,,,-,-__,...a..-.-..-_., ,.- - . , - - . . - . , . , - . - - - - - - .. , _ , - - , ,-
i
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REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 145 F, and:
- a. When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 50 F, or
- b. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recircula-tion loops is less than or equal to 50 F and the operating loop flow rate is less than or equal to 50% of rated loop flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4. ACTION: With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop. SURVEILLANCE REQUIREMENTS 4.4.1.4 The temperature differentials and flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idle recirculation loop. O LIMERICK - UNIT 1 3/4 4-6
REACTOR COOLANT SYSTEM ,( 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*# l 4 safety / relief valves @ 1130 psig +1% 5 safety /reliefvalves@1140psig11% 5 safety / relief valves @ 1150 psig +1% APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With the safety valve function of one or more of the above required
' safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- b. With one or more safety / relief valves stuck open, provided that suppres-sion pool average water temperature is less than 105 F, close the stuck open safety / relief valve (s); if unable to close the stuck open valve (s) within 2 minutes or if suppression pool average water temperature is 110 F or greater, place the reactor mode switch in the Shutdown position.
- c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and f] in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise level by performance of a:
- a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and a
- b. CHANNEL CALIBRATION at least once per 18 months **.
4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 18 months, and they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations tested at least once per 40 months.
*The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
**The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
- #Up to 2 inoperable valves may be replaced with spare OPERABLE valves with
( ,/ lower setpoints until the next refueling. ' ## Initial setting shall be in accordance with the manufacturer's recommendation. i Adjustment to the valve full open noise level shall be accomplished during the startup test program. ' LIMERICK - UNIT 1 3/4 4-7
REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant sy. i leakage detection systems shall be OPERABLE:
- a. The primary containment atmosphere gaseous radioactivity monitoring system,
- b. The drywell floor drain sump and drywell equipment drain tank flow monitoring system,
- c. The drywell unit coolers condensate flow rate monitoring system, and
- d. The primary containment pressure and temperature monitoring system.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.* ACTION: With only three 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 radioactive monitoring system, primary containment pressure and temperature monitoring system and/or the drywell unit coolers condensate flow rate monitoring system is ino;'erable; otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:
- a. Primary containment atmosphere gaseous radioactivity monitoring systems performance of a CHANNEL CHECK at least once per 12 hours, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
- b. The primary containment pressure shall be monitored at least once per 12 hours and the primary containment temperature shall be monitored at least once per 24 hours.
- c. Drywell floor drain sump and Drywell equipment drain tank flow monitor-ing system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.
- d. Drywell unit coolers condensate flow rate monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
CThe primary containment atmosphere gaseous radioactivity monitor is not required to be OPERABLE until OPERATIONAL CONDITION 2. LIMERICK - UNIT 1 3/4 4-8
[ 'S REACTOR COOLANT SYSTEM s ) OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
- a. No PRESSURE B0UNDARY LEAKAGE.
- b. 5 gpm UNIDENTIFIED LEAKAGE.
- c. 30 gpm total leakage.
- d. 25 gpm total leakage averaged over any 24-hour period,
- e. 1 gpm leakage at a reactor coolant system pressure of 950 110 psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.
, APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
/~% b. With any reactor coolant system leakage greater than the limits in b.
('" ) and/or c., above, reduce the leakage rate to within the limits within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 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 one other closed manual, deactivated automatic, or check
- valves, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- d. With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours; restore the inoperable monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
*Which have been verified not to exceed the allowable leakage limit at the last i refueling outage or after the last time the valve was disturbed, whichever is more recent.
O
> \
(l l LIMERICK - UNIT 1 3/4 4-9
SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
- a. Monitoring the primary containment atmospheric gaseous radioactivity at least once per 12 hours (not a means of quantifying leakage),
- b. Monitoring the drywell floor drain sump and drywell equipment drain tank flow rate at least once per 12 hours,
- c. Monitoring the drywell unit coolers condensate flow rate at least once per 12 hours,
- d. Monitoring the primary containment pressure at least once per 12 hours (not a means of quantifying leakage),
- e. Monitoring the reactor vessel head flange leak detection system at least once per 24 hours, and
- f. Monitoring the primary containment temperature at least once per 24 hours (not a means of quantifying leakage).
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
- a. At least once per 18 months, and
- b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3. 4.4.3.2.3 The high/ low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints set less than the allowable values in Table 3.4.3.2-1 by performance of a:
- a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- b. CHANNEL CALIBRATION at least once per 18 months.
O LIMERICK - UNIT 1 3/4 4-10
.. , --- . , . . - _ . - - .. ._ - . ~ - , . . . . . _ - . _- ., - _ . - .
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$ TABLE 3.4.3.2-1 1 5 i Q REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES h ALARM ALARM i 5 IST ISOLATION 2ND ISOLATION SETPOINT ALLOWABLE H
VALVE (S) NUMBER (S) VALVE (S) NUMBER (S) (psia) 'VALUE (psia) SERVICE ' HV-52-1F006A -HV-52-1F005 5 475 5 495 'A' Core Spray Injection HV-52-1F039A l ! HV-52-1F006B HV-52-108 < 475 < 495 'B' Core Spray /HPCI . HV-52-1F039B Injection I HV-51-1F041A HV-51-1F017A s.400 $ 420 'A' LPCI Injection t HV-51-142A
$ HV-51-1F041B HV-51-1F0178 5 400 $ 420 'B' LPCI Injection
- , HV-51-142B l HV-51-1F041C HV-51-1F017C i 400 5 420 'C' LPCI Injection 1 HV-51-142C HV-51-1F041D HV-51-1F017D $ 400 5 420 'D' LPCI Injection
] HV-51-142D I HV-51-1F022 HV-51-1F023 < 400 < 420 Head Spray , l t i HV-51-1F050A HV-51-1F015A -< 400 -< 420 'A' Shutdown Cooling i HV-51-151A Return to 'A' Recirc Loop ! HV-51-1F050B HV-51-1F015B -< 400 -< 420 'B' Shutdown Cooling > HV-51-151B Return to 'B' Recirc Loop 4 ! HV-51-1F009 HV-51-1F008 ~'< 125 -< 145 Shutdown Cooling Supply l From 'B' Recirc Loop 1 i
REACTOR COOLANT SYSTEM 3/4.4.4 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.4 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.4-1. APPLICABILITY: At all times. ACTION:
- a. In OPERATIONAL CONDITION 1:
- 1. With the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.4-1 for less than 72 hours during one continuous time interval and, for conductivity and chloride concentration, for less than 336 hours per year, but with the conductivity less than 10 pmho/cm at 25 C and with the chloride concentration less than 0.5 ppm, this need not be reported to the Commission and the provisions of Specification 3.0.4 are not applicable.
- 2. With the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.4-1 for more than 72 hours during one continuous time interval or with the conductivity and chloride concentration exceeding the limit specified in Table 3.4.4-1 for more than 336 hours per year, be in at least STARTUP within the next 6 hours.
- 3. With the conductivity exceeding 10 pmho/cm at 25 C or chloride concentration exceeding 0.5 ppm, be in at least HOT SHUTOOWN within 12 hours and in COLD SHUTOOWN within the next 24 hours,
- b. In OPERATIONAL CONDITION 2 and 3 with the conductivity, chloride concentration or pH exceeding,the limit specified in Table 3.4.4-1 for more than 48 hours during one continuous time interval, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- c. At all other times:
- 1. With the:
a) Conductivity or pH exceeding the limit specified in Table 3.4.4-1, restore the conductivity and pH to within the limit within 72 hours, or b) Chloride concentration exceeding the limit specified in Table 3.4.4-1, restore the chloride concentration to within the limit within 24 hours, or 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 structural integrity of the reactor coolant system remains acceptable for continued operation prior to proceeding to OPERATIONAL CONDITION 3.
- 2. .The provisions of Specification 3.0.3 are not applicable.
LIMERICK - UNIT 1 3/4 4-12
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS l 4.4.4 The reactor coolant shall be determined to be within the specified chemistry limit by:
- a. Measurement prior to pressurizing the reactor during each startup, if not performed within the previous 72 hours.
- b. Analyzing a sample of the reactor coolant for:
- 1. Chlorides at least once per:
a) 72 hours, and b) 8 hours whenever conductivity is greater than the limit in Table 3.4.4-1.
- 2. Conductivity at least once per 72 hours.
- 3. pH at least once per:
a) 72 hours, and b) 8 hours whenever conductivity is greater than the limit (O v) in Table 3.4.4-1.
- c. Continuously recording the conductivity of the reactor coolant, or, when the continuous recording conductivity monitor is inoperable for up to 31 days, obtaining an in-line conductivity measurement at least once per:
- 1. 4 hours in OPERATIONAL CONDITIONS 1, 2, and 3, ar.d
- 2. 24 hours at all other times.
- d. Performance of a CHANNEL CHECK of the continuous conductivity monitor with an in-line flow cell at least once per:
- 1. 7 days, and
- 2. 24 hours whenever conductivity is greater than the limit in in Table 3.4.4-1.
O U LIMERICK - UNIT 1 3/4 4-13
r-
% TABLE 3.4.4-1 x
y REACTOR COOLANT SYSTEM [ CHEMISTRY LIMITS 5
-4 - OPERATIONAL CONDITION CHLORIDES CONDUCTIVITY (pmhos/cm @25 C) pH, 1 1 0.2 ppm i 1. 0 5.6 5 pH i 8.6 2 and 3 1 0.1 ppm i 2.0 5.6 1 pH $ 8.6 At all other times 1 0.5 ppm i 10.0 5.3 i pH i 8.6 e
O -- O O
REACTOR COOLANT SYSTEM
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v/ 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:
- a. Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131, and
- b. Less than or equal to 100/E microcuries per gram.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4. ACTION:
- a. In OPERATIONAL CONDITION 1, 2, or 3 with the specific activity of the primary coolant;
- 1. Greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram, operation may continue for up to 48 hours provided that the cumulative operating time under these circumstances does not exceed 800 hours in any consecutive 12-month period. With the total cumulative operating
, time at a primary coolant specific activity greater than 0.2 micro-
) curie per gram DOSE EQUIVALENT I-131 exceeding 500 hours in any x~f consecutive 6-month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours of operation above this limit.
The provisions of Specification 3.0.4 are not applicable.
- 2. Greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or for more than 800 hours cumulative operating time in a consecutive 12-month period, or greater than 4 microcuries per gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours.
- 3. Greater than 100/E microcuries per gram, be in at least HOT SHUTDOWN with the main steamline isolation valves closed within 12 hours.
- b. In OPERATIONAL CONDITION 1, 2, 3, or 4, with the specific activity of the primary coolant greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.
, This report shall contain the results of the specific activity r analyses and the time duration when the specific activity of the (3) v' coolant exceeded 0.2 microcurie per gram DOSE EQUIVALENT I-131 together with the following additional information. LIMERICK - UNIT 1 3/4 4-15
REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- c. In OPERATIONAL CONDITION 1 or 2, with:
- 1. THERMAL POWER changed by more than 15% of RATED THERMAL POWER in 1 hour *, or
- 2. The off gas level, at the SJAE, increased by more than 10,000 microcuries per second in 1 hour during steady-state operation at release rates less than 75,000 microcuries per second, or
- 3. The off gas level, at the SJAE, increased by more than 15% in 1 hour during steady-state operation at release rates greater than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4.b) of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit. Prepare and submit to the Commission a Special Report pursuant to Specification 6.9.2 at least once per 92 days containing the results of the specific activity analysis together with the below additional information for each occurrence.
Additional Information
- 1. Reactor power history startinD 48 hours prior to:
a) The first sample in which the limit was exceeded, and/or b) The THERMAL POWER or off gas level change.
- 2. Fuel burnup by core region.
- 3. Clean-up flow history starting 48 hours prior to:
a) The first sample in which the limit was exceeded, and/or b) The THERMAL POWER or off gas level change.
- 4. Off gas level starting 48 hours prior to:
a) The first sample in which the limit was exceeded, and/or b) The THERMAL POWER or off gas level change. SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysia program of Table 4.4.5-1.
*Not applicable during the startup test program.
LIMERICK - UNIT 1 3/4 4-16
/~'N 1
r C g TABLE 4.4.5-1 n PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM e OPERATIONAL CONDITIONS
- E TYPE OF MEASUREMENT SAMPLE AND ANALYSIS IN WHICH SAMPLE i Z AND ANALYSIS FREQUENCY AND ANALYSIS IS REQUIRED
, 1. Gross Beta and Gamma Activity At least once per 72 hours 1, 2, 3 Determination
- 2. Isotopic Analysis for DOSE At least once per 31 days 1
. EQUIVALENT I-1 131 Concentration
- 3. Radiochemical for E Determination At least once per 6 months
- 1
- 4. Isotopic Analysis for Iodine a) At least once per 4 hours, 1**, 2**, 3**, 4**
i
; R
- whenever the specific activity exceeds a limit, l 7 as required by ACTION b.
! U
- b) At least one sample, between 1, 2 1
2 and 6 hours following the } change in THERMAL POWER or ! off gas level, as required
- by ACTION c.
! 5. Isotopic Analysis of an Off- At least once per 31 days 1 gas Sample Including Quantitative Measurements for at least Xe-133,
- Xe-135, and Kr-88 I
- 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 primary coolant system is restored to within its limits.
4
REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curves A and A' for hydrostatic or leak testing; (2) curves B and B' for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C and C' for operations with a critical core other than low power PHYSICS TESTS, with:
- a. A maximum heatup of 100 F in any 1-hour period,
- b. A maximum cooldown of 100 F in any 1-hour period,
- c. A maximum temperature change of less than or equal to 20 F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and
- d. The reactor vessel flange and head flange temperature greater than or equal to 80 F when reactor vessel head bolting studs are under tension.
APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTOOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the ab3ve required heatup and cooldown limits and to the right of the limit lines of figure 3.4.6'.1-1 curves A and A', B and B', or C and C' as applicable, at least once per 30 minutes. O LIMERICK - UNIT 1 3/4 4-18
REACTOR COOLANT SYSTEM
\ -
SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curves C and C' within 15 minutes prior to the withdrawal of control rods to
~
bring the reactor to criticality and at least once per 30 minutes during system heatup. 4.4.6.1.3 The reactor vessel material. surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1. The results of these examinations shall be used to update the curves of Figure 3.4.6.1-1. 4.4.6.1.4 The reactor flux wire specimens shall be removed at the first refueling outage and examined to determine reactor pressure vessel fluence as a function of time and power level and used to modify Figure B 3/4 4.6-1. The results of these fluence determinations in conjunction with Figure B 3/4 4.6-2, shall be used to adjust the curves of Figure 3.4.6.1-1, as required. 4.4.6.1.5 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80 F:
- a. OPERATIONAL CONDITION 4 when reactor coolant system temperature
- 1. 1 100 F, at least once per 12 hours.
- 2. 1 90 F, at least once per 30 minutes.
- b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.
\ LIMERICK - UNIT 1 3/4 4-19
l l l O A'A 8'8 C'C 0 $ ! A SYSTEM HYDROTEST LIMIT CORE BELTLINE , ITH FUEL IN VESSEL
^F " "
(NOT lMIT N 8 NO NUCLEAR HE ATING
; s g 1200 - C NUCLEAR (CORE CRITICAL) l l LIMIT BASED ON G.E. BWR l l y LICENSING TOPICAL REPORT NEDO.21778A Q
l I I 1000 - I I VESSE L g p DISCO ITINUITY l l A', B'.C' CORE BELTLINE AFTER AN a LIMITS / l l ASSUMED 36 F TEMP. SHIFT FROM h f AN INITI AL PLATE RTNDT OF 20 0F. w p CURVES ARE NOT LIMITING (SHOWN FOR INFORMATION ONLY) o ex - I I b I g 5 / : m o n i e / 5600- / NOTE: w f
$ / CURVES A.B.& C ARE PREDICTED 12 TO APPLY AS THE LIMITS FOR j
E 40 YEARS (32 EFPY) OF OPERATION
= /
400 - ! 312 IOCFR50 95'9 APPENDIX G BOLTUP LIMIT 800F 200 - llO'F l I l l l 0 100 200 300 400 500 MINIMUM RE ACTOR VESSEL METAL TEMPER ATURE *F MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 LIMERICK - UNIT 1 3/4 4-20
m i r-g TABLE 4.4.6.1.3-1
- E M REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE i
* .i j i j E CAPSULE VESSEL LEAD WITHDRAWAL TIME j- p NUMBER LOCATION FACTOR * (EFPY)
- s 117C 4944 G004 30* 1.20 10
! 117C 4944 G001 120* 1.20 30 I l 1 117C 4944 G001 300* 1.20 Spare . i i ! w-1 t N t I l P
*At 1/4 T. .
t I i r i l r
.------,----~w g4-e'erer-.--m , ++ w ' -- ?---y- -++%++-ma ---f --
r--T- - - r- r - - - - - - - - - - - - - + y-+mmg?-t+-m--v m w te-amw --=--we<wys-sv . No-+ e z- rF-
REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the react'or steam dome shall be less than 1020 psig. A_PPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*. A_CTION: With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours. P SURVEILLANCE REQUIREMENTS O 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig at least once per 12 hours.
*Not applicable during anticipated transients.
O LIMERICK - UNIT 1 3/4 4-22
f' 'g REACTOR COOLANT SYSTEM ( 1-
-^-' 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to 5 seconds.
4 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one or more MSIVs inoperable:
- 1. Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours, either:
a) Restore the inoperable valve (s) to OPERABLE status, or b) Isolate the affected main steam line by use of a deactivated MSIV in the closed position.
- 2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours
\ and in-COLD SHUTDOWN within the following 24 hours.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5. v) ( LIMERICK - UNIT 1 3/4 4-23
REACTOR COOLANT SYSTEM 3/4.4.8 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.8. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5. 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 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.
- 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.
SUR'!EILLANCE REQUIREMENTS 4.4.8 No requirements other than Specification 4.0.5. O LIMERICK - UNIT 1 3/4 4-24
REACTOR COOLANT SYSTEM i 3/4.4.9 RESIDUAL HEAT REMOVAL HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.1 Two* shutdown cooling mode loops of the residual heat removal (RHR)
. system shall'be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation ** ***
with each loop consisting of at least:
- a. One OPERABLE RHR pump, and
- b. One OPERABLE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint. ACTION:
- a. With less than the above required RHR shutdown cooling mode loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within 1 hour and at least once per 24 hours thereafter, demonstrate the operability
, of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop. Be in at least COLD V) /
SHUTDOWN within 24 hours.****
- b. With no RHR shutdown cooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible. Within 1 hour establish reactor coolant circu-lation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.
SURVEILLANCE REQUIREMENTS 4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
*0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours for surveillance testing provided the other loop is OPERABLE and in operation.
**The shutdown cooling pump may be removed from operation for up to 2 hours per 8-hour period provided the other loop is OPERABLE.
***The RHR shutdown' cooling mode loop may be removed from operation during
_ hydrostatic testing.
****Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
o LIMERICK-- UNIT 1 3/4 4-25
REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.2 Two* shutdown cooling mode loons of the residual heat removal (RHR) system shall be OPERABLE and, unless at aust one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation ** *** with each loop consisting of at least:
- a. One OPERABLE RHR pump, and
- b. One OPERABLE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITION 4. ACTION:
- a. With less than the above required RHR shutdown cooling mode loops OPERABLE, within 1 hour and at least once per 24 hours thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.
- b. With no RHR shutdown cooling mode loop in operation, within 1 hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.
SURVEILLANCE REQUIREMENTS 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
*0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours for surveillance testing provided the other loop is OPERABLE and in operation. **The shutdown cooling pump may be removed from operation for up to 2 hours per 8-hour period provided the other loop is OPERABLE. ***The shutdown cooling mode loop may be removed from operation during hydrostatic testing.
LIMERICK - UNIT 1 3/4 4-26
A- 3/4.5 EMERGENCY CORE COOLING SYSTEMS
'3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:
- a. The core spray system (CSS) consisting of two subsystems with-each subsystem comprised of:
' 1. Two OPERABLE CSS pump (s), and
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
- b. The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of four subsystems with each subsystem comprised of:
- 1. One OPERABLE LPCI pump, and
- 2. An OPERABLE flow path capable of taking suction from the O suppression chamber and transferring the water to the reactor i vessel.
- c. The high pressure coolant injection (HPCI) system consisting of:
- 1. One OPERABLE HPCI pump, and
- 2. An OPERABLE flow path capable of taking suction from the i suppression chamber and transferring the water to the reactor vessel.
- d. The automatic depressurization' system (ADS) with at least five OPERABLE ADS valves.
-APPLICABILITY: OPERATIONAL CONDITION 1, 2* ** #,'and 3* ** ##.
- *The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or' equal to 200 psig.
- **The ADS is not required to be OPERABLE when reactor steam dome pressure is
'less :than or equal to 100 psig.
#See Special Test Exception 3.10.6.
##Two LPCI subsystems of the RHR system may be inoperable in that they are aligned 4in the shutdown cooling mode when reactor vessel pressure is less than the RHR Shutdown cooling permissive setpoint.
v LIMERICK - UNIT 1 3/4 5-1 = -
EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION:
- a. For the core spray system:
- 1. With one CSS subsystem inoperable, provided that at least two LPCI subsystems are OPERABLE, restore the inoperable CSS subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With both CSS subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- b. For the LPCI system:
- 1. With one LPCI subsystem inoperable, provided that at least one CSS subsystem is OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With one RHR cross-tie valve (HV-51-182 A or B) open, or power not removed from one closed RHR cross-tie valve operator, close the open valve and/or remove power from the closed valves operator within 72 hours, or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- 3. With no RHR cross-tie valves (HV-51-182 A, 8) closed, or power not removed from both closed RHR cross-tie valve operators, or with one RHR cross-tie valve open and power not removed from the other RHR cross-tie valve operator, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- 4. With two LPCI subsystems inoperable, provided that at least one CSS subsystem is OPERABLE, restore at least three LPCI subsystems to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 5. With three LPCI subsystems inoperable, provided that both CSS subsystems are OPERABLE, restore at least two LPCI subsystems to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 6. With all four LPCI subsystems inoperable, be in at least HOT SHUIDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.*
- c. For the HPCI system, provided the CSS, the LPCI system, the ADS and the RCIC system are OPERABLE:
- 1. With the HPCI system inoperable, restore the HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to < 200 psig within the following 24 hours.
1 CWhenever both shutdown cooling subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. LIMERICK - UNIT 1 3/4 5-2 1
F EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- d. For'the ADS:
- 1. With one of the above required ADS valves inoperable, provided the HPCI system, the CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within
- 14. days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to 1 100 psig within the next 24 hours.
- 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours and reduce reactor steam dome pressure to 5 100 psig within the next 24 hours.
- e. With a CSS and/or LPCI header AP instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours or determine the ECCS header AP locally at'least once per 12 hours; otherwise, declare the associated CSS and/or.LPCI, as applicable, inoperable.
'O f. In the event an ECCS system is actuated and injects water into the Q reactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total l accumulated actuation cycles to date. The current value of the useage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70. e LIMERICK - UNIT 1 3/4 5-3
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:
- a. At least once per 31 days:
- 1. For the CSS, the LPCI system, and the HPCI system:
a) Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water. b) Verifying that each val"- m 1, power-operated, or automatic) in the f'- r atn 6. ' not locked, sealed, or otherwise secured position, 1. in its correct
- position.
- 2. For the LPCI systen erifying that both LPCI system subsystem cross-tie valves (HV-51-182 A, B) are closed with power removed from the valve operr' ors.
- 3. For the HPCI system, .erifying that the HPCI pump flow controller is in the correct position.
- 4. For the CSS and LPCI system, performance of a CHANNEL FUNCTIONAL TEST of the injection header AP instrumentation.
- b. Verifying that, when tested pursuant to Specification 4.0.5:
- 1. Each CSS pump in each subsystem develops a flow of at least 3175 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of 1 105 psid plus head and line losses.
- 2. Each LPCI pump in each subsystem develops a flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of 1 20 psid plus head and line losses.
- 3. The HPCI pump develops a flow of at least 5600 gpm against a test line pressure which corresponds to a reactor vessel pressure of 1000 psig plus head and line losses when steam is being supplied to the turbine at 1000, +20, -80 psig.**
- c. At least once per 18 months:
- 1. For the CSS, the LPCI system, and the HPCI system, performing a system functional test which inciudes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
*Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
- The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
LIMERICK - UNIT 1 3/4 5-4
EMERGENCY CORE COOLING SYSTEMS
./ m .\
. $h SURVEILLANCE REQUIREMENTS (Continued)
- 2. For the HPCI system, verifying that:
a) The system develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of > 200 psig plus head and line losses, when steam is being supplied to the turbine at 200 + 15, - 0 psig.** b) The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber - water level high signal.
- 3. Performing a CHANNEL CALIBRATION of the CSS, LPCI, and HPCI system discharge line " keep filled" alarm instrumentation.
- 4. Performing a CHANNEL CALIBRATION of the CSS header AP instru-mentation and verifying the setpoint to be 5 the allowable value of 4.4 psid.
- 5. Performing a CHANNEL CALIBRATION of the LPCI header AP instru-mentation and verifying the setpoint to be $ the allowable value of 3.0 psid.
- d. For the ADS:
s /^
- 1. At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the accumulator backup compressed gas system low pressure alarm system.
- 2. At least once per 18 months:
a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation, b) Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig** and observing that either:
- 1) The control valve or bypass valve position responds accordingly, or
- 2) There is a corresponding change in the measured steam flow, c) Performing a CHANNEL CALIBRATION of the accumulator backup compressed gas system low pressure alarm system and verifying
-an alarm setpoint of 90 1 2 psig on decreasing pressure.
**The provisions of Specification 4.0.4 are not applicable provided the (Ol V
surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
- LIMERICK - UNIT-1 3/4 5-5 i
l i EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN 9 LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:
- a. Core spray system (CSS) subsystems with a subsystem comprised of:
- 1. Two OPERABLE CSS pumps, and
- 2. An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel:
a) From the suppression chamber, or b) When the suppression chamber water level is less than the limit or is drained, from t.he condensate storage tank containing at least 135,000 available gallons of water, equivalent to a level of 29 feet.
- b. Low pressure coolant injection (LPCI) system subsystems with a subsystem comprised of:
- 1. One OPERABLE LPCI pump, and
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5*. ACTION:
- a. With one of the above required subsystems inoperable, restore at least two subsystems to OPERABLE status within 4 hours or suspend all operations with a potential for draining the reactor vessel.
- b. With both of the above required subsystems inoperable, suspend CORE ALTERATIONS and all operations with a potential for draining the reactor vessel. Restore at least one subsystem to OPERABLE status within 4 hours or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours.
- The ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.
O LIMERICK - UNIT 1 3/4 5-6
.. .- . . . - - . . - - . . . . _ . .. -- . _ ~ - ._
1 i i E EMERGENCY CORE COOLING SYSTEMS s SURVE lt3NCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per
- Surveillance Requirement 4.5.1.
4 4.5.2.2 The core spray system shall be determined OPERABLE at least once per 12 hours by verifying the condensate storage tank required volume when the condensate storage tank is required to be OPERABLE per Specification 3.5.2a.2.b). i i i l ( I LIMERICK - UNIT 1 3/4 5-7 . l- . _ . - - .
EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.5.3 The suppression chamber shall be OPERABLE:
- a. In OPERATIONAL CONDITIONS 1, 2, and 3 with a contained water volume of at least 122,120 fta, equivalent to a level of 22'0".
- b. In OPERATIONAL CONDITION 4 and 5* with a contained water volume of at least 88,815 ft3 , equivalent to a level of 16'0", except that the suppression chamber level may be less than the limit or may be drained provided that:
- 1. No operations are performed that have a potential for draining the reactor vessel,
- 2. The reactor mode switch is locked in the Shutdown or Refuel position,
- 3. The condensate storage tank contains at least 135,000 available gallons of water, equivalent to a level of 29 feet, and
- 4. The core spray system is OPERABLE per Specification 3.5.2 with an OPERABLE flow path capable of taking suction from the condensate storage tank and transferring the water through the spray sparger to the reactor vessel.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5*. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression chamber water level less than the above limit, restore the water level to within the limit within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION 4 or 5* with the suppression chamber water level less than the above limit or drained and the above required conditions not satisfied, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock the reactor mode switch in the Shutdown position. Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours.
- The suppression chamber is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppression pool, the spent fuel pool gates are removed when the cavity I is flooded, and the water level is maintained within the limits of l Specifications 3.9.8 and 3.9.9.
LIMERICK - UNIT 1 '3/4 5-8
)
m EMERGENCY CORE COOLING SYSTEMS j SURVEILLANCE REQUIREMENTS i 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying the water level to be greater than or equal to, as applicable:
- a. 22'0" at least once per 24 hours.
- b. -16'0" at least once per 12 hours.
4.5.3.2 With the suppression chamber level less than the above limit or drained in OPERATIONAL CONDITION 4 or 5*, at least once per 12 hours:
- a. Verify the required conditions of Specification 3.5.3b. to be satisfied, or
- b. Verify footnote conditions
- to be satisfied.
h. O i I l *The suppression chamber is not required to be OPERABLE provided that the }
-reactor vessel head is removed, the cavity is flooded or being flooded from f the suppression pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.
l LIMERICK - UNIT 1 3/4 5-9 i'
s
-3/4.6 CONTAINMENT SYSTEMS gm '
3/4.6.1 PRIMARY CONTAINMENT . PRIMARY CONTAINMENT-INTEGRITY
- LIMITING CONDITION FOR OPERATION l . 3.6.1.1. PRIMARY CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: OP'ERATIONAL CONDITIONS 1, 2*, and 3. ACTION: , Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:
- a. After each closing of each penetration subject to Type B testing,
- except the primary containment air locks, if opened following Type A or B test, by leak rate testing the seals with gas at P,, 44.0 psig,
.I
> k.,_ ) - arid verifying that when the measured leakage rate for these seals.is added to the leakage rates determined pursuant to Surveillance
- Requirement 4.6.1.2d. for all~other Type B and C penetrations ~, the combined leakage rate is less than or equal to 0.60 L,.
- b. At least once per 31 days by verifying that all primary containment penetrations ** not capable of being closed by.0PERABLE containment automatic isolation valves and required to be closed during accident
, conditions are closed by valves, blind flanges, or deactivated 1 automatic valves' secured in position, except as provided in Table 3.6.3-1 of Specification 3.6.3.
- c. By verifying the primary cantainment air lock is in compliance with i the requirements of Specification 3.6.1.3.
5 d. By verifying the suppression chamber is in compliance with the , requirements of Specifit:ation 3.6.2.1. l- *See Special Test Exception 3.10.1 i **Except-valves,-blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed, or otherwise secured in the !- closed positio.n. These penetrations shall be verified closed during each COLD-SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often O than once per 92 days.
- Q LIMERICK - UNIT 1 3/4 6-1 a
g -
,,. - v - , .-,,.,-e-- , - * .~,y- ,w , ,,e- ....w. ._~ .-~ , , , . _ , , . . - - , - - - - - - - ,-
i CONTAINMENT SYSTEMS PRIMARi w.,..u.uitNT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of less than or equal to L , 0.500 percent by weight of the containment air per 24 hours at a
P , 44.0 psig. a
- b. A combined leakage rate of less than or equal to 0.60 L f r all a
penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves
- and valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests when pressurized to P , 44.0 psig.
a
- c. *Less than or equal to 11.5 scf per hour for any one main steam line through the isolation valves when tested at P , 22.0 psig.
t
- d. A combined leakage rate of less than or equal to 1 gpm times the total number of containment isolation valves in hydrostatically tested lines which penetrate the primary containment, when tested at 1.10 Pa , 48.4 psig.
APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1. ACTION: With:
- a. The measured overall integrated priraary containmeni leakage rate exceeding 0.75 L,, or
- b. The measured combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves
- and valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests exceeding 0.60 La , r
- c. The measured leakage rate exceeding 11.5 scf per hour for any one main steam line through the isolation valves, or
- d. The measured combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 1 gpm times the total number of such valves, restore:
- a. The overall integrated leakage rate (s) to less than or equal to 0.75 L , and a
- Exemption to Appendix J of 10 CFR Part 50.
LIMERICK - UNIT 1 3/4 6-2
- CONTAINMENT SYSTEMS A LIMITING CONDITION FOR OPERATION (Continued) 4 l ACTION: (Continued)
, b. The combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves
- and i i valves which are_hydrostatically tested per Table 3.6.3-1, subject i to Type _B and C tests to less than or equa1 to 0.60 L,, and
- c. The leakage rate to less than or equal to'11.5 scf per hour for any 1 one main steam line through the isolation valves, and
- d. The combined leakage rate for all containment isolation valves in hydrostatically tested lines which pentrate the primary containment to less than or equal to 1 gpm times the total number of such valves,
~ prior to increasing reactor coolant system temperature above 200*F. SURVEILLANCE REQUIREMENTS
, 4.6.1.2 The primary containment leakage rates shall be demonstrated at the
- following test schedule and shall be determined in conformance with the criteria
' specified in Appendix J of 10 CFR Part 50 using the methods and provisions of-ANSI 45.4-1972 and BN-TOP-1 and verifying the result by the Mass Point Methodology described in ANSI N56.8-1981:
- a. Three Type A Overall Integrated Containment Leakage Rate tests shall
' fI be conducted at-40 1 10 month intervals during shutdown at P,, 44.0 psig,
.Q
'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. ,
4 -b. If any periodic Type A test fails to meet 0.75 L,, the test schedule . for subsequent Type A tests shall be reviewed and approved by the ' j Commission. If two_ consecutive Type A tests fail to meet 0.75 L,, i- a Type.A test shall be performed at least every 18 months until.two
- consecutive Type A tests meet 0.75 L,, at which time the above test schedule may be resumed.
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
1
~
- 1. Confirms the accuracy of the test by verifying that the difference-between the supplemental data and the Type A test data is within 0.25 L,. The formula to be used is: [L, + L, - 0. 25 L,-] < L c
. < [Lg + L,, + 0.25 L,] where Lc = supplemental test result; L, =
i superimposed leakage; L,, = measured Type A leakage.
- 2. Has duration sufficient to est'ablish accurately the change in I leakage rate between the Type A test and the supplemental test.
l
- 3. Requires the quantity of gas injected into the containment or j
bled from the containment during the supplemental test to be between 0.75 L, and 1.25 L,.
- Exemption to Appendix "J" to 10 CFR Part 50.
