ML20246C285
| ML20246C285 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/30/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1360, NUDOCS 8907100312 | |
| Download: ML20246C285 (534) | |
Text
{{#Wiki_filter:NUREG-1360 O m Technical Specifications Limerick Generating Station, Unit No. 2 Docket No. 50-353 Appendix "A" to License No. NPF-83 p
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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
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o 'U l NUREG-1360 Technical Specifications Limerick Generating Station, Unit No. 2 Docket No. 50-353 Appendix "A" to (O) License No. NPF-83 issued by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1989
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-{9_ J, SECTION y im 11.0 DEFINITIONS:
PAGE 1.1 'ACTI0N....c................................................... 1. 1.2 ' AVERAGE PLANAR EXP05URE....................................... 1-1 1.3 AVERAGE PLANAR LINEAR. HEAT GENERATION RATE..................... 1-1
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' -( sd ~1.4 CHANN E L CA LI B R AT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . 1-1 1.5 CHANNEL CHECK................................................ 1-1 1.6 CHANNEL FUNCTIONAL' TEST...................................... 1-1 11.7 CORE ALTERATION.............................................. 1-2 1.8 CRITICAL'P0WERLRATIO......................................... 1-2
- 1. 9 DOSE EQUIVALENT'I-131........................................ 1-2 p;, 1.105-AVERAGEDISINTEGRATIONENERGY.............................. 1-2
'\ 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME........... 1-2 1.12 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TINE.... ~ 1-2 1.13. FRACTION OF LIMITING POWER DENSITY........................... 1-3 1.14 FRACTION OF RATE 0 THERMAL P0WER.............................. 1-3 1.15 FREQUENCY N0TATION........................................... 1-3 1.16 I D ENTI FI ED LEA KAG E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.17 ISOLATION SYSTEM RESPONSE TIME............................... 1-3 1.18 LIMITING CONTROL ROD PATTERN............. ................... 1-3 1.-19 LINEAR HEAT GENERATION RATE.................................. 1-3 11.20 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-4 1.21 MAXIMUM FRACTION OF LIMITING POWER DENSITY..,................ 1-4 1
LIMERICK - UNIT 2- i a=-____-__________-_-___-_-__---_-_-____-__-_---_-__-
y _ l. INDEX DEFINITIONS i SECTION. DEFINITIONS (Continued) PAGE 1.22 MEMBER (S) 0F THE PUBLIC...................................... 1-4 1.23 MINIMUM CRITICAL POWER RATI0................................. 1-4 1.24 0FFSITE DOSE CALCULATION MANUAL............................... 1-4 l
'1.25 OPERABLE - OPERABILITY....................................... 1-4 ;
4 1 1.26 OPERATIONA'L CONDITION - CONDITION............................ 1-4 ; 1.27 PHYSICS TESTS................................................ 1-4 i 1.28 PRESSURE BOUNDARY LEAKAGE.................................... 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY................................ 1-5 i 1 1.30 PROCESS CONTROL PR0 GRAM...................................... 1-5 1.31 PURGE - PURGING.............................................. 1-5 l 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 R0D DENSITY.................................................. 1-7 1.38 SHUT 00WN 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 P0WER.......................................... ...... 1-8 1.44 UNIDENTIFIED LEAKAGE................................... ..... 1-8 O LIMERICK - UNIT 2 ii
4 i INDEX j L-
,~s L DEFINITIONS-
. .(c N,;) l SECTION DEFINITIONS (Continued) PAGE 1.45 UNRESTRICTED AREA..................................... .. ,, 1_g 1.46 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-g 1.47 VENTING....................................,,,,,,,,,,,,,,,,,, 3_g Table 1.1, Surveillance Frequency Notation...................,,,,, 1_g Table 1.2, Operational Conditions................................. 1-10
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m f s LIMERICK - UNIT 2 111 l
r INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS lf SEC7 ION 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 Protection System Instrumentation Setpoints......... 2-3 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints............... 2-4 BASES 2.1 SAFETY LIMITS S, THERMAL PDWER, Low Pressure or Low Flow........... ......... B 2-1 THERMAL POWER, High Pressure and High Flow............. .. . B 2-2 Intentionally Left Blank.................... ...... ..... . B 2-3 Intentionally Left Blank..................... ... .......... 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 i i
Ol i LIMERICK - UNIT 2 iv I
f 1 INDEX l 1 7m, i ( ( LIMITINGCONDITIf.SFOROPERATIONANDSURVEILLANCEREQUIREMENTS L/ SECTION PAGE f 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.......................................... 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability.................................. 3/4 1-3 5 Control Rod Maximum Scram Insertion Times................ 3/4 1-6 Control Rod Average Scram Insertion Times................ 3/4 1-7 Four Control Rod Group Scram Insertion Times............. 3/4 1-8 Control Rod Scram Accumulators........................... 3/4 1-9 Control Rod Drive Coupling............................... 3/4 1-11 h Control Rod Position Indication.......................... 3/4 1-13 Control Rod Drive Housing Support............... ........ 3/4 1-15 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Rod Worth Minimizer...................................... 3/4 1-16 Rod Block Monitor........................................ 3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............... ............ 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Solution Temperature / Concentration Requirements........................ 3/4 1-21 Figure 3.1.5-2 (LEFT BLANK INTENTIONALLY) . . . . . . . . . 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE............... 3/4 2-1 Figure 3.2.1-1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types BP8CRB278............ 3/4 2-2 i ( l l 1 LIMERICK - UNIT 2 v i _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________J
I s INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) Figure 3.2.1-2 Maximum Avera;;e Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types BP8CRB248.......... 3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types BP8CRB163.......... 3/4 2-4 Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus l Average Planar Exposure Initial i Core Fuel Types BP8CRB094.......... 3/4 2-5 l 1 Figure 3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CRB071.......... 3/4 2-6 Intentionally Left Blank..... .... 4/4 2-6a j 1 3/4.2.2 APRM SETP0lNTS..... .................................... 3/4 2-7 ! 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................ 3/4 2-8 l Table 3.2.3-1 Minimum Critical Power Ratio (MCPR) Versus Plant Operating Condition....... 3/4 2-8a Figure 3.2.3-la Minimum Critical Power Ratio (MCPR) l Versus I at Maximum Core Flow < 100% Rated....................T........ 3/4 2-10 Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR) Versus I at Maximum Core Flow < 105% Rated....................T........ 3/4 2-10a Figure 3.2.3-2 K f Factor.............................. 3/4 2-11 3/4.2.4 LINEAR HEAT 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 LIMERICK - UNIT 2 vi
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'INDEX LIMITING CONDITIONS FOR OPERATION'AND SURVEILLANCE REQUIREMENTS I :SECTION , i PAGE INSTRUMENTATION (Continued)-
. 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 Instrumentation Response Time ...................... 3/4 3-23 Tsble 4.3.2.1-1 Isolation Actuatiori Instrumentation 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 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 Cors Cooling System Actuation Instrumentation Surveillance Requirements....... . 3/4 3-40 3/4.3.4- RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION 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 Instrumentation Surveillance Requirements...................... 3/4 3-45 End-of-Cycle Recirculation Pump Trip System Instrumentation........................................ 3/4 3-a6 LIMERICK - UNIT 2 vii
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION ,PAGE I INSTRUMENTATION (Continued) Table 3.3.4.2-1 End-of-Cycle Recirculation Pump ) Trip System Instrumentation....... 3/4 3-48 f l Table 3.3.4.2-2 End-of-Cycle Recirculation Pump Trip Setpoints.................... 3/4 3-49 ( Table 3.3.4.2-3 End-Of-Cycle Recirculation Pump Trip System Response Time......... 3/4 3-50 Table 4.3.4.2.1-1 End-0f-Cycie Recirculation
, Pump Trip System Surveillance Requirements..... .......... ... 3/4 3-51 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION l INSTRUMENTATION......................................... 3/4 3-52 Table 3.3.5-1 Reactor Core Isolation Cooling System Actuation Instruments-tion............... ..... .......... 3/4 3-53 ,
Table 3.3.5-2 Reactor Core Isolation Cooling System Actuation Instrumentation Setpoints........................... 3/4 3-55 Table 4.3.5.1-1 Reactor Core Isolation Cooling System Actuation Instrumentation Surveillance Requirements....... .. 3/4 3-56 3/4.3.6 CONTROL R00 Bi.0CK INSTRUMENTAII0N....................... 3/4 3-57 Table 3.3.6-1 Control Rod Block Instruments-tion...... .......................... 3/4 3-58 Table 3.3.6-2 Control Rod Block Instruments- { tion Setpoints................. ..... 3/4 5-60 Table 4.3.6-1 Control Rod Block Instruments- i 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 O LII'Ei'ICK - UNIT 2 viii
m r I I r, ~aw '- 4 " -c INDEX ' l[po .LIMITINGCONDI{']NSFOROPERATIONAND'SURVEILLANCEREQUIREMENTS-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 2, Table 3.3.7.2-1 Seismic Monitoring L : Instrumentation.................... 3/4 3-69
. Table-4.3.7.2-1 Seismic' Monitoring i
Instrumentation Surv6111ance Requirements........................ 3/4 3-71 q. 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 L;( 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 Accident Monitoring Instrumentation..................... 3/4 3-84 Table 3.3.7.5-1 Accident Monitoring Instrumen-oh tation...... ..................... 3/4 3-85 Table 4.3.7.5-1 Accident Monitoring Instruments-tion Surveillance Requirements.... 3/4 3-87 Source Range Monitors................................... 3/4 3-88 Traversing In-Core Probe System............... ......... 3/4 3-89 Chlorine Detection System............................... 3/4 3-90 Toxic Gas Detection System.............................. 3/4 3-91 Fire Detection Instrumentation....... .............. ... 3/4 3-92 LIMERICK - UNIT 2 ix
g l IllDEX l 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- 1 tation.................................................. 3/4 3-98 Table 3.3.7.11-1 Radioactive Liquid Effluent l Monitoring Instrumentation....... 3/4 3-99 Table 4.3.7.11-1 Radioactive Liquid Effluent Monitoring Instrumentation
' Surveillance Requi rements. . . . . . . . 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/ Main Turbine 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 Instruments-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 2 x
INDEX LIMITING CONDITIONS'FOR'0PERATION'AND SURVEILLANCE REQUIREMENTS LJ SECTION PAGE. REACTOR COOLANT SYSTEM'(Continued) Figure 3.4.1.1-1 Thermal Power versus Core Flow... 3/4 4-3 Jet Pumps............................................... 3/4 4-4 g 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. Ope rati onal Lea kage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-9 Table 3.4.3.2-1 Reactor Coolant System Pressure 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
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.. SPECIFIC ACTIVITY............................... ....... 3/4 4-15 Table 4.4.5-1 Primary Coolant Specific Activity Sample and Analysis Program......... 3/4 4-17 s
3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................. 3/4 4-18 Figure 3.4.6.1-1 Minimum Reactor Pressure Vessel ! I Metal Temperature Vs. Reactor Vessel Pressure.................. 3/4 4-20
. Table 4.4.6.1.3-1 Reactor Vessel Material Surveil-lance Program - Withdrawal Schedule........................ 3/4 4-21 Reactor Steam Dome...................................... 3/4 4-22 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................ 3/4 4-23 O