LIMERICK - UNIT 1 3/4 6-3 1 i
.-~.-a-.. .--.. -- ---.--- - -. - - - , - -- _ ---. ,,, . - -- - _ .,. ,..,,.- . --. . , - - -
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- d. Type B and C tests shall be conduc ed with gas at P a
, 44.0 psig*,
at intervals no greater than 24 months except for tests involving:
- 1. Air locks,
- 2. Main steam line isolation valves,
- 3. Containment isolation valves in hydrostatically tested lines which penetrate the primary containment, and
- e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
- f. Main steam line isolation valves shall be leak tested at least once per 18 months.
- g. Containment isolation valves in hydrostatically tested lines which
. penetrate the primary containment shall be leak tested at least once per 18 months.
- h. The provisions of Specification 4.0.2 are not applicable to Specifica-tions 4.6.1.2a., 4.6.1.2b., 4.6.1.2c., 4.6.1.2d., and 4.6.1.2e.
*Unless a hydrostatic test is required per Table 3.6.3-1.
O LIMERICK - UNIl 1 3/4 6-4
' TN CONTAINMENT SYSTEMS k )
- O PRIMARY CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 The primary 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 a at Pa , 44.0 psig.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3. , ACTION: 1
- a. With one pri' mary 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 (s) 24 hours or lock the OPERABLE air lock door closed. . 2. Operation may then continue until performance of the 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 SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 4. The provisions of Specification 3.0.4 are not applicable.
i b. With the primary containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within i 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and i in COLD SHUTDOWN within the following 24 hours. 4 t i I
-^s *See Special Test Exception 3.10.1.
l (N~ l l t LIMERICK - UNIT 1 3/4 6-5 i
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS
- 4. 6.1. 3 The primary containment air lock shall be demonstrated OPERABLE:
- a. By verifying the seal leakage rate to be less than or equal to 5 scf per hour when the gap between the door seals is pressurized to 10 psig:
- 1. within 72 hours after each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours; and
- 2. prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air lock has been used and no maintenance has been performed on the air lock.**
- b. By conducting an overall air lock leakage test at Pa , 44.0 psig, and by verifying that the overall air lock leakage rate is within its limit:
- 1. At least once per 6 months,* and
- 2. Prior to establishing PRIMARY 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 the air lock can be opened at a time.***
*The provisions of Specification 4.0.2 are not applicable. ** Exemption to Appendix J, Paragraph III.D.2.(b)(ii) of 10 CFR Part 50.
- Except that the airlock doors need not be opened to verify interlock OPERA-BILITY when the primary containment is inerted, provided that the airlock doors' interlock is tested within 8 hours after the primary containment has been deinerted and provided the shield door to the airlock is maintained locked closed.
LIMERICK - UNIT 1 3/4 6-6
gy CONTAINMENT SYSTEMS
\ )
v' MSIV LEAKAGE CONTROL SYSTEM LIMITING CONDITICN FOR OPERATION 3.6.1.4 Two independent MSIV leakage control system (LCS) subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With one MSIV leakage control system subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.1.4 Each MSIV leakage control system subsystem shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
- 1. Starting the blower (s) from the control room and operating the blower (s) for at least 15 minutes.
,.~
( ') 2. Energizing the heaters and verifying a temperature rise indicat-
\, / ing heater operation on downstream piping.
- b. During each COLD SHUTDOWN, if not performed within the previous 92 days, by cycling each motor operated valve through at least one complete cycle of full travel.
- c. At least once per 18 months by:
- 1. Performance of a functional test which includes simulated actua-tion of the subsystem throughout its operating sequence, and verifying that each interlock and timer operates as designed, each automatic valve actuates to its correct position and the blower starts.
- 2. Verifying that the blower (s) develops at least the below required vacuum at the rated capacity:
a) Inboard valves, 15" H2 0 at 100 scfm. b) Outboard valves, 15" H2 0 at 200 scfm.
- d. By verifying the operating instrumentation to be OPERABLE by performance of a:
- 1. CHANNEL CHECK at least once per 24 hours,
- 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- 3. CHANNEL CALIBRATION at least once per 18 months.
(~N >\ i l LIMERICK - UNIT 1 3/4 6-7
\
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.5. APPLICABTLITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With the structural integrity of the primary containment not conforming to the above requirennents, re2 tore the structural integrity to within the limits , within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.1.5.1 The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test by a visual inspection of those surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. 4.6.1.5.2 Reports Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days. This report shall include a description of the condition of the liner and concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken. O LIMERICK - UNIT 1 3/4 6-8
CONIAINMENT SYSTEMS
; 1 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.6 Drywell and suppression chamber internal pressure shall be maintained between 0.0 and +2.0 psig.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. , ACTION: With the drywell and/or suppression chamber internal pressure outside of the specified limits, restore the internal pressure to within the limit within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
, ~3 \- / SURVEILLANCE REQUIREMENTS 4.6.1.6 The drywell and suppression chamber internal pressure shall be determined to be within the limits at least once per 12 hours.
i [7
\~.-
l . t l LIMERICK - UNIT 1 3/4 6-9 l t
CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.7 Drywell average air temperature shall not exceed 135 F. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With the drywell average air temperature greater than 135 F, reduce the average air temperature to within the limit within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average air temperature shall be the volumetric average O of the temperatures at the following locations and shall be determined to be within the limit at least once per 24 hours: Elevation Azimuth *
- a. 330' 45 , 90 , 225
- b. 320' 105 , 225 , 345
- c. 260' 50 , 165 , 285
- d. 248' 11 , 74 , 150 , 182 , 253 , 337
- At least one reading from each elevation is required for a volumetric average calculation.
LIMERICK - UNIT 1 3/4 6-10
7] CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM M MITING CONDITION FOR OPERATION
- 3. 6.1. 8 The drywell and. suppression chamber purge system may be in operation for up to 90 hours each 365 days with the supply and exhaust isolation valves in one supply line and one exhaust line open for inerting, deinerting, or pressure control.*
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With a drywell and/or suppression chamber purge supply and/or exhaust isolation valve open, except as permitted above, close the valve (s) within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours
.and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.1.8 Before being opened, the drywell and suppression chamber purge supply and exhaust butterfly isolation valves shall be verified not to have been open Q for more than 90 hours in the previous 365 days.* t i Q ^ Valves open for pressure control are not subject to the 90 hour per 365 day limit provided the 1-inch /2-inch bypass line is being utilized. 1 LIMERICK - UNIT 1 3/4 6-11
CONTAINMENT SYSTEMS 314.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERABLE with:
- a. The pool water:
- 1. Volume
- between 122,1203 ft and 134,600 fta, equivalent to a level between 22' 0" and 24' 3", and a -
- 2. Maximum average temperature of 95 F except that the maximum average temperature may be permitted to increase to:
a) 105 F during testing which adds heat to the suppression chamber. b) 110 F with THERMAL POWER less than or equal to 1% of RATED THERMAL POWER. c) 120 F with the main steam line isolation valves closed following a scram.
- b. Drywell-to-suppression chamber bypass leakage less than or equal to 10% of the acceptable A/[K design value of 0.0500 ft2 ,
- c. At least eight suppression pool water temperature instrumentation indicators.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With the suppression chamber water level outside the above limits, restore the water level to within the limits within 1 hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With the suppression chamber average water temperature greater than 95 F, restore the average temperature to less than or equal to 95 F within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, except, as permitted above:
- 1. With the suppression chamber average water temperature greater than 105 F during testing which adds heat to the suppression chamber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than 95 F within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With the suppression chamber average water temperature greater than:
a) 95 F for more than 24 hours and THERMAL POWER greater than 1% of RATED THERMAL POWER, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. b) 110 F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode. QIncludes the volume inside the pedestal. LIMERICK - UNIT 1 3/4 6-12
CONTAINMENT SYSTEMS i ) LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- 3. With the suppression chamber average water temperature greater than 120 F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours.
,c . With only one suppression chamber water level indicator OPERABLE and/or with less than eight suppression pool water temperature indicators, one in each of the eight locations OPERABLE, restore the inoperable indicator (s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least once per 12 hours.
- d. With no suppression chamber water level indicators OPERABLE and/or with less than seven suppression pool water temperature indicators covering at least seven locations OPERABLE, restore at least one water level indicator and at least seven water temperature indicators to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- e. With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200 F.
,m. SURVEILLANCE REQUIREMENTS
( \ 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:
- a. By verifying the suppression chamber water volume to be within the limits at least once per 24 hours.
- b. At least once per 24 hours by verifying the suppression chamber average water temperature to be less than or equal to 95 F, except:
- 1. At least once per 5 minutes 6 rk i testing which adds heat to the suppression chamber, W w i 'ing the suppression chamber average water temperatuto m mn or equal to 105 F.
- 2. At least once per hour when suppression chamber average water temperature is greater than or equal to 95 F, by verifying:
a) Suppression chamber average water temperature to be less than or equal to 110 F, and b) THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWER 12 hours after suppression chamber average water temperature has exceeded 95 F for more than 24 hours.
- 3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 95 F, by verifying suppression chamber average water temperature less than or equal to 120 F.
.j LIMERICK - UNIT 1 3/4 6-13
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- c. By verifying at least two suppression chamber water level indicators and at least 8 suppression pool water temperature indicators in at least 8 locations, OPERABLE by performance of a:
- 1. CHANNEL CHECK at least once per 24 hours,
- 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- 3. CHANNEL CALIBRATION at least once per 18 months, with the water level and temperature alarm setpoint for:
- 1. High water level 5 24'1 "
- 2. High water temperature:
a) First setpoint 5 95 F b) Second setpoint 5 105 F c) Third setpoint 5 110 F d) Fourth setpoint 5 120 F
- d. At least once per 18 months by conducting a drywell-to-suppression chamber bypass leak test at an initial differential pressure of 4 psi and verifying that the A/Jk calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test faits to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 9 months until two consecutive tests meet the specified limit, at which time the 18 month test schedule may be resumed.
O LIMERICK - UNIT 1 3/4 6-14
(~~N CONTAINMENT SYSTEMS t Y~)i SUPPRESSION POOL SPRAY LIMITING CONDITION FOR OPERATION 3.6.2.2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:
- a. One OPERABLE RHR pump, and
- b. An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and the suppression pool spray sparger(s).
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one suppression pool spray loop inoperable, restore the inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With both suppression pool spray loops inoperable, restore at least one loop to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN
- within the
, following 24 hours. V SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool spray mode of the RHR 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 each of the required RHR pumps develops a flow of at least 500 gpm on recirculation flow through the RHR heat exchanger and the suppression pool spray sparger when tested pursuant to Speci-fication 4.0.5.
*Whenever both RHR subsystems are inoperable, if unable to attain COLD p) t v
SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. LIMERICK - UNIT 1 3/4 6-15
CONTAINMENT SYSTEMS SUPPRESSiun .voc COOLING LIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression pool cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:
- a. One OPERABLE RHR pump, and
- b. An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. Wit.h one suppression pool cooling loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With both suppression pool cooling loops inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN
- within the next 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.2.3 The suppression pool cooling mode of the RHR 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 each of the required RHR pumps develops a flow of at least 10,000 gpm on recirculation flow through the RHR heat exchanger, the suppression pool and the full flow test line when tested pursuant to Specification 4.0.5. 1
- Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN I as required by this ACTION, maintain reactor coolant temperature as low as l practical by use of alternate heat removal methods. l 9l LIMERICK - UNIT 1 3/4 6-16
j CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The primary containment isolation valves and the reactor instrumentation line excess flow check valves shown in Table 3.6.3-1 shall be OPERABLE. with isolation 'imes less than or equal to those shown in Table 3.6.3-1. APPLICg:1ITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one or more of the primary containment isolation valves shown in Table 3.6.3-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours either:
- 1. Restore the inoperable valve (s) to OPERABLE status, or
- 2. Isolate each affected penetration by use of at least one de-activated automatic valve secured in the isolated position,* or
- 3. Isolate each affected penetration by use of at least one closed manua) valve or blind flange.*
- 4. The provisions of Specification 3.0.4 are not applicable provided p) that within 4 hours the affected penetration is isolated in accordance with ACTION a.2. or a.3. above, and provided that the associated system, if applicable, is declared inoperable and the appropriate ACTION statements for that system are performed.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.
- b. With one or more of the reactor instrumentation line excess flow check valves shown in Table 3.6.3-1 inoperable, operation may continue and the provisions of Specifications 3.0.3 and 3.0.4 are not applicable provided that within 4 hours either:
- 1. The inoperable valve is returned to OPERABLE status, or
- 2. The instrument line is isolated and the associated instrument is declared inoperable.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- Isolation valves closed to satisfy these requirements may be reopened on an intermittent basis under administrative control.
f3
? I C/
LIMERICK - UNIT 1 3/4 6-17
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE prior to returning the valve to service after mainte-nance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time. 4.6.3.2 Each primary containment automatic isolation valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE during COLD SHUTDOWN or REFUELING at least once per 18 months by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position. 4.6.3.3 The isolatiori time of each primary containment power operated or automatic valve shown in Table 3.6.3-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. 4.6.3.4 Each reactor instrumentation line excess flow check valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE at least once per 18 months by verifying that the valve checks flow. 4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying the continuity of the explosive charge.
- b. At least once per 18 months by removing the explosive squib from the explosive valve, such that each explosive squib in each explosive valve will be tested at least once per 90 months, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured ~ batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life and/or operating life, as applicable.
i Ol 1 LIMERICK - UNIT 1 3/4 6-18 l l
r% LJ .(< ;Om i TABLE 3.6.3-1 F PART A - PRIMARY CONTAINMENT'~ ISOLATION VALVES
'a E INBOARD OUTBOARD ISOL;
- FUNCTION ISOLATION ~ ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID 7 PENETRATION BARRIER TIME.IF APP. IF APP.
NUMBER BARRIER. ' E (SEC)(26) (20) Z .
~ 003B CONTAINMENT INSTRUMENT 59-1005B (CK) NA 59 GAS SUPPLY - HEADER 'B' HV59-129B .7 C,H,S 003D-2 CONTAINMENT INSTRUMENT 59-1112(CK) NA GAS SUPPLY TO ADS VALVES HV59-151B 45 M 59
? E&K I 007A(B,C,D) MAIN STEAM LINE HV41-1F022A 5* C,D,E,F,P,Q 6 41
'A'(B,C,D) (B,C,D) ! HV41-1F028A 5* C,D,E,F,P,Q 6 i
4 -(B,C,D)
- HV40-1F0018 45 EA '6 i (F,K,P) l5 (XV40-1018 NA 6,1 (F,K,P)
SEE PART B, THIS TABLE) 008 MAIN STEAM LINE DRAIN HV41-1F016 30 C,D,E,F,P,Q 4 41 i HV41-1F019 30 C,D,E,F,P,Q t 009A FEEDWATER 41-1F010A(CK) .NA 41
- HV41-1F074A(CK) NA i 41-1036A(CK) NA
! HV41-1308 45 HV41-133A 45 l HV41-109A NA 32 HV41-1F032A(CK) NA ll HV55-1F105 HV44-1F039(CK) NA 30 7 l' (X-98) 41-1016(X-9B, NA 31 X-44) i
TABLE 3.6.3-1 (Continued) PART A - PRIMARY CONTAINMENT ISOLATION VALVES E INBOARD OUTB0ARD ISOL.
- ' PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&IO NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
E (SEC)(26) (20) w r- 009B FEEDWATER 41-1F010B(CK) NA HV41-1F074B(CK) NA 41 41-1036B(CK) NA HV41-130A 45 HV41-133B 45 HV41-1098 NA 32 HV41-1F0323(CK) NA HV49-1F013 23 LFCC HV44-1F039(CK) NA (X-9A) R 41-1016(X-9A, NA 31 X-44) E$ 010 RCIC STEAM SUPPLY HV49-1F007 7.2* K, KA 5 49 HV49-1F008 7.2* K, KA HV49-1F076 45 K, KA 011 HPCI STEAM SUPPLY HV55-1F002 12* L, LA 5 55 HV55-1F003 12* L, LA HV55-1F100 45 L, LA 012 RHR SHUTDOWN COOLING HV51-1F009 100 A,V 9,22 51 SUPPLY PSV51-155 NA HV51-1F008 100 A,V 013A(B) RHR SHUTDOWN COOLING HV51-1F050A(B) NA A,V 9,22 51 RETURN (CK) HV51-151A(B) 20 A,V f:V51-1F015A(B) 45 A,V 014 RWCU - SUCTION HV44-1F001 10* B,J,Y 44 HV44-1F004 10* B,J,Y O O O
O O O TABLE 3.6.3-1 (Continued) L % PART A - PRIMARY CONTAINMENT ISOLATION VALVES 5 Q INBOARD OUTBOARD ISOL.
. PENETRATION FUNCTION ISOLATION ISOLATION MAX.~ ISO L. SIGNAL (S),. NOTES P&ID-
- c. NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
5 (SEC)(26) (20) w ~ 016A CORE SPRAY INJECTION HV52-1F006A(CK) NA 9,22 52 HV52-1F039A 7 9,22~ HV52-1F005 18 016B CORE SPRAY INJECTION HV52-1F006B(CK) NA 9,22 52 HV52-1F039B 7 9,22 HV52-108(CK) NA 017 RPV HEAD SPRAY HV51-1F022 60 A,V 4,9,22 51 w PSV51-122 NA 9,22
} HV51-1F023 '35
. A,V 021 SERVICE AIR TO DRYWELL 15-1140 NA 15
- 15-1139 NA 022 DRYWELL PRESSURE HV42-147C 45 10 42 INSTRUMENTATION 023 RECW SUPPLY TO HV13-106 40 .11,28, 13 RECIRC PUMPS 29 HV13-108 30 11,28 29 HV13-109 NA 11,13 024 RECW RETURN FROM HV13-107 40 11,28, 13 RECIRC PUMPS 29 HV13-111 30 11,28, 29 HV13-110 NA 11,13
TABLE 3.6.3-1 (Continued) h PART A - PRIMARY CONTAINMENT ISOLATION VALVES 9 ISOL. El INBOARD OUTBOARD FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID
- PENETRATION
' NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
E (SEC)(26) (20) --e HV57-121(X-201A) 5** B,H,S,U,W 3,11,14,25 57 r 025 DRYWELL PURGE SUPPLY 5** B,H,S,U,W 3,11,14,25 HV57-123 HV57-163 9 B,H,R,5 3,11,14 HV57-109 6** B,H,5,U,W 11,25 (X-201A) HV57-131 5** B,H,S,U,W 11,25 (X-201A) HV57-135 6** B,H,S,U,W 11,25 026 DRYWELL PURGE EXHAUST HV57-114 5** B,H,S,U,W 3,11,14,25 57 HV57-111 15** B,H,5,U 5,11,25 R* 9 B,H,R,5 3,11,14 HV57-161 SV57-139 5 10 i' y HV57-115 6** B,H,S,U,W 11,25 HV57-117 5** B,H,S,U 11,25 SV57-145 5 B,H,R,5 11 59-1128(CK) NA 59 027A CONTAINMENT INSTRUMENT GAS SUPPLY TO ADS VALVES HV59-151A 45 M H,M,&S 028A-1 RECIRC LOOP SAMPLE HV43-1F019 10 B,0 43 HV43-1F020 10 B,D 028A-2 DRYWELL H2/02 SAMPLE SV57-132 5 B,H,R,5 11 57 SV57-142 5 B,H,R,5 11 028A-3 DRYWELL H2/02 SAMPLE SV57-134 5 8,H,R,5 11 57 SV57-144 5 B,H,R,5 11 O O O
i TABLE 3.6.3-1 (Continued) C PART A - PRIMARY CONTAINMENT ISOLATION VALVES E INBOARD OUTBOARD _ ISOL. 7 PENETRATION FUNCTION ISOLATION ISOLATION ' MAX.ISOL. SIGNAL (S), NOTES ' P&ID c NUMBER BARRIER BARRIER TIME.IF APP. IF APP. 5 (SEC)(26) (20) w w 028B DRYWELL H2/02 SAMPLE SV57-133 5 B,H,R,5 11 57 SV57-143 5 B,H,R,5 11 SV57-195 5 B,H,R,5 11 030B-1 DRYWELL PRESSURE HV42-147A 45 10 42 INSTRUMENTATION 035A TIP PURGE 59-1056(CK) NA 59 (DOUBLE "0" RING) HV59-131 7 B,H,5 16 s 035C-G TIP DRIVES XV59-141A-E NA -B,H 11,16,21 59 i (DOUBLE "0" RING) E$ XV59-140A-E NA 11,16 037A-D CRD INSERT LINES BALL CHECK NA 12 47 HCU NA 12-038A-D CRD WITHDRAW LINES HCU NA 12 47 SDV VENTS & DRAINS XV47-1F010 25 30 XV47-1F180 30 30 XV47-1F011 25 30 XV47-1F181 30 30 039A(B) DRYWELL SPRAY HV51-1F021A(B) 160 4,11 51 HV51-1F016A(B) 160 11 040E DRYWELL PRESSURE HV42-147D 45 10 42 INSTRUMENTATION 040F-2 CONTAINMENT INSTRUMENT' HV59-101 45 C,H,S 5 59 GAS -SUCTION HV59-102 7 C,H,5
y TABLE 3.6.3-1 (Continued) { PART A - PRIMARY CONTAINMENT ISOLATION VALVES b INBOARD OUTB0ARD ISOL. 9 PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID i NUMBER BARRIER BARRIER TIME.IF APP. IF APP. g (SEC)(26) (20) 040G-1 ILRT DATA ACQUISITION 60-1057 NA 5,11 60 60-1058 NA 11 040G-2 ILRT DATA ACQUISITION 60-1071 NA 5,11 60 60-1070 NA 11 040H-1 CONTAINMENT INSTRUMENT 59-1005A(CK) NA 59 GAS SUPPLY - HEADER 'A' HV59-129A 7 C,H,5 042 STANDBY LIQUID CONTROL 48-1F007(CK) NA 48 (X-116) HV48-1F006A 60 29 043B MAIN STEAM SAMPLE HV41-1F084 10 8,D 41
? HV41-1F085 10 8,D 044 RWCU ALTERNATE 41-1017 NA 5,31 41 RETURN 41-1016(X-9A, NA X-98)
PSV41-112 NA 045A(B,C,D) LPCI INJECTION 'A'(B,C,D) HV51-1F041A(B,C, NA 9,22 51 D)(CK) HV51-142A(B,C,D) 7 9,22 HV51-1F017A 38 (B,C,D) 050A-1 DRYWELL PRESSURE HV42-1478 45 10 42 l INSTRUMENTATION 053 DRYWELL CHILLED WATER HV87-128 60 C,H 11 87 SUPPLY - LOOP 'A' HV87-120A 60 11,28, 29 l HV87-125A 60 11,28,29 l 9 O O
i O O O >
- TABLE 3.6.3-1 (Continued)
PART A -' PRIMARY' CONTAINMENT ISOLATION VALVES E ISOL'.
;Q INBOARD OUTBOARD
' ISOLATION -ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID
, PENETRATION FUNCTION BARRIER BARRIER- . TIME.IF APP. IF APP.
.c NUMBER (20) i z (SEC)(26) ,
l U HV87-129 .60' C,H 11 87 i ~ 054 DRYWELL CHILLED WATER 11,28, RETURN - LOOP 'A' HV87-121A 60 l 29 ! HV87-124A' 60 11,28, l 29 1 HV87-122 60 C,H- 11 '87: 055 DRYWELL CHILLED WATER = SUPPLY - LOOP 'B' HV87-120B 60 11,28, 29 ; HV87-1258 60 11,28,29 4 w ! 2 DRYWELL CHILLED WATER HV87-123 60 C,H 11 87 , j ,056 60 11,28,29 RETURN LOOP 'B' HV87-121B j 1 4
- HV87-124B 60 11,28,29 i
i
'A' NA 15 43
! 061-1 RECIRC PUMP SEAL 43-1004A(CK) (XV43-103A - NA 1 ! PURGE SEE PART B, , l 4 THIS TABLE) ; I 15 43 ;
- 061-2 RECIRC PUMP 'B' SEAL 43-1004B(CK) NA (XV43-103B - NA 1 i
PURGE - SEE PART B, THIS TABLE) ) SV57-150(X-220A) 5 B,H,R,5 11 57 l j 062 ORYWELL H2/02 SAMPLE SV57-159 5 B,H,R,S. 11 '
. RETURN, N2 MAKE-UP (X-220A)
HV57-116 30** B,H,R,5 11 : (X-220A) SV57-190 5 B,H,R,S' 11 l (X-220A) i 1 1
TABLE 3.6.3-1 (Continued) PART A - PRIMARY CONTAINMENT ISOLATION VALVES E INBOARD OUTBOARD ISOL. FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID 7 PENETRATION NUMBER BARRIER BARRIER TIME.IF APP. IF APP. c z (SEC)(26) (20)
- SV57-191 5 B,H,R,5 11 (X-220A) 116 STANDBY LIQUID CONTROL 48-1F007(CK) NA 48 (X-42) HV48-1F0068 60 29 117B-1 DRYWELL RADIATION SV26-190A 5 B,H,R,5 11 26 MONITORING SUPPLY SV26-190B 5 B,H,R,5 11
, 1178-2 DRYWELL RADIATION SV26-190C . B,H,R,S 11 26 g MONITORING RETURN SV26-190D 5 B,H,R,5 11 SUPPRESSION POOL PURGE HV57-124 5** B,H,S,U,W 3,11,14,25 57 [m201A SUPPLY HV57-131(X-25) 5** 3,H,S,U,W 3,11,14,25 HV57-164 9 1,H,R,5 3,11,14 HV57-109(X-25) 6** ;,H,5,U,W 11,25 HV57-147 6** 8,H,S,U,W 11,25 HV57-121(X 25) 5** B,H,S,U,W 11,25 202 SUPPRESSION P0OL PURGE HV57-104 5** B,H,S,U,W 3,11,14,25 57 EXHAUST HV57-105 15*' B,H,S,0 5,11,25 HV57-162 9 B,H,R,5 3,11,14 HV57-112 6** B,H,5,U,W 11,25 HV57-118 5** B,H,S,U 11,25 SV57-185 5 B,H,R,5 11 203A(B,C,D) RHR PUMP SUCTION HV51-1F004A(B, 240 4,22, 51 C,D) 19,29 PSV51-1F030A(B, NA 22 C,D) O O O
q < _fw TABLE 3.6.3-1 (Continued) C. g PART A - PRIMARY CONTAINMENT ISOLATION VALVES-x y INBOARD OUTBOARD MAX.ISOL. ISOL. SIGNAL (S), NOTES P&ID-
, PENETRATION FUNCTION ISOLATION ISOLATION BARRIER BARRIER TIME.IF APP. IF APP.
e NUMBER (20) z (SEC)(26) q .- HV51-125A(B) 180 '4,22,29 51
- 204A(B) RHR PUMP TEST LINE AND CONTAINMENT COOLING SUPPRESSION POOL SPRAY HV51-1F027A(B) 45 C,G 11- 51 205A(B) 206A(B,C,0) CS PUMP SUCTION HV52-1F001A 160 4,22,29 52 (B,C,0)
.CS PUMP TEST AND FLUSH HV52-1F015A(B) 23 C,G 5,22 52 207A(B)
CS PUMP MINIMUM RECIRC HV52-1F031B 45 LFCH 5,22,29 52
$208B 209 HPCI PUMP SUCTION HV55-1F042 160 L,LA 4,22 55 HPCI TURBINE EXHAUST HV55-1F072 120 4,22,29 55 210 212 HPCI PUMP TEST AND FLUSH HV55-1F071 40 B,H 4,22 55 RCIC PUMP SUCTION HV49-1F031 60 4,22,29 49 214 RCIC TURBINE EXHAUST HV49-1F060 80 4,22,29 49 215 216 RCIC MINIMUM FLOW HV49-1F019 8 LFRC 5,22 49 i
b
TABLE 3.6.3-1 (Continued) F
.x PART A - PRIMARY CONTAINMENT ISOLATION VALVES 9
E INBOARD OUTBOARD ISOL. FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID 7 PENETRATION BARRIER BARRIER TIME.IF APP. IF APP. NUMBER (SEC)(26) (20) l $ i
- 217 RCIC VACUUM PUMP DISCH HV49-1F002 60 5,29 49 49-1F028(CK) NA 218 INSTRUMENT GAS TO 59-1001(CK) NA 59 VACUUM RELIEF VALVES HV59-135 7 C,H,5 219A INSTRUMENTATION - --
HV55-121 45 10 55 i SUPPRESSION P00L l LEVEL INSTRUMENTATION - HV55-120 45 10 55 M 219B SUPPRESSION P00L T LEVEL Yn 220A H2/02 SAMPLE RETURN SV57-191(X-62) 5 B,H,R,5 11 57 l l SV57-190(X-62) 5 B,H,R,5 11 ! HV57-116(X-62) 30** B,H,R,5 11 SV57-150(X-62) 5 B,H,R,5 11 l SV57-159(X-62) 5 B,H,R,S 11 i INSTRUMENTATION - SV57-101 5 10 57 i 220B l SUPPRESSION POOL PRESSURE l SUPPRESSION POOL LEVEL l 221A WETWELL H2/02 SAMPLE SV57-181 5 B,H,R,5 11 57 ! SV57-141 5 B,H,R,5 11 SV57-184 5 B,H,R,5 11 221B WETWELL H2/02 SAMPLE SV57-183 5 B,H,R,5 11 57 SV57-186 5 B,H,R,S 11 O O O
4
\ f i TABLE 3.6.3-1 (Continued)
C PART A - PRIMARY CONTAINMENT ISOLATION VALVES ' ISOL. E INBOARD OUTBOARD ISOLATION ISOLATION- MAX.ISOL. SIGNAL (S), NOTES P&ID 7 PENETRATION FUNCTION BARRIER BARRIER TIME.IF APP. IF APP.
- c NUMBER (SEC)(36) (20) -
~'
z
~
RHR VACUUM RELIEF SUCTION HV51-130 60 B,H 4,11 51-
- 225 l HV51-131 60 8,H 11 i
RHR MINIMUM RECIRC HV51-105A 40 4,22,29 51 226A RHR MINIMUM RECIRC HV51-105B 40 4,22,29 51 2268 60-1073 NA 5 60 227 ILRT DATA ACQUISITION , SYSTEM 60-1074 NA > HPCI VACUUM RELIEF HV55-1F095 40 H,LA 4,11,24 55
$228D HV55-1F093 40 H,LA 11,24 j ,
INSTRUMENTATION - DRYWELL HV61-102 45 1,23,29 61
$230B SUMP LEVEL HV61-112 45 23,29 i
J HV61-132 ~ 45 23,29 w 231A DRYWELL FLOOR DRAIN -HV61-110 30 B,H 11,22 61 l 11,22 1 SUMP DISCHARGE HV61-111 30 B,H-231B HV61-130 30 B,H 11,22 61 i DRYWELL EQUIPMENT DRAIN TANK DISCHARGE HV61-131 30 8,H 11,22 CS PUMP MINIMUM RECIRC HV52-1F031A 45 5,22,29 52 235 , HPCI PUMP MINIMUM RECIRC HV55-1F012 15 LFHP 5,22 55 236 I l l
TABLE 3.6.3-1 (Continued) M PART A - PRIMARY CONTAINMENT ISOLATION VALVES E ISOL. p INBOARD OUTB0ARD SIGNAL (S), NOTES P&ID PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. NUMBER BARRIER BARRIER TIME.IF APP. IF APP. e z (SEC)(26) (20) _ O
~ 237-1 SUPPRESSION POOL CLEANUP HV52-127 60 B,H 4,11,22 52 PUMP SUCTION PSV52-127 NA 11,22 HV52-128 60 B,H 11,22 HV52-139 45 10 52 237-2 SUPPRESSION P0OL LEVEL INSTRUMENTATION SV52-139 6 10 RHR RELIEF VALVE HV-C-51-1F104B 18 C,G 51 238 DISCHARGE PSV51-106B NA 19 PSV51-1F0558 NA 19 PSV51-1018 NA 19 HV-C-51-1F103A 18 C,G 51
[ca239 RHR RELIEF VALVE DISCHARGE PSV51-106A NA 19 PSV51-1F055A NA 19 PSV51-101A NA 19 RHR RELIEF VALVE PSV51-1F097 NA 19 51 f 240 DISCHARGE RCIC VACUUM RELIEF HV49-1F084 40 H,KA 4,11,24 49 241 HV49-1F080 40 H,KA 11,24 1 l l l 1 O O O
( - TABLE 3.6.3-1 (Continued) C M x PART B - PRIMARY CONTAINMENT-ISOLATION EXCESS FLOW CHECK VALVES INB0ARD' OUTBOARD ISOL. Q MAX.ISOL. SIGNAL (S), NOTES P&ID
. PENETRATION FUNCTION ISOLATION ISOLATION.
c- NUMBER BARRIER, BARRIER. TIME.IF APP. IF APP. i'i (SEC)(26) (20) -4 "" 003A-1 INSTRUMENTATION 'D' -- XV41-1F070D 1 41 MAIN STEAM LINE FLOW XV41-IF073D 003A-2 INSTRUMENTATION 'A' -- XV43-1F003A 1 ~43 RECIRC PUMP SEAL PRESSURE 003C-l' INSTR. - HPCI STEAM FLOW -- XV55-1F024A 1 55 003C-2 INSTR. - HPCI STEAM FLOW -- XV55-1F024C 1 55
'A' MAIN STEAM XV41-1F070A~ 1 41
$003D-1 INSTR. -- m LINE FLOW XV41-1F073A 007A(B,C,0) INSTR 'A'(B,C,D) MAIN (HV41-1F022A(B,- -5 C,D,E,F,P,Q 6 41 STEAM LINE PRESSURE -C,D) SEE PART A (HV41-1F028A 5 C,D,E,F,P,Q 6
-THIS TABLE) .(B,C, D) SEE PART A THIS-TABLE)
(HV40-1F001B 1 e (F,K,P) SEE PART A THIS TABLE) XV40-101B(F,. 1 K,P) 020A-1 INSTR - RPV LEVEL -- XV42-1F045B 1 42 020A-2 INSTR 'B' LPCI DELTA P -- XV51-102B 1 51 020A-3 INSTR 'D'~LPCI DELTA P -- XV51-103B 1 51 020B-1 INSTR - RPV LEVEL -- XV42-1F045C 1 42 020B-2 INSTR 'C' LPCI DELTA P -- XV51-102C 1 51
TABLE 3.6.3-1 (Continued) PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES E INBOARD OUTBOARD ISOL. FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID 7 PENETRATION NUMBER BARRIER BARRIER TIME.IF APP. IF APP. E (SEC)(26) (20)
-4 s- 027B-1 INSTR - HPCI FLOW --
XV55-1F024B 1 55 0278-2 INSTR - HPCI FLOW -- XV55-1F0240 1 55 029A INSTR - RPV FLANGE -- XV41-1F009 1,27 41 LEAKAGE 029B INSTR - CS DELTA P -- XV52-1F018A 1 52 030A INSTR 'D' MAIN STEAM -- XV41-1F071D 1 41 FLOW XV41-1F0720 { i 030B-2 INSTR 'C' MAIN STEAM -- XV41-1F071C 1 41 U LINE FLOW XV41-1F072C 031A INSTR - JET PUMP FLOW -- XV42-1F0598 1 42 (JP1) XV42-1F0590 (JP2) XV42-1F059F (JP3) 031B INSTR - JET PUMP FLOW -- XV42-1F059H . 1 42 (JP4) XV42-1F051B (JPS) XV42-1F0538 (JP6) 1 i l O O O
_ _ . . _ _ . . . _ . _ . . . _ . . _ , ~ . . . _ . _ _ _ . _ . . - _, _ _ , . _ . . .
'\ .
l y
. TABLE 3.6.3-1 (Continued) - .% PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES S
rs INBOARD OUTBOARD ISOL. . 7 PENETRATION FUNC110N ISOLATION ISOLATION MAX.ISOL. SIGNAL (S),- NOTES P&ID i c NUMBER BARRIER BARRIER TIME.IF APP, IF APP. .
= (SEC)(26) (20) w 032A INSTR - JET PUMP FLOW --
XV42-1F059M 1 42 (JP6) XV42-IF059P ' l (JP7)
- XV42-1F0595 (JP8) j 0328 INSTR - JET PUMP FLOW --
XV42-1F0590 1 42 (JP9) ; XV42-1F051D (JP10)
- um XV42-1F0530 .-
3 (JP10) 52 033A-1 INSTR-PRESSURE AB0VE -- XV42-1F055 1 42 ) CORE PLATE. XV42-1F076 i
- 033A-2 INSTR-PRESSURE BELOW --
XV42-1F061 1 42 * ! CORE PLATE i t t j 033B INSTR-RCIC STEAM FLOW -- XV49-1F044A,C 1 49 [ i - 1 034A INSTR 'C' MAIN STEAM -- XV41-1F070C 1 42
- LINE FLOW
- XV41-1073C ,
i 034B-1 INSTR - RECIRC FLOW -- XV43-1F009C 1 43 : XV43-1F0100 1 034B-2 INSTR - RECIRC FLOW -- XV43-1F009D 1 43 XV43-1F010C i k l I. _ _ _ _ . . _ _ _ , _ , _
TABLE 3.6.3-1 (Continued) PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES E INBOARD OUTBOARD ISOL. FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&IO l 7 PENETRATION BARRIER BARRIER TIME.IF APP. IF APP. NUMBER E (SEC)(26) (20) w XV42-1F059L 1 42 I - 040A INSTR - JET PUMP FLOW -- (JP15) XV42-1F059N (JP17) XV42-1F059R (JP18) INSTR - JET PUMP FLOW -- XV42-1F059G 1 42 040B (JP14) XV42-1F051A R (JP16) XV42-1F053A T (JP16)
% 1 42 040C INSTR - JET PUMP FLOW --
XV42-1F059A (JP11) XV42-1F059C (JP12) XV42-1F059E (JP13) XV42-1F057 1 42 040D-1 INSTR - PRESSURE BELOW CORE PLATE XV44-170 1 44 0400-2 INSTR - RWCU BOTTOM DRAIN FLOW XV44-171 O O O
,n .
\ . _
TABLE 3.6.3-1 (Continued) C g PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES 5 ISOL. . p INBOARD OUTBOARD ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID
, PENETRATION FUNCTION ISOLATION BARRIER BARRIER TIME.IF APP. IF APP.
e NUMBER (SEC)(26) (20) 3
-4 XV49-1F044B 1 49
- 040F-1 INSTR - RCIC STEAM FLOW XV49-1F044D XV87-156B 17 87 040H-2 INSTR 'B' RECIRC PUMP COOLER FLOW XV87-1578
-- XV44-102A,8 1 44 041-1 INSTR - RWCU FLOW XV51-103A 1 51 041-2 INSTR 'A' LPCI DELTA P --
-- XV43-1F040A,C 1 43
$043A INSTR - RECIRC LOOP m 'A' DELTA P XV44-102D 1 44 047 INSTR - RWCU FLOW XV42-1F065B 1 42 INSTR - RPV LEVEL 048A-1 XV42-1F0478 XV52-1F018B 1 52 048A-2 INSTR - CS DELTA P XV42-1F065A 1 42 048B INSTR - RPV LEVEL XV42-1F047A XV41-1F071A,B 1 41 049A,8 INSTR 'A' AND 'B' MAIN --
STEAM LINE FLOW XV41-1F072A,8 XV43-1F011A 1 43 050A-2 INSTR 'B' RECIRC FLOW XV43-1F0128
1 TABLE 3.6.3-1 (Continued) PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES y INBOARD OUTBOARD ISOL.
, PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID c NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
=, (SEC)(26) (2C) - 050A-3 INSTR 'B' RECIRC FLOW -- XV43-1F0118 1 43 XV43-1F012A 0508-1 INSTR 'A' RECIRC PUMP -- XV43-1F004A 1 43 SEAL PRESSURE 050B-2 INSTR 'A' RECIRC PUMP -- XV87-156A 17 87 COOLER FLOW XV87-157A 051A-1 INSTR 'A' RECIRC LINE -- XV43-1F009A 1 43 $ FLOW XV43-1F010B i 051A-2 INSTR 'A' RECIRC LINE -- XV43-1F009B 1 43 $ FLOW XV43-1F010A 051B INSTR - JET PUMP FLOW -- XV42-1F059T . 42 (JP19) XV42-1F051C (JP20) - XV42-1F053C (JP20) 052A INSTR 'B' MAIN STEAM -- XV41-1F0708 1 41 LINE FLOW XV41-1F073B 052B-1 INSTR 'B' RECIRC -- XV43-1F011C,D 1 43 LINE FLOW 0528-2 INSTR 'B' RECIRC -- XV43-1F012C,0 1 43 LINE FLOW 057 INSTR - RWCU FLOW -- XV44-102C 1 44
~
O O O
._ m . . _ . . - . ._ - _ . . . _. _ _ _ . - . _ , . __ . . _ _ _ ,_..m. ,_ . . _ .- .
l
/ t i
j TABLE 3.6.3-1 (Continued) ^ 3
- r-PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES i
E INBOARD OUTBOARD ISOL. 7 PENETRATION FUNCTION ISOLATION. . ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID , c NUMBER BARRIER BARRIER TIME.IF APP. IF APP. j z (SEC)(26) (20) l
- - 058A INSTR - RECIRC LOOP --
XV43-1F0408 -1 43 l
- 'B' DELTA P , 061-1 RECIRC PUMP SEAL (43-1004A(CK) - 15 43
- PURGE See Part A of XV43-103A 1
?, this table) l 061-2 RECIRC PUMP SEAL (43-1004B(CK) 15 43 PURGE See Part A of XV43-103B 1 - m this table) I k
, 063-1 INSTR - RECIRC LOOP 'B' --
XV43-1F0400 1 43
; DELTA P ,
4 % i 063-2 INSTR 'B' RECIRC PUMP -- XV43-1F004B 1 43 ' SEAL PRESSURE XV43-1F0038 065A - INSTR - RPV PRESSURE -- XV42-1F043B 1 42 065B INSTR - RPV PRESSURE -- XV42-1F049A 1 42 066A-1 INST-RPV LEVEL -- XV42-1F0450 _
- 1 42 ;
i 066A-2 INSTR 'B' LPCI DELTA P -- XV51-102D 1 51 q XV51-103D I j 066B-1 INST - RPV LEVEL -- XV42-1F045A - 1 42 t j 0668-2 INST 'A' LPCI DELTA P -- XV51-102A 1 51 j XV51-103C 4 i 067A INSTR - RPV PRESSURE -- XV42-1F049B 1 42 i l i s
s TABLE 3.6.3-1 (Continued) b PART B - PRIMARY CONTAINMENT ISOLAION EXCESS FLOW CHECK VALVES E$ INBOARD OUTBOARD ISOL.
$' PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
EE (SEC)(26) (20)
~ 0678-1 INSTR - RPV PRESSURE --
XV42-1F043A 1 42 0678-2 INSTR - RPV LEVEL -- XV42-1F041 1 42 102A INST - JET PUMP, REACTOR -- XV42-185A(JP16) 1 42 LEVEL 107 INST. - JET PUMP, REACTOR -- XV42-18SB(JPS) 1 42 LEVEL M a a O O O -
7 -~3.. . ( b Q TABLE 3.6.3-1 (Continued) PART C - PRIMARY CONTAINMENT PENETRATIONS (TYPE B) E5 INBOARD -0UTBOARD ISOL.' PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID '[ NUMBER BARRIER BARRIER TIME.IF APP. IF APP. !E (SEC)(26) (20) -e sa NA DRYWELL HEAD FLANGE DOUBLE 0-RING -- -- -- 2 60 001 EQUIPMENT ACCESS. DOOR DOUBLE 0-RING -- -- -- 2 60 -
~
002 EQUIPMENT ACCESS DOOR DOUBLE 0-RING -- -- -- 2;18 60 AND PERSONNEL LOCK 004 HEAD ACCESS MANHOLE DOUBLE 0-RING -- -- -- 2 60 - ,,006 CRD REMOVAL HATCH DOUBLE 0-RING -- -- -- 2 60 , s NEUTRON MONITORING SYSTEM CANISTER -- -- -- 8 60 ll100A-D E$101A-D RECIRC PUMP POWER CANISTER -- -- -- 8 60 103A,B TEMPERATURE AND CANISTER -- -- -- 8 60 LOW LEVEL SICNALS ' 104A-D CR0 POSITION INDICATOR CANISTER -- -- -- 8 60 105A-E MISCELLANEOUS LOW- CANISTER -- -- -- 8 60 VOLTAGE CONTROL POWER - 106A-C LOW-VOLTAGE CONTROL ' - CANISTER -- -- --
~8 60
TABLE 3.6.3-1 (Continued) C g PART C - PRIMARY CONTAINMENT PENETRATIONS (TYPE B) 5 g INBOARD OUTBOARD ISOL.
, PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID c NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
3 (SEC)(26) (20) - 200A,B ACCESS HATCH DOUBLE 0-RING -- -- -- 2 60 222 INDICATION AND CONTROL CANISTER -- -- -- 8 60 230A STRAIN GAUGE INSTR. CANISTER -- -- -- 8 60 M = 8 O O O
TABLE 3.6.3-1 A PRIMARY CONTAINMENT ISOLATION VALVES NOTATION (G) NOTES
- 1. Instrumentation line isolation provisions consist of an orifice and expess flow-check valve or remote manual isolation valve. The excess flow-check valve is subjected to operability testing, but no Type C test is performed or required. The line does not isolate during a LOCA and can leak only if the line or instrument should rupture. Leaktightness of the line is verified during the integrated leak rate test (Type A test).
- 2. Penetration is sealed by a blind flange or door with double 0-ring seals.
These seals are leakage rate tested by pressurizing between the 0-rings.
- 3. Inboard butterfly valve tested in the reverse direction.
- 4. Inboard gate valve tested in the reverse direction. -
- 5. Inboard globe valve tested in the reverse direction.
- 6. The MSIVs and this penetration are tested by pressurizing between the valves.
J. - Testing of the inboard valve in the reverse direction tends to unseat the valve and is therefore conservative. The valves are Type C tested at a test pressure of 22 psig. p
- 7. Gate valve tested in the reverse direction.
- 8. Electrical penetrations are tested by pressurizing between the seals,
- 9. The isolation provisions for this penetration consist of two isolation valves and a closed system outside containment. Because a water seal is maintained in these lines by the safeguard piping fill system, the inboard valve may be tested with water. The outboard valve will be pneumatically tested.
- 10. The valve does not receive an isolation signal but remains open to measure containment conditions post-LOCA. Leaktightness of the penetra-tion is verified during the Type A test. Type C test is not required.
- 11. All isolation barriers are located outside containment.
- 12. Leakage monitoring of the control rod drive insert and withdraw line is provided by Type A leakage rate test. Type C test is not required.
- 13. The motor operators on HV-13-109 and HV-13-110 are not connected to any power supply.
- 14. Valve is provided with a separate testable seal assembly, with double concentric 0-ring seals installed between the pipe flange and valve flange facing primary containment. Leakage through these seals is included within the Type C leakage rate for this penetration.
.Dn)
( LIMERICK - UNIT 1 3/4 6-41
TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES NOTATION NOTES (Continued)
- 15. Check valve used instead of flow orifice.
- 16. Penetration is sealed by a flange with double 0-ring seals. These seals are leakage rate tested by pressurizing between the 0-rings. Both the TIP Purge Supply (Penetration 35A) and the TIP Drive Tubes (Penetration 35C-G) are welded to their respective flanges. Leakage through these seals is included in the Type C leakage rate total for this penetration. The ball valves (XV-141A-E) are Type C tested. It is not practicable to leak test the shear valves (XV-140A-E) because squib firing is required for closure.
Shear valves (XV-140A-G) are normally open. 17 Instrument line isolation provisions consist of an excess flow check valve. Because the instrument line is connected to a closed cooling water system inside containment, no flow orifice is provided. The line does not isolate during a LOCA and can leak only if the line or instrument should rupture. Leaktightness of the line is verified during the integrated leak rate test (Type A test).
- 18. In addition to double "0" ring seals, this penetration is tested by pres-surizing volume between doors per Specification 4.6.1.3.
- 19. The RHR system safety pressure relief valves will be exempted from the initial LLRT. The relief valves in these lines will be exposed to contain-ment pressure during the initial ILRT and all subsequent ILRTs. In addi-tion, modifications will be performed at the first refueling to facilitate local testing or removal and bench testing of the relief valves during sub-sequent LLRTs. Those relief valves which are flanged to facilitate removal will be equipped with double 0-ring seal assemblies on the flange closest to primary containment by the end of the first refueling outage. These seals will be leak rate tested by pressurizing between the 0-rings, and the results added into the Type C total for this penetration.
- 20. See Specification 3.3.2, Table 3.3.2-1, for a description of th&PCRVICS isolation signal (s) that initiate closure of each automatic isolation valve.
In addition, the following non-PCRVICS isolation signals also initiate closure of selected valves: EA Main steam line high pressure, high steam line leakage flow, low MSIV-LCS dilution air flow LFHP With HPCI pumps running, opens on low flow in associated pipe, closes when flow is above setpoint LFRC With RCIC pump running, opens on low flow in associated pipe, closes when flow is above setpoint LFCH With CSS pump running, opens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve fully closed or RCIC turbine stop valve fully closed All power operated isolation valves may be opened or closed remote manually. LIMERICK - UNIT 1 3/4 6-42
7m TABLE 3.6.3-1 l ) PRIMARY CONTAINMENT ISOLATION VALVES C/ N0IATION NOTES (Continued)
- 21. Automatic isolation signal causes TIP to retract; ball valve closes when probe is fully retracted.
- 22. Isolation barrier remains water filled or a water seal remains in the line post-LOCA. Isolation valve may be tested with water. Isolation valve leakage is not included in 0.60 La total Type B & C tests.
- 23. Valve does not receive an isolation signal. Valves will be open during lype A test. Type C test not required.
- 24. Both isolation signals required for valve closure.
- 25. Isolation capability upon refueling floor high radiation (signal R in Speci-fication 3.3.2, Table 3.3.2-1) and low differential pressure (signal T in Specification 3.3.2, Table 3.3.2-1) will be added by the end of the first refueling outage.
- 26. Valve stroke times listed are maximum times verified by testing per Speci-fication 4.0.5 acceptance criteria. The closure times for isolation valves in lines in which high-energy line breaks could occur are identified with a single asterisk. The closure times for isolation valves in lines which provide an open path from the containment to the environs are identified
/ with a double asterisk.
(]/ 27. The reactor vessel head seal leak detection line (penetration 29A) excess flow check valve is not subject to OPERABILITY testing. This valve will not be exposed to primary system pressure except under the unlikely con-ditions of a seal failure where it could be partially pressurized to reactor pressure. Any leakage path is restricted at the source; therefore, this valve need not be OPERABILITY tested.
- 28. Automatic isolation logic to be added by the end of the first refueling outage.
- 29. Valve may be open during normal operation; capable of manual isolation from control room. Position will be controlled procedurally.
- 30. Valve normally open, closes on scram-signal.
- 31. Valve 41-1016 is an outboard isolation barrier for penetrations X-9A, B and X-44. Leakage through valve 41-1016 is included in the total for penetration X-44 only.
- 32. Feedwater long path recirculation valves are sealed closed whenever the reactor is critical and reactor pressure is greater than 600 psig. The valves are expected to be opened only in the following instances:
- a. Flushing of the condensate and feedwater systems during plant startup.
- b. Reactor pressure vessel hydrostatic testing, which is conducted follow-l ing each refueling outage prior to commencing plant startup.
i,A. Therefore, valve stroke timing in accordance with Specification 4.0.5 is not required. l (v) LIMERICK - UNIT 1 3/4 6-43
CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION FOR OPERATION i 3.6.4.1 Each pair of suppression chamber - drywell vacuum breakers shall be OPERABLE and closed. APPLICABILITY: OPERATIONAL CONP'TIONS 1, 2, and 3. ACTION:
- a. With one or more vacuum breakers in one pair of suppression chamber -
drywell vacuum breakers inoperable for opening but known to be closed, restore the inoperable pair of vacuum breakers to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With one suppression chamber - drywell vacuum breaker open, verify the other vacuum breaker in the pair to be closed within 2 hours; restore the open vacuum breaker to the closed position within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- c. With one position indicator of any suppression chamber - drywell vacuum breaker inoperable:
- 1. Verify the other vacuum breaker in the pair to be closed within 2 hours and at least once per 15 days thereafter, or
- 2. Verify the vacuum breaker (s) with the inoperable position indicator to be closed by conducting a test which demonstrates that the AP is maintained at greater than or equal to 0.7 psi for one hour without makeup within 24 hours and at least once per 15 days thereafter.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. O LIMERICK - UNIT 1 3/4 6-44
= . . . - -. _ . - _ . _ . . -- _. .
+ l CONTAINMENT SYSTEMS G SURVEILLANCE REQUIREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be:
- a. Verified closed at least once per 7 days. -
- b. Demonstrated OPERABLE:
. 1. At least once per 31 days and within 2 hours after any discharge of steam to the suppression chamber from the safety / relief valves, by cycling each vacuum breaker through at least one complete - cycle of full travel.
- 2. At least once per 31 days by verifying both position indicators -
OPERABLE by observing expected valve movement during the cycling test.
- 3. At least once per 18 months by; a) Verifying each valve's opening'setpoint, from the closed position, to be 0.5 psid i 5%, and b) Verifying both position indicators OPERABLE by performance L O c) of a CHANNEL CALIBRATION.
Verifying that each outboard valve's~ position indicator is capable of detecting disk-displacement >0.050", and each inboard valve's position indicator
- is capable of detecting disk displacement >0.120".
l V LIMERICK'- UNIT 1 3/4 6-45
,, . - . . - - . ~ _ - . - - - - . _ . - _ - . _ - _ . . - . . _ - _ _ . - . - .
CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.3 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained. APPLICA3ILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon-strated by:
- a. Verifying at least once per 24 hours that the pressure within the reactor enclosure secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.
- b. Verifying at least once per 31 days that:
- 1. All reactor enclosure secondary containment equipment hatches and blowout panels are closed and sealed.
- 2. At least one door in each access to the reactor enclosure secondary containment is closed.
- 3. All reactor enclosure secondary containment penetrations not capable of being closed by OPERABLE secondary containment auto-matic isolation dampers / valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers / valves secured in position.
- c. At least once per 18 months:
- 1. Verifying that one standby gas treatment subsystem will draw down the reactor enclosure secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than or equal to 121 seconds with the reactor enclosure recirc system in operation, and
- 2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inch of vacuum water gauge in the reactor enclosure secondary containment at a flow rate not exceeding 1250 cfm.
O LIMERICK - UNIT 1 3/4 6-46
[]
)
/
CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: OPERATIONAL CONDITION *. ACTION: Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, susperid handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specifica-tion 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:
- a. Verifying at least once per 24 hours that the pressure within the
[ ^s refueling area secondary containment is greater than or equal to
! 0.25 inch of vacuum water gauge.
- b. Verifying at least once per 31 days that:
- 1. All refueling area secondary containment equipment hatches >and ,
blowout panels are closed and sealed.
- 2. At least one door in each access to the refueling area secondary containment is closed.
- 3. All refueling area secondary containment penetrations not capable of being closed by CPERABLE secondary containment automatic iso-lation dampers / valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers / valves secured in position.
.c . At least once per 18 months:
Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal.to 0.25 inch of vacuum water gauge in the refueling area secondary containment at a flow rate not exceeding 764 cfm. f
*When irradiated fuel is being handled in the refueling area secondary contain-
. ment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
('] , t V) LIMERICK - UNIT 1 3/4 6-47
REACTOR CONTAINMENT SYSTEMS REACTOR Ektwaunt SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.1 The reactor enclosure secondary containment ventilation system auto-matic isolation valves shown in Table 3.6.5.2.1-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.2.1-1. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With one or more of the reacto secondary containment ventilation system automatic isolation valve, shown in Table 3.6.5.2.1-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours either:
- a. Restore the inoperable valves to OPERABLE status, or
- b. Isolate cach affected penetration by use of at least one deactivated valve secured in the isolation position, or
- c. Isolate each affected penetration by use of at least one closed manual valve or blind flange.
Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.6.5.2.1 Each reactor enclosure secondary containment ventilation system automatic isolation valve shown in Table 3.6.5.2.1-1 shall be demonstrated OPERABLE:
- a. 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 cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
- b. At least once per 18 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.
- c. By verifying the isolation time to be within its limit at least once per 92 days.
O LIMERICK - UNIT 1 3/4 6-48
/]/ TABLE 3.6.5.2.1-1 REACTOR ENCLOSURE SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES REACTOR ENCLOSURE (ZONE I) MAXIMUM ISOLATION TIME ISOLATIgN)
VALVE FUNCTION (Seconds) SIGNALS
- 1. Reactor. Enclosure Ventilation Supply Valve HV-76-107 5 B,H,S,U
- 2. ' Reactor Enclosure Ventilation Supply Valve HV-76-108 5 B,H,S,U
- 3. Reactor Enclosure Ventilation Exhaust Valve HV-76-157 5 B,H,S,0
- 4. Reactor Enclosure Ventilation Exhaust Valve HV-76-158 5 B,H,5,0
- 5. Reactor Enclosure Equipment Compartment Exhaust Valve HV-76-141 5 B,H,S,U
- 6. Reactor Enclosure Equipment Compartment Exhaust Valve HV-76-142 5 B,H,S,U
- 7. Drywell Purge Exhaust Valve HV-76-030(D) 5 B,H,S,0
- 8. Drywell Purge Exhaust Valve HV-76-031(D) 5 B,H,S,U
.p -i ) \.J .
(a)See Specification 3.3.2, Table 3.3.2-1, for isolation signals that operate each automatic valve. (b) Isolation capability upon refueling floor high radiation (Signal R in Specification 3.3.2, Table 3.3.2-1) and low differential pressure (Signal T V) ( in Specification 3.3.2, Table 3.3.2-1) will be added prior to handling irradiated fuel in the refueling area secondary containment. LIMERICK UNIT-1 , 3/4 6-49
CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shown in Table 3.6.5.2.2-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.2.2-1. APPLICABILITY: OPERATIONAL CONDITION *. ACTION: With one or more of the refueling area secondary containment ventilation system automatic isolation valves shown in Table 3.6.5.2.2-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours either:
- a. Restore the inoperable valves to OPERABLE status, or
- b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or
- c. Isolate each affected penetration by use of at least one closed manual valve or blind flange.
Otherwise, in Operational Condition *, suspend handling of irradiated fuel in the refueling area secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shown in Table 3.6.5.2.2-1 shall be demonstrated OPERABLE:
- a. 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 cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
- b. At least once per 18 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position,
- c. By verifying the isolation time to be within its limit at least once per 92 days.
*When irradiated fuel is being handled in the refueling area secondary contain-meat and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
LIMERICK - UNIT 1 3/4 6-50
l TABLE 3.6.5.2.2-1 ( ) REFUELING AREA SECONDARY CONTAINMENT VENTILATION SYSTEM
\w/ AUTOMATIC ISOLATION VALVES REACTOR ENCLOSURE (ZONE III) MAXIMUM ISOLATION TIME ISOLATIg)
VALVE FUNCTION (Seconds) SIGNALS
- 1. Refueling Area Ventilation Supply Valve HV-76-117 (Unit 1) 5 R,T
- 2. Refueling Area Ventilation Supply Valve HV-76-118 (Unit 1) 5 R,T
- 3. Refueling Area Ventilation Exhaust Valve HV-76-167 (Unit 1) 5 R,T
- 4. Refueling Area Ventilation Exhaust Valve HV-76-168 (Unit 1) 5 R,T
- 5. Refueling Area Ventilation Supply Valve HV-76-217 (Unit 2)** 5 R,T
- 6. Refueling Area Ventilation Supply Valve HV-76-218 (Unit 2)** 5 R,T
- 7. Refueling Area Vdntilation Exhaust Valve HV-76-267 (Unit 2)** 5 R,T
- 8. Refueling Area Ventilation Exhaust 77 Valve HV-76-268 (Unit 2)** 5 R,T t \
- 9. Drywell Purge Exhaust Valve HV-76-030(D) 5 B,H,S,U
- 10. Drywell Purge Exhaust Valve HV-76-031(D) 5 8,H,S,U
*The provisions of Specification 3.0.4 are not applicable.
**These lines are blanked off during Unit 1 operation / Unit 2 construction.
(a)See Specification 3.3.2, Table 3.3.2-1, for isolation signals that operate each automatic isolation valve. ( ) Isolation capability upon refuelirig floor high radiation (Signal R in Specification 3.3.2, Table 3.3.2-1) and low differential pressure (Signal T /'~N. in Specification 3.3.2, Table 3.3.2-1) will be added prior to handling (v ) irradiated fuel in the refueling area secondary containment. LIMERICK - UNIT 1 3/4 6-51
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and *. ACTION:
- a. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:
- 1. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.
- 2. In Operational Condition * , suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.
- b. With both standby gas treatment subsystems inoperable in Operational Condition *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS or operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3.
are not applicable. SURVEILLANCE REQUIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:
- a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.
*When irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
O LIMERICK - UNIT 1 3/4 6-52
CONTAINMENT SYSTEMS t a
> j SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or cha> coal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:
- 1. Verifying that the subsystem 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.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm i 10%.
- 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 0.175%; and
- 3. Verifying a subsystem flow rate of 3000 cfm i 10% during system operatior,when tested in accordance with ANSI N510-1980.
- c. After every 720 hours of charcoal adsorber operation by verifying
/^ i within 31 days after removal that a laboratory analysis of a repre-( / sentative carbon sample obtained in accordance with Regulatory N# 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, for a mathyl iodide penetration of less than 0.175%.
- 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 4.8 inches water gauge while operating the filter train at a flow rate of 3000 cfm i 10%.
- 2. Verifying that the fan starts and isolation valves necessary to draw a suction from the refueling area
- or the reactor enclosure recirculation discharge open on each of the following test signals:
a) Manual initiation from the control room, and b) Simulated automatic initiation signal.
- 3. Verifying that the standby gas treatment system can be placed in the cooldown mode of operation from 'the control room.
)
- Capability to draw a suction from the refueling area will be added prior to e/ handling irradiated fuel in the refueling area secondary containment.
LIMERICK - UNIT 1 3/4 6-53
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 9
- 4. Verifying that the temperature differential across each heater is > 15 F when tested in accordance with ANSI N510-1980.
- e. After each complete or partial vsplacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 3000 cfm i 10%.
- f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and 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 3000 cfm i 10%. O i O LIMERICK - UNIT 1 3/4 6-54
/] CONTAINMENT SYSTEMS REACTOR ENCLOSURE RECIRCULATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one reactor enclosure recirculation subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With both reactor enclosure recirculation subsystems inoperable, be in at least H0T SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS O 4.6.5.4 OPERABLE: Each reactor enclosure recirculation subsystem shall be demonstrated
- a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates properly.
- 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 subsystem by:
- 1. Verifying that the subsystem 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.S.a. C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 60,000 cfm i 10%.
- 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%; and
- 3. Verifying a subsystem flow rate of 60,000 cfm i 10% during system
[] V operation when tested in accordance with ANSI N510-1980. LIMERICK - UNIT 1 3/4 6-55
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 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, for a methyl iodide penetration of less than 1%.
- d. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the filter train at a flow rate of 60,000 cfm i 10%, verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.
- 2. Verifying that the filter train starts and the isolation valves which take suction on and return to the reactor enclosure open on each of the following test signals:
- a. Manual initiation from the control room, and
- b. Simulated automatic initiation signal.
- 3. Vcrifying that the reactor enclosure recirculation system can be placed in the cooldown mode from the Control Room,
- e. Af ter each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 60,000 cfm i 10%.
- f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 60,000 cfm i 10%.
O' LIMERICK - UNIT 1 3/4 6-56 i i
( CONTAINMENT SYSTEMS 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL PRIMARY CONTAINMENT HYDR 0 GEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.5.6.1 Two independent primary containment hydrogen recombiner systems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With one primary containment hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.6.6.1 Each primary containment hydrogen recombiner system shall be demon-strated OPERABLE:
- a. At least once per 6 months by performance of:
- 1. A CHANNEL CHECK of all Control Room Recombiner Instrumentation.
- 2. A Trickle Heat Circuit check, p 3. A Heater Coil Check.
V 4. A verification of valve operation by stroking all the valves to their proper positions.
- b. At least once per 18 months by:
- 1. Performing a CHANNEL CALIBRATION of all control room recombiner instrumentation and control circuits.
- 2. Verifying the integrity of all heater electrical circuits by perform-ing a resistance to ground test within 30 minutes following the below ,
required functional test. The resistance to ground for any heater phase shall be greater than or equal to 100 megohms. -
- 3. 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.
- 4. Verifying during a recombiner sysMm functional test that the minimum heater outlet gas temperature increases to greater than or equal to 1150 F within ]?0 minutes and maintained for at least one' hour.
- c. By measuring the system leakage rate:
- 1. As a part of the overall integrated leakage rate test required by Specification 3.6.1.2, or
- 2. By measuring the leakage rate of the system outside of the contain-ment isolation valves at P ,a 44.0 psig, on the schedule required by Specification 4.6.1.2, and including the measured leakage as a part d of the leakage determined in accordance with Specification 4.6.1.2.
LIMERICK - UNIT 1 3/4 6-57
CONTAINMENT SYSTEMS ORYWELL HYDR 0 GEN MIXING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.2 Four independent drywell unit cooler hydrogen mixing subsystems (1AV212, IBV212, 1GV212, 1HV212) shall be OPERABLE with each subsystem consist-ing of one unit cooler fan. APPLICABILITY: OPERATIONAL CONDITIONS 1 and ^ ACTION: With one drywell unit cooler hydrog. . nixing subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hou SURVEILLANCE REQUIREMENTS 4.6.6.2 Each drywell unit cooler hydrogen mixing subsystem shall be demonstrated OPERABLE at least once per 92 days by:
- a. Starting the system from the control room, and
- b. Verifying that the system operates for at least 15 minutes.
1 Ol LIMERICK - UNIT 1 3/4 6-58
-- - - . - -_ - . ~ . . - .- ..- --
-D 7 CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER OXYGEN CONCENTRATION t LIMITING CONDITION FOR OPERATION 3 3.6.6.3 The drywell and suppression chamber atmosphere oxygen concentration ! shall be less than 4% by volume. 4- APPLICABILITY: OPERATIONAL CONDITION 1*, during the time period:
- a. Within 24 hours ** after THERMAL POWER is greater than 15% of RATED THERMAL POWER, following startup, to
- b. Within 24 hours ** prior to reducing THERMAL POWER to less than 15% of RATED THERMAL POWER, preliminary to a scheduled reactor shutdown.
ACTION: With the drywell and/or suppression chamb'er oxygen concentration exceeding the. limit, restore the oxygen concentration to within the limit . A within 24 hours or be in at least STARTUP within the next 8 hours. I. t SURVEILLANCE REQUIREMENTS , l 4.6.6.3 'The drywell and suppression chamber oxygen concentration shall be verified to be within the limit within 24 hours after THERMAL POWER is i greater than 15% of RATED THERMAL POWER and at least once per 7 days thereafter. L I I
*See Special Test Exception 3.10.5.
** Specification 3.6.1.8 is applicable during this 24 hour period.
1 LIMERICK - UNIT 1 .3/4 6-59 -
3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.1 At least the following independent residual heat removal service water (RHRSW) system subsystems, with each subsystem comprised of:
- a. Two OPERABLE RHRSW pumps, and
- b. An OPERABLE flow path capable of taking suction from the RHR service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water through one RHR heat exchanger, shall be OPERABLE:
- a. In OPERATIONAL CONDITIONS 1, 2, and 3, two subsystems.
- b. In OPERATIONAL CONDITIONS 4 and 5, the subsystem (s) associated with systems and components required OPERABLE by Specification 3.4.9.1, 3.4.9.2, 3.9.11.1, and 3.9.11.2.
(' APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5. ACTION:
- a. In OPERATIONAL CONDITION 1, 2, or 3:
- 1. With RHRSW pump A or B inoperable, verify the capability to power RHRSW pump C or D, as applicable, from the applicable Unit 1 diesel generator bus within 2 hours and at least once per 12 hours there-after; restore the inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With RHRSW pump C or D inoperable, restore the inoperable pump to OPERABLE status within 92 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 3. With RHRSW pumps A and B inoperable, restore RHRSW pump A or B to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
With one RHRSW pump in each subsystem inoperable, restore at least !. one of the inoperable RHRSW pumps to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD l SHUTDOWN within the following 24 hours.
- 4. With one RHRSW subsystem (RHRSW pumps A&C or B&D) inoperable, n
l[ j restore the inoperable subsystem to OPERABLE status with at least y/ one OPERABLE RHRSW pump within 72 hours or be in at least HOT l SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. l LIMERICK - UNIT 1 3/4 7-1
PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- 5. With both RHRSW subsystems inoperable, restore at least one subsystem to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours,and in COLD SHUTDOWN
- within the following 24 hours.
- b. In OPERATIONAL CONDITION 3 or 4 with the RHRSW subsystem (s), which is associated with an RHR loop required OPERABLE by Specification 3.4.9.1 or 3.4.9.2, ine:erable, declare the associated RHR loop inoperable and take the ACTION required by Specification 3.4.9.1 or 3.4.9.2, as applicable.
- c. In OPERATIONAL CONDITION 5 with the RHRSW subsystem (s), which is associated with an RHR loop required OPERABLE by Specification 3.9.11.1 or 3.9.11.2, inoperable, declare the associated RHR system inoperable and take the ACTION required by Specification 3.9.11.1 or 3.9.11.2, as applicable.