v 3/4.4.8 STRUCTURAL INTEGRITY.................................... 3/4 4-24 LIMERICK - UNIT 2 xi l .. - - _ - - - _ . . _ _ - _ _ _ - _
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REACTOR COOLANT SYSTEMS (Continued) 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
- i. 3/4.5.2 ECCS - SHUTD0WN......................................... 3/4 5-6 l 3/4.5.3 SUPPRESSION CHAMBER..................................... 3/4 5-8 l
l 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 Ai r 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 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 LIMERICK - UNIT 2 x;i l 1 l I O__m____.___ ___. _ . _ __ _ . _ . _ -
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS v/ SECTION PAGE CONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEF Suppression Chamber - Drywell Vacuum Breakers........... 3/4 6-44 3/4.6.5 SECONDARY CONTAINMENT , i Reactor Enclosure Secondary Containment Integrity....... 3/4 6-46 Refueling Area Secondary Containment Integrity.......... 3/4 6-47 Reactor Enclosure Secondary Containment Automatic Isolation Va1ves........................................ 3/4 6-48 Table 3.6.5.2.1-1 Reactor Enclosure Secondary Containment Ventilation System Automatic Isolation Valves.......... ............... 3/4 6-49 Refueling Area Secondary Containment Automatic Isolation 3/4 6-50 [nJ v Valves........................................ 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 - Common System............ 3/4 6-52 Reactor Enclosure Recirculation System.................. 3/4 6e-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 SY3TEMS Residual Heat Removal Service Water System - Common System.................... ..... .... .................. 3/4 7-1 Emergency Service Water System - Common System.. ...... 3/4 7-3
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Ul ti ma te He a t S i n k. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-5 LIMERICK - UNIT 2 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 - COMMON SYSTEM.................................................. 3/4 7-6 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM...............,.... 3/4 7-9 l 3/4.7.4 SNUBBERS................................................ 3/4 7-11 Figure 4.7.4-1 Sample Plan 2) For Snubber l Functional Test.................... 3/4 7-16 l 3/4.7.5 SEALED SOURCE CONTAMINATION............................. 3/4 7-17 i 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System........................... 3/4 7-19 Spray and/or Spri nkler Systems. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-22 l l CO 2 Systems............................................. 3/4 7-24 Halon Systems............. ............ ................ 3/4 7-25 Fire Hose Stations...................................... 3/4 7-26 Table 3.7.6.5-1 Fire Hose Stations................ 3/4 7-27 Yard Fire Hydrants and Hose Cart Houses ................ 3/4 7-29 Table 3.7.6.6-1 Yard Fire Hydrants and Hose Cart 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 l 3/4.8.2 D.C. SOURCES D. C. Source s - Ope rati ng. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-10 l LIMERICK - UNIT 2 xiv i
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w - ( L INDEX l[ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
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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
..3/4.8.3. ONSITE POWER D' DISTRIBUTION SYSTEMS 1
Distribution - Operating................................ 3/4 8-15 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 Table 3.8.4.1-1 Primary Containment Penetration Conductor Overcurrent Protective Devices........................... 3/4 8-23 Motor-Operated Valves Thermal Overload Prot'ection. . . . . . . 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 3/4.9.3 CONTROL ROD P0SITION.................................... 3/4 9-5 I 3/4.9.4 DECAY TIME.............................................. 3/4 9-6 3//f.9.5 COMMUNICATIONS.......................................... 3/4 9-7 3/4.9.6 REFUELING PLATF0RM...................................... 3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL . STORAGE P00L. . . . . . . . . . . . . .. . . . . 3/4 9-10 3/4.9.8 WATER LEVEL - REACTOR VESSEL............................ 3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L. . . . . . . . . . . . . . . . . . . 3/4 9-12 LIMERICK - UNIT 2 xv
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.10 CONTROL R0D REMOVAL Single Control Rod Remova1.............................. 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 Leve1......................................... 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 WORTH MINIMIZER..................................... 3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.......................... 3/4 10-3 3/4.10.4 RECIRCULATION L00PS..................................... 3/4 10-4 3/4.10.5 0XYGEN CONCENTRATION.................................... 3/4 10-5 3/4.10.6 TRAINING STARTUPS....................................... 3/4 10-6 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING.......... 3/4 10-7 3/4.11 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 Progrsm........................ 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 2 xvi
3 li i i INDEX y% ,
}- LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
- \ J ;
^
SECTION PAGE RADI0 ACTIVE 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 Dose - Iodine-131, Iodine-133, Tritium, and. Radionuclides in Particulate Form..................... 3/4'11-13 Ventilation Exhaust Treatment System.................... 3/4 11-14 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 00SE.............................................. 3/4 11-20 t
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... 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 i-3/4.12.3 INTERLABORATORY COMPARISON PROGRAM.......... ........... 3/4 12-14 l
LIMERICK - UNIT 2 xvii
INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY............................................ B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN....................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B '/4 1-1 3/4.1.3 C O NTRO L R0 D S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-2 3/4.1.4 CONTROL R0D PROGRAM CONTROLS. . . . . . . . . . . . . . . . . . . . . . . . . B 3/ 4 1-3 1 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 i Parameters to the Loss of-Coolant Accident Analysis...... B 3/4 2-3 3/4.2.2 APRM SETP0lNTS....... ................................ 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 j 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION.. .. ... ........... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION l Radiation Monitoring Instrumentation.......... ... ... B 3/4 3-4 LIMERICK - UNIT 2 xviii
.______ __- _ a
.l l
(:
'f 1
1 INDEX i 4 A BASES SECTION: PAGE f ' INSTRUMENTATION (Continued) 1 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 Fire Detection Instrumentation.......................... B 3/4 3-6
. Loose-Part Detection System............................. B 3/4 3-6 Radioactive Liquid Effluent Monitoring 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 Bases Figure B 3/4.3-1 Reactor Vessel Water Leve1...................... B 3/4 3-8 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 i
LIMERICK - UNIT 2 xix i
INDEX 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/A.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 RESIDUAL HEAT REM 0 VAL................................... 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 Contai nment Integrity. . . . . . . . . . . . . . . . . . . . . . B 3/4 6-1 Primary Containment Leakage........................ B 3/4 6-1 Primary Containment Air Lock...... ................ 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 i Drywell and Suppression Chamber Purge System...... B 3/4 6-2 ] 3/4.6.2 DEPRESSURIZATION SYSTEMS..................... ..... B 3/4 6-3 l LIMERICK - UNIT 2 xx
'-'.l[ y'
)
1 INDEX Q jl yys BASES' LV' R SECTION. PAGE l CONTAINMENT SYSTEMS (Continued) L , ? 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 SECONDARY CONTAINMENT.'............................. B 3/4 6-5
, 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL... ......... B 3/4 6-7
'3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS - COMMON SYSTEMS............. B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -
COMMON SYSTEM...................................... B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM.............. B 3/4 7-la 3/4.7.4 SNUBBERS........................................... B 3/4 7-2
.s [~' 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 R4TED ASSEMBLIES.............................. B 3/4 7-4 l 3/4.8- ELECTRICAL PDWER SYSTEMS 3/4.8.1, 3/f. 8.2, and 3/4.8.3 /..C SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS...... ........................ B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PR0 LLC 11VE DEVICES............ B 3/4 8-3 3/4.9 REFUELING OPERATIONS l 3/4.9.1 REACTOR MODE SWITCH................................ B 3/4 9-1 1
L 3/4.9.2 INSTRUMENTATION.................................... B 3/4 9-1 1 3/4.9.3 CONTROL ROD P0SITION............................... B 3/4 9-1 3/4.9.4 DECAY TIME............. ........................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS..................................... P 3/4 9-1 l LIMERICK - UNIT 2 xxi
INDEX BASES
)
SECTION PAGE 1 1 REFUELING OPERATIONS (Continued) l 4
, 3/4.9.6 REFUELING PLATF0RM................................ B 3/4 9-2 )
3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGF P00L............ B 3/4 9-2 I 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL ' l and WATER LEVEL - SPENT FUEL STORAGE P00L... ..... B 3/4 E-2 , 3/4.9.10 CONTROL R0D 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 EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY..................... B 3/4 10-1 3/4.10.2 R0D WORTH MINIMIZER........... ................... B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.... ........ .... . B 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS......... ....... ............. B 3/4 10-1 3/4.10.5 0XYGEN CONCENTRATION.............................. B 3/4 10-1 , 1 1 3/4.10.6 TRAINING STARTUPS................................. B 3/4 10-1 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING.... B 3/4 10-1 1 i 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 O LIMERICK - UNIT 2 xxii
t INDEX
/_'Ti ;
i BASES
<j
- i. SECTION PAGE RADI0 ACTIVE EFFLl!CNTS (Continued) {
Explosive Gas Mixture.............................. B 3/4 11-4 Main Condenser..................................... B 3/4 11-5 Venting er Purging................................. B 3/4 11-5 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 ENVIR0t4 MENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM................................. B 3/4 12-1 3/4.12.2 LAND USE CENSUS........................... ........ B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARIS0N PR0 GRAM................. B 3/4 12-2 O V rs ( k LIMERICK - UNIT 2 xxiii
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 GOUNDARY 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 5.1.3-1b Map Defining UNRESTRICTED 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 Sec e.da ry Co ntai nme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 5.3 REACTOR CORE Fuel Assemblies.............................................. 5-7 Control Rod Assemblies....................................... 5-7
- 5. 4 REACTOR COOLANT SYSTEM l
Design Pressure and Temperature...................... ....... 5-7 Volume................. .... .......... .................... 5-8 l l 5.5 FUEL STORAGE " Criticality................. ................ .............. 5-8 i LIMERICK - UNIT 2 xxiv ! 3 l
l INDEX J,m
! i DESIGN FEATURES '\ J SECTION PAGE FUEL STORAGE (Continued)
Drainage................................................,,,,, 5.g l V Capacity...................................................., s.g 5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT.......................... 5-8 Table 5.6.1-1 Component Cyclic or Transient Limits..... 5-9 f
/x \
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o I l l l l l l O
/
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LIMERICK - IINIT 2 xxv