- d. In all OPERATIONAL CONDITIONS, if any connection between the RHRSW and Limerick Unit 2 is open, the appropriate subsystem shall be declared inoperable. The inoperable subsystem shall be restored to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.1 At least the above required residual heat removal service water system subsystem (s) shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve 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 the isolation function occurs on a radiation test signal.
RWhenever both RHRSW subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. O LIMERICK - UNIT 1 3/4 7-2
--. PLANT SYSTEMS / i
( l EMERGENCY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:
- a. Two OPERABLE emergency service water pumps, and
- b. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water to the associated safety related equipment, shall be OPERABLE:
- a. In OPERATIONAL CONDITIONS 1, 2, and 3, two loops.
- b. In OPERATIONAL CONDITIONS 4, 5, and *, one loop.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2, or 3:
.T'N 1. With one emergency service water pump inoperable, restore the d j-inoperable pump to OPERABLE-status within 45 days or be in a least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With one emergency service water pump in each loop inoperable, restore at least one inoperable pump to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 3. With one emergency service water system loop inoperable, align all the diesel generators to the .svailable loop ** and declare all equipment aligned to the inoperable loop inoperable. Restore the inoperable loop to OPERABLE status with at least one OPERABLr, pump within 72 hours or be in at least HOT SHUTDOWN within che next 12 hours and in COLD SHUT 00WN within the following 24 hours.
- b. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours or:
- 1. In OPERATIONAL CONDITION 4 or 5, declare the associated safety related equipment inoperable and take the ACTION required by Specifications 3.5.2 and 3.8.1.2.
p *When handling irradiated fuel in the secondary containment.
**Until after completion of confirmatory flow testing, all the diesel l 'v} generators shall not be aligned to the available loop and those not aligned shall have their equipment declared inoperable.
LIMERICK - UNIT 1 3/4 7-3
PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
- 2. In Operational Condition *, verify adequate coolir.g remains available for the diesel generators required to be OPERABLE or declare the associated diesel generator (s) inoperable and take the ACTION required by Specification 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.
- c. In all OPERATIONAL CONDITIONS, if any connection between ESW and Limerick Unit 2 is open, tha appropriate subsystem shall be declared inoperable. The inoperable system shall be restored to OPERABLE status within 8 hours, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.2 At least the above required emergency service water system loop (s) shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. At least once par 18 months during shutdown, by verifying that:
- 1. Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.
- 2. Each pump starts automatically when its associated diesel generator starts.
O LIMERICK - UNIT 1 3/4 ,-4
PLANT SYSTEMS o I S ULTIMATE HEAT SINK . V LIMITING CONDITION FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:
- a. A minimum pond water level at or above elevation 250' Mean Sea Level, (single-unit operation) and
- b. A pond water temperature of less than or equal to 88 F.
APPLICABILITY: OPERATIONAL CONDITI,0NS 1, 2, 3, 4, 5, and *. ACTION: With the requirements of the above specification not satisfied:
', a . In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
- b. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.
- c. In Operational Condition *, declare the emergency service water system p) g v
inoperable and take the ACTION required by Specification 3.7.1.2. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:
- a. By verifying the pond water level to be greater than its limit at least once per 24 hours.
- b. By verifying the water surface temperature (within the upper two feet of the surface) to be less than or equal to 88 F:
- 1. at least once per 4 hours when the spray pond temperature is greater than or equal to 80 F; and
- 2. at least once per 2 hours when the spray pond temperature is greater than or equal to 85 F; and
- 3. at least once per 24 hours when the spray pond temperature is greater than 32*F.
- c. By verifying all piping above the frost line is drained within 1 hour after being used.
O
*When handling
- irradiated fuel in the secondary containment.
LIMERICK - UNIT 1 3/4 7-5
PLANT SYSTEMS 3/4.7.2 no0M EMERGENCY FRESH AIR SUPPLY SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE. APPLICABILI1Y: All OPERATIONAL CONDITIONS and
- ACTION:
- a. In OPERATIONAL CONDITION 1, 2, or 3 with one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION 4, 5, or *:
- 1. With one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.
- 2. With both control room emergency fresh air supply subsystems inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
- c. The provisions of Specification 3.0.3 are not applicable in Operational Condition *.
SURVEILLANCE REQUIREMENTS 4.7.2 Each control room emergency fresh air supply subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying the control room air tempera-ture to be less than or equal to 85 F effective temperature.
- b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HED". filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.
- c. At least once per 18 months or (1) af ter any structural maintenance on the ,HEPA filter or charcoal adsorter housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:
- 1. Verifying that the subsystem 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 3000 cfm + 10%.
- When irradiated fuel is being handled in the secondary containment. O LIMERICK - UNIT 1 3/4 7-6
, PLANT SYSTEMS 'I )
(/ _ SURVEILLANCE REQUIREMENTS (Continued)
- 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon. sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2,. March 1978, for a methyl iodide penetration of less than 1%; and
- 3. Verifying a subsystem flow rate of 3000 cfm 1 10% during subsystem operation when tested in accordance with ANSI N510-1980,
- d. 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 accord:nce 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, for a methyl iodide penetration of less than 1%.
- e. At least once per 18 months by:
fG 1. Verifying that the pressure drop across the combined prefilter, t V) upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm i 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.
- 2. Verifying that on each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches to the chlorine isolation mode of operation and the isolation ,
valves close within 5 seconds: a) Outside air intake high chlorine, and b) Manual initiation from the control room.
- 3. Verifying that on each of the below radiation isolation mode actuation test signals, the subsystem automatically switches to the radiation isolation mode of operation and the control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine enclosure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor air flow rate less than or equal to 525 cfm:
a) Outside air intake high radiation, and b) Manual initiation from control room. v LIMERICK - UNIT 1 3/4 7-7
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetra-tion and bypass leakage testing acceptance criteria of less than 0.05%
in accordance with ANSI N510-1980 while operating the system at a flow rate of 3000 cfm 1 10%.
- g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass 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 3000 cfm 210%.
O O LIMERICK - L' NIT 1 3/4 7-8 l
.gm PLANT SYSTEMS i )
(./ 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. ACTION:
- a. With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours.
- b. In the event the RCIC system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
[\ Q. SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
- 1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.
- 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.
- 3. Verifying that the pump flow controller is in the correct position.
- b. At least once per 92 days by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1000 + 20, - 80 psig.*
, *The-provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is l (D - l adequate to perform the test. LIMERICK - UNIT 1 3/4 7-9
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) O
- c. At least once per 18 months by:
- 1. Performing a system functional test which includes simulated automatic actuation and restart and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded.
- 2. Verifying that the system will develop a flow of greater than or equal to 600 gpm in the test flow path when steam is supplied to the turbine at a pressure of 150 + 15, - O psig.*
- 3. Verifying that the suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal.
- 4. Performing a CHANNEL CALIBRATION of the RCIC system discharge line " keep filled" level alarm instrumentation.
O
- The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the tests.
O LIMERICK - UNIT 1 3/4 7-10
O \ / PLANT SYSTEMS V 3/4.7.4 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.4 All snubbers shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE in thoso OPERATIONAL CONDITIONS. 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 per Specification 4.7.4g on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. I b .v a. Inspection Types As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity,
- b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed after 4 months but within 10 months of commencing POWER OPERATION and shall include all snubbers. If all snubbers of each type on any system are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection of that system shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given system shall be performed in accordance with the following schedule:
U LIMERICK - UNIT 1 3/4 7-11
PLANT SYSTEMS , SURVEILLANCE REQUIREMENTS (Continued) No. Inoperable Snubbers of Each Type on Any System Subsequent Visual per Inspection Period Inspection Period *# 0 18 months 1 25% 1 12 months i 25% 2 6 months 25% 3,4 124 days i 25% 5,6,7 62 days i 25% 8 or more 31 days i 25%
- c. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that:
(1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type on that system that may be generically susceptible; and/or (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifications 4.7.4f. For those snubbers common to more than one system, the OPERABILITY of such snubbers shall be considered in assessing the surveillance schedule for each of the related systems.
- d. Transient Event Inspection An inspection shall be performed of all snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients, as determined from a review of operational data or a visual inspection of the systems, within 72 hours for accessible systems and 6 months for inaccessible systems following this deter-mination. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following: (1) manually induced snubber movement; or (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.
- The inspection interval for each type of snubber on a given system shall not be lengthened more than one step at a time unless a generic problem has been identified and corrected; in that event the inspection interval may be lengthened one step the first time and two steps thereaf ter if no inoperable snubbers of that type are found on that system.
- The provisions of Specification 4.0.2 are not applicable.
LIMERICK - UNIT 1 3/4 7-12
I e PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 'T
- e. Fdnctional Tests T -During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers shall be tested using one of the following sample plans for each type of snubber. The sample plan shall be selected prior to the test period '
and cannot be' changed during the test period. The NRC Regional Admin-istrator shalljbe notified in writing of the sample plan selected
' prior to the test period or the sample plan used in the prior test period shall be implemented:
i'
- 1) . At least 10% of the total of each type of snubber shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria'of Specification 4.7.4f., an additional 10% of that type of snubber shall be functionally tested until no more failures are '
found or until all snubbers of that type have been functionally tested; or
- 2) A representative sample of each type of snubber shall be functionally tested in accordance with Figure 4.7.4-1. "C" is the total number of snubbers of a type found not meeting the acceptance' requirements of Specification 4.7.4f. The cumulative
- number of snubbers of a type tested is denoted by "N'. At the end of each day's testing, the new values of "N" and "C" (previous
.( day's total plus current day's increments) shall be plotted on Figure 4.7.4-1. If at any time the point plotted falls on or above the " Reject" line all snubbers of that type shall be functionally I tested. If at any time the point plotted falls on or below the
" Accept" line, testing of snubbars of that type may be terminated. '
When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall_be tested until the point falls in ~the 'l Accept" region or the " Reject" region, or all the snubbers of that type have been tested. Testing equipment failure-l- during functional testing may. invalidate that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the failed equipment during the day of , , equipment failure are retested; or n 3) An initial representative sample of 55 snubbers of.each type shall it be functionally tested. For each snubber type which does not meet-the functional' test acceptance criteria, another sample of at least , one-half the size of the initial sample shall be tested until the total number tested is ' equal to the initial sample size multiplied by the factor, 1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. - The results from this sample plan shall be plotted using an " Accept" line which follows the equation N = 55(1 + C/2). Each snubber l point should be plotted as soon as the snubber is tested. If the F point plotted falls on or below the " Accept" line, testing of that
- jN type of snubber may be terminated. If the point plotted falls above the " Accept" line, testing must continue until the point I (' ') falls on or below the " Accept" line or all the snubbers of that type have been tested.
. LIMERICK - UNIT 1 3/4 7-13
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) The representative sample selected for the function test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure as far as practical that they are representative of the various configu-rations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same locations as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan, and failure of this functional test shall act be the sole cause for increasing the sample size under the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional testing results shall be reviewed at the time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.
- f. Functional Test Acceptance Criteria The snubber functional test shall verify that:
- 1) Activation (restraining action) is achieved within the specified range in both tension and compression;
- 2) Snubber bleed, or release rate where required, is present in both tension and compression, within the specified range (hydraulic snubbers only);
- 3) For mechanical snubbers, the force required to initiate or main-tain motion of the snubber is within the specified range in both directions of travel; and
- 4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.
Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be corre-lated to the specified parameters through established methods.
- g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
For the snubbers found inoperable,'an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service. LIMERICK - UNIT 1 3/4 7-14
O \ PLANT SYSTEMS i
'N SURVEILLANCE REQUIREMENTS (Continued)
If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen-in place, the cause will be evaluated and. if caused by manufacturer or design deficiency all snubbers of the same type subject to the same defect shall be functionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.4e. for snubbers not meeting the functional test acceptance criteria.
- h. Functional Testing of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test result shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.
- i. Snubber Service Life Replacement Program The service life of all snubbers shall be monitored to ensure that
[-) the service life is not exceeded between surveillance inspections. V The maximum expected service life for various seals, springs, and other critical parts shall be extended or-shortened based on moni-tored test results and failure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be. documented and the documentation shall be retained in accordance with Specification 6.10.3. O 1 v) i . LIMERICK - UNIT 1 3/4 7-15
-m _ ,. _ _ _ _ - . _ -
9 10 9 8 7 REJECT 6 , Q I e , ,# , ' A 4 CONTINUE
/ TesTiNo
/
/ /
2
/ /s/
',/ acceer r
0 / 20 30 4h .0 60 70 80 90 100 10 N FIGURE 4.7.4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST LIMERICK - UNIT 1 3/4 7-16
[ FLANT SYSTEMS V 3/4.7.5 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.5 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 limit, 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.
p b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. (v) SURVEILLANCE REQUIREMENTS 4.7.5.1 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. 4.7.5.2 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 material:
- 1. With a half-life greater than 30 days, excluding Hydrogen 3, and
/O
( ) 2. In any form other than gas.
\J LIMERICK - UNIT 1 3/4 7-17 g - - . . - - - - . ,,e- , +-- ---w --- , , v+---ey , - - - - - g-
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (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 certificatb indicating the last test date shall be tested prior to being placed into use.
- c. Startup sources and fission detectors - Each sealed starttp 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.
4.7.5.3 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 LIMERICK - UNIT 1 3/4 7-18
PLANT SYSTEMS 3/4.7.6 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 The fire suppression water system shall be OPERABLE with:
- a. Two OPERABLE fire suppression pumps, one electric motor driven and one diesel engine driven, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header,
- b. Separate fire water supplies, each with a minimum contained volume of 311,000 gallons, and
- c. An OPERABLE flow path capable of taking suction from the Unit 1 Cooling Tower Basin and the Unit 2 Cooling Tower Basin and transfer-ring the water through distribution piping with OPERABLE sectional-izing control or isolation valves to the yard hydrant curb valves, the'last valve ahead of the water flow alarm device on each wet pipe sprinkler system and the last valve ahead of the deluge valve on each deluge, spray, or pre-action sprinkler system and the last valve ahead of the fire hose stations required to be OPERABLE per Specifica-tions 3.7.6.2, 3.7.6.5, and 3.7.6.6.
/ ~'N - APPLICABILITY: At all times. ACTION:
- a. With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or provide an alternate backup pump or supply. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- b. With the fire suppression water system otherwise inoperable, establish a backup fire suppression water system within 24 hours.
SURVEILLANCE REQUIREMENTS 4.7.6.1.1 The fire suppression water system shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying the minimum contained water supply volume.
- b. At least once per 31 days by starting the electric motor-driven fire suppression pump and operating it for at least 15 minutes on recirculation flow.
O c. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position. LIMERICK - UNIT 1 3/4 7-19
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) O
- d. At least once per 12 months by performance of a system flush.
- e. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
- 1. Verifying that each fire suppression pump develops at least 2500 gpm at a system head of 125 psig,
- 2. Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and
- 3. Verifying that each fire suppression pump starts to maintain the fire suppression water system pressure greater than or equal to 95 psig.
- g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.
4.7.6.1.2 The diesel-driven fire suppression pump shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
- 1. Verifying the fuel day tank contains at least 330 gallons of fuel.
- 2. Starting the diesel-driven pump from ambient conditions and operating for greater than or equal to 30 minutes on recirculation flow,
- b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-75, is within the acceptable limits specified in Table 1 of ASTM 0975-77 when checked for viscosity, water, and sediment.
- c. At least once per 18 months by subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.
O LIMERICK - UNIT 1 3/4 7- N
. . .- .. . _ . ... . . - _ _ _ _ _ - _= . .- -- . . _- - . . . - . ..-
, _ , PLANT SYSTEMS
. SURVEILLANCE REQUIREMENTS (Continued) 4.7.6.1.3 The diesel-driven fire pump starting 24-volt battery bank and charger shall-be demonstrated OPERABLE: i
- a. At least once per 7 days by verifying that:
- 1. The electrolyte level of each cell is above the plates, t
- 2. The pilot cell specific gravity, corrected to 77 F and full electrolyte level is greater than or equal to 1.260, and
- 3. The overall battery voltage is greater than or equal to 24 volts.
- b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.
i
- c. At least once per 18 months by verifying that:
- 1. The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and
, (g 2. Battery-to-battery and terminal connections are clean, tight, ( j free of corrosion, and coated with anticorrosion material. 4 2 t I l. . ts_ f LIMERICK - UNIT 1 3/4 7-21
;,-,a,----. _ _.-___ __ -, -_ _ _.--_-__.__.._-__ __.--__ _ _ _.,____.-.._._ _
PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.2 The following spray and sprinkler systems shall be OPERABLE: Fire Zone Description Reactor Enclosure Hatchway Water Curtains:
- 1. EL 253'
- 2. EL 283'
- 3. EL 313' Fire Area Separation Water Curtains:
48A 1. Area 602, EL 313' 45A 2. Area 402, EL 253' 44 3. Area 304, EL 217' (2 curtains) 22 Cable Spreading Room, Room 450, EL 254', 27 Control Structure Fan Room, EL 304' 27 CREFAS System Filters, EL 304' 28B SGTS Filters, Compartment 624 and SGTS Access Area 625, EL 332' 33 RCIC Pump Room, Room 108, EL 177' 34 HPCI Pump Room, Room 109, EL 177' 41* RECW Area, EL 201' 42A* Safeguard System Access Area 200, EL 201' 44 Safeguard System Access Area 304, EL 217' (Partial) (2 systems) 45A CRD Hydraulic Equipment Area 402, Reactor Enclosure, EL 253' (Partial) 45B Neutron Monitoring System Area 406, El 253' (Partial) 47A General Equipment Area 500 and Corridor 506, Reactor Enclosure, EL 283' (Partial) 51A & B Reactor Enclosure Recirculation System Filters, EL 331' 79,80,81,82 Diesel Generator cells (4 Cells) APPLICABILITY: Whenever equipment protected by the spray and/or sprinkler systems is required to be OPERABLE. ACTION:
- a. With one or more of the above required spray and/or sprinkler systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged;.for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
*Not required to be OPERABLE until prior to exceeding 5% of RATED THERMAL POWER.
LIMERICK - UNIT 1 3/4 7-22
c
/N PLANT SYSTEMS
( ) v SURVEILLANCE REQUIREMENTS 4.7.6.2 Each of the above required spray and sprinkler systems 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 is in its correct position.
- b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.
- c. At least once per 18 months:
- 1. By performing a system functional test which includes simulated automatic actuation of the system, and:
a) Verifying that the automatic valves in the flow path actuate to their correct positions on a test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel. . O) ( v
- 2. By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity, and
- 3. By a visual inspection of each sprinkler nozzle's spray area to verify that the spray pattern is not obstructed.
- d. At least once per 3 years by performing an air or water flow test through each open head spray and sprinkler header system and verifying each open head spray nozzle and sprinkler header system is unobstructed, except the charcoal filter system spray nozzles which only need to be visually inspected and verified to be unobstructed each time the charcoal is changed.
O LIMERICK - UNIT 1 3/4 7-23
PLANT SYSTEMS CO2 SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.3 The following low pressure CO2 system shall be OPERABLE:
- a. Control Room Entrance, Hose Rack OHR601 and OHR 602.
APPLICABILITY: Whenever equipment protected by the CO2 systems is required to be OPERABLE. ACTION:
- a. With the above required CO2 system inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.6.3.1 The above required low pressure CO2 system shall be demonstrated OPERABLE at least once per 7 days by verifying the C02 storage tank level to be greater than 25% and pressure to be greater than 265 psig. 4.7.6.3.2 The above required CO2 system shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or auto-matic) in the flow path is in its correct position. O i LIMERICK - UNIT 1 3/4 7-24
PLANT SYSTEMS ' O.I
- \.
U HALON SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.4 The following Halon systems shall_be OPERABLE with the storage tanks having at least 95% of full charge weight and 90% of full charge pressure:
- a. Remote Shutdown Panel Area 540, EL 289' (Raised Floor), and
- b. Auxiliary Equipment Room 542, El 289' (Raised Floor).
APPLICABILITY: 'Whenever equipment protected by the Halon systems is required , , to be OPERABLE. ACTION:
- - a. With one' or more of the above required Halon systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression t equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
g SURVEILLANCE REQUIREMENTS w 4.7.6.4 Each of the above required Halon systems 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 is in its correct position.
- b. At least once per 6 months by verifying Halon storage tank weight '
and pressure.
- c. At least once per 18 months by:
- 1. Performance of a functional test of the general alarm circuit
' and associated alarm and interlock devices, and
- 2. Performance of a system flow test to assure no blockage.
i O 4 1 LIMERICK - UNIT 1 3/4 7-25 i
PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION F0 H.dATION 3.7.6.5 The fire hose stations shown in Table 3.7.6.5-1 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. ACTION:
- a. With ono'or more of the fire hose stations shown in Table 3.7.6.5-1 inoperable, pravide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the standard length of hose provided at the hose station. The second outlet of the wye shall be t.onnected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station.
Where it can be demonstrated that the physical routing of the fire hose would result in e. recognizable hazard to opcrating technicians, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station. Signs shall be mounted above the gated wye (s) to identify the proper hose to use. The above ACTION shall be accomplished within 1 hour if the inoperable fire hose is the prima,ay means of fire suppression; otherwise route the additional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RECUIREMENTS 4.7.6.5 Each cf the fire hose stations shown in Table 3.7.6.5-1 shall be demonstrated OPERABLE:
- a. At least once per 31 days by a visual inspection of the fire hose stations accessible during plant operation to assure all required equipmtat is at the station,
- b. At least once per 18 months by:
- 1. Visual Fespection of the fire hose stations not accessible during plant operation to assure all required equipment is at the station.
- 2. Removing the hose for inspection and reracking, and
- 3. Inspecting all gaskets and replacing any degraded gaskets in the couplings.
c At least once per 3 years by:
- 1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.
- 2. Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above the maximum fire main operating pressure, whichever is greater.
LIMERICK - UNIT 1 3/4 7-26
, ._ . _- .. -. .- - - _ ~ . . . _ _ . - . . .
TABLE 3.7.6.5-1 V) s FIRE H0SE STATIONS H0SE RACK LOCATION ELEVATION HENTIFICATION 4 1. Control
Enclosure:
Stairwell 350' 1PR-141-Stairwell, Outside SGTS Room 332' 1HR-140 Stairwell, Outside Fan Room 304' 1HR-103 , Outside 13kV Switchgear Room 217' 1HR-116 ., Stairwell,.Outside Aux Equip Rm 289' '1HR-130 l Stairwell, Outside Cable Spreading Rm 254' 1HR-250 5 Wall, Outside 4kV Switchgear & Battery. ' Rooms 239' 1HR-251 Corridor 448, South' Side of 4kV Switchgear & Battery Rooms 239' 1HR-124 -
. Wall, Corridor 265 200' 1HR-120 Wall, Corridor 164 180' 1HR-121 . 2. Refueling Area:
SW Corner Refuel Floor 352' 1HR-201 NW Corner Refuel Floor 352' 1HR-202 North Wall-Center 352' 1HR-203 i South Wall-Center 352' 1HR-204-1
- 3. Reactor
Enclosure:
SW Corner Reactor Enclosure 331' 1HR-205 SW Corner Reactor Enclosure
- (RERS Fan Area) 313' 1HR-207 I
NW Corner Reactor Enclosure (Laydown Area 601) 313' 1HR-208 SE Corner Reactor Enclosure (Near' Refuel Floor Exh. Fans) 313' 1HR-209 NE Corner Reactor Enclosure (Near D124 Load Center) 313' IHR-210
. West Wall Reactor Enclosure
] (Corridor 506) 283' 1HR-215 I NW Corner Reactor Enclosure , (Corridor 506) 283' 1HR-216 a LIMERICK - UNIT 1 3/4 7-27
TABLE 3.7.6.5-1 (Continued) FIRE HOSE STATIONS HOSE RACK LOCATION ELEVATION IDENTIFICATION
- 3. Reactor
Enclosure:
(Continued) SE Corner Reactor Enclosure (SLC Pumps Area 500) 283' 1HR-217 NE Corner Reactor Enclosure 283' 1HR-218 West Wall Reactor Enclosure (Area 402A, Near CRD Repair) 253' 1HR-223 NW Corner Reactor Enclosure (Near Drywell Equip Hatch) 253' 1HR-224 SE Corner Reactor Enclosure (Near Drywell Personnel Lock) 253' 1HR-225 East Wall Reactor Enclosure (Near TIP Mach'ines) 253' 1HR-226 West Wall Reactor Enclosure (Near HPCI Equip Hatch) 217' 1HR-232 NW Corner Reactor Enclosure (Near Supp Pool Access Hatch) 217' 1HR-233 East Wall Reactor Enclosure (Near Equipment Airlock 300) 217' 1HR-234 NE Corner Reactor Enclosure (Near MCC D124-R-G) 217' 1HR-235 West Wall Reactor Enclosure (Near MCC 0134-R-H) 201' 1HR-240 NW Corner Reactor Enclosure (Near MCC D134-R-H1) 201' 1HR-241 East Wall Reactor Enclosure l (Near RECW Heat Exchangers) 201' 1HR-242 NE Corner Reactor Enclosure (Near RECW Pumps) 201' 1HR-243 i SW Corner Reactor Enclosure 177' 1HR-252 NW Corner Reactor Enclosure 177' 1HR-253 NE Corner Reactor Enclosure 177' 1HR-142 l l 1 O LIMERICK - UNIT 1 3/4 7-28
J
, -ss PLANT SYSTEMS l ) '
Ns _,/ YARD FIRE HYDRANTS AND HOSE CART HOUSES LIMITING CONDITION FOR OPERATION 3.7.6.6 The yard fire hydrants and hose cart houses shown in Table 3.7.6.6-1 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is requirol to be OPERABLE. ACTION:
- a. With one or more of the yard fire hydrants or hose cart houses shown in Table 3.7.6.6-1 inoperable, within 1 hour have sufficient additional lengths of 2 1/2 inch diameter hose located in an adjacent OPERABLE hose cart house to provide cervice to the unprotected area (s) if the inoper-able fire hydrant or hose cart house is the primary means of fire suppression; otherwise provide the additional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
;O i V
4.7.6.6 Each of the yard fire hydrants and hose cart houses shown in Table 3.7.6.6-1 shall be demonstrated OPERABLE:
- a. At least once per 31 days by visual inspection of the hose cart house to assure all required equipment is at the hose house.
- b. At least once per 6 months, during March, April, or May and during September, October, or November, by visually inspecting each yard fire hydrant und verifying that the hydrant barrel is dry and that the hydrant is not damaged.
- c. At least once per 12 months by:
- 1. Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above the maximum fire main operating pressure, whichever is greater.
- 2. Replacement of all degraded gaskets in couplings.
- 3. Performing a flow check of each hydrant.
.A (v)
LIMERICK - UNIT 1 3/4 7-29
,a , , , - - , - - - , , -- - . , - - - -
- - - - - - , - - - - - , , - , , - , , --.-. , - - - -. , , - - - - , . - ~ , , -
TABLE 3.7.6.6-1 YARD FIRE HYDRANTS AND HOSE CART HOUSES LOCATION HYDRANT NUMBER West of Diesel Generator Enclosure FH #7 South of Diesel Generator Enclosure FH #8 LOCATION HOSE CART HOUSE NUMBER West of Diesel Generator Enclosure HCH #1 O O LIMERICK - UNIT 1 3/4 7-30
p PLANT SYSTEMS l 1 V 3/4.7.7 FIRE RATED ASSEM8 LIES , LIMITING CONDITION FOR OPERATION 3.7.7 All fire rated assemblies, including walls, floor / ceilings, cable tray
, enclosures and other fire barriers, separating safe shutdown fire areas or separating portions of redundant systems important to safe shutdown within a fire area, and all sealing devices in fire rated assembly penetrations, including fire doors, fire windows, fire dampers, cable, piping and ventilation duct penetration seals and ventilation seals, shall be OPERABLE.
i APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within 1 hour establish a continuous fire watch on at least one side of the affected assembly (s) and/or sealing device (s) or verify the OPERABILITY of fire detectors on at
-least one side of the inoperable assembly (s) and sealing device (s) and establish an hourly fire watch patrol,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- SURVEILLANCE REQUIREMENTS 4.7.7.1 Each of the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE at least once per 18 months by performing a visual inspection of:
- a. The exposed surfaces of each fire rated assembly.
- b. Each fire window, fire damper, and associated hardware.
- c. At least 10% of each type of sealed penetration, except internal conduit seals. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10% of each type of sealed penetration shall be made. This inspection process shall continue until a 10% sample with no apparent changes in appearance or abnormal degradation is found. Samples shall be selected such that each penetration seal will be inspected at least once per 15 years.
i j i l LIMERICK - UNIT 1 3/4 7-31 4 l
1 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) O-4.7.7.2 Each of the above required fire doors which are not electrically supervised shall be verified OPERABLE by inspecting the closing mechanism and latches at least once per 6 months, and by verifying:
- a. That each locked-closed fire door is closed at least once per 7 days.
- b. That each unlocked fire door without electrical supervision is closed at least once per 24 hours.
4.7.7.3 Each of the above required fire doors which are electrically supervised shall be verified OPERABLE:
- a. By verifying that each locked-closed fire door is closed at least once per 7 days.
- b. By verifying the OPERABILITY of the fire door supervision system for each electrically supervised fire door by performing a CHANNEL FUNCTIONAL TEST at least once per 31 days.
- c. By inspecting the closing mechanism and latches at least once per 6 months.
O LIMERICK - UNIT 1 3/4 7-32
fS 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 =A.C. SOURCES-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. Four separate and independent diesel generators, each with: ;
- 1. A separate day. tank containing a minimum of 200 gallons of fuel,
- 2. A separate fuel storage system containing a minimum of 33,500 gallons of fuel, and
- 3. A separate fuel transfer pump.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.
sources by performing Surveillance Requirements 4.8.1.1 la.'and 4.8.1.1.2a.4., for one diesel generator at a time, within 24 hours and at least once per 7 days thereafter; restore the inoperable diesel generator to OPERABLE status within 92 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. ~ With two diesel generators of.the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.
sources by performing Surveillance Requirements 4.8.1.1.la and 4.8.1.1.2a.4. , for one diesel generator at a time, within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoper-able diesel generators:to OPERABLE status'within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- c. With three diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a. and 4.8.1.1.2a.4. , for one diesel gcnerator at a time, within 1 hour and at least once per 8 hours thereafter; restore at least one of the inop-
-erable diesel generators to OPERABLE status within 2 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the' following 24 hours.
- d. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY A of the remaining A.C. sources by performing Surveillance Requirements
! 4.8.1.1 la. and 4.8.1.1.2a.4. within 1 hour and at least once per
\ 8 hours thereafter. Restore at least two offsite circuits and at -LIMERICK - UNIT 1 3/4 8-1
i ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) least three of the above required diesel generators to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- e. With two diesel generators of the above required A.C. electrical power sources inoperable, in addition to ACTION b., above, verify within 2 hours that all required systems, subsystems, trains, components, and devices that depend on the remaining diesel generators as a source of emergency power are also OPERABLE; otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- f. With one offsite circuit of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.
sources by performing Surveillance Requirements 4.8.1.1.la. and 4.8.1.2a.4, for one diesel generator at a time, within 1 hour and at least once per 8 hours thereafter; restore at least two offsite circuits to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 h:1rs.
- g. Witt, two of the above required offsite circuits inoperable, demonstrate the OPERABILITY of all of the above required diesel generators by performing Surveillance Requirement 4.8.1.1.2a.4., for one diesel generator at a time, within 1 hour and at least once per 8 hours there-after, unless the diesel generators are already operating; restore at least one of the inoperable offsite circuits to OPERABLE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours. With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- h. With one offsite circuit and two diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.la. and 4.8.1.1.2a.4. within 1 hour and at least once per 8 hours thereafter; restore at least one of the above required inoperable A.C. sources to OPERABLE status within 12 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. Restore at least two offsite circuits and at least three of the above required diesel generators to OPERABLE status within 72 hours from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
LIMERICK - UNIT 1 3/4 8-2
1 i ELECTRICAL POWER SYSTEMS
- &O) v 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: L a. Determined OPERABLE at least_once per 7 days by verifying correct
- breaker alignments and indicated power availability, and i-- 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 of the above required diesel generators shall be demonstrated OPERABLE: In accordance with the frequency specified in Table 4.8.1.1.2-1 on a
- 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 fuel tank.
- 4. Verifying the diesel starts from ambient conditions
- and accel-erates to at least 882 rpm in less than or equal to 10 seconds.
,j The generator voltage and frequency shall reach 4285 1 420 volts ,4 and 60 1 1.2 Hz within 10 seconds after the start signal. The 'N diesel generator shall be started for this test by using one of the following signals: . a) Manual.** b) Simulated loss-of-offsite power by itself. c) Simulated loss-of-offsite power in conjunction with an ESF
- actuation test signal.
d) An ESF actuation test signal by itself.
- 5. Verifying the diesel generator is synchronized, loaded to greater than or equal to 2850 kW in less than or equal to j 200 seconds, and operates with this load for at least 60 minutes.
I 6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
- 7. Verifying the pressure in all diesel generator air start receivers to be greater than or equal to 225 psig.
*The diesel generator start (10 sec) and subsequent loading (200 sec) from
.. ambient conditions shall be performed at least once per 184 days in these surveillance tests. All other engine starts and loading for the purpose of this' surveillance testing may be preceded by an engine prelube period and/or other warmup procedures recommended by the manufacturer so that-mechanical D stress and wear on the diesel engine is minimized.
, **If diesel generator started manually from the control room, 10 seconds after . the automatic prelube period. LIMERICK - UNIT 1 3/4 8-3
- . . . . . , _ _ _ . _ _ , ~ . _..,,...,-,...,.,,,my,. .,.m... , __.m., . . , . . , , , , _ , , . ~ , , . , , . - . _ . , , , _ , - - . . - , _ , , _ _ , , . _ _ , . _ _ - m,.
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. By removing accumulated water:
- 1) From the day tank at least once per 31 days and after each occa-sion when the diesel is operated for greater than 1 hour, and
- 2) From the storage tank at least once per 31 days.
- c. By sampling new fuel oil in accordance with ASTM D4057-81 prior to addition to the storage tanks and:
- 1) By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:
a) An API Gravity of within 0.3 degrees 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 spccific gravity at 60/60 F of greater than or equal to 0.83:but less than or equal to 0.89 or an API gravity at 60 F of greater than or equal to 27 degrees but less than or equal to 39 degrees. b) A kinematic viscosity at 40 C of greater than or equal to 19 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 equal to or greater than 125 F, and d) A clear and bright appearance with proper color when tested in accordance with ASTM 04176-82.
- 2) By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-31 are met when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM 01552-79 or ASTM D2622-82.
- d. At least once every 31 days by obtaining a sample of fuel oil from the storage tanks 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.
- e. At least once per 18 months, during shutdown, by:
- 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. Verify the diesel generator capability to reject a load of greater than or equal to that of the RHR Pump Motor (992 Kw) for each diesel generator while maintaining voltage at 4285 420 volts and frequency at 60 i 1.2 hz.
LIMERICK - UNIT 1 3/4 8-4
r
- r3 ELECTRICl.c.__'0WER SYSTEMS V SURVEILLANCE REQUIREMENTS (Continued)
- 3. Verifying the diesel generator capability to reject a load of
< 2850 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:
a) Verifying deenergization of the emergency busses and load shedding from the emergency busses. b) Verifying-the diesel generator starts on the auto-start signal, energizes the emergency busses within 10 seconds, energizes the auto-connected loads through the individual load timers 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 4285 1 420 volts and 60 1 1.2 Hz during this test.
- 5. Verifying that on an ECC$ 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 (q , 5 minutes. The generator voltage and frequency shall reach V 4285 1 420 volts and 60 i 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 ECCS actuation test signal, and:
a) Verifying deenergization of the emergency busses and load shedding from the emergency busses. b) Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses within 10 seconds, energizes the auto-connected shutdown loads through the individual lead timers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. Af ter energ!? stion, the steady-state voltage and frequency of the emergency busses shall be maintained at 4285 1 420 volts and 60 1 1.2 Hz during this test.
- 7. Verifying that all automatic diesel generator trips, except engine overspeed and generator differential over-current are automatically bypassed upon an ECCS actuation signal.
kO) v LIMERICK - UNIT 1 3/4 8-5 L - _ -. - -
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 8. Verifying the diesel generator operates for at least 24 hours.
During the first 2 hours of this to-t, the diesel generator shall be loaded to greater than or squal to.3135 kW and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 2850 kW. The generator voltage and frequency shall reach 4285 1 420 volts and 60 1 1.2 Hz within 10 seconds ** after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform Surveillance Requirement 4.8.1.1.2e.4.b).*
- 9. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3100 kW.
- 10. Verifying the diesel generator's capability to:
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 statts.
- 11. Verifying that with the diesel generator operating in a test mode and connected to its bus, a simulated ECCS actuation signal overrides the test mode by (1) returning the diesel generator to standby operation, and (2) automatically energizes the emergency loads with offsite power.
- 12. Verifying that the automatic load sequence timers are OPERABLE with the interval between each load block within i 10% of its design interval.
*If Surveillance Requirement 4.8.1.1.2e.4.b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated at 2850 kW for 1 hour or until operating temperature has stabilized.
a*If diesel generator started manually from the control room, 10 seconds after the automatic prelube period. O LIMERICK - UNIT 1 3/4 8-6
N ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 13. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a) Control Room Switch In Pull-To-Lock (With Local / Remote Switch in Remote) b) Local / Remote Switch in Local. 1 c) Emergency Stop
- f. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all four diesel generators simultaneously, during shutdown, and verifying that all four diesei generators accelerate to at least 882 rpm in less than or equal to 10 seconds.
- g. 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 or equivalent solution, and g3
- 2. Perf,orming a pressure test of those portions of the diesel fuel (V) oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section XI Article IWD-5000.
4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 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, Revision 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.
.O LIMERICK - UNIT 1 3/4 8-7
TABLE 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN LAST 100 VALID TESTS
- TEST FREQUENCY
$1 At least once per 31 days 2 At least once per 14 days 3 At least once per 7 days
>4 At least once per 3 days
- Criteria for determining 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, where the last 100 tests are determined on a per nuclear unit basis. For the purposes of this test schedule, only valid tests conducted after the OL issuance date shall be included in the computation of the "last 100 valid tests." Entry into this test schedule shall be made at the 31-day test frequency.
O LIMERICK - UNIT 1 3/4 8-8
ge g ELECTRICAL POWER SYSTEMS i V)
.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 distribution system, and
- b. Two diesel generators each with:
- 1. A day fuel tank containing a minimum of 200 gallons of fuel.
- 2. A fuel storage system containing a minimum of 33,500 gallons of fuel.
- 3. A fuel transfer pump.
APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *.
/
g ) ACTION: O a. With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment, operations with a potential for draining the reactor vessel and crane operations over the spent fuel storage pool when fuel assemblies are stored therein. In addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.
- b. The provisions of Specification 3.0.3 are not applicable.
l SURVEILLANCE REQUIREMENTS l l 4. 8.1. 2 At least the above required A.C. electrical power sources shall be l demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2, l and 4.8.1.1.3, except for the requirement of Specification 4.8.1.1.2a.5. l l 1. t . l s s _,, *When handling irradiated fuel in the secondary containment. LIMERICK - UNIT 1 3/4 8-9 l t
ELECTRICAL POWER SYSTEMS 3/4.8.2 0.C. SOURCES D.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical power sources shall be OPERABLE:
- a. Division 1, Consisting of:
- 1. 125-Volt Battery 1A1 (1A1D101).
- 2. 125-Volt Battery 1A2 (1A2D101).
- 3. 125-Volt Battery Charger 1BCA1 (1 AID 103).
- 4. 125-Volt Battery Charger 1BCA2 (IA20103).
- b. Division 2, Consisting of:
- 1. 125-Volt Battery 1B1 (1B1D101).
- 2. 125-Volt Battery 1B2 (1820101).
- 3. 125-Volt Battery Charger 1BCB1 (1B1D103).
- 4. 125-Volt Battery Charger 1BCB2 (182D103).
- c. Division 3, Consisting of:
- 1. 125-Volt Battery IC (ICD 101).
- 2. 125-Volt Battery Charger 1BCC (ICD 103).
- d. Division 4, Consisting of:
- 1. 125-Volt Battery 10 (1DD101).
- 2. 125-Volt Battery Charger 1BCD (100103).
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With any battery and/or charger of the above required D.C. electrical power sources inoperable, restore the inoperable division battery to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.8.2.1 Each of the above required division batteries and chargers shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1. The parameters in Table 4.8.2.1-1 meet the Category A limits, and
- 2. Total battery terminal voltage for each 125-volt battery is greater than or equal to 131 volts on float charge.
LIMERICK - UNIT 1 3/4 8-10
. p)
\
v ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 105 volts or battery overcharge with battery terminal voltage above 150 volts, by verifying that:
- 1. The parameters in Table 4.8.2.1-1 meet the Category B limits,
- 2. There is no visible corrosion at either terminals or connectors, I or the connection resistance of these items is less than 150 x 10 8 ohm, and
- 3. The average electrolyte temperature of each sixth cell is > 60 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, free of corrosion and coated with anticorrosion material,
- 3. The resistance of each cell-to-cell and terminal connection is s less than or equal to 150 x 10 6 ohm excluding cable intercell
, connections, and
- 4. The battery chargers will supply the currents listed below at a minimum of 132 volts for at least 8 hours:
Charger Current (Amperes) IBCA1 300 1BCA2 300 18C81 300 18C82 300 1BCC 75 1BCD 75
- d. At least once per 18 months, during shutdown, by verifying that either:
- 1. The battery capacity is adequate to supply and maintain in OPERABLE status all of the actual emergency loads for the design duty cycle when the battery is subjected to a battery service test, or
- 2. The battery capacity is adequate to supply a dummy load of the
,. following profile while maintaining the battery terminal voltage greater than or equal to 105 volts for the nominal 125-volt batteries and 210 volts for the nominal 125/250-volt batteries:
d LIMERICK - UNIT 1 3/4 8-11
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) LOAD CYCLE (amps) Division Battery 0-1 Min. 1-239 Min. 239-240 Min. I 1A1 546 168 187 1A2 449 129 147 II 181 889 158 321 182 823 119 282 III IC 193 31 31 IV 1D 169 21 21 Each 125/250-volt battery is rated at 1500 ampere-hours at an 8-hour discharge rate, based on a terminal voltage of 1.75 volts-per-cell at 77 F. Each 125-volt battery is rated at 250 ampere-hours at an 8-hour discharge rate, based on a terminal voltage of 1.75 volts per-cell at 77 F.
- 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. At this once per 60 month interval, this performance discharge test may be performed in lieu of the battery service test (Specification 4.8.2.1.d).
- f. At least once per 18 months during shutdown performance discharge tests of battery capacity shall be given 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.
O LIMERICK - UNIT 1 3/4 8-12
!O TABLE 4.8.2.1-1 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2)
Parameter Limits for each Limits for each Allowable ( ) designated pilot connected cell value for each cell connected cell Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, l and i " above and 5 " above and not maximum level maximum level overflowing
; indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (4) > 2.07 volts Not more than 0.020 below the Specific average of all Gravity ( ) ~> 1.195(6) ~> 1.190 connected cells O
Average of all Average of all
- connected cells connected cells
> 1.200 > 1.190( )
( )For any Category A parameter (s) outside the limit (s) shown, the battery
- may be considered OPERABLE provided that within 24 hours all the Category 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.
( )For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are l 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 4 inoperable battery.
-( )May be corrected for average electrolyte temperature.
( ) Corrected for electrolyte temperature of 77*F and full level. ( )Or battery charging current is less than 1 amperes when on float charge. j LIMERICK - UNIT 1 3/4 8-13
ELECTRICAL POWER SYSTEMS D.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C. electrical power sources system shall be OPERABLE with:
- a. Division 1, Consisting of:
- 1. 125-Volt Battery 1A1 (IA10101).
- 2. 125-Volt Battery 1A2 (1A2D101).
- 3. 125-Volt Battery Charger 1BCA1 (1A10103).
- 4. 125-Volt Battery Charger 1BCA2 (1A2D103).
- b. Division 2, Consisting of:
- 1. 125-Volt Battery 181 (181D101).
- 2. 125-Volt Battery 182 (182D101).
- 3. 125-Volt Battery Charger 1BCB1 (181D103).
- 4. 125-Volt Battery Charger 1BCB2 (182D103).
- c. Division 3, Consisting of:
- 1. 125-Volt Battery 1C (ICD 101).
- 2. 125-Volt Battery Charger 1BCC (1C0103).
- d. Division 4, Consisting of:
- 1. 125-Volt Battery ID (10D101).
- 2. 125-Volt Battery Charger 1BCD (1DD103).
APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *. ACTION:
- a. With less than two divisions of the above required D.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.
- When handling irradiated fuel in the secondary containment.
LIMERICK - UNIT 1 3/4 8-14
i ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following power distribution system divisions shall be. energized:
- a. A.C. power distribution:
- 1. Division 1, Consisting of:
L a) 4160-VAC Bus: 011 (10A115) b) 480-VAC Load Center: 0114 (10B201) c) 480-VAC Motor Control Centers: D114-R-C1 (108219) D114-R-C (10B213) 0114-S-L (00B519) D114-R-G (10B211) D114-R-G1 (108215) D114-D-G (108515) 4 d) 120-VAC Distribution Panels: 10Y101 10Y206 01Y501
- 2. Division 2, Consisting of:
,, s a) 4160-VAC Bus: D12 (10A116)
(s /) b) 480-VAC Load Center: D124 (10B202) c) 480-VAC Motor Control Centers: D124-R-C1 (10B220) D124-R-C (10B214) D124-5-L (008520) D124-R-G (10B212) 0124-R-G1 (108216) D124-D-G (108516) d) 120-VAC Distribution Panels: 10Y102 10Y207 02Y501
- 3. Division 3, Consisting of:
a) 4160-VAC Bus: D13 (10A117) b) 480-VAC Load Center: D134 (108203) c) 480-VAC Motor Control Centers: D134-R-H1 (10B221) D134-R-H (10B217) D134-R-E (10B223) - D134-C-B (00B131) D134-D-G (108517) D234-5-L (00B521) d) 120-VAC Distribution Panels: 10Y103 10Y163 03Y501
- 4. Division 4, Consisting of:
7 ~'g ( ) a) b). 41.60-VAC Bus: 480-VAC Load Center: 014 (10A118) D144 (10B204) LIMERICK - UNIT 1 3/4 8-15
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c) 480-VAC Motor Control Centers: D144-R-G (108222) 0144-R-H (108218) D144-R-E (108224) D144-C-B (00B132) D144-D-G (10B518) D244-5-L (008522) d) 120-VAC Distribution Panels: 10Y104 10Y164 04YS01
- b. D.C. Power Distribution Panels
- 1. Division 1, Consisting of:
a) 250-V DC Fuse Box: 1FA (1AD105) b) 250-V DC Motor Control Centers: 1DA (100201) c) 125-V DC Distribution Panels: 1 PPA 1 (1AD102) 1 PPA 2 (1AD501) IPPA3 (1AD162)
- 2. Division 2, Consisting of:
a) 250-V DC Fuse Box: IFB (1BD105) b) 250-V DC Motor Control Centers: 1D8-1 (10D202) 10B-2 (10D203) c) 125-V DC Distribution Fanels: 1 PPB 1 (18D102) 1 PPB 2 (1BD501) 1 PPB 3 (IBD162)
- 3. Division 3, Consisting of:
a) 125-V DC Fuse Box: 1FC (1C0105) b) 125-V DC Distribution Panels: 1PPC1 (ICD 102) IPPC2 (1CD501) 1PPC3 (ICD 162)
- 4. Division 4, Consisting of:
a) 125-V DC Fuse Box: 1FD (1DD105) b) 125-V DC Distribution Panels: 1 PPD 1 (100102) 1 PPD 2 (10D501) IPPD3 (1DD162) APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one of the above required A.C. distribution system divisions not energized, reenergize the division within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
O LIMERICK - UNIT 1 3/4 8-16
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
- b. With one of the above required D.C. distribution system divisions not energized, reenergize the division within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.8.3.1 Each of the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct breaker alignment and voltage on the busses /MCCs/ panels. O O a LIMERICK - UNIT 1 3/4 8-17
ELECTRICAL POWER SYSTEMS DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, 2 of the 4 divisions of the power distribution system shall be energized with:
- a. A.C. power distribution:
- 1. Division 1, Consisting of:
a) 4160-VAC Bus: Dll (10A115) b) 480-VAC Load Center: D114 (108201) c) 480-VAC Motor Control Centers: D114-R-C1 (108219) D114-R-C (108213) Dll4-S-L (008519) D114-R-G (108211) D114-R-G1 (10B215) D114-D-G (108515) d) 120-VAC Distribution Panels: 10Y101 10Y206 01Y501
- 2. Division 2, Consisting of:
a) 4160-VAC Bus: D12 (10A116) b) 480-VAC Load Center: 0124 (108202) c) 480-VAC Motor Control Centers: D124-R-Cl (10B220) D124-R-C (108214) D124-S-L (00B520) D124-R-G (108212) D124-R-G1 (108216) D124-D-G (108516) d) 120-VAC Distribution Panels: 10Y102 10Y207 02Y501
- 3. Division 3, Consisting of:
a) 4160-VAC Bus: D13 (10A117) b) 480-VAC Load Center: 0134 (108203) c) 480-VAC Motor Control Centers: D134-R-H1 (108221) D134-R-H (108217) D134-R-E (108223) D134-C-B (00B131) D134-D-G (108517) 0234-S-L (00B521) d) 120-VAC Distribution Panels: 10Y103 10Y163 03Y501
- 4. Division 4, Consisting of:
a) 4160-VAC Bus: D14 (10A118) b) 480-VAC Load Center: 0144 (10B204) LIMERICK - UNIT 1 3/4 8-18
7 ' '] ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c) 480-VAC Motor Control Centers: D144-R-G (10B222) D144-R-H (108218) D144-R-E (108224) 0144-C-B (00B132) D144-D-G (108518) D244-S-L (008522) d) 120-VAC Distribution Panels: 10Y104 10Y164 04Y501
- b. D.C. power distribution:
- 1. Division 1, Consisting of:
a) 250-V DC Fuse Box: 1FA (1AD105) b) 250-V DC Motor Control Center: 10A (10D201) c) 125-V DC Distribution Panels: IPPA1 (1AD102) 1 PPA 2 (1AD501) IPPA3 (1AD162)
- 2. Division 2, Consisting of:
a) 250-V DC Fuse Box: IFB (1BD105)
,-s b) 250-V DC Motor Control Centers: 10B-1 (10D202)
( ) 10B-2 (10D203)
\s,_ / c) 125-V DC Distribution Panels: IPPB1 (18D102) 1 PPB 2 (1BD501)
IPPB3 (18D162)
- 3. Division 3, Consisting of:
a) 125-V DC Fuse Box: IFC (1C0105) b) 125-V DC Distribution Panels: 1PPCl (ICD 102) IPPC2 (ICD 501) IPPCS (ICD 162)
- 4. Division 4, Consisting of:
a) 125-V DC Fuse Box: 1FD (1D0105) b) 125-V DC Distribution Panels: 1 PPD 1 (1DD102) 1 PPD 2 (IDD501) 1 PPD 3 (10D162) APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and *. ACTION:
- a. With less than two divisions of the above required A.C. distribution system energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
- b. With less than two divisions of the above required D.C. distribution system energized, suspend CORE ALTERATIONS, handling of irradiated
/'~'N fuel in the secondary containment and operations with a potential for draining the reactor vessel.
(x.-) i *When handling irradiated fuel in the secondary containment. LIMERICK - UNIT 1 3/4 8-19
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.8.3.2 At least the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct breaker alignment and voltage on the busses /MCCs/ panels. 9 l I l l 1 O LIMERICK - UNIT 1 3/4 8-20
rN ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 All primary containment penetration conductor overcurrent protective devices ~shown in Table 3.8.4.1-1 shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one or more of the above required containment penetration conductor overcurrent devices shown in Table 3.8.4.1-1 inoperable:
- 1. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping and locking, racking out, or removing the alternate device or racking out or removing the inoperable device within 72 hours, and
- 2. Declare the affected system or component inoperable, and
- 3. Verify at least once per 7 days thereafter the alternate device is tripped and locked, racked out, or removed, or the inoperable device is racked out or removed.
( Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. The provisions of Specification 3.0.4 are not applicable to overcurrent devices which have the inoperable device racked out or removed or, which have the alternate device tripped, racked out, or removed.
SURVEILLANCE REQUIREMENTS 4.8.4.1 Each of the primary containment penetration conductor overcurrent protective devices shown in Table 3.8.4.1-1 shall be demonstrated OPERABLE:
- a. At least once per 18 months:
- 1. By verifying that the medium voltage 4.16 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers and performing:
a) A CHANNEL CALIBRATION of the associated protective relays, and b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and overcurrent control circuits function as designed. c) For each circuit breaker found inoperable during these func-tional tests, an additional representative sample of at least
- 10% of all the circuit breakers of the inoperable type shall
- \ also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
LIMERICK - UNIT 1 3/4 8-21
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2. By selecting and functionally testing a representative sample of at least 10% of each type of the 480 VAC circuit breakers.
Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing *of these circuit breakers shall consist of injecting a current 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, and verifying that the circuit breaker operates within the time delay band-width for that current specified by the manufacturer. The instantaneous element shall be tested by injecting a current equal to 120% of the pickup value of the element and verifying that the circuit breaker trips instantaneously with no inten-tional time delay. Molded case circuit breaker testing shall also follow this procedure except that generally no more than two trip elements, time delay and instantaneous, will be involved. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
- 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.
O LIMERICK - UNIT 1 3/4 8-22
TABLE 3.8.4.1-1
.k PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
- 1. 4160-VOLT CIRCUIT BREAKERS CIRCUIT -
SYSTEMS OR BREAKER NO. LOCATION EQUIPMENT POWERED 152-20101 10A201 1A Reactor Recirc Pump
'A' RPT Breaker 152-20102 10A201 1A Reactor Recirc Pump
'B' RPT Breaker 152-20201 10A202 IB Reactor Recirc Pump
'A' RPT Breaker 152-20202 10A202 IB Reactor Recirc Pump
'B' RPT Breaker
- 2. 480-VOLT HOLDED CASE BREAKERS *
- Primary and backup breakers have the same device numbers and are located in the same Motor Control Center cubicle.
CIRCUIT SYSTEMS OR
- BREAKER NO. LOCATION TYPES EQUIPMENT POWERED 52-21198 D114-R-G IM HFB100 1A1 Drywell Area Unit TM HFB100 Cooler 1A1V212
.52-21109 D114-R-G IM HFB100 1El Drywell Area Unit TM HFB100 Cooler 1E1V212 52-21110 D114-R-G IM HFB100 1C1 Drywell Area Unit l TM HFB100 Cooler 1C1V212 52-21111 D114-R-G IM HFB100 1G1 Drywell Area Unit TM HFB100 Cooler 1G1V212 4
52-21124 0114-R-G IM HFB25 RHR S/D C1g. Suction Inbrd TM HFB100 Isol V1v HV-51-1F009 52-21125 D114-R-G IM HFB25 Rx Head Spray Inbrd TM HFB40 Isol Viv HV-51-1F022 52-21126 D114-R-G IM HFB50 RWCU Inbrd
.TM HFB100 Isol V1v HV-44-1F001 52-21138 D114-R-G IM HFB25 Mn Stm Line Drain Inbrd TM HFB40 Isol V1v HV-41-1F016 52-21141 D114-R-G IM HFB25 Inst Gas Compr Suct Line TM HFB40 Inbrd Isol V1v HV-59-101 LIMERICK - UNIT 1 3/4 8-23 t.
, , . , . ~ . - - . ~ ,- - .._.. ,.- ,-. , , . _ . . _ _ . _ _ . _ . _ , . . . . - , - ,
TABLE 3.8.4.1-1 (Continued) PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
- 2. 480-VOLT MOLDED CASE BREAKERS (Continued)
CIRCUIT SYSTEMS OR BREAKER NO. LOCATION TYPES EQUIPMENT POWERED 52-21208 D124-R-G IM HFB100 181 Drywell Area Unit TM HFB100 Cooler 181V212 52-21209 D124-R-G IM HFB100 IF1 Drywell Area Unit TM HFB100 Cooler 1F1V212 52-21210 D124-R-G IM HFB100 1D1 Drywell Area Unit TM HFB100 Cooler 1D1V212 52-21211 D124-R-G IM HFB100 1H1 Drywell Area Unit TM HFB100 Cooler 1H1V212 52-21216 D124-R-G IM HFB25 1B Reactor Recirc Pump TM HFB100 Suction Vlv HV-43-1F023B 52-21309 D114-R-C IM HFB50 Feedwater Line ' A' Inbrd TM HFB150 Maint V1v HV-41-1F011A 52-21707 D134-R-H IM HFB100 1C2 Drywell Area Unit TM HFB100 Cooler 1C2V212 52-21708 D134-R-H IM HFB100 1G2 Drywell Area Unit TM HFB10) Cooler 1G2V212 52-21807 D144-R-H IM HFB100 102 Drywell Area Unit TM HFB100 Cooler 102V212 52-21808 0144-R-H IM HFB100 IF2 Drywell Area Unit TM HFB100 Cooler 1F2V212 52-22310 D134-R-E IM HFB100 1A2 Drywell Area Unit TM HFB100 Cooler 1A2V212 52-22311 D134-R-E IM HFB100 1E2 Drywell Area Unit TM HFB100 Cooler 1E2V212 52-22313 D134-R-E IM HFB25 RCIC Mn Stm Supply Inbrd TM HFB40 Isol Vlv HV-49-1F007 52-22314 D134-R-E IM HFB50 Feedwater Line 'B' Inbrd. TM HFB100 Maint Vlv HV-41-1F011B LIMERICK - UNIT 1 3/4 8-24
~~
TABLE 3.8.4.1-1 (C;ntinued) ib PRIMARY CONTAINMENT PEFdTRATI0N CONDUCTOR OVERCURRENT PRCfECTIVE DEVICES
- 2. 480-V0LT MOLDED CASE BREAKERS (Continued)
CIRCUIT SYSTEMS OR BREAKER NO. LOCATION TYPES EQUIPMENT POWERED 52-22410 D144-R-E IM HFB100 182 Drywell Area Unit TM HFB100 Cooler IB2V212 52-22411 D144-R-E IM HFB100 1H2 Drywell Area Unit TM HFB100 Cooler 1H2V212 52-22418 D144-R-E IM HFB50 HPCI Mn Stm Supply Inbrd TM HFB150 Isol V1v HV-55-1F002 52-22516 114B-R-C IM HFB25 1A Reac Recirc Pump TM HFB100 Suction VLV HV-43-1F023A 52-22518 1148-R-C IM HFB25 1A Reac Recirc Pump y TM HFB100 Discharge VLV HV-43-1F031A
\. 52-22520 1148-R-C IM HFB25 Reactor Bottom Head Drain VLV TM HFB40 HV-44-1F100
.52-22536 1148-R-C IM HFB25 Reactor Bottom Head Drain VLV TM HFB40 HV-C-44-1F105 52-22534 1148-R-C IM HFB25 Reactor Vessel Head Vent-TM HFB40 HV-41-1F001 52-22535 1148-R-C IM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-1F005 52-22537 1148-R-C TM HFB15 Disposal Cask Removal Cart TM HFB20 Hoist 10H236 52-22538 1148-R-C TM HFB15 Control Rod Drive Platform TM HFB20 Hoist 10H229 52-22608 124B-R-C TM HFB15 CRD Equipment Handling IM HFB20 Platform 10N22608 l 52-22618 1248-R-C IM HFB25 IB Reac. Recirc. Pump l~ TM HFB100 Discharge VLV HV-43-1F0318 O, *52-22622 124B-R-C TM HFB125 Permanent Plant In-Containment i Welding System 10NW201 LIMERICK - UNIT 1 3/4 8-25
TABLE 3.8.4.1-1 (Continued) PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
- 2. 480-V0LT MOLDED CASE BREAKERS (Continued)
CIRCUIT 5YSTEMS OR BREAKER NO. LOCATION TYPES EQUIPMENT POWERED
*52-22626 1248-R-C TM HFB50 Unit 1 Reactor Enclosure 1L36 (Main Breaker) IL36 EB3090** Lighting XFMR 1X28 *52-22630 1243-R-C TM HFB20 1A Reac. Recirc. Pump TM HFB20 Motor Hoist 1AH203 *52-22631 124B-R-C TM HFB20 1B Reac. Recirc. Pump TM HFB20 Motor Hoist 1BH203 52-22634 1248-R-C IM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-1F002 *52-22707 114C-R-A TM HFB15 Mn Stm Relief V1v Removal TM HFB15 Hoist 10H232 *52-22708 114C-R-A TM HFB15 Mn Stm Relief Vlv Removal TM HFB15 Hoist 10H230 *These breakers shall be administratively maintained open in OPERATIONAL CONDITIONS 1, 2 and 3 and are not required to be tested.
- 208 VAC circuit breaker
,AB_b_REVI ATIONS: TM Thermal Magnetic IM I:1stantaneous Magnetic O LIMERICK - UNIT 1 3/4 8-26
/ ) ELECTRICAL POWER SYSTEMS '\._/
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION LIMITING C3NDITION FOR OPERATION 3.8.4.2 The thermal overload protection of all Class 1E motor operated valves shall be either:
- a. Continously bypassed for all valves with maintained position control switches; or,
- b. Bypassed only under accident conditiens for all valves with spring-return-to-normal control switches.
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 not bypassed continuously or only under accident conditions, as applicable, restore the thermal overload bypass within 8 hours or declare the affected valve (s) inoperable and apply the appropriate ACTION statement (s)
/N for the affected system (s).
(v) SURVEILLANCE REQUIREMENTS 4.8.4.2.1 The thermal overload protection for the above required valves which are continuously bypassed and temporarily placed in force only when the valve motor is undergoing periodic or maintenance testing shall be verified to be bypassed following periodic or maintenance testing during which the thermal overload protection was temporarily placed in force. 4.8.4.2.2 At least once per 18 months, a CHANNEL FUNCTIONAL TEST of all those valves which are bypassed only under accident conditions (valves with spring-return-to-normal control switches) shall be perfor'med to verify that the thermal overload protection will be bypassed under accident conditions. t 3 LIMERICK - UNIT 1 3/4 8-27
ELECTRICAL POWER SYSTEMS REACTOR PROTECTION SYSTEM ELECTRICAL POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.3 Two reactor protection system (RPS) electric power monitoring channels for each inservice RPS Inverter or alternate power supply shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one RPS electric power monitoring channel for an inservice RPS Inverter or alternate power supply inoperable, restore the inoperable power monitoring channel to OPERABLE status within 72 hours or remove the associated RPS Inverter or alternate power supply from service.
- b. With both RPS electric power monitoring channels for an inservice RPS Inverter or alternate power supply inoperable, restore at least one electric power monitoring channel to OPERABLE status within 24 hours or remove the associated RPS Inverter or alternate power supply from service.
SURVEILLANCE REQUIREMENTS 4.8.4.3 The above specified RPS electric power monitoring channels shall be determined OPERABLE:
- a. At least once per six months by performance of a CHANNEL FUNCTIONAL TEST.
- b. At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage, and uaderfrequency protective instrumenta-tion by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic, and output circuit breakers and verifying the following setpoints.
- 1. Overvoltage < 132 VAC,
- 2. Undervoltage 1 109 VAC,
- 3. Underfrequency 1 57 Hz.
O LIMERICK - UNIT 1 3/4 8-28
3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH LIMITING CONDITION FOR OPERATION 3.9.1 The reactor mode switch shall be OPERABLE and locked in the Shutdown or Refuel position. When the reactor mode switch is locked in the Refuel position:
- a. A control rod shall not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE.
- b. CORE ALTERATIONS shall not be performed using equipment associated with a Refuel position interlock unless at least the following Refuel position interlocks associated with that equipment are OPERABLE:
- 1. All rods in.
- 2. Refuel platform position.
- 3. Refuel platform hoists fuel-loaded.
- 4. Service platform hoist fuel-loaded.
[]Y \ APPLICABILITY: OPERATIONAL CONDITION 5* **. ACTION:
- a. With the reactor mode switch not locked in the Shutdown or Refuel position as specified, suspend CORE ALTERATIONS and lock the reactor mode switch in the Shutdown or Refuel position,
- b. With the one-rod-out interlock inoperable, lock the reactor mode switch in the Shutdown position.
- c. With any of the above required Refuel position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment interlock.
*See Special Test Exceptions 3.10.1 and 3.10.3.
**The reactor shall be maintained in OPERATIONAL CONDITTON 5 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
O) (v LIMERICK - UNIT 1 3/4 9-1
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS
- 4. 9.1.1 The reactor mode switch shall be verified to be locked in the Shutdown or Refuel position as specified:
- a. Within 2 hours prior to:
- 1. Beginning CORE ALTERATIONS, and
- 2. Resuming CORE ALTERATIONS when the reactor mode switch has been unlocked.
- b. At least once per 12 hours.
4.9.1.2 Each of the above required reactor mode switch Refuel position interlocks
- shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST within 24 hours prior to the start of and at least once per 7 days during control rod withdrawal or CORE ALTERATIONS, as applicable.
4.9.1.3 Each of the above required reactor mode switch Refuel position interlocks
- that is affected shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or CORE ALTERATIONS, as applicable, following repair, maintenance or replacement of any component that could affect the Refuel position interlock.
- The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that all control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
O LIMERICK - UNIT 1 3/4 9-2 l l
e 'N REFUELING OPERATIONS i ) 'v' 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least two source range monitor * (SRM) channels shall be OPERABLE and inserted to the normal operating level with:
- a. Continuous visual indication in the control room, )
- b. At least one with audible alarm in the control room,
- c. One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and
- d. Unless adequate shutdown margin has been demonstrated, the shorting links shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn.**
APPLICABILITY: OPERATIONAL CONDITION 5. ACTION: With the requirements of the above specification not satisfied, immediately 'v') suspend all operations involving CORE ALTERATIONS and insert all insertable control rods. SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:
- a. At least once per 12 hours:
- 1. Performance of a CHANNEL CHECK,
- 2. Verifying the detectors are inserted to the normal operating level, and
- 3. During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant.
*The use of special movable detectors during CORE . TERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors O are connected to the normal SRM circuits. ) **Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 1 3/4 9-3
i REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) i
- b. Performance of a CHANNEL FUNCTIONAL TEST:
- 1. Within 24 hours prior to the start of CORE ALTERATIONS, and
- 2. At least once per 7 days.
- c. Verifying that the channel count rate is at least 3.0 cps:*
- 1. Prior to control rod withdrawal,
- 2. Prior to and at least once per 12 hours during CORE ALTERATIONS, and
- 3. At least once per 24 hours.
- d. Verifying, within 8 hours prior to and at least once per 12 hours during, that the RPS circuitry " shorting links" have been removed during:
- 1. The time any control rod is withdrawn,** or
- 2. Shutdown margin demonstrations.
*May be reduced to 0.7 cps provided the signal-to-noise ratio is > 2.
- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 1 3/4 9-4
REFUELING OPERATIONS 3/4.9.3 CONTROL R0D POSITION LIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be inserted.* APPLICABILITY: OPERATIONAL CONDITION 5, during CORE ALTERATIONS.** ACTION: With all control rods not inserted, suspend all other CORE ALTERATIONS, except that one control rod may be withdrawn under control of the reactor mode switch Refuel position one-rod-out interlock. SURVEILLANCE REQUIREMENTS 4.9.3 All control rods shall be verified to be inserted, except as above V specified:
- a. Within 2 hours prior to:
- 1. The start of CORE ALTERATIONS.
- 2. The withdrcwal of one control rod under the control of the reactor mode switch Refuel position one-rod-out interlock.
- b. At least once per 12 hours.
*Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.
**See Special Test Exception 3.10.3.
O U LIMERICK - UNIT 1 3/4 9-5
REFUELING OPERATIONS 3/4.9.4 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.4 The reactor shall be subcritical for at least 24 hours. APPLICABILITY: OPERATIONAL CONDITION 5, during movement of irradiated fuel in the reactor pressure vessel. ACTION: With the reactor subcritical for less than 24 hodrs, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. O SURVEILLANCE REQUIREMENTS 4.9.4 The reactor shall be determined to have been subcritical for at least 24 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel. O LIMERICK - UNIT 1 3/4 9-6
REFUELING OPERATIONS
' %(']).
3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communication shall be maintained between the control room and refueling floor personnel. APPLICABILITY: OPERATIONAL CONDITION 5, during CORE ALTERATIONS.*
-ACTION:
When direct communication between.the control room and refueling floor-personnel cannot be maintained, immediately suspend CORE ALTERATIONS.* O SURVEILLANCE REQUIREMENTS 4.9.5 Direct communication between the control room and refueling floor personnel shall be demonstrated within 1 hour prior to the start of and at least once'per 12 hours during CORE ALTERATIONS.*
*Except movement of incore instrumentation and control rods with their normal drive system.
t-i LIMERICK - UNIT 1 3/4 9-7 L
REFUELING OPERATIONS 3/4.9.6 REFUELING PLATFORM LIMITING CONDITION FOR OPERATION 3.9.6 The refueling platform shall be OPERABLE and used for handling fuel assemblies or control rods within the reactor pressure vessel. APPLICABILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel. ACTION: With the requirements for refueling platform OPERABILITY not satisfied, suspend use of any inoperable refueling platform equipment from operations involving the handling of control rods and fuel assemblies within the reactor pressure v:ssel after placing the load in a safe condition. SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling platform main hoist used for handling of control rods or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 7 days prior to the start of such operations by:
- a. Demonstrating operation of the overload cutoff on the main hoist when the load exceeds 1150 1 50 pounds.