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY............................................... 6-1 6.2 ORGANIZATION..................................,.............. 6-2 6.2.1 0ffsite................................................. 6-1 Figure 6.2.1-1 Offsite Organization............... 6-3 6.2.2 Unit Staff.............................................. 6-1 ' i Figure 6.2.2-1 Organization for Conduct of I Plant Operations................... 6-4 Table 6.2.2-1 Minimum Shift Crew ! Composition......................... 6-5 t 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) , Function .............................................. 6-6 Composition............................................ 6-6 Responsibilities....................................... 6-6 Records.................. ............................. 6-6 6.2.4 SHIFT TECHNICAL ADVIS0R................................ 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 i Quorum............................... ................. 6-7 Responsibilities ................................... .. 6-8 i Records................................................ 6-9 LIMERICK - UNIT 2 xxvi
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INDEX. n j'~'N ADMINISTRATIVE CONTROLS ! i v) .SECTION PAGE 6.5.2 NUCLEAR REVIEW BOARD (NRB) Functioc .............................................. 6-9 Composition ............................................ 6-9 Alternates.............................................. 6-10 Consultants............................................ 6-10 Mee ti ng F r eq ue ncy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 Quorum................................................. 6-10 Review................................................. 6-10 Audits................................................. 6-11 Records..................................... .......... 6-12 6.6 REPORTABLE EVENT ACTI0N..................................... 6-12 m [ \ 6.7 SAFETY LIMIT VIOLATION...................................... 6-12
.Q 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 ........................................ 5-15
' Monthly Operating Reports.............................. 6-16 Annual Radiological Environmental Operating Report..... 6-16 Semiannual Radioactive Effluent Release Report......... 6-17 6.9.2 SPECIAL REP 0RTS........................................ 6-18 6.10 RECORD RETENTION........................................... 6-19 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-20 6.12 HIGH RADIATION AREA............................. .......... 6-20 LIMERICK - UNIT 2 xxvii
l INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.13 PROCESSCONTROLPROGRAM(PCPl............................... 6-21 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)..................... 6-22 I' 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS....... 6-22 l 1 I l 1 l l l -- 1 i
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l l 1 i l O O l LIMERICK - UNIT 2 xxviii
I er i SECTION 1.0 DEFINITIONS l
')
9
F 1 l 1.0 DEFINITIONS
) The following terms are defined so that uniform interpretation of these i "' specifications may be achieved. The defined terms appear in capitalized type
~
and shall be applicable ti,coughout these Technical Specifications. ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial 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. AVERAGE PLANAR LINEAR HEAT GENERATION RATE 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 HEAI 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. Ch SNEL CALIBRAT19N 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 O 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 total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 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 instrunent channels , measuring the same parameter. { CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.
I b. Bistable channels - the injection of a simulated signal into the sensor 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 2 1-1
DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, rel,.__ tion or movement of fue'1, 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 ALTERATI0H. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position. l CRITICAL POWER RATIO i 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 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be thase 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 radionuclides 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 ( inciuoe diesel generator starting and sequence loading delays where j applicable. The response time may be measured by any series of sequential, l overlapping or total steps such that the entire response time is measured. l l 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 2 1-2
y , 1 e' 1 l DEFINITIONS , _ f, This: total system response time consists of two components, the instrumen- l [y \'~
.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 p that bundle-type.
FRACTION OF RATED THERMAL POWER' 1.14 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measurc$ 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: s 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 I b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation I 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 l the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. 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 R0D PATTERN 1.18 A LIMITING CONTROL R0D 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 g transfer area associated with the unit length. LIMERICK - UNIT 2 1-3
1 I i DEFINITIONS , LOGIC SYSTEM FUNCTION L EST I 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 l value of the FLPD which exists in the core. NEMBER(S) 0F THE PCBtIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupational 1.y 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 : 1e MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l exists in the core (for each class of fuel). OFFSITE DOSE CALCULATION MANUAL 1.24 The 0FFSITE DOSE CALCULATION MANUAL (0DCM) 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, subsystem, 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 (3) 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 2 1-4
o 3 e, h DEFINITIONS PRESSURE BOUNDARY LEAKAGE
- 1.28 PRESSURE BOUNDARY 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 i 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.
, f. The sealing mechanism associated with each primary containment i 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 l, results in a waste package with properties that meet the minimum and i 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, 2H O content, solids content ratio of.solidificat bn 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, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. LIMERICK - UNIT 2 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, slide gate camper or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.5.2.1-1 of Specific 6 tion 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 within the reactor enclosure secondary containment is less than or equal to the value required by Specification 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 de-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, slide gate damper or du.:tivated 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 2 1-6 l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l
s -- 1 i DEFINITIONS- : i [ ({p-}
' REFUELING FLOOR' SECONDARY CONTAINMENT-INTEGRITY (Continued)
]
- b. lAll 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. At least one coor in each access to the. refueling floor _ secondary
~
'd.
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 tne refueling floor secondary containment is
.less than or equal to the_value required by Specification 4.6.5.1.2a.
1 REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in LSection 50.73 to 10 CFR Part 50. R0D 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. 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 condition; 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-la.
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. 1 LIMERICK - UNIT 2 1-7
1 1 l { i DE,FINITIONS STAGGERED TEST BASI l 1.42 A STAGGERED TEST BASIS shall consist of: I
- a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.
l b. The testing of one system, subsystem, trair., or other designated l component at the beginning of each subintertal. THERMAL POWER I 1.43 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. UNIDENTIFIED LEAKAGE 1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTT.FIED LEAKAGE. UNRESTRICTED AREA i 1.45 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l access to which is not controlled by the licensee for purposes of protec-
- i. tion of individuals from exposure to radiation and radioacti9e materials, l or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
VENTILATION EXHAUST TREATMENT SYSTEM j 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 particulate 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 l 1.47 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, co, centration 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. 1 l l 9 LIMERICK - UNIT 2 1-8
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SURVEILLANCE FREQUENCY NOTATION. , 0 .; y ' -! b ' NOTATION. FREQUENCY' em M S' At least once per 12 hours.
.Ec 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. E f-SA'- At'least once per 184 days. L A At least once per 366 days.
~R.- At least once per 18 months =(550 days).
b S/U. Prior to each reactor startup. (
, .P Prior to each radioactive release.
N. A. Not applicable.
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_t LIMERICK - UNIT 2 1-9 u=-----_--__- - _ _ _ _ - . _ - _ _ _ - _ _ _ _ - . _ _ - _ -
DEFINITIONS TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE
- 1. POWER ODERATION Run Any temperature
- 2. STAR 10P Startup/ Hot Standby Any temperature
- 3. HOT SHUTUOWN Shutdown # *** > 200 F
- 4. COLD SHUTDOWN Shutdown # ## *** $ 200 F
- 5. REFUELING
- Shutdown or Refuel ** # $ 140 F t
#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 r. 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 j 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.
l
**See Special Test Exceptions 3.10.1 and 3.10.3.
***The reactor mode switch may be placed in the Refuel positio, while a single cor, trol rod is being recoupled provided that the one-rod-out interlock is l CPERABLE.
LIMERICK - UNIT 2 1-10
1
.O L
L SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O O
1 g. I 2.0 ~ SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS A ! d 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 l f reactor vessel steam dome pressure less than 785 psig or core flow less than '
-10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIO,NS 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 greater than 10% of rated flow.
p 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 CONDITION 5 1, 2, 3, and 4. ACTION: With the reactor coolan+, 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. IMERICK - UNIT 2 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, l after depressurizing the reactor vessel, if required. Comply with the l requirements of Specification 6.7.1. 9 O, LIMERICK - UNIT 2 2-2
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- SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
,; -~Q . l . 2. 2 LIMITING SAFETY SYSTEM SETTINGS l- 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. s
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LIMERICK - UNIT 2 2-3
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i O BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O
O I i l l l M 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 I l l I l l
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i ! U i 2.1 ' SAFETY LIMITS- .l
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D BASES W
' 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 integrity. 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 perfor6tions 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 fission product migration from cladding perforation is just as measurable as
.that from use related cracking, the thermally caused cladding perforations
, 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 103 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. LIMERICK - UNIT 2 B 2-1
1 SAFET,Y LIMITS BASES Ol 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 pa'rameters 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 precedures 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 a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. Calculation of the Safety Limit MCPR is described in Reference 1. O
Reference:
- 1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A .