- b. Demonstrating operation of the hoist loaded control rod block interlock on the main hoist when the load exceeds 485 50 pounds.
- c. Demonstrating operation of the redundant loaded interlock on the main hoist when the load exceeds 550 + 0, - 115 pounds.
- d. Demonstrating operation of the uptravel interlock when uptravel brings the top of the active fuel to 8 feet 6 inches below the normal water level.
O LIMERICK - UNIT 1 3/4 9-8
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 4.9.6.2 The refueling platform frame-mounted auxiliary hoist used for handling of control rods within the reactor pressure vessel shall be demon-strated OPERABLE within 7 days prior to the use of such equipment by:
- a. Demonstrating operation of the overload cutoff on the frame mounted hoist when the load exceeds 1000 1 50 pounds.
- b. Demonstrating operation of the uptravel mechanical stop on the frame mounted hoist when uptravel brings the top of active fuel to 8 feet 6 inches below the normal fuel storage pool water level.
- c. Demonstrating operation of the control rod block interlock on the
. frame mounted hoist when the load exceeds 400 1 50 pounds.
4.9.6.3 The refueling platform monorail mounted auxiliary hoist used for handling of control rods within the reactor pressure vessel shall be demonstra-ted OPERABLE within 7 days prior to the use of such equipment by:
- a. Demonstrating operation of the overload cutoff on the monorail hoist when the load exceeds 1000 1 50 pounds,
- b. Demonstrating operation of the uptravel mechanical stop on the monorail hoist when uptravel brings the top of active fuel to-8 feet 6 inches below the normal fuel storage pool water level.
- c. Demonstrating operation of the control rod block interlock on the monorail hoist when the load exceeds 400 1 50 pounds.
O d LIMERICK - UNIT 1 3/4 9-9
. - _ . _ - . - - - - + -m--
l REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 1200 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage pool racks. APPLICABILITY: With fuel assemblies in the spent fuel storage pool racks. ACTION: With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks which prevent crane travel over fuel assemblies in the spent fuel storage pool racks shall be demonstrated OPERABLE within 7 days prior to and at least once per 7 days during crane operation. O LIMERICK - UNIT 1 3/4 9-10
,m /
L') REFUELING OPERATIONS 3/4.9.8 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.8 At least 22 feet of water shall be maintained over the top of the reactor pressure vessel flange. APPLICABILITY: During handling of fuel assemblies or control rods within the reactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel after placing all fuel assemblies and control rods in a safe condition. O SURVEILLANCE REQUIREMENTS r 4.9.8 The reactor vessel 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 during handling of fuel assemblies or control rods within the reactor pressure vessel. m LIMERICK - UNIT 1 3/4 9-11
REFUELING OPERATIONS 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.9 At least 22 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel storage pool. ACTION: With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.9 The water level in the spent fuel storage pool shall be determined to be at least at its minimum required depth at least once per 7 days. O LIMERICK - UNIT 1 3/4 9-12
4 l p) REFUELING OPERATIONS 1.) ' 3/4.9.10 CONTROL R0D REMOVAL SINGLE CONTROL R0D REMOVAL LIMITING CONDITION FOR OPERATION I 3.9.10.1 One control rod and/or the associated control rod drive mechanism may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is fully inserted in the core.
- a. The reactor mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Table 1.2 and Specification 3.9.1.
- b. The source range monitors (SRM) are OPERABLE per Specification 3.9.2.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed;
,-- 1. May be assumed to be the highest worth control rod required to
( ; he assumed to be fully withdrawn by the SHUTDOWN MARGIN test, (_,/ and
- 2. Need not be assumed to be immovable or untrippable.
- d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
- e. All other control rods are inserted.
APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5. ACTION: With the requirements of the above specification not satisfied, suspend removal of the control rod and/or associated control rod drive mechanism from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements. r'~s i / v 9 LIMERICK - UNIT 1 3/4 9-13
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.10.1 Within 4 hours prior to the start of removal of a control rod and/or the associated control rod drive mechanism from the core and/or reactor pressure vessel and at least once per 24 hours thereafter until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that:
- a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1.
- b. The SRM channels are OPERABLE per Specification 3.9.2.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied per Specification 3.9.10.1c.
- d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
- e. All other control rods are inserted.
O LIMERICK - UNIT 1 3/4 9-14
C'\
- t REFUELING OPERATIONS
\ /
MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.2 Any number of control rods and/or control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the Core.
- a. The reactor mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Specification 3.9.1, except that the Refuel position "one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below,
- b. The source range monitors (SRM) are OPERABLE per Specification 3.9.2.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
('~') d. All other control rods are either inserted or have the surrounding
\s_,/ four fuel assemblies removed from the core cell.
- e. The four fuel assemblies surrounding each control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
APPLICABILITY: OPERATIONAL CONDITION 5. ACTION: With the requirements of the above specification not satisfied, suspend removal of control rods and/or control rod drive mechanisms from the core and/or reactor i pressure vessel and initiate action to satisfy the above requirements. 4 + lO l xs l i ! LIMERICK - UNIT 1 3/4 9-15
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.10.2.1 Within 4 hours prior to the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and at least once per 24 hours thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core, verify that:
- a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position per Specification 3.9.1.
- b. The SRM channels are OPERABLE per Specification 3.9.2.
- c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
- d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.
- e. The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.
4.9.10.2.2 Following replacement of all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had been bypassed. O LIMERICK - UNIT 1 3/4 9-16
+
REFUELING OPERATIONS f'O).
'"~'
3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.1 At least one shutdown cooling mode loop of the residual heat removal (RHR) system shall be OPERABLE and in operation
- with at least:
- a. One OPERABLE RHR pump, and
- b. One OPERABLE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet above the top of the reactor pressure vessel flange. ACTION:
- a. With no RHR shutdown cooling mode loop OPERABLE, within 1 hour and at least once per 24 hours thereafter, demonstrate the OPERABILITY S
of at least one alternate method capable of decay heat removal. [' Otherwise, suspend all operations involving an increase in the ( reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours.
- b. With no RHR shutdown cooling mode loop in operation, within 1 hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.
SURVEILLANCE REQUIREMENTS 4.9.11.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
*The shutdown cooling pump may be removed from operation for up to 2 hours .
per 8-hour period.
! ks
- LIMERICK - UNIT 1 3/4 9-17
REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and at least one loop shall be in operation,* with each loop consisting of at least:-
- a. One OPERABLE RHR pump, and
- b. One OPERABLE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet above the top of the reactor pressure vessel flange. ACTION:
- a. With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within 1 hour and at least once per 24 hours thereafter, demonstrate the OPERABILITY of at least one alternate method capable of decay heat removal for each inoperable RHR shut-down cooling mode loop.
- b. With no RHR shutdown cooling mode loop in operation, within 1 hour establish reactor coolant circulation by an alternate method and
, monitor reactor coolant temperature at least once per hour.
SURVEILLANCE REQUIREMENTS 4.9.11.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
- The shutdown cooling pump may be removed fr6m operation for up to 2 hours per 8-hour period.
O LIMERICK - UNIT 1 3/4 9-18
C' 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.10.1 The provisions of Specifications 3.6.1.1, 3.6.1.3, and 3.9.1 and Table 1.2 may be suspended to permit the reactor pressure vessel closure head and the drywell head to be removed and the primary containment air lock doors to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER and reactor coolant temperature less than 200 F. APPLICABILITY: OPERATIONAL CONDITION 2, during low power PHYSICS TESTS. ACTION: With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 200 F, immediately place the reactor mode switch in the Shutdown position. m (v SURVEILLANCE REQUIREMENTS 4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified to be within the limits at least once per hour during low power PHYSICS TESTS. O LIMERICK - UNIT 1 3/4 10-1
SPECIAL TEST EXCEPTIONS 3/4.10.2 R0D SEQUENCE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod worth minimizer (RWM) per Specification 3.1.4.1 and by the rod sequence control system (RSCS) per Specification 3.1.4.2 may be suspended by means of bypass switches for the following tests provided that control rod movement prescribed for this testing is verified by a second licensed operator or other technically qualified member of the unit technical staff present at the reactor console:
- a. Shutdown margin demonstration, Specification 4.1.1.
- b. Control rod scram, Specification 4.1.3.2.
- c. Control rod friction measurements.
- d. Startup Test Program with the THERMAL POWER less than 20% of RATED THERMAL POWER.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With the requirements of the above specification not satisfied, verify that the RWM and the RSCS are OPERABLE per Specifications 3.1.4.1 and 3.1.4.2, respectively. SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RSCS and/or RWM are bypassed, verify;
- a. That movement of control rods from 75% R0D DENSITY to the RSCS preset power level is blocked or limited to the approved control rod with-drawal sequence during scram and friction tests.
- b. That movement of control rods during shutdown margin demonstrations is limited to the prescribed sequence per Specification 3.10.3.
- c. Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff. l l
9 LIMERICK - UNIT 1 3/4 10-2
SPECIAL TEST EXCEPTIONS
.g)
(
\
3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CONDITION FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3, and Table 1.2 may be suspended to permit the reactor mode switch to be in the Startup position and to allow core than one control rod to be withdrawn for shutdown margin demonstration, provided that at least the following requirements are . satisfied.
- a. The source range monitors are OPERABLE with the RPS circuitry " shorting links" removed per Specification 3.9.2.
- b. The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and is programmed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.
- c. The " continuous rod withdrawal" control shall not be used during out-of-sequence movement of the control rods.
[]
\ )
- d. No other CORE ALTERATIONS are in progress.
v OPERATIONAL CONDITION 5, during shutdown margin demonstrations. APPLICABILITY: ACTION: With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refuel position. SURVEILLANCE REQUIREMENTS 4.10.3 Within 30 minutes prior to and at least once per 12 hours during the performance of a shutdown margin demonstration, verify that;
- a. The source range monitors are OPERABLE per Specification 3.9.2,
- b. The rod worth minimizer is OPERABLE with the required program per Specification 3.1.4.1 or a second licensed operator or other techni-cally qualified member of the unit technical staff is present and verifies compliance with the shutdown margin demonstration procedures, and
, c. No other CORE ALTERATIONS are in progress.
v) LIMERICK - UNIT 1 3/4 10-3
SPECIAL TEST EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 that recirculation loops be in operation may be suspended for up to 24 hours for the performance of:
- a. PHYSICS TESTS, provided that THERMAL POWER does not exceed 5% of RATED THERMAL POWER, or
- b. The Startup Test Program.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, during PHYSICS TESTS and the Startup Test Program. ACTION:
- a. With the above specified time limit exceeded, insert all control rods,
- b. With the above specified THERMAL POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.
SURVEILLANCE REQUIREMENTS 4.10.4.1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours at least once per hour during PHYSICS TESTS and the Startup Test Program. 4.10.4.2 THERMAL POWER shall be determined to be less than 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. O LIMERICK - UNIT 1 3/4 10-4
,e SPECIAL TEST EXCEPTIONS \, 3/4.10.5 OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.10.5 The provisions of Specification 3.6.6.3 may be suspended during the performance of the Startup Test Program until either the required 100% of RATED THERMAL POWER trip tests have been completed or the reactor has operated for 120 Effective Full Power Days. APPLICABILITY: OPERATIONAL CONDITION 1. ACTION With the requirements of the above specification not satisfied, be in at least STARTUP within 6 hours. SURVEILLANCE REQUIREMENTS 4.10.5 The Effective Full Power Days of operation shall be verified to be less than 120, by calculation, at least once per 7 days during the Startup Test Program. O LIMERICK - UNIT 1 3/4 10-5
SPECIAL TEST EXCEPTIONS 3/4.10.6 TRAINING STARTUPS LIMITING CONDITION FOR OPERATION 3.10.6 The provisions of Specification 3.5.1 may be suspended to permit one RHR subsystem to be aligned in the shutdown cooling mode during training startups provided that the reactor vessel is not pressurized, THERMAL POWER is less than or equal to 1% of RATED THERMAL POWER and reactor coolant temperature is less than 200 F. APPLICABILITY: OPERATIONAL CONDITION 2, during training startups. ACTION: With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position. SURVEILLANCE REQUIREMENTS O 4.10.6 The reactor vessel shall be verified to ba unpressurized and the THERMAL POWER and reactor coolant temperature shall be verified to be within the limits at least once per hour during training startups. O LIMERICK - UNIT 1 3/4 10-6
j'~'N 3/4.11.1 LIQUID EFFLUENTS I CONCENTRATION-LIMITING CONDITION FOR OPERATION 1 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1.3-1) 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 microcuries/mi total activity. APPLICABILITY: At all times. ACTION: i With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concen-
'tration to within the above limits.
SURVi!LLANCE REQUIREMENTS 'I (' 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table .4.11.1.1.1-1. l 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. l 1 7 l i l N. LIMERICK - UNIT 1 3/4 11-1
TABLE 4.11.1.1.1-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRA_M LOWER LIMIT MINIMUM TYPE OF 0FDETEC{ ION LIQUID RELEASE SAMPLING ANALYSIS ACTIVITY (LLD) TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/mL) A. Batch Waste P P PrincipafGamma 5x10 7 Release Each Batch Each Batch Emitters b Tanks I-131 1x10 8
- 1. Floor Drain P M Dissolved and 1x10 8 Sample Tank One Batch /M Entrained Gases No. 2 (Gamma Emitters)
- 2. Laundry P M H-3 1x10 8 Drain Sample Composite d
Each Batch Tank Gross Alpha 1x10 7 P Q Sr-89, Sr-90 5x10 8 d x Each Batch Composite Fe-55 1x10 8
/
B. Continuogs W W PrincipafGamma 5x10 7 Release Grab Sample Emitters I-131 1x10 8
- 1. RHR Service W W Dissolved and 1x10 8
~ Water System Grab Sample Entrained Gases Effluent Line 7 (Gamma Emitters)
- 2. S e'Wa r W M d
H-3 1x10 s System ra ample Composite Effluent Gross Alpha 1x10 7 W Q d Sr-8 M D' Sx10 8 Grab Sample Composite Fe-55 1x10 8 O LIMERICK - UNIT 1 3/4 11-2
. . _ . - -_ ~ - - . _- .-
TABLE 4.11.1.1.1-1 (Continued) () TABLE NOTATIONS a The LLD is defined, for purposes of these specifications, as the smallest
- concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation: 4.66s b LLD = E V 2.22 x 108 " Y exp (-AAt) Where: LLD is the a priori lower limit of detection as defined above (as microcuries per unit mass or volume), s h is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute), J ? E is the counting efficiency, as counts per disintegration, ( p) V V is the sample size, in units of mass or volume, 2.22 x 108 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and At for the plant effluenLs is the elapsed time between the midpoint of sample collection and time of counting. Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a costeriori (after the fact) limit for a particular measurement. d LIMERICK - UNIT 1 3/4 11-3
TABLE 4.11.1.1.1-1 (Continued) TABLE NOTATIONS b A batch release is the oischarge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure 'epresentative sampling. c The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-annual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8. d A composite sample is one in which the quantity of liquid sampled is propor-tional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. "A continuous release is the discharge of liquid wastes of a nondircrete volume, e.g., from a volume of a system that has an input flow during the continuous release. I Whenever effluent releases are in excess of the monitor's setpoint. O LIMERICK - UNIT 1 3/4 11-4
d 7- s RADI0 ACTIVE EFFLUENTS
/ \
i s_ ,/ 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 reactor unit to UNRESTRICTED AREAS (See Figure 5.1.3-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to the total 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 total body and to less than or equal to 10 mrems to any organ.
4 APPLICABILITY: At all times, o 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 Specifica-tion 6.9.2, a Special Report which identifies the cause(s) for f%g exceeding the limit (s) and defines the corrective actions that have 1 ('/ 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. This Special Report shall also include the radiological impact on finished drinking water supplies at the nearest downstream drinking water source.
i b. The provisions of specification 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. l f L) LIMERICK - UNIT 1 3/4 11-5
, . . - , ~ , . - . . , _ - _.-.m.,-- -,., _ -_ . . ,-- , , - - .-_.. --, - . _ . . . . ~ . _.- --
RADI0 ACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE and appropri-ate portions of the system shall be used to reduce the radioactive materials in liquid waste prior to their discharge when the projected doses due to the liquid effluent, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1.3-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31-day period. APPLICABILITY: At all times. ACTION:
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-systems, 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 reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM. 4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2. O LIMERICK - UNIl 1 3/4 11-6
r
/7 4 RADI0 ACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outside temporary tanks 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 of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and 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 Report,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above
' tanks shall be determined to be within the above limit by analyzing a repre-sentative sample of the tank's contents at least once per 7 days when radio-active materials are being added to the tank.
LIMERICK - UNIT 1 3/4 11-7
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS 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 or beyond the SITE BOUNnARY (see Figure 5.1.3-1) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and
- b. For iodine-131, for iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days; Less than or equal to 1500 mrems/yr to any organ. (Inhalation pathways only.)
APPLICABILITY: At all times. ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limits.
- b. .The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 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 of the ODCM. 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 of the ODCM by obtain!ng representative samples and performing analyses in accordance with the sampling and analysis program specified in Tabie 4.11.2.1.2-1. O LIMERICK - UNIT 1 3/4 11-8
k / k
) l
%) (V) /.
J-TABLE 4.11.2.1.2-1 C iR RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM l 5 9 i MINIMUM LOWER LIMIT OF 5 SAMPLING ANALYSIS TYPE OF DETECTION (LLD)a Q GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/mL) w P P A. Containment Purge Each Purge Each Purge Principal Gamma Emitters *'I 1x10 4 (Pretreatment) Grab Sample -i B. North Stack and M b gb Principal Gamma Emitters
- 1x10 4 South Stack Grab Sample H-3 1x10 6 C. Hot Maintenance Continuous d yc I-131 1x10 12 O Shop Ventilation Charcoal I-133 1x10 1 a Sample Exhausth and All Release Types ;
- Listed in B Continuous d yc Principal Gamma Emitters
- 1x10 11 j above Particulate (I-131, Others)
Sample d Continuous Q Gross Alpha 1x10 11 l Composite Par-l ticulate Sample l d l Continuous Q Sr-89, Sr-90 1x10 11 l Composite Par-
- ticulate Sample i
d D. All Release Types Continuous Noble Gas . Noble Gases 1x10 8 Listed in B above Monitor Gross Beta or Gamma (Based on and the Main Xe-133) Condenser Off Gas Pretreatment Radio-activity Monitor i t
TABLE 4.11.2.1.2-1 (Continued) TABLE NOTATIONS a The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a
'real" signal.
For a particular measurement system, (which may include radiochemical separation): 4' $ D LLD = E V 2.22 x 106 Y exp (-Aat) Where: LLD is the a priori lower limit of detection as defined above (as microcuries per unit mass or volume), s b is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute), E is the counting efficiency (as counts per disintegration), V is the sample size (in units of mass or volume), 2.22 x 106 is the number of disintegrations per minute per mic.*ocurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples) The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. Typical values of E, V, Y, and at shall be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a_ posteriori (after the fact) limit for a particular measurement. O LIMERICK - UNIT 1 3/4 11-10
i
, ~)i TABLE 4.11.2.1.2-1 (Continued)
V TABLE NOTATIONS b Sampling and analyses shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the main condenser offgas pre-treatment radioactivity monitor shows that effluent activity has not increased more than a factor of 3. c Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour and analyses completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not incteased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3. d The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation (~1 made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.
- i. )
U e The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Da-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 for par-ticulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks which are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report, pursuant to Specifica-tion 6.9.1.8. f Under the provisions of footnote e. above, only noble gases need to be considered. g Roquired only when handling or storing irradiated fuel in the secondary containment.
- h. Required for the hot maintenance shop ventilation exhaust only during opera-tion of the hot maintenance shop ventilation exhaust system.
O v LIMERICK - UNIT 1 3/4 11-11
RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) 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 ca'lculated 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 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releces and the proposed corrective actions to be taken to assure than subsequent releases will be in compliance with the above limits.
- b. The provisions of Specification 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 LIMERICK - UNIT 1 3/4 11-12
RADI0 ACTIVE EFFLUENTS (]) b DOSE - 10 DINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES 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 reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) 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, p tritium, and radionuclides in particulate form with half-lives I greater than 8 days, in gaseous effluents exceeding any of the above (d limits, prepare and st.bmit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure than 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 ODCM at least once per 31 days. O t v/ LIMERICK - UNIT 1 3/4 11-13
RADI0 ACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM sball be OPERABLE and appropriate portions of the system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from each reactor unit to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) when averaged over 31 days would exceed 0.3 mrem to any organ in a 31-day period. APPLICABILITY: At all times. ACTION:
- a. With 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 6.9.2, a Special Report which includes the following information:
- 1. Explanation of why gaseous radwaste was being discharged without 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 4.11.2.4.1 Doses due to gaseous releases from each reactor unit to areas at and beyond the SITE B0UNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM. 4.11.2.4.2 The VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by meeting Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. O LIMERICK - UNIT 1 3/4 11-14
RADIOACTIVE EFFLUENTS
-~3
[ l
\_/
s EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be limited to less than or equal to 4% by volume. APPLICABILITY: Whenever the main condenser air ejector system is in operation. ACTION:
- a. With the concentration of hydrogen in the main condenser offgas treatment system exceeding the limit, restore the concentration to within the limit within 48 hours.
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE. REQUIREMENTS 4.11.2.5 The concentration of hydrogen in the main condenser offgas treatment
- system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system with O the hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specifica-tion 3.3.7.12.
O G LIMERICK - UNIT 1 3/4 11-15
RADI0 ACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.6 The rate of the sum of the activities of the noble gases Ar-85m, Kr-87, Kr-88, Xe-133, Xe-135, and Xe-138 measured at the recombiner after-condenser discharge shall be limited to less than or equal to 330 millicuries /second. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3*. ACTION: With the rate of the sum of the activities of the specified noble gases at the recombiner after-condenser discharge exceeding 330 millicuries /second, restore the gross radioactivity rate to within its limit within 72 hours or be ' in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.11.2.6.1 The rate of the sum of the activities of noble gases at the recombiner after-condenser discharge shall be continously monitored in accor-dance with Specification 3.3.7.12. 4.11.2.6.2 The rate of the sum of the activities of the specified noble gases from the recombiner after-condenser discharge shall be determined to be within the limits of Specification 3.11.2.6 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the recombiner after condenser discharge:
- a. At least once per 31 days,
- b. Within 4 hours following an increase, as indicated by the Main Condenser Off-Gas Pretreatment Radioactivity Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level or air in-leakage, in the nominal steady-state fission gas release from the primary coolant.
- c. The provisons of Specification 4.0.4 are not applicable.
*When the main condenser air ejector is in operation.
O LIMERICK - UNIT 1 3/4 11-16
/~ RADI0 ACTIVE EFFLUENTS i]
V VENTING OR PURGING LIMITING CONDITION FOR OPERATION 3.11.2.7 VENTING or PURGING of the Mark II containment shall be through the standby gas treatment system. APPLICABILITY: Whenever the containment is vented or purged.* ACTION:
- a. With the requirements of the above specification not satisfied, suspend all VENTING and PURGING of the containment.
- b. The provisions of Specifications 3.0.3 and'3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
/'~'s 4.11.2.7.1 The containment shall be determined to be aligned for VENTING or ' ( ,)
PURGING through the standby gas treatment system within 4 hours prior to start of and at least-once per 12 hours during VENTING or PURGING of the containment. 4.11.2.7.2 Prior to use of the purge system through the standby gas treatment 4 system assure that:
- a. Both standby gas treatment system trains are OPERABLE whenever the purge system is in use, and
- b. .Whenever the purge system is in use during OPERA 7 TONAL CONDITION 1 or 2 or 3, only one of the standby gas treatnent system trains may be used.
*Except for the one inch /two inch vent valves to the Reactor Enclosure Equipment Compartment Exhaust Filters when used for containment pressure control and nitrogen make-up operations. ,s
!V l LIMERICK - UNIT 1 3/4 11-17
RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADWASTE TREATMENT 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 rece.ived 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 the SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1) test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and (2) take appropriate administrative action to prevent recurrence.
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
'l SURVEILLANCE REQUIREMENTS 4.11.3.1 The PROCESS CONTROL PROGRAM shall be used to verify that the properties of the packaged waste meet the minimum stability requirements of 10 CFR Part 61 and other requirements for transportation to the disposal site and receipt at the disposal site. O LIMERICK - UNIT 1 3/4 11-18
. (o
\
~
RADI0 ACTIVE EFFLUENTS SURVEILLANCE REQUIREMENTS (Continued) 4.11.3.2 If the SOLIDIFICATION method is used, the PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, and sodium' sulfate solutions).
- a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-TION of the batch under test shall be suspended until such time as additonal 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 le'ast three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided
[- } in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste. 1
'\s) ss \d )
LIMERICK - UNIT 1 3/4 11-19
RADIGACTIVE 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 radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total 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 reactor units and from outside storage tanks to determine 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 6.9.2, 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 20.405c, 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) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentra-tions. If the estimated dose (s) excerds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance 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 If the cumulative dose contributions exceed the limits defined in 3.11.4, ACTION a, Cumulative dose contributions from direct radiation from unit operation shall be determined in accordance with the methodology and parameters in the ODCM. LIMERICK - UNIT 1 3/4 11-20
/
(%.) ) 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12.1-1. APPLICABILITY: At all times. ACTION:
- a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report per Specification 6.9.1.7, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant
,ss to Specification 6.9.2, a Special Report that identifies the cause(s)
( ) for exceeding the limit (s) and defines the corrective actions to be
\s_ / taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) reporting level (1) , concentration (2) reporting level (2) + ***> 1.0 When radionuclides other than those in Table 3.12.1-2 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 is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 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 described in the Annual Radiological Environmental Operating Report.
/~'s v
LIMERICK - UNIT 1 3/4 12-1
RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12.1-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Specifica-tion 6.9.1.8, identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).
- d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12.1 2, Lne detection capabilities required by Table 4.12.1-1. O LIMERICK - UNIT 1 3/4 12-2
s V V TABLE 3.12.1-1 E RADIOLOGICAL ENVIRONMENTAL' MONITORING PROGRAM 9 h NUMBER OF REPRESENTATIVE SAMPLING AND i EXPOSURE PATHWAY SAMPLES AND TYPE AND FREQUENCY c AND/0R SAMPLE SAMPLE LOCATIONS (3) COLLECTION FREQUENCY OF ANALYSIS Z 1. DIRECT RADIATION (b) 40 routine monitoring stations At least Quarterly. Gamma dose at least s either with two or more dosimeters quarterly. or with one instrument for measuring and recording dose rate continuously placed as follows: (1) An inner ring of stations, one in each meteorological sector, in the general area of the SITE BOUNDARY; (2) An outer ring of stations, one in each meteorological sector, in the 3-to 9-mile range from the site; (3) The R balance of the stations placed in special interest areas such as U population centers, nearby resi-O dences, schools, and in 1 or 2 areas to serve as control stations.
- 2. AIRBORNE Radioiodine and Samples from 5 locations: Continuous sampler Radiciodine Cannister:
Particulates operation with sample I-J" analysis weekly.
- a. 3 samples from close to the collection weekly, or 3 SITE BOUNDARY locations (in more frequently if different sectors) of the required by dust Particulate Sampler:
- highest calculated annual loading.c Gross beta radio average groundlevel X/Q. activity analysis
- b. 1 sample from the vicinity following filter community having one of the d l ha highest calculated annual 3 l Ga 1s topic I highest groundlevel X/Q. analysis of composite (by l
' location) at least quarterly. l
TABLE 3.12.1-1 (Continued) C M RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 5 9 NUMBER OF REPRESENTATIVE c- EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY SAMPLE LOCATIONS a) Z AND/0R SAMPLE COLLECTION FREQUENCY OF ANALYSIS ~ 2. AIRBORNE (Continued) 1 sample from a control loca tion, as for example 15-30 km distant and in the least pre valent wind direction.
- 3. WATERBORNE
- a. Surface # 1 sample upstream. Camma isotopic analysis' 1 sample downstream. Compositesamplgover 1-month period. monthly. Composite for w tritium analysis quarterly.
A g b. Ground Samples from 1 or 2 sources Qu rterly. Gamma isotopic" and tritium only if likely to be affected.b analysis quarterly.
- c. Drinking 1 sample of each of 1 to 3 Monthly. (Composite) I-131 analysis on each of the nearest water supplies composite when the dose that could be affected by its calculated for the con discharge. sumption of the water is greater than 1 mrem per year. i Composite 1 sample from a control for gross beta and gamma location. isotopic analyses' monthly.
Composite for tritium analysis quarterly.
- d. Sediment 1 sample from downstream area Semiannually. Gamma isotopic analysis' from with existing or potential semiannually.
shoreline recreational value. O - O O
m TABLE 3.12.1-1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM f h t E h NUMBER OF REPRESENTATIVE
' EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY g AND/0R SAMPLE SAMPLE LOCATIONS (3) _ COLLECTION FREQUENCY OF ANALYSIS U 4. INGESTION
- a. Milk a. Samples from milking animals Semimonthly when ani Gamma isotopic and I-131 in 3 locations within 5 km mais are on pasture, analysis semimonthly when distance having the highest monthly at other times. animals are on pasture; dose potential. If there are monthly at other times.
none, then,1 sample from milk-I ing animals in each of 3 areas between 5- to 8-km distance where doses are calculated to be q greater than 1 mrem per yr.I l
- 1 sample from milking animals 7
at a control location (15-30 km km distant and in the least prevalent wind direction).
- b. Fish and a. 1 sample of two recreationally Sample in season, or Gamma isotopic Invertebrates important species in vicinity semiannually if they analysis on edible
! of plant discharge area. are not seasonal. portions. 1 sample of same species in areas not influenced by plant discharge. i 2
TABLE 3.12.1-1 (Continued)
- o y RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM I
E NUMBER OF REPRESENTATIVE Z EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY w AND/0R SAMPLE SAMPLE LOCATIONS (a) COLLECTION FREQUENCY OF ANALYSIS
- 4. INGESTION (Continued)
- c. Food Products Samples of 3 different Monthly when available, Gamma isotopic
- and kinds of broad leaf if milk sampling is I-131 analysis.
vegetation grown nearest not performed. each of two different offsite locations of highest y predicted annual average a ground level D/Q if milk sampling is not performed.
- 1 sample of each of the similar Monthly when available, Gamma isotopic" and broad leaf vegetation grown if milk sampling is I-131 analysis.
15-30 km distant in the least not performed, prevalent wind direction if milk sampling is not performed. O O O
a l RADIOLOGICAL ENVIRONMENTAL MONITORING h[qi TABLE 3.12.1-1 (Continued) TABLE NOTATIONS
- Specific parameters of distance and direction sector from the centerline of the two reactors and additional description where pertinent, shall be provided i for each and every sample location in Table 3.12.1-1 in a table and figure (s) in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
It is recognized that, at times, it may not be possible or practicable to con-tinue to obtain ramples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitori'ig program. Pursuant to Specification 6.9.1.8, identify the cause of the unavailability of samples f'or that pathway and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).
/O b One or more instruments, such as a pressurized ion chamber, for measuring and b) recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation.
c Methodology for recovery of radioiodine shall be described in the ODCM. d Airborne particulate sample filters shall be analyzed for gross beta radio-activity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
' Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone. O V LIMERICK - UNIT 1 3/4 12-7
RADIOLOGICAL ENVIRONMENTAL MONITORING TABLE 3.12.1-1 (Continued) TABLE NOTATIONS 9A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g. , hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. h Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination. I The dose shall be calculated for the raaximum organ and age group, using the methodology and parameters in the 00CM. O O LIMERICK - UNIT 1 3/4 12-8
) hT i
) O V
[ TABLE 3.12.1-2 M
$ REPORTING LEVELS FOR RAOI0 ACTIVITY CONCENTRATIONS.IN ENVIRONMENTAL SAMPLES R
, Reporting Levels E
G g WATER AIRBORNE PARTICULATE FISH MILK. F000 PRODUCTS ANALYSIS (pCf/L) or GASES (pCi/m3 ) -(pC1/kg, wet) -(pCi/L) (pCi/kg, wet) H-3 20,000* Mn-54 1,000 30,000 Fe-59 400 10,000 w Co-58 1,000 30,000 A y Co-60 300 10,000 i 7
- Zn-65 300 20,000 Zr-Nb-95 400**
I-131 2 0. 9 3 100 Cs-134 30 10 1,000 60 1,000 i Cs-137 50 20 2,000 70 2,000 i Ba-La-140 200** 300 l *For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/L may be used.
** Total for parent and daughter.
i
C M TABLE 4.12.1-1 x h DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (a) LOWER LIMIT OF DETECTION (LLD){ )(C) 5
-8 W
WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS SEDIMENTS ANALYSIS (pCi/L) OR GASES (pCi/m3 ) (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry) Gross Beta 4 0.01 H-3 2000 Mn-54 15 130 w Fe-59 30 260 h
- Co-58, 60 15 130 1 7 Zn-65 30 260 a; Zr-95 30 Nb-95 15 I-131 1(d) 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 13 80 180 l Ba-140 60 60 La-140 15 15 1
l l 9 O O
~ TABLE 4.12.1-1 (Continued) TABLE NOTATIONS (a)This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable at 95% confidence level, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radio-logical Environmental Operating report pursuant to Specification 6.9.1.7. (b) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13. (c)The LLD is defined, for purpose of these specifications, as the smallest con-centration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): 4.66 s D m LLD = E V 2.22 Y exp(-AAt) where . LLD is the "a priori" lower limit of detection as defined above (as picocuries per unit mass or volume), shis the standard deviation of the background counting rate or of the cDunting rate of a blank sample as appropriate (as counts per minute), E is the counting efficiency (as counts per disintegration), V is the sample size (in units of mass or volume), 2.22 is the number of disintegrations per minute per picacurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and i At for environmental samples is the elapsed time between sample collection (or end of the sample collection period) and time of counting. Typical values of E, V, Y, and At should be used in the calculation. t ( v LIMERICK - UNIT 1 3/4 12-11
Table 4.12.1-1 (Continued) TABLE NOTATIONS It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measuremeiit system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contrib-uting factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7. (d)LLD for drinking water samples. O O LIMERICK - UNIT 1 3/4 12-12
( f] RADIOLOGICAL ENVIRONMENTAL MONITORING
- 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 m 2 (500 ft 2) producing broad leaf vegetation.
APPLICABILITY: At all times. ACTION:
- a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new loca-tion (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.8.
- b. With a land use census identifying a location (s) that yields a calculated 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 loca-
/^'s tion (s) to the radiological environmental monitoring program within
(') 30 days. 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 6.9.1.8, identify the new
. location (s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7.
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors p with the highest predicted D/Qs in lieu of the garden census. Specifications .
y) ( for broad leaf vegetation sampling in Table 3.12.1-1 item 4.c. shall be followed, including analysis of control samples. LIMERICK - UNIT 1 3/4 12-13
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive m'aterials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. 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 6.9.1.7.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7. O LIMERICK - UNIT 1 3/4 12-14
o s sJ m a L., , is &+ gA .- A $.-
,)
BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION i AND l SURVEILLANCE REQUIREMENTS i
-e I
h t-l [
i l l 9 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 0' l 1 l
.g ( 3/4.0 APPLICABILITY I l BASES The specifications of this section provide the general requ.irements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements.within Section 3/4. 3.0.1 This specification states the applicability of each specification in terms of defined OPERATIONAL CONDITION or other specified applicability condition and is provided to aelineate specifically when each specification is applicable. 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 This 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 the specification. For example, Specifi-cation 3.7.2 requires two control room emergency filtration subsystems to be OPERABLE,and provides explicit ACTION requirements if one subsystem is
' inoperable. Under the requirements of Specification 3.0.3, if both of the required subsystems are inoperable, within one hour measures must be initiated to place the unit in at least STARTUP within the next 6 hours, in at least HOT SHUTDOWN within the following 6 hours and in COLD SHUTDOWN within the subsequent 24 hours. As a further example, Specification 3.6.6.1 requires two primary containment hydrogen recombiner systems to be OPERABLE and provides explicit ACTION requirements if one recombiner system is inoperable. Under the require-ments of Specification 3.0.3, if both of the required systems are inoperable, within 1 hour measures must be initiated to place the unit in at least STARTUP within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
3.0.4 This specification provides that entry into an OPERATIONAL CONDITION must be made with (a) the full complement of required systems, equipment or components' OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements. The intent of this provision is to ensure that unit operation is not initiated with either required equipment or systems inoperable or other 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 Q safety. These exceptions are stated in the ACTION statements of the appropriate specifications. LIMERICK - UNIT 1 B 3/4 0-1 s
APPLICABILITY BASES 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 CONDITIONS or other conditions for which the Limiting Conditions for Operation are applicable. Provisions.for additional surveillance activities to be performed without regard to the applicable OPERATIONAL CONDITIONS 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 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allow-able tolerance; instead, it permits the more frequent performance of surveillance activities. The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval. 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under this criteria, equipment, systems, or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems, or components OPERABLE, when sud items are found or known to be inoperable although still meeting the Surveillance Requirements. I 4.0.4 This specification ensures that surveillance activities associated with a Limiting Conditions for Operation have been performed within the specified time interval prior to entry into an applicable OPERATIONAL CONDITION or other specified applicability condition. The intent of this provision is to ensule that surveillance 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 outage, the applicable surveillance activ-ities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status. LIMERICK - UNIT 1 8 3/4 0-2
p)
\
v APPLICABILITY BASES 4.0.5 This specification ensures that 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 will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR Part 50, Section 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications. This specification includes a clarification of the frequencies of performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure V ssel Code and applicable Addenda. This clarifica-tion is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an [m\ OPERATIONAL CONDITION or other specified applicability condition takes , (,,/ precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. And for example, 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 declared inoperatele and takes precedence over the ASME Boiler and Pressure Vessel provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. h LIMERICK - UNIT 1 B 3/4 0-3
[ U 3/4.1 REACTIVITY CONTROL SYSTEMS BASES
'/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated 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.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38% a k/k or R + 0.28% a k/k, as appropriate. The 0.38% A k/k includes uncertainties and calculation biases. The value of R in units of % A k/k is the difference between the calculated value of minimum shutdown margin during the operating cycle and the calculated shutdown margin at the time of the shutdown margin test at the beginning of cycle. The value of R must be positive or zero and must be determined for each fuel loading cycle. Two different values are supplied in the Limiting Condition for Ooeration [ N to provide for the different methods of demonstration of the SHUTDOWN MARGIN. The highest worth rod may be determined analytical 1. or by test. The SHUTDOWN MARGIN is demonstrated by (an insequence) control rod withdrawal at the beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure. Observation of subcriticality in this condition assures subcriticality with the most reactive control rod fully withdrawn. This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is inaipble of insertion. 3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditicas is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns. Since the comparisons are easily done, frequent checks are not an imposition on normal operations. A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated. A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients. , \. ,I LIMERICK - UNIT 1 B 3/4 1-1
I REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis. Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods. Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements. The number of control rods permitted to be inoperable could be more than the eight allowed by the spacification, but the occurrence of eight inoperable rods could be indicative af a generic problem and the reactor must be shutdown for investigation and resolution of the problem. The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than 1.06 during the limiting power transient analyzed in Section 15.2 of the FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than 1.06. The occurrence of scram time's longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem. The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required. Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods Gven under the most unfavorable depressurization of the reactor. LIMERICK - UNIT 1 B 3/4 1-2
REACTIVITY CONTROL SYSTEMS k BASES CONTROL'R005 (Continued) Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel _ position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be nerformed prior to'ichieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demon-
~s tration.
In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position
~ indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no press ee to act as a driving force to rapidly eject a drive housing. 'I
' The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL R0D PROGRAM CONTROLS t ..
, Control rod withdrawal and insertion sequences are established to assure
- .that the maximum insequence individual control rod or control rod segments which
- are withdrawn at any time during the fuel cycle could not be worth enough to
- ~ result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control i
rod drop accident. The specified sequences are characterized by homogeneous, i scattered patterns of control rod withdrawal. When THERMAL-POWER is greater j than 20% of RATED THERMAL' POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RSCS and RWM to be OPERABLE when
- THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.
The RSCS and RWM provide automatic supervision to assure that out-of-l sequence rods will not be withdrawn or inserted. [ The analysis of the rod drop accident is presented in Section 15.4.0 of
. the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References '2 and 3.
l l The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power g operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods. LIMERICK - UNIT 1 B 3/4 1-3 p
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and improper mixing, this concentration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 4,620 gallons of sodium pentaborate , solution containing a minimum of 5,500 lbs of sodium pentaborate. This quan-tity of solution is a net arount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected. The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permis-sible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution. With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. The SLCS system consists of three separate and independent 100% capacity pumps and explosive valves. Only two of the separate and independent pumps and explosive valvec are required to meet the minimum requirements of this technical specification and satisfy the single failure criterion. Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours assures that the solution is available for use. Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges. I i
- 1. C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Anab ~is for Large BWR's," G. E. Topical Report NE00-10527, March 1972.
- 2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NED0-10527, July 1972.
- 3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, " Exposed Cores,"
Supplement 2 to NED0-10527, January 1973. LIMERICK - UNIT 1 B 3/4 1-4 .)
e A ( ) b/ 3/4.2 POWER DISTRIBUTION LIMITS BASES The specif'. cations of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46. 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secon-darily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than,the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting
! value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and G'; 3.2.1-5.
The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) cal-culational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses can be broken down as follows.
- a. Input Changes
- 1. Corrected Vaporization Calculdt f od - Coefficients in Me vaje idtl40 correlation used in the REFLOOD F9th were corrected.
- 2. Incorporated more accurate bypass ain s - The bypass arthl5 ln the top guide were realculated Usi/@ (store aW!rdth technique.
- 3. Corrected guide tube thermal resi6(au m I
- 4. Correct heat C#pacity of f Mrlor internof s heat nodes.
<]
LIML9 fry - UNIT 1 0 3/4 2-1 A H
1 POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)
- b. Model Change
- 1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
- 2. Incoporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL. These changes are listed below.
- a. Input Change
- 1. Break Areas - The DBA break area was calculated more accurately.
- b. Model Change
- 1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation. i A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3/4.2.1-1.
3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a poder distribution -nich would yield the design LHGR at RATED THERMAL POWER. The flow biased n d mn flux-upscale scrain trip setpoint and flow biased neutron flux-upsca. ; trol rod block functions of the APRM instruments must be adjusted to ensure r t the MCPR dces not become less than 1.06 or that > 1% plastic strain does not c cur in the c" graded situation. The scram and rod block setpoints are adjust e in accorca ce with the formula in this sce-cification when the combination of THEPM 0WER and CMFLPD indicates a higher peaked power distribution to ensure thz. 1 LHGR transient would not be increased in the degradec condition. LIMERICK - UNIT 1 '/4 2-2
- n ,
I I a.
~P0WER DISTRIBUTION LIMITS l BASES TABLE B 3/4.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS Plant Parameters:
Core THERMAL POWER .................... 3430 MWt* which corresponds to 105% of rated steam flow , Vessel Steam Output .... .. . . ........ 14.86 x 105 inm/h which cc tesponds. to 105% of rated su =n flow Vessel Steam Dome Pressure. .. 1055 csia Desi a Basis Recirculation Line Break Area for:
- a. Large Break < 4.1 ft 2, 1. ft2
- b. Small Breaks 1.0 ft2, 0.07 ft2 0') ft 2 , 0.02 f t' i cue' Paramet<
u- iLCHNIC;u INIE i~ ,lFICATIC9 L E ! :// MINIM.
- NEAR HEAT 4XIAL CRITIL .
iUEL BUN 0tE A RATION RATE' PEAKING P0wi FU TYPE GECMEisi (kW/'t) FACTOR RAT! Core cx8 _ 13.4 1.4 1.20 A more etailed listin;; of inpu - each model anti its source is presented in Sect n II of Referroce 1 ar: s+ction 15,0.2 of the FSAR.
*This pc4 r level meets the Appe K requirement td 1: < core heatup e alculation assumes a bunc. power asistent with operation of the i ghest powered rod at 102% of < Tect ical Specification LINEAR HEA ;ENERATION RATE limit.
t
~~
LIMERICK - UNIT 1 B 3/4 2-3 a __
l POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients. For any abr.ormal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of
'!cw, inct ease in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCF; When added to the Safety Limit MCPR of 1.06, the required minimum ope ning limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-1.
The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-2 that are input tu a GE-core dynamic beha ;or transient computer rrogrg The code used to evaluate pressurization evenn is cescribed in NEDO-241M and the program used in non pressurization
.ents is cescribed in NED0-10M/g g The outputs of this program along with the initial MCPR fcrm the input for further analyses of the thermally limiting bundlewithtgsicalechannel transient thermal hydraulic TASC code described in NEDE 25149 The principal result of this evaluation is the reduction in MCPR caused by the transient.
The purpose of the fK factor of Figure 3.2.3-2 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and the Kf factor. The K f factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor generatcr speed control failure. The K factors may be applied to both manual and automatic flow control modes. f The K factors values shown in Figure 3.2.3-2 were developed generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The Kf factors were f derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow. For the manual flow control mode, the K7 factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was ad. justed until the MCPR changes with different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the K7. LIMERICK - UNIT 1 B 3/4 2-4 l
s o) G( POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow. , The K, factors shown in Figure 3.2.3-2 are conservative for the General Electric B6iling Water Reactor plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the or,iginal 1.20 operating limit MCPR used for the generic derivation of Kf. At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control. rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement hg for calculating MCPR when THERMAL POWER is greater than or equal to 25% of i (',/ RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The require-ment for calculating MCPR when a limiting control rod pattern'is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit. 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in l any rod is less than the design linear heat generation even if fuel pellet l densification is postulated. 1
References:
- 1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.
- 2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NE00-10802, February 1973.
- 3. Qualification of the One Dimensional Core Transient Moael for Boiling Water Reactors, NED0-2415a, October 1978.
- 4. TASC 01-A Computer Program for the Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.
O LIMERICK - UNIT 1 8 3/4 2-5
[ v
-3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR' PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
- a. Preserve the integrity of the fuel cladding.
- b. Preserve the integrity of the reactor coolant system.
- c. Minimize the energy which must be absorbed following a loss-of-coolant accident, and
- d. Prevent inadvertent criticality.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of main-4
-(
- tenance. When necessary, one channel may be made inoperable for brief intervals
() to conduct required surveillance. The reactor-protection system is made up of two independent trip systems. There are usually _ four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for. nuclear power plant protection systems. The bases for the trip settings of the RPS-are discussed in the bases for Specification 2.2.1. [ The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-pleted within the time. limit. assumed in the safety analyses. No credit was taken 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 measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be
- demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certifited response times.
] f M LIMERICK - UNIT 1 B 3/4 3-1
INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The set-points of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating rar.ge to prevent inadvertent actuation of the systems involved. Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators. In this evenL, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power estab-lishment will establish the response time for the isolation functions. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the de_ign protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specificallly allocated for each trip in the safety analyses. O LIMERICK - UNIT 1 B 3/4 3-2
INSTRUMENTATION v BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDO-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR. The end-of-cycle recirculation pump trip (E0C-RPT) system is a supplement to the reactor trip. During turbine trip and generator load rejection events, the EOC-RPT will reduce the likelihood of reactor vessel level decreasing to level 2. Each E0C-RPT system trips both recirculation pumps, reducing coolant
-flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system. Similarly, h
\
a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPT system. For each E0C-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the E0C-RPT system and trip both recirculation pumps. Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room. The E0C-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e., 175 ms. Included in this time are: the response time of the sensor, the time allotted for breaker arc suppression, and the response time of the system logic. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the dif f-i erence between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. t > A ( [ I C) LIMERICK - UNIT 1 B 3/4 3-3
INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor' isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of the emergency core cooling equipment. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the dif-ference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64. 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring 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 capability is required to permit comparison of the measured response to that used in the design basis for the unit. 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radia-tion doses to the public as a result of routine or accidental release of radio-active 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. This instrumentation is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February,1972. LIMERICK - UNIT 1 B 3/4 3-4
i
, -INSTRUMENTATION BASES METEOROLOGICAL MONITORING INSTRUMENTATION (Continued)
Site data compiled since January 1972 provide correlation between Eleva-tion 1 (Tower 1) and Elevation 1 (Tower 2), and between Elevation 2 (Tower 1) and Elevation 2 (Tower 2). This correlation serves as justification for the use of the appropriate Tower 2 instrument as a back-up to the Tower 1 instrument as shown in Table 3.3.7.3-1. 3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and mainte-nance of HOT SHUTOOWN of the unit 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 10 CFR Part 50, Appendix A. 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant para.neters to monitor and assess important variables following an accident. This capability is con-g
'j sistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the
' status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels,. reactivity additions shall not be made without this flux level information available to the operator. When the inter-mediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.
3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core. - The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e., detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. i V) The OPERABILITY of individual detectors to be used for monitoring is demon-
.strated by comparing the detector (s) output in the resultant heat balance calcu-lation (P-1) with data obtained during a previous heat balance calculation (P-1).
LIMERICK - UNIT 1 B 3/4 3-5
INSTRUMENTATION BASES 3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room p'ersonnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection. Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection. The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95 " Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975. 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge extin-guishing agent in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program. Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification. Consequently, the minimum number of OPERABLE fire detectors must be greater. The loss of detection capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1.133, '? Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. 3/4.3.7.11 RADI0 ACTIVE LIQUID EFFLUENT MONI.TORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures 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 instru-mentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFh Part 50. LIMERICK - UNIT 1 B 3/4 3-6
O INSTRUMENTATION O BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.12 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 effluents during actual or potential releases of gaseous effluents. The alarm / 4 trip setpoints for these instruments shall be calculated in accordance with
; the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mix-
- tures in the off gas system. 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.
3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEf1 This specification is provided to ensure that the turbine overspeed
, ' protection system 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 p damage safety related components, equipment or structures.
I 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP. SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand. i s 4
. v LIMERICK - UNIT 1 B 3/4 3-7 i
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3/4.4 REACTOR COOLANT SYSTEM > t (,/ BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable is prohibited until an evaluation of the performance of the ECCS during one loop operation has been performed, evaluated and determined to be acceptable. An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a pre-scribed schedule for significant degradation. Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In order to prevent undue stress on the vessel nozzles and bottom head region, the reci fcu?ation loop temperatures shall be within 50 F of each other prior to startup otian fdle loop. The loop temperature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper b regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145 F. The objective of GE BWR plant and fuel design is to provide stable opera-tion with margin over the normal operating domain. However, at the high _ power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod pattern, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region. Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1. Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case the degree of conservatism can be reduced since plant to plant variability would be eliminated. In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8. d LIMERICK - UNIT 1 B 3/4 4-1
REACTOR COOLANT SYSTEM BASES RECIRCULATION SYSTEM (Conti.iued) Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores tipically operate with neutron flux noise caused by random boiling and flow noise. TyW1 neutron flux noise levels of 1-12% of rated power (peak-to peak) have .been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence. In addition, stability tests at operating BWRs have demonstrated that when stability related neutron f, lux limit cycle oscillations occur they result in peak-to peak neutron flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations. Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern) with two reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows. To maintain a reasonable variation between the low flow and high flow end of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation. Data from tests and operating plants indicate that a range of 20% y of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops. Baseline data should be taken near the maximum rod line at which the majority of operation will occur. However, baseline data taken at lower rod lines (i.e. lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow. 3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Coce. A total of 11 OPERABLE safety / relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient. Demonstration of the safety / relief valve lift settings will occur only during shutdown. The safety / relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency. O LIMERICK - UNIT 1 8 3/4 4-2
REACTOR COOLANT SYSTEM p] BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided 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. In conformance with Regulatory Guide 1.45, the channel calibration tests will verify the ability to detect a 1 gpm leak in less than 1 hour and an atmospheric gaseous radioactivity system sensitivity of 10 6 pC/cc. 3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also con-sidered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow 7 rapidly. However, in all cases, if the leakage rates exceed the values specified C) or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby r9ducing the probability of gross valve failure and consequent inters'rstem LGCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE a.ad will be considered as a portion of the allowed limit. 3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration.of chlorides is not considered harmful during these periods. Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.
? ) The surveillance requirements provide adequate assurance that concentrations kJ in excess of the limits will be detected in sufficient time to take corrective action.
LIMERICK - UNIT 1 B 3/4 4-3 . j
i REACTOR COOLANT SYSTEM BASES 3/4.4.5 SPECIFIC ACTIVITY The Ifmitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as SITE BOUNDARY location and meteoro-logical conditions, were not considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4 micro-curies per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram DOSE EQ1IVALENT I-131 must be restricted to no more than 800 hours per year, approximately 10% of the unit's yearly operating time, since these activity levels increase the 2-hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam line rupture. The reporting of cumulative operating time over 500 hours in any 6-month consecutive period with greater than 0.2 micro-curie per gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800-hour limit. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained. Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outt.ide containment. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in suf ficient time to take corrective action. 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduccd by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. LIMERICK - UNIT 1 B 3/4 4-4
l h A\ REACTOR COOLANT SYSTEM [ . x.) 1
^
BASES-4 PRESSURE / TEMPERATURE LIMITS (Continued) The operating' limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section III, Appendix G. The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, " Fracture Toughness." The reactor vessel materials have been tested to determine their initial RT NDT. The results of these tests are shown in Table B 3/4.4.6-1. Reactor
- operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, phosphorus content and copper content of the material in question, can be preilicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, " Effects of Residual Elements on Predicted ,
- Radiation Damage to Re' actor Vessel Materials." The pressure / temperature limit curve, Figure 3.4.6.1-1, curves A', B' and C', includes an assumed
. shift in RT NDT f r the end of life fluence. O The actual shift in RT NDTc ,f the vessel material will be established period-ically during operation by removing and evaluating, in accordance with 10 CFR Part 50, Appendix H, irradiated reactor vessel flux ' ire specimens installed 3 near the inside wall of the reactor vessel in the co e area. Since the neutron spectra at the flux wires and vessel inside radius are essentially identical, the irradiated flux wires can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the flux wire data and recommendations of Regulatory Guide 1.99, Revision 1. The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves i C, and C', and A and A', for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing. The. number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are pro-i vided in Table 4.4.6.1.3-1 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. 4 l LIMERICK - UNIT 1 B 3/4 4-5 i
. ~ . - - , . - . - . .-. .-- - - - - - - - _ - - , - . - _ - - ,-- .- - _.- - - , - . ., - ,
REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Deuble isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges. 3/4.4.8 STRUCTURAL INTEGRITY Tl.e inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. 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 1971 Edition and Addenda through Winter 1972. The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed 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 oy the NRC pursuant to 10 CFR 50.55a(g)(6)(i). 3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera- ( ture indication, however, single failure considerati<ms require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation. O LIMERICK - UNIT 1 8 3/4 4-6
\} ,/ '%_l BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS
} MIN. UPPER 5 . HEAT / SLAB HIGHEST STARTING MAX.
- SHELF MAX.
Q BELTLINE WELD SEAM I.D. OR RT LRT RT . COMPONENT OR MAT'l TYPE HEAT / LOT CU (%) P (%) NDT ( F) NDT (*F) (LFT-LBS) NDT ( F) Plate SA-533 Gr B CL.1 C 7677-1 .11 .016 +20 +36 NA +56 Weld SF,^ 5.5, 662A746/ .03 .021 -20 +35 NA +15 (E 80T8-G) H013A27A NOTE:* These values are given only for the benefit of calculating the end-of-life (EOL) RTNDT HIGHEST STARTING NON-BELTLINE Mi'L TYPE OR HEAT / SLAB OR RT COMPONENT WELD STEAM I.D. HEAT / LOT NDT ( F) Shell Ring SA 533, Gr. B, CL. 1 C7711-1 +20
" +12 C'
Bottom Head Dome C7973-1 t) Bottom Head Torus " C7973-1 +12
- Top Head Dome A6834-1 +10
? Top Head Torus " B1993-1 410 " Top Head Flange SA-508, CL. 2 123B195-289 0 Vessel Flange " 2V1924-302 -30 Feedwater Nozile " Q2Q22W-412 -10 Weld Non-Beltline All 0 LPCI Nozzle
- SA-508, CL. 2 Q2Q25W -6 Closure Studs SA-540, Gr. B-24 All Meet requirements of 45 ft-lbs and 25 mils Lat. Exp. at +10 F
- The design of the LPCI nozzles results in their experiencing an E0L fluence in excess of 1017 N/Cm2 which predicts an EOL RTNDT f +14 F.
l REACTOR COOLANT SYSTEM i 1.2 S 1.0 X , k 1
- 0.8 2 A
E
~
E
$ 0.6 7
e O g 0.4 b 5 0.2 - o I 10 20 30 40 Service Life (Years *) FAST NEUTRON Fli;ENCE (E>l MeV) AT \ T AS A FUNCTION OF SERVICE LIFE
- BASES FIGURE B 3/4.4.6-1
- At 90% of RATED THERMAL POWER and S0% availability.
O LIMERICK - UNIT 1 B 3/4 4-8
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3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN The core spray system (CSS), together with the LPCI mode of the RHR system, is provided to assure that the core is adequately cooled following a loss-of- l coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS. The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized trA a source for flooding of the core in case of accidental draining. The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment. The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS. The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment. The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HCPI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling. The capacity of the system is selected to provide the required core cooling. The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactor pressures between 1141 and 200 psig. Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool.into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water. LIMERICK - UNIT 1 B 3/4 5-1
EMERGENCY CORE COOLING SYSTEM BASES ECCS - OPERATING and SHUTOOWN (Continued) With the HPCI system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the CS and LPCI systems. In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems and the RCIC system. The surveillance requirements provide adequate assurance that the HPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment. Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automa-tically causes selected safety / relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200 F. ADS is conserva-tively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig. This pressure is substantially below that for which the low pressure core cool-ing systems can provide adequate core cooling for events requiring ADS. ADS automatically controls five selected safety-relief valves although the safety analysis only takes credit for four valves. It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability. 3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCI, CS and LPCI systems in the event of a LOCA. This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber in OPERATIONAL CONDITION 1, 2, or 3 is also required by Specification 3.6.2.1. Repair work might require making the suppression chamber inoperable. This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5. In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200 F. Since pressure suppression is not required below 212 F, the minimum water volume is based on NPSH, recirculation volume and vortex prevention plus b a safety margin for conservatism. LIMERICK - UNIT 1 B 3/4 5-2
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the. release of radioactive mate-rials 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 limits of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE , The limitations on primary 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 of 44.02 psig, P . As an added conserva-tism, the measured overall integrated leakage rate is f6rther limited to less during performance of the periodic tests to account for than or equal possible to 0.75 of degradation L,the containment leakage barriers between leakage tests. Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore ti'.e special requirement for testing these valves. i The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for leak te: ting of the main steam isolation valves, the airlock, TIP shear valves and RhR relief valves. 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2. The specification makes allowances for the fact that there may be l ! long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock j is required to maintain the integrity of the containment. ! 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main j steamline isolation valves in the postulated LOCA situations wuld be a small l fraction of the 10 CFR Part 100 guidelines, provided the main steam line system l from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage c requirements have not always been maintained continuously. The requirement for the leakage control system will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required. LIMERICK - UNIT 1 B 3/4 6-1
4 CONTAINMENT SYSTEMS BASES , 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is' required to ensure that the cc.1tainment will withstand the maximum pressure of 44.02 psig in the event of a LOCA. A visual inspection in conjucction with Type A leakage tests is sufficient to demonstrate i this capability. l 3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 44.02 psig does not exceed the design pressure of 55 psig during LOCA conditions or that the external pressure differ-ential does not exceed the design maximum external pressure differential of
- 5.0 psid. The limit of - 1.0 to + 2.0 psig for initial positive containment i
pressure will limit the total pressure to 44.02 psig which is less than the design pressure and is consistent with the safety analysis. 3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the con-
- tainment peak air temperature does not exceed the design temperature of 340 F l during steam line break conditions and is consistent with the safety analysis.
3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM l The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, deinerting and pressure control. The 90 hours per 365 day limit on purge valve operation is imposed to protect the integrity of the SGTS filters. Analysis indicates that shoulc a LOCA occur while this pathway is being utilized, i the associated pressure surge through the (18 or 24") purge lines will adversely , affect the integrity of SGTS. This limit is not imposed, however, on the subject valves when pressure control is being performed through the 2-inch bypass line, since a pressure surge through this line does not threaten the OPERABILITY of j SGTS. I 4 1 i . ) i LIMERICK - UNIT 1 8 3/4 6-2 i
. , , - _ _ _ . . _ . . . - -__m._....._.-..m.., . _ _ .._ _.,___ ,_ . . _ , , _ , , , , _ . . . . _ _ _ , _ . . - . - . _ , _ - , . - - . _ , , ,._,._m _. ., . , , . _ , , -,m ,.
CONTAINMENT SYSTEMS BASES 3/4.6.2. OEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary system blowdown from full operating pressure. The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1040 psig. Since all of the gases in the drywell are purged into the suppres-sion chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged tothe suppression chamber. Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is approximately 30 psig which is below the design pressure of 55 psig. Maximum water volume of 134,600 ft3 results in a downcomer submergence of 12'3" and the minimum volume of 122,120 fta results in a submergence approximately 2'3" less. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170 F. Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3. Under full power operating conditions, blowdown through safety / relief valves assuming an initial suppression chamber water temperature of 95 F results in a bulk water temperature of approximately 136*F immediately following blowdown which is below the 190 F bulk temperature limit used for complete condensation via T quencher devices. At this temperature and atmospheric pressure, the avail-able NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations. Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 200*F during any period of relief valve operation for T quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings. LIMERICK - UNIT 1 B 3/4 6-3
CONTAINMENT SYSTEMS I BASES i DEPRESSURIZATION SYSTEMS (Continued) Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. . In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppres-sion pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety / relief valve to assure mixing and uniformity of energy insertion to the pool. 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABILITY of the primary 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 GDC 54 through 57 of Appendix A of 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 environ-ment will be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 VACUUM RELIEF l l' Vacuum relief valves are provided to equalize the pressure between the suppression chamber and drywell. This system will maintain the structural integrity of the primary containment under conditions of large differential pressures. l The vacuum breakers bett;een the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the i suppression pool in case of an accident. There are four pairs of valves to l provide redundancy so that operation may continue for up to 72 hours with no i more than one pair of vacuum breakers inoperable in the closed position. ( Each vacuum breaker valve's position indication system is of great enough j sensitivity to ensure that the maximum steam bypass leakage coefficient of ! b l A 8 = 0.05 ft 2 l for the vacuum relief system (assuming one valve fully open) will not be exceeded. LIMERICK - UNIT 1 B 3/4 6-4 l
CONTAINMENT SYSTEMS BASES 3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containinent during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified. Establishing and maintaining a vacuum in the reactor enclosure secondary containment with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment. The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA or refueling accident (SGTS only). The reduction in containment iodb e inventory reduces the resulting SITE B0UNDARY radiation doses associated rii.n containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA and refueling accident analyses. Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters. Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 135 seconds, these surveillance require-ments specify a draw down time of 121 seconds. This 14 second difference is due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement. The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 2800 cfm. The surveillance requirements satisfy this assumption by starting the drawdown from ambient conditions and at the SGTS rated flow rate of 3000 cfm. i j 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL i The OPERABILITY of the systems required for the detection and conf.rol of hydrogen combustible mixtures of hydrogen and oxygen ensures that these systems will be available to maintain the hydrogen concentration within the primary containment below the lower flammability limit during post-LOCA conditions. The primary containment hydrogen recombiner is provided to maintain the oxygen concentration below the lower flammability 1.imit. The combustible gas analyzer is provided to continuously monitor, both during normal operations and post-LOCA, ( the hydrogen and oxygen concentrations in the primary containment. The primary containment atmospheric mixing system is provided to ensure adequate mixing of i the containment atmosphere to prevent localized accumulations of hydrogen and l oxygen from exceeding the lower flammability limit. The hydrogen control l system is consistent with the recommendations of Regulatory Guide 1.7, " Control l of Combustible Gas Concentrations in Containment Following a LOCA," March 1971. I LIMERICK - UNIT 1 B 3/4 6-5
l l 3/4.7 PLANT SYSTEMS BASES 3/4.7.2 SERVICE WATER SYSTEMS The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits. 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM The OPERABILITY of the control room emergen'y c fresh air supply system ensures that the control room will remain habitable for operations personnel during and following all design basis accident conditions. Constant purge of the system at 1 cfm is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. .This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50. 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM ( The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event'of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling system equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure ex-ceeds 150 psig. This pressure is substantially below that for which low pressure core cooling systems can provide adequate core cooling. The RCIC system specifications are applicable during OPERATIONAL CONDITIONS , 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the l primary non-ECCS source of emergency core cooling when the reactor is
- j. pressurized.
With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCI system and justifies the specified 14 day out-of-service
- j. period.
The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping
, is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment. -
i l LIMERICK - UNIT 1 B 3/4 7-1
PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snub-bers excluded from this inspection program 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. 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 Plant Operations Review Committee. 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 snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50. The visual inspection frequency is based upon maintaining a constant level of snubber protection to each safety-related system. Therefore, the required inspection interval varies inversely with the observed snubber failures on a i given system and is determined by the number of inoperable snubbers found during !. an inspection of each system. In order to establish the inspec. tion frequency l for each type of snubber on a safety-related system, it was assumed that the , frequency of snubber failures and initiating events is constant with time and that the failure of any snubber on that system could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has i elapsed (nominal time less 25%) may not be used to lengthen the required inspec-l tion interval. Any inspection whose results required a shorter inspection ! interval will override the previous schedule. The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers. L l LIMERICK - UNIT 1 B 3/4 7-2
PLANT SYSTEMS BASES SNUBBERS (Continued) To provide assurance of snubber functional reliability one of three functional testing methods is used with the stated acceptance criteria:
- 1. Functionally test 10% of a type of snubber with an additional 10%
tested for each functional testing failure, or
- 2. Functionally test a sample size and de'termine sample acceptance or reiection using Figure 4.7.4-1, or
- 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.
Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in Quality Control and Industrial Statistics" by Acheson J. Duncan. 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 exempticns. The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and asso-ciated installation and maintenance records (i.e., newly installed snubber, seal 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 statis-tical bases for future consideration of snubber service life. 3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plu-tonium. 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 requirements 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 devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism. a LIMERICK - UNIT 1 B 3/4 7-3
PLANT SYSTEMS BASES 3/4 7.6 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equi,pment is located. The fire suppression system consists of the water system, spray and/or sprinkler systems, CO 2 systems, Halon systems, and fire hose stations. The collective capability of the fire suppression systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program. In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the in-operable fire fighting equipment is intended for use as a backup means of fire supp-rassion, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. The surveillance requirements provide assurances that the minimum OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying the w2ight and pressure of the tanks. The source of water for the fire protection system is two cooling tower basins that have a capacity of 7,200,000 gallons each, for a total capacity of 14 r400,000 gallons. For a system pumping capacity of 5000 gpm, this allows continuous operation of both fire pumps for 48 hours. If one cooling tower basin or supply line is not available, the remaining water source provides both fire pumps with a 24-hour supply of water. Water for the fire pumps is taken from oither Unit 1 or Unit 2 cooling tower water basins through connections to the circulating water lines. One cooling tower will be out of service for up to 30 days each refueling outage on each unit, to remove the accumulated mud deposits. The ninimum contained volume of 311,000 gallons is based on the CMEB BTP 9.5-1 rsquirement of 500 gpm for manual hose streams plus the largest design demand of any sprinkler or deluge system for a period of 2 hours. The largest plant sprinkler system flow is 2090 gpm for the turbine condenser compartment. The minimum fuel supply of 330 gallons for the diesel driven fire pump is based on providing fuel for 24 hours of fall load operation. In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. 3/4.7.7 FIRE RATED ASSEMBLIES . The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited. These design features minimize the possibility g of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodic-ally inspected to verify their OPERABILITY. LIMERICK - UNIT 1 B 3/4 7-4
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 SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution 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 distribution 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 are con-sistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least two of the onsite A.C. and the corresponding D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. or D.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 two diesel generators are inoperable, there is an additional ACTION requirement to verify that all required systems, susbsystems, trains, components, and devices, that depend on the remaining OPERABLE diesel generators as a source of emergency power, are also OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete 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 administrative 1y check by examining logs or other information to determine 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 minimum 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 Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977 and Regulatory Guide 1.137 " Fuel-Oil Systems for Standby Diesel Generators," Revision 1, October 1979. LIMERICK - UNIT 1 B 3/4 8-1
ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued) The surveillance requirements for demonstrating the OPERABILITY of the unit batteries are in accordance with 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 4.8.2.1-1 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 4.8.2.1-1 is permitted for up to 7 days. During this 7-day period: (1) the allowable value 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 c 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 l cell's float voltage, greater than 2.07 volts, ensures the battery's capa-l bility to perform its design function. 6 LIMERICK - UNIT 1 B 3/4 8-2
ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary containment electrical penetrations and penetration conductors are protected by either de-energizing circuits not required during reactor operation or demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers by periodic surveillance. The surveillance requirements applicable to lower voltage circuit breakers provides assurance of breaker reliability by testing at least one representative sample of each manufacturers brand of circuit br'eaker. Each manufacturer's molded case circuit breakers are grouped into representative samples which are than tested on a rotating basis to ensure that all breakers are tested. The bypassing of the motor operated valves thermal overload protection continuously.by integral bypass devices ensures that the thermal overload pro-tection will not prevent safety related valves from performing their function. The Surveillance Requirements for demonstrating the bypassing of the thermal overload protection continuc" sly are met by functionally testing the automatic operation of the motor operated valve and ensuring that the motor thermal overload protection design does not change and is in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves", Revision 1, March 1977. LIMERICK - UNIT 1 8 3/4 8-3
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 REACTOR MODE SWITCH Locking the OPERABLE reactor mode switch in the Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal and refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor inte'rnals or fuel assemblies, and exposure of personnel to excessive radioactivity. 3/4.9.2 INSTRUMENTATION The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. 3/4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during other CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control rod. 3/4.9.4 CECAY TIME The minimum requirement for reactor subcriticality prior to fuel movement ensures that sufficient time has elapsea to allos the radioactive decay of the short lived fission products. This decay time is consistent with the assump-tions used in the accident analyses. 3/4.9.5 COMMUNICATIONS The recuirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel. LIMERICK - UNIT 1 B 3/4 9-1
REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations, and (4) inadvertent criticality will not occur due to fuel being loaded into a unrodded cell. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL - The restriction on movement of loads in excess of the nominal weight of a fuel assembly and associated lifting device 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 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL STORAGE P0OL 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. This minimum water depth is consistent with the assumptions of the accident analysis. 3/4.9.10 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn. 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1) suf-ficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during REFUELING, and 2) sufficient coolant circulation would be available through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system. The requirement to have two shutdown cooling mode loops OPERABLE,when there is less than 22 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of resid-ual heat removal capability. With the reactor vessel head removed and 22 feet h of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core. LIMERICK - UNIT 1 B 3/4 9-2
3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS. , 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints on control rod movement. The additional surveillance requirments ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis. 3/4.10.3 SHUTDOWN M RGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed i requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO. 3/4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels. 3/4.10.5 OXYGEN CONCENTRATION Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase of operation. Without this access the startup and test program could be restricted and delayed. 3/4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system. 4 LIMERICK - UNIT 1 B 3/4 10-1
3/4.11 RADI0 ACTIVE EFFLUENTS 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 to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limits for dissolved or entrained noble gases are based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protection (ICRP) Publication 2. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedules Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detec-tion and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Opera-tion implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time imple-ment the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the require-ments in Section III.A of Appendix I that conformance with the guides of Appendix 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 equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid efflu-ents are consistent with the methodology provided in Regulatory 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 1.113, " Estimating Aquatic Dispersion 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 reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system. LIMERICK - UNIT 1 B 3/4 11-1
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implcments the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appro-priate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. , 3/4/11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification 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 specified tanks provides assurance that in the event of an uncontrolled release of the tanks' 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 dose associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in 'g~aseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR 20.106(b)(1)). 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 sufficiently low to com-pensate 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 BOUNDARY to less than or equal to 500 mrems/ year to the total 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.
This specification applies to the release of radioactive materials in gaseous effluents from all reactors at the site. I LIMERICK - UNIT 1 B 3/4 11-2
r RADI0 ACTIVE EFFLUENTS BASES DOSE RATE (Continued) i The required detection capability for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., "Detec-tion Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.2.2 DOSE - N0BLE GASES This specification is provided to implement the requirements of Sections II.8, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix 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 gaseous etfluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on I models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calcula-tion of Annual Doses to Man from Routine Releases cf 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 with site specific dispersion curves and deposition methodology. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions. 3/4.11.2.3 DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIONUCLIDES IN l PARTICULATE FORM r This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions 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 time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix 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 LIMERICK - UNIT 1 B 3/4 11-3
I RADI0 ACTIVE EFFLUENTS BASES DOSE-IODINE-131, 10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM (Continued) unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory 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 with site specific dispersion curves and deposition methodology. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man in areas at and beyond the SITE B0UNDARY. The pathways which were examined in the develop-ment of these calculations were: (1) individual inhalation of airborne radio-nuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man. 3/4.11.2.4 VENTILATION EXHAUST TREATMENT SYSTEM The requirement that the appropriate portions of this system 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 II.D 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, 10 CFR Part 50, for gaseous effluents. 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the main condenser offgas treat-ment system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen below its flammability limit 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. b l LIMERICK - UNIT 1 B 3/4 11-4
RADI0 ACTIVE EFFLUENTS l BASES 3/4.11.2.6 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50. ' 3/4.11.2.7 VENTING OP PURGING This specification provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for UNRESTRICTED AREAS. 3/4.11.3 SOLID RADWASTE TREATMENT The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. This specification implements the requirements i 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, waste pH, waste / liquid /solidifi-cation agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing tin,es. 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 CFR 18525. The specification requires the preparation and submittal of a Special Report whenever the calcu-lated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. 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 exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles 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 conditions 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 considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. 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. LIMERICK - UNIT 1 8 3/4 11-5
_ _ - _ - - - - - . _ ~ . 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring progrgm required by this specifica-tion provides representative measurements of radiation and of radioactive
, materials in those exposure pathways and for those radionuclides that lead to 3
the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting i from the station operation. This monitoring program implements Section IV.B.2 i of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent , monitoring program by verifying that the measurable concentrations of radioactive j materials and levels of radiation are not higher than expected on the basis of l the effluent measurements and modeling of the environmental exposure pathways. j 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 (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measurements i in industrial laboratories. It should be recognized that the LLD is defined i as an a priori (before the fact) limit representing the capability of a measure-
- ment system and not as an a posteriori (after the fact) Ifmit for a particular l
t measurement. I Detailed discussion of the LLD, and other detection limits, can be found ! i in HASL Procedures Manual, HASL-300 (revised annually); Currie, L. A., " Limits ; for Qualitative Detection and Quantitative Determination - Application to ; Radiochemistry" Anal. Chem. 40, 586-93 (1968); and Hartwell, J. K., " Detection j Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford F j Company Report ARH-SA-215 (June 1975). ! 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas I, at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census. The best information from the door-to-door survey, aerial survey or consulting i with local agricultural authorities or any combination of these methods shall
- be used. This census satisfies the requirements of Section IV.8.3 of Appendix I
! to 10 CFR Part 50. Restricting the census to gardens of greater than 500 ( square feet provides assurance that significant exposure pathways via leafy i 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. To determine
- this minimum garden size, the following assumptions were used
- (1) that 20% of
- the garden was used for growing broad leaf vegetation (i.e., similar to lettuce
{ and cabbage), and (2) a vegetation yield of 2 kg/ square meter. 1 LIMERICK - UNIT 1 8 3/4 12-1 . f 4
RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an 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 environmental monitoring in order to demonstrate that the results are reasonably valid for the purpose of Section IV.B.2 of Appendix I to 10 CFR Part 50. f I LIMERICK - UNIT 1 B 3/4 12-2
0 SECTION 5.0 DESIGN FEATURES
5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shcwn in Figure 5.1.1-1. LOW POPULATION ZONE 5.1. 2 The low population zone shall be as sho.wh in Figure 5.1.2-1. MAPS DEFINING UNRESTRICTED AREAS AND SIfE BOUNDARY FOR RADIOACTIVE GASEOUS 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 defini-tion of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBER OF THE PUBLIC, shall be as shown in Figures 5.1.3-la and 5.1.3-lb. METEOROLOGICAL TOWER LOCATION 5.1. 4 The meteorological towers shall be located as shown on Figure 5.1.4-1. i 5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel lined reinforced concrete structure consisting of a drywell and suppression chamber. The drywell is a steel-lined reinforced concrete vessel in the shape of a truncated cone on top of a water filled suppression chamber and is separated by a diaphragm slab and connected to the suppression chamber through a series of downcomer vents. The drywell has a maximum free air volume of 243,580 cubic feet at a minimum suppression pool level of 22 feet. The suppression chamber has a maximum air region of 159,540 cubic feet and a minimum water region of 122,120 cubic feet. DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:
- a. Maximum internal pressure 55 psig.
- b. Maximum internal temperature: drywell 340 F. .
suppression pool 220 F.
- c. Maximum external pressure 5 psig,
- d. Maximum floor differential pressure: 30 psid, downward.
20 psid, upward. LIMERICK - UNIT 1 5-1
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_'., - . {0f,ER , ,, c FIGURE 5.1.4-1 h METEOROLOGICAL TOWER LOCATION l LIMERICK - UNIT 1 5-6
DESIGN FEATURES SECONDARY CONTAINMENT 5.2.3 The secondary contain.?ent consists of three distinct isolatable zones. Zones I and II are the Unit 1 and Unit 2 reactor enclosures respectively. Zone III is the common refueling area. Each zone has an independent normal ventilation system which is capable of providing secondary containment zone isolation as required. Each reactor enclosure (Zone I or II) completely encloses and provides secondary containment for its corresponding primary containment and reactor auxiliary or service equipment, and has a minimum free volume of 1,800,000 cubic feet. The common refueling area (Zone III) completely encloses and provides secondary containment for the refueling servicing equipment and spent fuel storage facilities for Units 1 and 2, and has a minimum free volume of 2,200,000 cubic feet. 5.3 REACTOR CORE FUEL ASSEMBLIES
- 5. 3.1 The reactor core shall contain 764 fuel assemblies with each fuel assembly containing 62 fuel rods and two water rods clad with Zircaloy-2.
Each fuel rod shall have a nominal active fuel length of 150 inches. The initial core loading shall have a maximum average enrichment of 1.90 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading. CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing 143 inches i of boron carbide, B 4C, powder surrounded by a cruciform shaped stainless steel sheath. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 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 applicable Surveillance Requirements,
! LIMERICK - UNIT 1 5-7
DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE (Continued)
- b. For a pressure of:
- 1. 1250 psig on the suction side of the recirculation pump.
- 2. 1500 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.
- 3. 1500 psig from the discharge shutoff valve to the jet pumps.
- c. For a temperature of 575'F.
VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal steam dome saturation temperature of 547 F. 5.5 FUEL STORAGE CRITICALITY 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. A k,ff equivalent to less than or equal to 0.95 when flooded with unborated water, including all calculational uncertainties and biases as described in Section 9.1.2 of the FSAR.
- b. A nominal 6.625 inch center-to-center distance between fuel assemblies placed in the storage racks.
5.5.1.2 The k,ff for 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 5.5.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 346'0". CAPACITY 5.5.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2040 fuel assemblies. 5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.6.1 The components identified in Table 5.6.1-1 are designed and shall be b maintained within the cyclic or transient limits of Table 5.6.1-1. s LIMERICK - UNIT 1 5-8
C E TABLE 5.6.1-1 5 Q COMPONENT CYCLIC OR TRANSIENT LIMITS CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT w Reactor 120 heatup and cooldown cycles 70*F to 560 F to 70*F 80 step change cycles Loss of feedwater heaters 180 reactor trip cycles 100% to 0% of RATED THERMAL POWER 130 hydrostatic pressure and Pressurized to > 930 and leak tests < 1250 psig T D 9 a
3 SECTION 6.0 ADMINISTRATIVE CONTROLS l
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Superintendent shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1. 2 The Shift Supervisor, 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-Electric
~
Production shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION 0FFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown on Figure 6.2.1-1. UNIT STAFF 6.2.2 The unit organization shall be as shown on Figure 6.2.2-1 and:
- a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1;
- b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2, or 3, 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 reactor;
- d. ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
- e. A site fire brigade of at least five members shall be maintained on site at all times *. The fire brigade shall not include the Shift Superintendent, the Shift Technical Advisor, nor the two other members of the minimum shift crew necessary for safe shutdown of the I unit and any personnel required for other essential functions during l a fire emergency; and
*The Health Physics Technician and fire brigade composition may,be less than l the minimum requirements for a period of time not to exceed 2 hours, in order to accommodate unexpected absence, provided immediate action is taken to fill l the required positions.
LIMERICK - UNIT 1 6-1
ADMINISTRATIVE CONTROLS UNI'T STAFF (continued)
- f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions (e.g. , licensed Senior Operators, licensed Operators, health physi-cists, auxiliary operators, and key maintenance personnel).
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shut-4 down for refueling, major maintenance, or major unit modifications, on a temporary basis. the following guidelines shall be followed:
- 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
- 2. An individual should not be permitted to work more than 16 hours 4
in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period, all excluding shift turnover time.
- 3. A break of at least 8 hours should be allowed between work periods, including shift turnover time.
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the Station Superintendent or his deputy, (for operating personnel) or the Superintendent - Maintenance Division or the Engineer-In-Charge, Station Testing Section or their designees (for key maintenance personnel), or higher levels of management, in accordance with established proce-dures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Station Superintendent, or the Superintendent - Maintenance Division or Engineer In-Charge, Station Testing Section or their designees to assure that excessive hours i_ have not been assigned. Routine deviation from the above guidelines is not authorized. LIMERICK - UNI? 1 6-2
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TABLE 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMON CONTROL ROOM WITH UNIT (2) IN CONDITION 4 OR 5 OR DEFUELED POSITION NUMBER OF INDIVIOUALS REQUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SS 1* 1* SRO 1 1* R0 2 1 NLO 2 > 2** STA 1 None WITH UNIT (2) IN CONDITION 1, 2, OR 3 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SS 1* 1* SRO 1* 1* R0 2** 1 NLO 2** 1 STA 1* None TABLE NOTATIONS l
- Individual may fill the same position on Unit 2.
**0ne of the two required individuals may fill the same position on Unit 2.
SS - Shift Superintendent or Shift Supervisor with a Senior Operator license ! on Unit 1. l SR0 - Individual with a Senior Operator license on Unit 1. l RO - Individual with an Operator license on Unit 1. l NLO - Non-licensed operator properly qualified to support the unit to which h assigned. STA - Shift Technical Advisor Except for Shift Supervision (SS), the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 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 compo-sition to within the minimum requirements of Table 6.2.2-1. This provision i does not permit any shift crew position to be unmanned upon shift change due I to an oncoming shift crewman being late or absent. During any absence of Shift Supervision (SS) from the control r,oom while the unit is in OPERATIONAL CONDITION 1, 2, or 3, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of Shift Supervision from the control room while the unit is in OPERATIONAL CONDITION 4 l or 5, an individual with a valid Senior Operator license or Operator license i shall be designated to assume the control room command function. , LIMERICK - UNIT 1 6-5 l
ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) FUNCTION 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience information, including units of simi-lar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to the Engineer-In-Charge, Nuclear Safety Section. COMPOSITION 6.2.3.2 The Limerick ISEG shall be composed of at least three, dedicated, full-time engineers, including the ISEG Supervisor, located onsite. Each shall have a bachelor's degree in engineering or related science and at least two years professional level experience in his or her field. The Limerick ISEG Supervisor shall have at least six years of experience in the nuclear field. The corporate ISEG shall be composed of two dedicated full time engineers each with a Bachelors degree in engineering or related science and at least 2 years professional level experience in his or her field, at least 1 year of which experience shall be in the nuclear field. The Corporate ISEG Supervisor is the Engineer-In-Charge, Nuclear Safety Section. RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification
- that these activities are per-formed correctly and that human errors are reduced as much as practical.
RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to the Engineer-In-Charge, Nuclear Safety Section. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to Shift Supervision in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineer-ing discipline and shall have received specific training in the response and anal-ysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room. 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI /ANS 3.1-1978 for comparable positions, except for the Senior Health Physicist who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The licensed Operators and Senior Operators shall als'o meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees. b
- Not responsible for sign-off function.
LIMERICK - UNIT 1 6-6 , w u w- - , . - - ~ Y Q ao
m ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Superintendent, Nuclear Training Section, shall meet or exceed the requirements of ANSI /ANS 3.1-1978 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience. 6.5 REVIEW AND AUDIT . 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) FUNCTION 6.5.1.1 The PORC shall function to advise the Station Superintendent on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The PORC shall be composed of the: Chairman: Station Superintendent Member: Engineer - Technical Member: Engineer - Operations Member: Engineer - Maintenance Member: Senior Health Physicist Member: I & C Engineer Member: Reactor Engineer Member: Shift Superintendent Member: Regulatory Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities at any one time. MEETING FREQUENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate. QUORUM 6.E 1.5 The quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates. LIMERICK - UNIT 1 6-7
i I t ADMINISTRATIVE CONTROLS I RESPONSIBILITIES l l 6.5.1.6 The PORC shall be responsible for: l a. Review of (1) all procedures required by Specification 6.8 and changes i thereto, (2) all programs required by Specification 6.8 and changes l thereto, and (3) any other procedures or changes thereto as determined l by the Station Superintendent to affect nyclear safety; ! b. Review of all proposed tests and experiments that affect nuclear safety; l c. Review of all proposed changes to Appendix A Technical Specifications;
- d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety; l e. Review of the safety evaluations for procedures and changes thereto l completed under the provisions of 10 CFR 50.59.
j f. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence, to the Superintendent - ( Nuclear Generation Division and to the Nuclear Review Board;
- g. Review of all REPORTABLE EVENTS; l h. Review of unit operations to detect potential hazards to nuclear safety;
! i. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Station Superintendent or the Nuclear i Review Board; l j. Review of the Security Plan and implementing procedures and submittal ! of recommended changes to the Nuclear Review Board; and
- k. Review of the Emergency Plan and implementing procedures and submittal of the recommended changes to the Nuclear Review Board.
i 1. Review of every unplanned onsite release of radioactive material to l the environs including the preparation and forwarding of reports cover-l ing evaluation, recommendations and disposition of the corrective ( action to prevent recurrence to the Superintendent - Nuclear l Generation Division and to the Chairman of the Nuclear Review Board.
- m. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.
6.5.1.7 The PORC shall: i
- a. Recommend in writing to the Station Superintendent approval or dis-approval of items considered under Specification 6.5.1.6a. through
! d. prior to their implementation.
- b. Render determinations in writing with regard to whether or not each
' item considered under Specification 6.5.1.6a. through f. constitutes an unreviewed safety question. l I l l l l LIMERICK - UNIT 1 6-8 1
w ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued)
- c. Provide written notification within 24 hours to the Superintendent -
Nuclear Generation Division and the Nuclear Review Board of disagree-ment between the PORC and the Station Superintendent; however, the Station Superintendent shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1. RECORDS 6.5.1.8 The PORC shall maintain written minute,s of each PORC meeting that, at a minimum, document the results of all PORC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Superintendent - Nuclear Generation Division and the Nuclear Review Board. 6.5.2 NUCLEAR REVIEW BOARD (NRB) FUNCTION 6.5.2.1 The NRB 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 NRB shall report to and advise the Vice President - Electric Production on those areas of responsibility in Specifications 6.5.2.7 and 6.5.2.8. COMPOSITION 6.5.2.2 The Chairman, members, and alternates of the NRB shall be appointed in writing by the Vice President - Electric Production, and shall have an academic degree in an engineering or physical science field; and in addition, shall have a minimum of 5 years technical experience, of which a minimum of 3 years shall be in one or more areas given in Specification 6.5.2.1. The NRB shall be composed of no less than eight and no more than 12 members. l The members and alternates of the NRB will be competent in the area of Quality Assurance practice and cognizant of the Quality Assurance requirements of 10 CFR l Part 50, Appendix B. Additionally, they will be cognizant of the corporate l Quality Assurance Program and will have the corporate Quality Assurance organization available to them. l LIMERICK - UNIT 1 6-9
ADMINISTRATIVE CONTROLS ALTERNATES 6.5.2.3 All alternates shall be appointed in writing by the NRB Chairman to serve on a continuing basis. They shall receive correspondence sent to NRB members with regard to NRB activities and shall be invited to attend all NRB meetings. Alternates shall vote only in the absence of those members for whom they are the alternate. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NRB Chairman to provide expert advice to the NRB. < MEETING FREQUENCY 6.5.2.5 The NRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter. QUORUM 6.5.2.6 The quorum 6f the NRB necessary for the performance of the NRB review and audit functions of these Technical Specifications shall consist of the Chairman or a designated alternate and at least four but not less than one half of the voting NRB members. No more than a minority of the quorum shall have line responsibility for operation of the facility. REVIEW 6.5.2.7 The NRB shall review:
- a. The safety evaluations for (1) changes to procedures, equipment, facilities 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; I- b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10 CFR 50.59;
- c. Proposed tests or experiments which involve an unreviewed safety l question as defined in 10 CFR 50.59; l
- d. Proposed changes to Technical Specifications or this Operating l License; i
- e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructi,ons having nuclear safety significance; j f. Significant operating abnormalities or deviations from normal and I expected performance of unit equipment that affect nuclear safety;
- g. All REPORTABLE EVENTS; LIMERICK - UNIT 1 6-10
ADMINISTRATIVE CONTROLS REVIEW (Continued)
- 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
- 1. Reports and meeting minutes of the PORC.
AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the NRB. These audits shall encompass:
- a. The conformance of unit operation 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 unit staff at least once per 12 months;
- c. The results of actions taken to correct deficiencies occurring in unit 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 Emergency Plan and implementing procedures at least once per 12 months.
- f. The Security Plan and implementing procedures at least once per 12 months.
- g. Any other area of unit operation considered appropriate by the NRB or the Vice President - Electric Production.
I F
- h. The Fire Protection Program and implementing procedures at least once l per 24 months.
l
- 1. An independent fire protection and loss prevention inspection and l audit shall be performed at least once per 12 months utilizing either l qualified offsite licensee personnel or an outside fire protection firm.
- j. An inspection and audit of the fire protection and loss prevention l
program shall be performed by an outside qualified fire consultant at intervals no greater than 36 months. l LIMERICK - UNIT 1 6-11 l
ADMINISTRATIVE CONTROLS AUDITS (Continued)
- k. The radiological environmental monitoring program and the results thereof at least once per 12 months.
- 1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.
- m. The PROCESS CONTROL PROGRAM and implementing procedures at least once per 24 months.
- n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December, 1977, at least once per 12 months. >
RECORDS 6.5.2.9 Records of NRB activities shall be prepared, approved, and distributed as indicated below:
- a. Minutes of each NRB meeting shall be prepared, approved, and forwarded to the Vice President - Electric Production within 14 days following each meeting.
- b. Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President - Electric Production within 14 days following completion of the review.
- c. Audit reports encompassed by Specification 6.5.2.8 shall be forwarded to the Vice President - Electric Production and to the management positions responsible for the areas audited within 30 days after com-pletion of the audit by the auditing organization.
6.6 REPORTABLE EVENT ACTION 6.6.1 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 PORC and submitted to the NRB and the Superintendent - Nuclear Generation Division.
- 6. 7 SAFETY LIMIT VIOLATION 6.7.1 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' - Electric Production and the NRB shall be notified within 24 hours.
b LIMERICK - UNIT 1 6-12
ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued)
- b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the NRB. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon unit 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 NRB, and the Vice President - Electric Production within 14 days of the violation.
- d. Critical operation of the unit shall not be resumed until authorized by the Commission. >
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
- a. The appiicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.
- b. The applicable procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737,
- c. Refueling operations.
I
- d. Surveillance and test activities of safety-related equipment.
- e. Security Plan implementation.
- f. Emergency Plan implementation.
- g. Fire Protection Program implementation.
- h. PROCESS CONTROL PROGRAM implementation.
- i. OFFSITE DOSE CALCULATION MANUAL implementation.
- j. Quality Assurance Program for effluent and environmental monitoring, l using the guidance of Regulatory Guide 4.15, February 1979.
6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed in accordance with Specification 6.5.1.6 and shall be approved by the Station Superintendent prior to implementation and reviewed periodically as l set forth in administrative procedures. 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided:
- a. The intent of the original procedure is not altered;
- b. The change is approved by two members of the unit 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 PORC, and approved by the Station Superintendent within 14 days of implementation.
LIMERICK - UNIT 1 6-13
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4 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 outside
, containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the core spray, high pressure coolant injection, reactor core isolation cooling, residual heat removal, post-accident sampling system, safeguard piping fill system, control rod drive scram . discharge system, and containment air monitor systems. The program shall include the following:
- 1. Preventive maintenance and periodic visual inspection requirements, and
- 2. Integrated leak test requirements for each system at refueling cycle intervals or less,
- b. In-Plant Radiation Monitoring A program which 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 i 3. Provisions for maintenance of sampling and analysis equipment.
l
- c. Post-accident Samplina*
A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous efflu-
~
ents, and containment atmosphere samples under accident conditions. l 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.
(
*Not required until prior to exceeding S% of RATED THERMAL POWER. ,
b i l LIMERICK - UNIT 1 6-14
ADMINISTRATIVE CONTROLS ^ 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 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 6.9.1.1 A summary report of plant 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 unit. 6.9.1.2 The startup report shall address each of the tests identified in Sub-section 14.2.12 of the Final Safety Analysis Report and shall include a descrip-tion 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 conditions based on other commitments shall be g included in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence-ment 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 submitted at least every 3 months until all three events have been completed. ANNUAL REPORTS
- 6.9.1.4 Annual reports covering the activities of the unit 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. 6.9.1.5 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 mrea/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
*A single submittal may be made for a multiple unit station. **This tabulation supplements the requirements of 520.407 of 10 CFR Part 20.
LIMERICK - UNIT 1 6-15
ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) - dosimeter, thermoluminescent dosimeter (TLD), or film badge measure-ments. 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. Documentation of all challenges to safety / relief valves; and
- c. Any other unit unique reports required on an annual basis.
MONTHLY OPERATING REPORTS ! 6. 9.1. 6 Routine reports of operating statistics and s'hutdown experience, includ-ing documentation of all challenges to the the main steam system safety / relief valves, 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. ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- i
! 6. 9.1. 7 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted , prior to May 1 of each year. The initial report shall be submitted prior to
- May 1 of the year following initial criticality.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological i environmental surveillance activities for the report period, including a com-parison (as appropriate), with preoperational studies, operational controls and previous environmental surveillance reports and an assessment of the ob-served impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.12.2. i The Annual Radiological Environmental Operating Reports shall include the results of all radiological environmental samples and of all environmental radiation measurements taken during the report period pursuant to the locations specified in the tables 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 Branch Technical Position, Revision 1 ) November 1979. In the event that some individual results are not available for j inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted 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 legible maps **
*A single submittal may be made for a multiple unit station.
**0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations. b
- LIMERICK - UNIT 1 6-16
ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT covering all sampling 1ccations keyed to a table giving distances and directions from the centerline of the reactor plant; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3; discussion of all deviations from the Samplin(, Schedule of Table 4.12.1-1; and discussion of all analyses in which the LLD required by Table 4.12.1-1 was not achievable. SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT
- 6.9.1.8 Routine Semiannual Radioactive Efflueqt Release Reports covering the operation of the unit during the previous 6 months of operation shall be sub-mitted 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 sh:11 include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-active 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. The Semiannual Radioactive Effluent Release Report.to be submitted 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 direc-tion and atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, atmospheric stability.** This same report shall include an assessment 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 gase-ous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figures 5.1.3-la and 5.1.3-lb) during the report period. All assump- !. tions used in making these assessments (i.e. , specific activity, exposure time i and location) shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the methodology and parameters of the OFFSITE DOSE CALCULATION MANUAL (0DCM). I
*A single submittal may 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.
**In lieu of submission with the first half year Semiannual Radtoactive 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 upon request.
LIMERICK - UNIT 1 6-17
ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) The Semiannual Radioactive Effluent Release Report to be submitted 60 days after 4 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 path-ways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear 3 Power Operation. Acceptable methods for calculating the dose contribution from l liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977. The Semiannual Radioactive Effluent Release Reports shall include the following information for each type of solid waste (as defined in 10 CFR Part 61) shipped offsite during the report period:
- a. Container volume,
- b. Total curie quantity (specify whether determined by measurement or estimata,
- c. Principal radionuclides (specify whether determined by measurement or estimate),
- d. Source of waste and processing employed (e.g., dewatered spent resin,
; compacted dry waste, evaporator bottoms),
- e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and i
- f. SOLIDIFICATION agent or absorbent (e.g., cement; urea formaldehyde).
The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio- ) active 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 (PCP) and to the ODCM, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2. SPECIAL REPORTS
.6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
! I
- LIMERICK - UNIT 1 6-18
ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 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. 6.10.2 The following records shall be retained for at least 5 years:
- a. Records and logs of unit 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 Specification 6.8.1.
- f. Records of radioactive shipments.
4
- g. Records of sealed source and fission detector leak tests and results.
- h. Records of annual physical inventory of all sealed source material of record.
6.10.3 The following records shall be retained for the duration of the unit Operating License:
- a. Records and drawing changes reflecting unit 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 l.
assembly burnup histories.
- c. Records of radiation exposure for all individuals entering radiation j control areas.
l L l i I l l LIMERICK - UNIT 1 6-19 l
ADMINISTRATIVE CONTROLS' F
' RECORD RETENTION (Continued)
- d. Records of gaseous and liquid radioactive material released to the environs.
- e. _ Records of transient or operational cycles for those unit components identified in Table 5.6.1-1.
- f. Records of reactor tests and experiments.
- g. Records of training and qualification for current members of the unit 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 not listed in Section 6.10.2.
- j. Records of reviews performed for changes made to procedures or equip-ment or reviews of tests and axperiments pursuant to 10 CFR 50.59.
~
- k. Records of meetings of the PORC and the NRB.
- 1. Records of the service lives of all snubbers including the date at which the service life commences and associated installation and maintenance records.
- m. Records of analysis required by the Radiological Environmental Monitoring Program that would permit evaluation of the' accuracy of the analysis at a later date. "
6.11 -RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel. radiation protection shall be prepared con-sistent with the requirements of'10 CFR Part 20 and shall be approved,' main-tained, and adhered to for all operat' ions involving personnel radiation exposure.
. 6.12~ HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /h but less than 1000 mrem /h 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)*. 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 which continuously indicates the radiation dcse rate in the-area. .
, t
- Health physics personnel or personnel. escorted by health physics personnel _,
shall be exempt from the RWP issuance requirement during the performance of ( their assigned radiation protection duties, provided they.are otherwise following plant radiation protection procedures for entry into high radiation areas. LIMERICK - UNIT 1 6-20
4 ADMINISTRATIVEIONTROLS HIGH RADIATION AREA (Continued)
- b. A radiation monitoring device which 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 estab-lished and personnel have been made knowledgeable of them.
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activi-ties within the area and shall perform periodic radiation surveil-lance at the frequency specified by the Health Physicist in the RWP.
6.12.-2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour a dose greater than 1000 mrems shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of Shift Supervision on duty and/or the health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in ( that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour a dose in excess of 1000 mrems* that are located within large areas, such as the containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted, and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, continuous surveillance direct or remote (such as use of closed circuit TV cameras), may be made by personnel qualified in radiation protection procedures to provide . positive exposure control over the activities within the area. 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation. 6.13.2 Licensee-initiated changes to the PCP:
- a. 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.
- Measurement made at 18 inches from source of radioactivity.
LIMERICK - UNIT 1 6-21
1 ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM'(Continued)
- 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 change has been reviewed and found acceptable by the PORC.
- b. Shall become effective upon review and acceptance by the PORC.
6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) < 6.14.1 The ODCM shall be approved by the Commission prior te implementation. 6.14.2 Licensee-initiated changes to the ODCM:
- a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the changds) was made effective. This submittal shall contain:
- 1. Sufficiently detailed information to totally support the rationale for the change without benefit of addition 1 or supple-mental information. Information submitted should consist of a package of those pages of the 00CM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);
- 2. A determir.ation 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 Engineer-In-Charge, Nuclear and Environmental Section and the PORC.
- b. Shall become effective upon review and acceptance by the Engineer-In-Charge, Nuclear and Environmental Section and the PORC.
6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS 6.15.1 Licensee-initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):
- a. Shall be reported to the Commission in the Semiannual Radioacti.e
' Effluent Release Report for the period in which the change was made effective. 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 accordan x with 10 CFR 50.59; h
LIMERICK - UNIT 1 6-22
m' . 3 L
, ' ADMINISTRATIVE CONTROLS 4 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Continued)
- 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 plant 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 pre-dicted in the license application and amendments thereto;
- 5. An evaluation of the change which shows the expected maximum exposures to individual 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 changes are 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 PORC.
- b. Shall be reviewed and accepted by the PORC prior to implementation.
LIMERICK - UNIT 1 6-23
g,,o== u s uca24. tumo. co ...o* > ,caf~uw. - ..r,oc.,,- ... , E*1 E BIBLIOGRAPHIC DATA SHEET NUREG-ll49 Sit 145f auCTaONS ON T-E mEVEnst 2 TsTLE A%o SUGTITLE 3 LE AVE BLANn Technical Specifications for Limerick Generating Station, Unit No. 1 . oAte ...oai coo,atio
.o~T. . A.
j
. Auf-oas. June 1985 6 Daf t atPoRT isSL,to Robert E. Martin "
June l 1985 7 PimF Omus%G OR ^.ANi2 AfiO4 NAME ANo MasLING AoomE55 f#serwar le Cears a enoJECT T ASE . ORE UNel hum 9tR Office of Nuclezr Reactor Regulation Division of Licensing . . ,~ o ca A ~ T ~u . . . U. S. Nuclear Regulatory Commission Washington, D. C. 20555 - 10 SPomsome%G omOANil ATsON mava ANo MAILING AoomE55 tracbeele caser 11. Y vet o, REPohY Same as 7. above . ,a aloo cow.a a o ,,,,, , 12 EUPPLEUtNT Amy NOTts Appendix "A" to License No. NPF-39 Docket No. 50-352 13 A857R ACT (J00 ereres or 'ess, The Limerick Generating Station, Unit No.1, Technical Specifications were prepared by the U. S. Nuclear Regulatory Commission to set forth the limits, operating conditions and other requireinents applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public.
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Unlimited
, '6 SEcuniTV CLA55i8l CATION o r . ,,,
. ,oi~T.. 5 o.a. .o Ti. 5 Unelassified
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Unclassified
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UNITED STATES
,ou.r ct.ss o.n NUCLEAR REGULATORY COMMISSION 'oS2 *G' 6 "8 5 o WASHINGTON, D.C. 20666 wfs*f"o'c
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OFFICIAL tuSWESS PENALTY FOR PftiVATE USE. 6300 0
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