(latest approved revision). l l l LIMERICK - UNIT 2 B 2-2 i l
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l [ ! SAFETY: LIMITS.. l
?b l b -BASES !
i 2.1. 3 REACTOR COOLANT SYSTEM PRESSURE , 1 The Safety Limit for the reactor coolant system pressure has been j 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 L. 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 L 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 coolant system is designed to the ASME. Boiler and Pressure Vessel Code, 1977 Edition, including Addenda through Summer 1977 for the reector 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.
] 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 i x . 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 tooling capability could lead to elevated cladding i temperatures and clad perforation in the event that the water level became less ; than two-thirds of the core heignt. The Safety Limit has been established at i the top of the active irradiated fuel to provide a point which can be monitored j and also provide adequate margin for effective action. ] I I I 4 l l i l LIMERICK - UNIT 2 B 2-5 i l
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PP.0TECTION 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 i and reactor coolant system are prevented from exceeding their Safety Limits I
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 i allowance assumed for each trip in the safety analyses. l l 1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip 1 systems. The IRM is a 5 decade 10 range instrument. The trip seipoint of 120 l 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 l also ranged up. The IRM instruments provide for overlap with both the APRM l 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 reyuired 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 l 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. l Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of 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 RWM. Of all the passible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.
O LIMERICK - UNIT 2 B 2-6
I I LIMITING SAFETY SYSTEM SETTINGS ( ( V ). BASES I i REACTOP 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 change 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 Neutren Flux-Upscale flow bias setpoint; i.e, for a power increase, the THERMAL PO4 R of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat g transfer associated with the fuel. i \ V 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 specified 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 td o setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin tc 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. r [ ( LIMERICX - UNIT 2 B 2-7
s 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 I
l f ar 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. l Main Steam Line Isolation Valve-Closure 5. l The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIVs L 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 dttectors 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 isolation 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 Pressure-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 0 B 2-8
r , I i c !
, LIMITING SAFETY SYSTEM SETTING i j
) BASES REACTOR PROTECTION SYSTEM INSTRUME!,TATION SETPOINTS (Continued)
- 8. .' Scram Discharge Volume Water Level-High The ceram 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 25.58 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.
- 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 i hydrauiic 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 i 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 I
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.
'\
LIMERICK - UNIT 2 B 2-9
r; i
~'qf j
SECTIONS 3.0 and 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS lO O
'3/4.0 APPLICABILITY i !
7
) LIMITING CONDITION FOR OPERATION LJ 3.0.1 Compliance 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 Limitir.'g Coadition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 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. I 1
- 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.
l Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for
/ -
Operation. Exceptions to these requirements are stated in the individual Speci-h fications. l 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 when the conditions fc. the Limiting Condition for Operation are not met, and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL CONDITION or other specified condition may be made in accordance with the ACTION I requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements are stcted in the individual Specifications. l l 1 LIMERICK - UNIT 2 3/4 0-1
i APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual Limiting Conditinns for Operation unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performec, 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 p shall not exceed 3.25 times the specified surveillance interval.
i 4.0.3 Failure to perform a Surveillance Requirement within the allowed l surveillance interval defined by Specification 4.0.2, shall constitute j noncompliance with the OPERABILITY requirements for a Limiting Condition for ) Operation. The time limits of the ACTION requirements are applicath at the { time it is identified that a Surveillance Requirement has not been i performed. The ACTION requirements may be delayed for up to 24 hours to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours. Surveillance requirements do not have to be performed 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. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply l with ACTION requiremcats. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, & 3 (omponents shall be applicable as follows:
- a. Inservice irspection of ASME Code Class 1, 2, and 3 amponents and inservice tisting of ASME Code Class 1, 2, and 3 puhps and valves shall be per formed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g) (6) (i). '
- 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:
O LIMERICK - UNIT 2 3/4 0-2
APPLICABILITY j j SURVEILLANCE' REQUIREMENTS (Continued) , w: 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 I Quarterly' or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days
- c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testitig activities,
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requiremvits of any Technical .5 specification.
/,. i L)
LIMERICK - UNIT 2 3/4 0-3
"a W 3/4.1 REACTIVITY-CONTROL SYSTEMS k ,[
3/4.1.1 SHUTDOWN MARGIN f 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
- b. 0.28% Ak/k with.the highest worth rod determined by test.
APPLICABILITY: OPERATIONAL CONDITIONS 3, 2, 3, 4, and 5. ACTION: With the SHUTOOWN 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 !
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. ! l
- c. In'0PERATIONAL CONDITION 5, suspend CORE ALTERATIONS and other
'O activities that could reduce the SHUTDOWN MARGIN and insert all insertable control rods within 1 hour. Establish SECONDARY CONTAIN-j
'l I
MENT INTEGRITY within 8 hours. J
)
I SURVEILLANCE REQUIREMENTS l 4.1.1 The SHUTDOWN MARGIN :shall be determined to be equal to or greater than i specified at any time during the fuel cycle: !
- a. By measurement, prior to or during the first startup after each !
refueling.
- 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 .I withdrawn worth of the immovable or untrippable control rod. l LIMERICK - UNIT 2 3/4 1-1
REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES [ LIMITING CONDITION FOR OPERATION , 3.1.2 The reactivity equivalence of the difference between the actual ROD DENSITY and the predicted R0D DENSITY shall not exceed 1% Ak/k. l 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. ) i
- b. 'Otherwise, be in at least HOT SHUTDOWN within the next 12 hours. l
{ 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: I
- a. During the first startup following CORE ALTERATI0flS, and
- b. At least once per 31 effective full power days during POWER OPERATION.
l I O LIMERICK - UNIT 2 3/4 1-2
'j i 1 4
/ REACTIVITY CONTROL SYSTEMS
_Op 3/4.1.3 CONTROL RODS s CONTROL ROD OPERABILITY- lI
' LIMITING CONDITION FOR OPERATION 3.l.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:
g 1. Within 1 hour: a) Verify that the inoperable control rod, if withdrawn, is separated from all other inoperable 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 drive water and exhaust water isolat:on valves.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- 2. Restore the inoperable control rod to OPERABLE status within g 48 hours or be in at least HOT SHUTDOWN 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 I 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 the 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: 1 a) Electrically, or b) Hydraulically by closing the drive water and exhaust water i isolation valves. , *The inoperable control rod may then be withdrawn to a position no further I withdrawn than its position when found to be inoperable. l
**May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
LIMERICK - UNIT 2 3/4 1-3
I REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- 2. If.the inoperable control rod (s) is inserted, within I hour disarm the associated directional control valves ** either:
l a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves. l Otherwise, be in at least HOT SHUTDOWN within the next 12 hours. L'
- 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. 1 ,, SURVEILLANCE REQUIREMENTS l 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, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch: I
- 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. , i l l
*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.
O LIMERICK - UNIT 2 3/4 1-4
7 O$k , i na l REACTIVITY CONTROL. SYSTEMS j ijs-~ , YYi a . l PA > ' SURVEILLANCE ~ REQUIREMENTS (Continued) l ll, - l 1 l e"' 4.1.3.1.4..The scram' discharge volume shall be determined OPERABLE by j demonstrating , +,- 1
- a. The scram discharge volume drain and vent valves OPERABLE,'when l
. control rods are scram' tested from a normal control rod configura-
~
j tion of-less .than or equal to 50% R0D DENSITY at least once per .l 18 months, by verifying.that the drain and vent valves: i
- 1. Close within 30 seconds after receipt of a signal for control rods to scram, and
- 2. Open when the scram signal is reset.
['I 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. ]
)
,sm q 1 l l 1 te I l I
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i LIMERICK - UNIT 2 3/4 1-5 l 1 i 1
REACTIVITY CONTROL SYSTEMS CONTROL ROD 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. i APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: j I i a. With the maximum scram insertion time of one or more control rods ! l exceeding 7 seconds: '
- l. Declare the control rod (s) with the slow insertion time inoperable, and j
- 2. Perform the Surveillance Requirements of Specification 4.1.3.2c.
at least once per 60 days when operation is continued with three 1 or more control rods with maximum scram insertion times in excess j of 7.0 seconds. 3 Otherwise, be in at least HOT SHUTDOWN within 12 hours. I
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 1 1 4.1.3.2 The maximum scram insertion time of the control rods shall be demon- l 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 ALTERATIONS or after a reactor i shutdown that is greater than 120 days. l
- 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 e specific control rods, and
- c. For at least 10% of the control rods, on a rotating basis, at least L once per 120 days of POWER OPERATION.
l l 9 LIMERICK - UNIT 2 3/4 1-6
REACTIVITY CONTROL SYSTEMS
- s. ! I
'V' CONTROL ROD 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 not 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 tire exceeding any of the above limits, be in
,q at least HOT SHUTDOWN within 12 hours.
k 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. n LIMERICK - UNIT 2 3/4 1-7
l J REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD GROUP SCRAM INSERTION TIMtS 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 nt>t 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 ar. 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 tedting from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2. O LIMERICK - UNIT 2 3/4 1-8
L , i REACTIVITY CONTROL SYSTEMS I
,m (
CONTROL ROD SCRAM ACCUMULATORS l (V) ,[IMITING CONDITION FOR OPERATION 1 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 one control rod scram accumulator inoperable, within 8 hours:
a) Restore the inoperable accumulator to OPr.RABLE 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 I least one withdrawn control rod at least one notch or place the reactor mode switch in the Shutdown position. p b) Insert the inoperable control rods and disarm the associated 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 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 closing the drive water and exhaust water isolation valves.
- 2. With more than one withdrawn control rod with the associated scram accumulator inoperable or no control rod drive pump oper-ating, immediately place the reactor mode switch in the Shutdown position.
- c. The provisions of Specification 3.0.4 are not applicable.
O *At least the accumulator associated with each withdrawn control rod. Not
'V) applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 2 3/4 1-9
I J REACTIVITY CONTROL SYSTEMS 1 SURVEILLANCE REQUIREMENTS 4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:
- a. At least once per 7 days by verifying that the indicated pressure is greater than or egeal to 955 psig unless the control rod is 1 l
inserted and disarmed or scrammed. 1
- b. At least once per 18 months by:
1 1. Performance of a: I l I a) CHANNEL FUNCTIONAL TEST of the leak detectors, and l l b) CHANNEL CALIBRATION of the pressure detectors, and j verifying an alarm setpoint of 970 + 15, psig on decreasing pressure.
- 2. Measuring and recording the time for up to 10 minutes that each l individual accumulator check valve maintains the associated accumulator pressure above the alarm set point with no control rod drive pump operating.
1 O l l l l I O I l LIMERICK - UNIT 2 3/4 1-10 l a_________
l 0 p, y REACNVITYCONTROLSYSTEMS
- 7 % .
L( ) CONTROL ROD DRIVE COUPLING ! l' V I LIMITING CONDITION FOR OPERATION i
-3.1.3.6 'All control rods shall be coupled to their drive mechanisms.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*. l ACTION: 1 1 i- 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, insert the control rod drive mechanism
.to accomplish recoupling and verify recoupling by withdrawing the control. rod, and:
a) Observing any indicated response of the nuclear instruments-tion, and b) Demonstrating that the control rod will not go to the over-travel position. Otherwise, be in at least' HOT SHUTDOWN within the next 12 hours. 2.- If recoupling is not accomplished on the first attempt or, if not permitted by the RWM, then until permitted by the RWM, declare the control rod inoperable, insert the control rod and disarm the associated directional control valves ** either: D a) Electrically, or b) . Hydraulically by closing the drive water and exhaust water isolation valves. Otherwise, be in at least H0T SHUTDOWN within the next 12 hours.
- b. In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours either:
- 1. Insert the control rod to accomplish recoupling and verify recoup-ling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel position, or
- 2. If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves ** either:
a) Electrically, or b) Hydraulically by closing the drive water and exhaust water isolation valves.
- 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.
**May be rearmed intermittently, under administrative control, to permit testing associated with restoring the cont.rol rod to OPERABLE status.
LIMERICK - UNIT 2 3/4 1-11
i' I REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.6 Each affected control rod shall be demonstrated to be coupled to its I 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. ;
l l 1 0 O LIMERICK - UNIT 2 3/4 1-12
i REACTIVITY CONTROL SYSTEMS x l (j)~ 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 using an alternate method, 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
- 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 RWM, declare the control rod inoperable. b) Greater than the preset power level of the RWM, 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 H0T SHUTDOWN 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 coiltrol rods removed per Specification 3.9.10.1 or 3.9.10.2.
**May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
O LIMERICK - UNIT 2 3/4 1-13
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS l 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 control 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 l rod position indicated by the " Full out" position indicator when performing Surveillance Requirement 4.1.3.6b.
1 0 L $ LIMERICK - UNIT 2 3/4 1-14 u--_---_-.
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i i l REACTIVITY CONTROL' SYSTEMS-I
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77 K.Q) . CONTROL ROD DRIVE HOUSING SUPPORT: LIMITING CONDITION FOR OPERATION l l p
- 3.1.3.8 The control rod drive housing support shall be in place. ;
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
.1
. ACTION:
With the control rod drive housing support not in place, be in at least HOT l p SHUTDOWN within 12 hours and.in COLD SHUTDOWN within the following 24 hours. j p { 1 SURVEILLANCE REQUIREMENTS '! i 4.1.3.8 The control rod' drive housing support shall be verified to be in place j 7-by a visual inspection prior to startup any time it has been disassembled or i when maintenance has been performed in the control' rod drive housing support j
. area. 1 l
l l i 3 l (
- LIMERICK - UNIT 2 3/4 1-15
__ _ _ __m__m__.______m_____.___________m_. _ _ _ _ _ _ _ _ _ , - _ _ _ .
i REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS R00 WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*, **, when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER. l ACTION: i i I l a. With the RWM inoperable after the first 12 control rods are fully withdrawn, operation may continue provided that control rod movement l and compliance with the prescribed control rod pattern are verified by a second licensed operator or technically qualified member of the unit technical staff.
- b. With the RWM inoperable before the first 12 control rods are fully withdrawn, one startup per calendar year may be performed provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or technically qualified member of the unit technical staff.
- c. Otherwise, with the RWM inoperable, control rod movement shall not be permitted except by full scram.***
l
*See Special Test Exception 3.10.2.
** 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.
*** Control rods may be moved, under administrative control, to permit testing associated with demonstrating OPERABILITY of the RWM.
LIMERICK - UNIT 2 3/4 1-16
q
- REACTIVITY CONTROL SYSTEMS i" O l b'd 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 I hour after RWM 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 I hour after RWM automatic initiation when reducing THERMAL POWER, by verifying the rod block funct-ion by demonstrating inability to withdraw an out-of'-sequence i 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. I 3.1.4.2 Deleted.
4.1.4.2 Deleted. i 1 i j l LIMERICK - UNIT 2 3/4 1-17 l
1 REACTIVITY CONTROL SYSTEMS I ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION i l l 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.
l l Otherwise, place the inoperable rod block monitor channel in the l tripped condition within the next hour.
- b. With bc 7 liBM channels inoperable, place at least one inoperable rod l 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 CHANNEL 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.
l O l LIMERICK - UNIT 2 3/4 1-18 l l t__________
REACTIVITY CONTROL SYSTEMS. Cl Y 3/4.1.5 STANDBY LIQUIS' CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1. 5 The standby. liquid control system consisting of a minimum of-two pumps and corresponding flow paths, shall be OPERABLE. 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, restore one inoperable pump and corresponding 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.. In OPERATIONAL' CONDITION 5*:
- 1. With only one pump and corresponding explosive-valve OPERABLE,
. restore one inoperable pump and corresponding explosive valve-to OPERABLE status within 30 days or insert all insertable control rods within the next hour.
- 2. With the standby liquid control system otherwise inoperable, insert all ir.sertable control rods within 1 hour.
. SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:
- a. At least once per 24 hours by verifying that;
- 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 at least ;
4537 gallons.
- 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.
*With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 2 3/4 1-19
1 i REACTIVITY CONTROL SYSTEMS , SURVEILLANCE REQUIREMENTS (Continued) 9<l
- b. At least once per 31 days by:
- 1. Verifying the continuity of the explosive charge.
- 2. Determining by chemical analysis and calculation
- that the l available weight of sodium pentaborate is greater than or equal l to 5389 lbs; the concentration of sodim pentaborate in solution
! is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied: C Q x 13% wt. 86 gpm -,y where C = Sodium pentaborate solution (% by weight) Q = Two pump flowrate, as determined per surveillance requirement 4.1.5.c.
- 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. ;
l c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure 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 i loops, including an explosive valve, and verifying thct a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge fnr 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 3 operating cycles.
- 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.
LIMERICK - UNIT 2 3/4 1-20
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LIMERICK - UNIT 2 3/4 1-21
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3/4.2 POWER DISTRIBUTION LIMITS
.i 3/4.2.1 l AVERAGE PLANAR LINEAR HEAT GENERATION RATE ii
--E LIMITING CONDITION FOR OPERATION o
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 Figures 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. SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2.'l-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 or 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.
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LIMERICK - UNIT 2 3/4 2-6a
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- g. . POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SET 90INTS LIMITING CONDITION FOR OPERATION 3.2.2 -The APRM flow biased neutron flux-upscale scram trip setpoint (S) and
^
flow biaseo neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships: TRIP SETPOINT ALLOWABLt VALUE S < (0.58W + 59%)T S < (0.58W + 62%)T Sf5(0.58W+50%)T R Sj$(0.58W+53%)T R where: S and S are in percent of RATED THERMAL POWER, DB W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million 1bs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY. T is applied only if less than or equal to 1.0. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: With the APRM flow biased neutron flux-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the A)10wable Value column for S or S as above determined, initiate corrective action within 15 minutes
.f' 3 a$,adjustSand/ ors u to be consistent with the Trip Setpoint values
- within 6 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL-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 c 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 THERMAL' POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times 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.
LIMERICK - UNIT 2 3/4 2 7
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 determined using the appropriate figure taken from i shown in Figure 3.2.3-2, provided that Table 3.2.3-1, times the end-of-cycle the K, ion pump trip (E0C-RPT) system is OPERABLE per recirculat l Specification 3.3.4.2, with: 7_ (Iave IB) T ~I A B where: T A = 0.86 seconds, control rod average scram insertion i time limit to notch 39 per Specification 3.1.3.3, N 1 T ] (0.052), B = 0.688 + 1.65[ " N. I 1 i=1 I t ave = i=1 N$ Tj , 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, Tj = average scram time to notch 39 of all rods measured in the i O surveillance test, and N total number of active rods measured in Specification y = 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 2 3/4 2-8
TABLE 3.2.3-1
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s_ / Minimum Critical Power Ratio (MCPR) l l'ersus Plant Operating Condition Rated-Feedwater Maximum Core MCPR Temperature Reduction Flow (% of rated) Figure No. From the Nominal, delta T* ( F) 0 5 100 3.2.3-la 5 60 5 105 3.2.3-1b
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*This delta T refers to the planned reduction of feedwater temperature at rated conditions from nominal rated feedwater temperature during the prolonged re-moval of feedwater heaters from service.
( LIMERICK - UNIT 2 3/4 2-8a
I POWER DISTRIBUTION LIMITS
=r A i G/ 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 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 the appro-priate figure taken from Table 3.2.3-1, for E0C-RPT inoperable curve, times the Kf shown in Figure 3.2.3-2.
- b. With MCPR less than the applicable MCPR limit as identified in ACTION a above, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours or reduce THERMAL F9WER to less than 25% of RATED THERMAL POWER within the next 4 hours.
SURVEl!LAN(,E 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 f(c)- V within 72 hours of the conclusion of each scram 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 the appropriate figure taken from Table 3.2.3-1 times the Kf shown in Figure 3.2.3-2.
- 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 ROD PATTERN for MCPR.
- d. The provisions of Specification 4.0.4 are not applicable.
1 i s LItiERICK - UNIT 2 3/4 2-9
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'lNCLUDES FHOOS 1.22 1.20 0 0.1 0.2 0.3 04 0.5 0.6 0.7 0.8 0.9 1.0 T
DEFINITIONS:
.TCF - INCREASED CORE FLOW (UP TO 105% RATED)
FHOOS - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO 60 DEG. F TEMP REDUCTION; ACHIEVED BY REMOVAL OF FEEDWATER HEATER (S)) FFWTR FINAL FEEDWATER TEMPERATURE REDUCTION AT END OF-CYCLE (UP TO 60 DEG. F TEMP REDUCTION; ACHIEVED BY REMOVAL OF ALL 6TH STAGE HEATERS) Minimum Critical Power Ratio (MCPR) Versus T at Maximum Core Flows 105% Rated and Maximum Feedwater Temperature Reduction s 60*F at Rated Conditions \ l Figure 3.2.3-1b LIMERICK UNIT 2 3r4 210e
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J POWER DISTRIBUTION LIMITS I 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION l l l 1 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kW/ft. 1 APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: 1 1 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 1 reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next l' 4 hours. 1 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.
- d. The provisions of Specification 4.0.4 are not applicable.
l I l LIMERICK - UNIT 2 3/4 2-12
3/4.3 INSTRUMENTATION !! \ 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION lC 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 l 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 I 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 (n) 4.3.1.1 Each reactor protection system instrumentation channel shat? be V 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 19 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.
LIMERICK - UNIT 2 3/4 3-1 i
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TABLE _3.3.1-1 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 - Be in at least HOT SHUTDOWN within 12 hours. . ACTION 2 - Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour. l ACTION 3 - Suspend all operations involving CORE ALTERATIONS and insert l all insertable control rods within 1 hour. ACTION 4 - Be in at least STARTUP within 6 hours. 1 ACTION 5 - Be in STARTUP with the main steam line isolation valves closed l 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. l l ACTION 7 - Verify all insertable ccntrol 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. l LIMERICK - UNIT 2 3/4 3-4 j
TABLE 3.3.1-1 (Continued) REAC'JR PROTECTION SYSTEM INSTRUMENTATION
%.J 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 channel 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 vessel head is removed per Specification 3.10.1. N (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. (j) This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER. (k) Also actuates the EOC-RPT system.
*Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 2 3/4 3-5
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( ) 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION v LIMITING CONDITION FOR OPERA'/ ION 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, I 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.
D 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 Tv.ble 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 condit % n; if both systems have the same numwer of inoperable channels, place either trip system in the tripped condition.
LIMERICK - UNIT 2 3/4 3-9 E_-_____.___.
4 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. l 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months. l 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function l 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 2 3/4 3-10 , i I
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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 % D 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 isclation 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. j i ACTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours or closa the affected system isolation valves within the raxt hour and declare the affected system iaoperable 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.
# During operation of the associated Unit 1 or Urit 2 ventilation exhaust system.
(a) See Specification 3.6.3, Table 3.6.3-1 for primary containment isolation valves which are actuated by these isolation signals. (b) A channel may be placed in an inoperable status fcr 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 outtlo.ard, as applicable, in each line is OPERABLE and all
~
required actuatioin instrumentation for that valve is OPERABLE, one channel mdy 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 secondary containment isolation valves shown in Table 3.6.5.2.1-1 and/or 3.6.5.2.2-1 and signals B, H, S, V, R and T also start the standby gas treatment system. (d) RWCU system inlet outboard isolation valve closes on SLCS "B" initiation. RWCU system inlet inboard isolation valve closes on SLCS "A" or SLCS "C" , initiation. LIMERICK - UNIT 2 3/4 3-16
l i. F TABLE 3.3.2-1 (Continued)
/ ') TABLE NOTATIONS L) '
(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 tne temperature - high setpoint may be raised by 50 F for a period not to exceed 30 minutes to permit restora-tion 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 isolated. (g) Wide' range accident monitor per Specification 3.3.7.5. i l l G l 1 i 1 f% LIMERICK - UNIT 2 3/4 3-17
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TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME (Q)- TRIP FUNCTION RESPONSE TIME (Seconds)#
- 1. MAIN STEAM LINE ISOLATION
- a. Reactor Vessel Water Level
- 1) Low, Low - Level 2 < 13(a)tt
- 2) Low, Low, Low - Level 1 31.0*
b. Main Steam Radiation - Higb Line (b) 5 1.0*/5 13(a),w
- c. Main Steam Line Pressure - Low $ 1.0*/5 13(a),a
- d. Main Steam Line Flow - High 5 0.5*/5 13(a),a
- e. Condenser Vacuum - Low N.A.
- f. Outboard MSIV Room Temperature - High N.A.
- g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
- h. Manual Initiation N.A.
- 2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION c
A a. Reactor Vessel Water Level
-(,-) Low - Level 3 5 13(a)
- b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High N. A.
- c. Manual Initiation N.A.
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. RWCS A Flow - High 5 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 5 13(a)
- f. Manual Initiation N.A.
I i (/ LIhERICK - UNIT 2 3/4 3-23
BBLL3.3.2-3(Continued) , 1501ATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION I
- a. HPCI Steam Line o Pressure - High 1 13(a)
- b. HPCI Steam Supply Pressure - Low $ 13(a)
- c. HPCI Turbine Exhaust Diaphragm Pressure - High N.A.
I
- 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. RCIC Steam Line A Pressure - High 5 13(a)
- b. RCIC Steam Supply Pressure - Low $ 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 - P;gh N.A.
- g. Manual Initiation N.A.
O LIMERICK - UNIT 2 3/4 3-24 I l l
o i TABLE 3.3.2-3f(Continued) (~~'k N ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME
' '/
TRIP FUNCTION RESPONSE TIME (Seconds)#
- 6. PRIMARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level '
- 1) Low,' Low - Level 2 < 13(a)
- 2) Low, Low, Low - Level I h13(a)
- b. Drywell Pressure - High < 13(a)
- c. North Stack Effluent Radiation - High N. A.
- d. Deleted
- e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High N.A.
- f. Outside Atmosphere To Reactor Enclosure A Pressure - Low N. A.
- g. Deleted f'~'g h. Drywell Pressure - High/
\ / Reactor Pressure - Low N. A. %/
- i. Primary Containment Instrument Gas to N.A.
Drywell A Pressure-Low
- j. Manual Initiation N.A.
- 7. SECONDARY CONTAINMENT ISOLATION
- a. Reactor Vessel Water Level Low, low - Level 2 N.A.
- b. Drywell Pressure - High N.A.
c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High N.A.
- 2. Refueling Area Unit 2 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.
f i
\
LIMERICK - UNIT 2 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. Reactor Enclosure Manual Initiation N.A.
- h. Refueling Area 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 2 3/4 3-26
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) L d L e I u E L n V LA i R EN t U NOT n S NIS M M M M M M R M o ATE C N HCT
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L WQ E ARR NO OFE 33 3 3 3 3 3 3 3 I C TSN ,, , , , , , , , ANA 22 2 2 2 2 2 2 2 ROL EIL ,, , , , , , , , PTI 11 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 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 R o ATE C N HCT ( O CN I U 1 T F
- A 1 T
. N 2 E
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) L d L e I d u E L n V LA d i R EN t U NOT n S NIS M M M M M M M R R o ATE t n
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I C i i e T t t n s A ah ah nh r i U l g l g og o l e - T C A N N O I T l e v e h ii tH V n e-ii tH V n e-ii tH a l - i t R cw ao eL f R e u e l a h t i n O A L g n n t n - u no o ow n l - 1 o 2 o I L i l T O r H i i ei Te To a a e A S e tt tt Vt r L M u u L I t2 - i a i a a eu e n f _ O a ni ni ei rs r- e a _ S T Wl e Ud Ud rd es e r M d I N e r a a ua h e h e u e E l v u aR aR sR pr pr s a t _ M ee s e e o sP su o e a N sL s rt rt l t o os l r i ms _ I s e Ac Ac cc mA c A d A e- r u u nu t t e nn n a _ T V P gD gD ED Ae Ar E o go r _ N w n n r P i ni r ro eu e O l it it rt rt i t i C oL l l s l s os ds dA oa l a t e eu eu t u io i ti ei g Y c , w ua ua ca sl sa ct ut n N O R A aw eo y r fh ex fh ex ah ex t c un t e ur ai en fi en i l I D RL D RE RE RE OE OA RI RI d T N n C O . . h a N C 1 2 U E . . . . . . . . s] F S a b c d e f g h n e P h I W e R .
- T 7 CM5R -
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i 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 SHUTOOWN 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 j 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 l 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. O1 LIMERICK - UNIT 2 3/4 3-32 l L__ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - -.
T 00 123 0011 3 445513 C 33 333 3333 3 333333 A 5 5 5 5 5 L * * * *
- EAS 4 4 4 4 4 LNN BOO , -
AII 33 3 - 3 3333 3 333333 CTT ',5
- IAI , , , , , , , , , , , , , ,
LRD 22 2 2 2222 2 222222 PEN ' PPO , , ,* , , , , , , , , , , , AOC 11 141 - 1111 1 111111 N O E I L - T BR A AE ) T RP a pp N E ( m E PS N mm uu) e v e M OLPO ) ) ) t U EII ppb
- l c d s R MNRT //( ( a ( ( y T UNTC 22 6 2 222v 1 44224s S MA N / /
N IH U 1 1 I NC F I N M O I 1 1 T A l
- l 1 U e
- e
- T v
- v 3 C e M e 2
. A L E l 3 ) T ) ) l 8
. M , e S ,
ewe e 3 E w v Y w vov v l T o i S o i Li e e E S L s L s - s L v L Y s R ses e B S w i H w i ri w - L A o m r R o mum rsr UM o wh og T G L L L , N e F ese Lih Hg I w P O w P eP E w L o ( o ( r( T o - i O L E L P S L H O w D w Y l C - o O - ol S - el - L M L a ve E l l i N l evl R e - N e - t O e L ee O v O v n I v L v C eh e I eh ee T ehk e L g r T L grr C L gnrL Y i u C iue E iae C rH s E rHsf J rHTt r N e s J e sf N e ae E t - e N t - ei I t - eWt G a r n I a rD n a g an R We P o WeP o T Weal Wo E
- r i T r e i N rro i M
- l u l t N l ul v t A l uool t E
- es e a A esel a L estP ea M ss s i L sssa i O ssS si E s e s t O sesV t O se nst T er e i O ere i C ereoei S VP V n C VPVn n VPtiVn Y I o I E as I S rl r E rl ri R rl ssr ol o l R ol ot l U ol neol Y t e t a U t et c a S t eert a A
R cw ay a c u n S S cwce ayaj u n S E cwd pcu aynpan P er e a E eren a R erouea S RD R M R RDRI M P RDCSRM P E H O N R O . . . . W O . . . . G I . . . . . . I C ab c d L ab cd. e H abcd ef T C N U . . . F 1 2 3 P I R T
- E i e% N { T$
l
C 3333 3333 O A I T 6 7 C 3 3 A L 5 5 EAS LNN , , BOO *
- AII 3333 3333 L *
- CTT EAS 4 4 IAI , , , , , , , , LNN LRD 2222 2222 BOO , ,
PEN AII 3 3 PPO , , , , , , , , CTT AOC 1111 1111 I AI , , LRD 2 2 PEN PPO , , N AOC 1 1 O E I L T BR / A AE ) SE e T RP a MLL c r N E ( UEB E PS N MNA s u M OLPO I NR u o U EII NAE b ss R MNRT 2212 4122 I HP / /u T UNTC MCO 1 1b S MA N N I H U ) ) I NC F e e / .
) I v v e m d N M i i S c e e O s s LP r t u I 1 s s EI s u s n T i i NR u o y i A l m m NT b ss s t U e r r A / / u b n T v e e HO 1 1b u o C e P P CT s C A l (h (
( g r M , hi 3 ) e 1 E w gH f p
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3 S L H e OS s
. Y e v NL i 3 S w - r e E n
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3 G L es AN e o N rs , r TA c i E I w ue w e OH r t L L o sr o m TC s u c B O L sP L i u o n A O e T F b ss u T C - re - 0 / /u F
- P g s 1 1b E
- l r l s p R
- e ea e a i O # v gh v p r C M eh rc e y - - T E lg as L B r r Y T i hi e e) r C S rH cD r e d)d e e N Y e s e r neng P E S t - ip t u U gU a G a D m ans a t s R
E N O Wer pP u Wos ie st sl ul uo l e M I l u m )l t r B oBV n E T es ueeeaP V n A ss Pd vsi y yd a Z se oistl cf ce h I er yM s e il nond C R VP ra sVne e gsgr ea U erIi I w E S rl m pC mr y rsrg L S ol iSP rol r eoee B E t eT L et aD mL mD A R P cw a yS e rR( anS P cu R E(E( R E E erd oH eaD E V eve P O RDACR RMA W k gk g O O a a C P 6t6t m N I 1l 1l u O T . . . . . . . . F . o .o m I A abcd e f gh O 4 v4 v i n T M C O S i N T S M U U O . . F A L 1 2 e h P T I
- R . .
- T 4 5
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TABLE 3.3.3-1 (Continued)
,/ m -l EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE NOTATIONS j (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 channel in the same trip system i is monitoring that parameter. I (b) hiso provides input to actuation logic for the associated emergency diesel generators. (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 used 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 taken together for Item 2.
-s 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.
,~
\
l C LIMERICK - UNIT 2 3/4 3-35 L_ _ _ _ _ _ _ _ _
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 l in the tripped condition within 1 hour or declare the l
associated system inoperable.
- b. With more than one channel inoperable, declare the associated system inoperable.
l 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 { I 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: Ol<I l
- a. For one channel inoperable, place the inoperable channel <
in the tripped condition within I hour or declare the HPCI I 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 Trip 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 einergency 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 bour; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST. l l O LIMERICK - UNIT 2 3/4 3-36
i i s i i , s sp ss a a aa e e4 s ee r r8 e rr c c h cc d e d e<- sn c nn i i s ( s (d ei s s(( s e e n s h e d e h g , h g ,a egc3 s h gn , ,h cig ci g hi n e ci o g g c E nsi nsid csith n s cii n L ips p i psi np ec i pessi BE ps i 3 en spp AU 68 68 p 8 .fi 68 0 WL 5 385 OA LV 3 1 8.3
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i pss np ec i pessi R E pp i 8en spp T S 98 98 8 .fi 98 5 S 265 2654 867 26555 . N P 1 5 . 1 57 . 3 644 . 1 0422
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S , L v h g A E w Y w e l l gi T T o S o - L e e iH: S L L v v H - Y R e , e e el S w H w r w l L - re o R o u o wh uv G L L s L og , , es e N F s M Lih w rsL I w O w e E w H g o ue L o o r T o - i L sr , O L E L P S L H sP w O w D w Y l w e o C - o O - ol S - el - M o reL L M L a ve E L P g-E l l i N l evl T rl R e - N e - t O e L ee S w eae O v O v n I v l v Y o gh v C eh e I eh ee T ehk e S Lh g rce L gr T L grr C L gnrL asL Y iu C iue E iae N i hi C rHs E rHsf J rHTt r O H cD r N e s J e sf N e ae I l s e E t - e N t - ei I t - eWt T e- i pt G a rn I a rD n a g an A v Dmans R E WeP o i r T WeP r ei o T N Weal Wo Z i rro I ee L r pP uWoi M l ul t N l ul vt A l uool t R u m l t E esea A esel a L estP ea U r. ueeaP M sssi L sssai O ssS si S er Pd si E sest O sesVt O se nst S t e ostl T erei O ere i C ereoei E ar yM e il S VPV n C VPVnn VPtiV n R WP ra Vne Y I oI E as I P erI I S rl r E rl ri R rl ssr E rl m pC r ol ol R ol otl U ol neol D ol iSP ol r Y t et a U t et ca S t eert a t eT Lt aD N A R cwcu ayan S S cwceu a,ajn S E cwd pcu C aynpan I cw a yS e rR anS cu O P erea E erena R erouea T erd oH eaD I S RDRM R RDRI M P RDCSRM A RDACRRMA T P M C E H O N R W G T U O . . . . O . . . . I . . . . U . F C ab cd L ab cde. H ab cd. ef. A ab. cd . . . ef. g P I R . . . . T 1 2 3 4
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- X s oO T
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- T 1 1 1 1 1 1 1 1 1 t 3C Y 1 Y 1 1 - - - - - - l
.A A - A - - Y Y Z Y Z Z o 3 L 7 L 7d2 7d7 7d2 7d2 v
.M E 2 E 2 n0 2 n2 2n6 2n6 3E R 1 R 1a1 1a1 1 a1 1a1 e T h ES T LY BS A
TG y N e e a I g g l L a a e O t t r O l l C o o e v v g E r r a R e e t O d d l C n n o U U v Y C s s y N u u a E B B l G ) R y) ye d E ce cg M ng na E ea et gt gl i rl ro t eo eV mV m E Ed R f e E Vo Vd W k k a O s r P 6 s 6g i N 1o 1 e O F .L .D I O 4( 4( T C S N s i U o . . F L a b i P I h R . T 5
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!J i TABLE 3.3.3-3' M~ ' ' '
- j. EMERGENCY' CORE' COOLING' SYSTEM' RESPONSE TIMES
'ECCS RESPONSE TIME (Seconds)-
'..t .' CORE SPRAY' SYSTEM- L127 i
- 2. LOW PRESSURE COOLANT INJECTION MODE.
CF RHR SYSTEM. 5 40 0
- 3. AUTOMATIC DEPRESSURIZATION SYSTEM- N.A.. ,
4.. HIGH PRESSURE COOLANT INJECTION SYSTEM- < 30
- 5. LOSS OF POWER N.A.
1 3
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l i l i I i
)
i I l 1 1 LIMERICK - UNIT 2 3/4 3-39
HD CE IR HI WU L Q * * * * ** ARE 5 55 5 55 NOR S OF , , , , , , T I E * ** * ** 4
- TSC 4 44 4 44
_ E ANN _ M ROA , , , , , , E EIL 3333 333 33 33 3 3 33 R PTL _ I OII _ U DE 2222 222 22 22 2 2 22 Q NV E OR , , , , , , , , , , , , , , ,
?< CU 1111 111 11 11 1 1 11 S
E C N A L E L LI N O F.RRN A. RRR RN A. RR R R R A. _ V ET N R NA _ U NR _ S AB _ HI _ N CL _ O A I C T A _ T _ N _ E V L 1 U LA _ - R EN 1 T NO
. S NIT 3 N ATS
_ . I HCE 3 CNT MMMR MMM MR MM M M MR N U 4 O F _ I M E T E L A T B U S A T Y T C . S . . A L A A EK . R A. . M NC SSSN H SS5 SN
- SS S 5 SN E NE R
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I INSTRUMENTATION j 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 (ATW5-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. 4 APPLICABILITY: OPERATIONAL CONDITION 1. l ACTION: l 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. l 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 I 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 system 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 q 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 2 3/4 3-42
TABLE 3.3.4.1-1 v3 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION -(L-[- MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP SYSTEM *
- 1. Reactor Vessel Water Level - l Low Low, Level 2 2
- 2. Reactor Vessel Pressure - High 2
('~'s
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v: i I 1 l
*0ae channel may be placed in an inoperable status for up to 2 hours for required surveillance provided the other channel is OPERABLE. ~
LIMERICK - UNIT 2 3/4 3-43
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 > -38 inches *
;; -45 inches
- 2. Reactor Vessel Pressure - High i 1993 psig 5 1108 psig l
l l O l *See Bases Figure B3/4.3-1. l O l LIMERICK - UNIT 2 3/4 3-44 l _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ ____ __. I
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TABLE 4.3.4 1-1-.
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ATWS RECIRCULATION' PUMP TRIP ACTUATION INSTRUMENTATION
,\
n. SURVEILLANCE REQUIREMENTS (, - I d.i. . CHANNEL- CHANNEL FUNCTIONAL - CHANNEL'
' TRIP FUNCTION < CHECK ' TEST CALIBRATION h 1. Reactor Vessel Water Level -
Low Low, Level 2 S M .R
' 2. . Reactor Vessel Pressure - High S M R
, 1 1
6-i LIMERICK - UNIT 2 3/4 3-45 1'
; a .. .
INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of cycle recirculation pump 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-0F-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 Specificatica 3 2.3.
f I Ol LIMERICK - UNIT 2 3/4 3-46 I I
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- 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 CHAMNEL 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-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of- 1 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 measured time shall be added to the most ; recent breaker arc suppression time and the resulting END-OF-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 I at least once per 60 months. 1 i 1 i O LIMERICK - UNIT 2 3/4 3-47
l LADLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM { OPERABLE CHANNELS l TRIP FUNCTION PER TRIP SYSTEM
- l
- 1. Turbine Stop Valve - Closure 2**
- 2. Turbine Control Valve-Fast Closure 2**
1 1 ? j l 9 l
*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 i O LIMERICK - UNIT 2 3/4 3-48
TABLE 3.3.4.2-2 [) 'w-END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS ALLOWABLE l TRIP FUNCTION TfQPSETPOINT VALUE
- 1. Turbine Stop Valve-Closure 5 5% closed 5 */% closed 1
- 2. Turbine Control Valve-Fast Closure 1 500 psig 1 465 psig O
O LIMERICK - UNIT 2 3/4 3-49
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 i 175
- 2. Turbine Control Valve-Fast Closure 5 175 l
O O LIMERICK - UNIT 2 3/4 3-50
- a. :: , :
1 [:- ,
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p- . TABLE 4.3.4.2.1-1 (eg.:_
~/
j ;. END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS
! CHANNEL FUNCTIONAL CHANNEL'
. TRIP FUNCTION TEST CALIBRATION
- 1. Turbine Stop. Valve-Closure M* 'R
~
- 2. Turbine Contro1 Valve-Fast Closure -M* R 4
- Including trip system logic testing.
'lV 1
LIMERICK - UNIT 2 3/4 3-51 ! i L--_.-_-._----_-.____--_-.___._-_
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 cnannels shown in Table 3.3.5-1 shall be OPERABLE with their i trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2. - l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. < ACTION: I
- a. With a RCIC system actuation instrumentation channel trip setpoint i less conservative than the value shown in the Allowable Values i 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 i inoperable, take the ACTION required by Table 3.3.5-1. l 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 2 3/4 3-52
'l- h
- TABLE 3.3.5-1 j% .
F( ) REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION-
%)
MINIMUM OPERABLE CHANNELS FUNCTIONAL UNITS PER TRIP FUNCTION
- ACTION
. a. ' Reactor Vessel Water Level -
Low i nwe 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
*A channel may be placed in an inoperable status for up to 2 hours for.
(j, required survelliance 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.
LIMERICK - UNIT 2 3/4 3-53 l
l I ( 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 l 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 i system inoperable.
- b. With more than one channel inoperable, declare the RCIC system inoperable.
1 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 conditior within I 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 2 3/4 3-54 l J
TABLE 3.3.5-2
,x REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETP01 HTS
')
%J ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT VALUE
- a. Reactor Vessel Water Level - ;
Low Low, Level 2 >-45 inches
>-38 inches *
- b. Reactor Vessel Water Level -
High, Level 8 1 54 inches 5 60 inches
- c. Condensate Storage Tank Level -
Low > > _ 132.3 inches _ 135.8** inches l
- d. Manual Initiation N.A. N. A.
*See Bases Figure B 3/4.3-1.
** Corresponds to 2.3 feet indicated. j i
j i (h j l i
's_/ j i
LIMERICK - UNIT 2 3/4 3-55
1 1 TABLE 4.3.5.1-1 REACTOR CORE ISOLATION SYSTEM ACTUATION INSTRUMENTATION I SURVEILLANCE REQUIREMENTS i CHANNEL CHANNEL FUNCTIONAL CHANNEL i FUNCTIONAL UNITS CHECK TEST CALIBRATION i
- a. Reactor Vessel Water Level - 1 Low Low, Level 2 S M R
]i
- b. Reactor Vessel Water Level - l High, Level 8 S M R l 1
- c. Condensate Storage Tank j Level - Low S M R
- d. Manual Initiation L.A. R N.A.
O l O LIMERICK - UNIT 2 3/4 3-56
INSTRUMENTATION e, .
'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. _ I ! l APPLICABILITY: As shown in Table 3.3.6-1. l ACTION: i
- 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 Minim'um OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.
.SURVE1LLANCE 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 l 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. I
-l l
l l l
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LIMERICK - UNIT 2 3/4 3-57 _ _ ___ m.____--_____-_________-_-________-______m___m_.__ _ _ - _ _
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. TABLE 3.3.6-1 (Continued) .j CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION
- 1 QJ -1 ACTION STATEMENTS ACTION 60 -
Declare the RBM inoperable.'and take the ACTION required by Specification 3.l A.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 j Minimum OPERABLE Channels pet 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.
[ With more than one control rod withdrawn. Not applicable to control rods ( removed per Specification 3.9.10. r- 3.9.10.2.
*** These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.
(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 2 3/4 3-59 L-__
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. . . A A.A.A NNNN.
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. . A.A A.A NNNN.
. A. W A.
O A N L NNN K HC - F C C O N L p O B u I t T D r A O a L R - t U S C L x R O u , S I R l e R C T F l O E N a n T n R O n c i I i C o s N E M r p S l O l M E t U R l M l U h T u O u u L g S e - T f E f O i Y R N I G V H S O x N t N t - T e d e u O o e A o e E l T e I v e v l M n v R n v G e N v N i e s ei e F i e ie R v A io O tl al tl E r tl E r tl A e L tt M eaa iaaan G oeaa T oeaa H L O eaa l rc Bcrco N tl rc A tl rc C C l rr N K aes sesr A caes I caes S r C aea O C cpn wppnt R ecpn D ecpn I e cpp I T L O sow pno l oUowu noe E epno M epno t s ow E t sow D t R som a O pno C B UID F I DN C DUID R DUIL M W T UIC N M R E A C U D R U T R A F O . . . P . . . . O . . . . N . . . . C . E . . . R ab c A a b cd S ab cd I abcd S a R abc P
~ I R . . . . . .
T 1 2 3 4 5 6 yEgMx
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TABLE 4.3.6-1 (Continued) CONTROL ROD 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 I rods removed per Specification 3.9.10.1 or 3,9.10.2.
1 O i i 1 O LIMERICK - UNIT 2 3/4 3-62
1 l i INSTRUMENTATION l i ' (r3 3/4.3.7 MONITORING INSTRUMENTATION ' RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The ridiation monitoring instrumentation channels shown it: 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 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 Specification 3.0.3 are not applicable.
O V 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. O LIMERICK - UNIT 2 3/4 3-63
N O I 0 1 3 2 9 T 7 7 7 7 C A
)
b ( _ h
/ )
_ R b m ( - c c 0 2 d n u
/
i d o r _ C n g P p a k IT c RN TI
/O MP RT 5
0 1 h
/
R m
)
b ( B a x _ AE x 5 A. 3 _ LS _ A 1 >, N 5 . N s s - O e e - I ES m m _ T LN i i A BO T T T N AI CT 5, l l _ E M II LD 3,
- l A
l A 1 U R PN PO 2, dn ) a t t AC 1a ( A A
- T
_ 1 S
. N 7 I 3
3 E L B G N I R O T I S 9 A N L T O E M N N N AE O HL I CB T A A 4 2 1 1 I U M D A R MM I N I M n g o s l i r t n at o c i r ea ri t i l o ro er l o ot od n l o it oi N a o eP Di C n R M u n t m F e mo eM oy y g oM r ol t t a o un R p i nr R n so p s l eo o oi l u r a pt l i l t oS o c SS ot ca N r t i ra ni O t r i t ti Ed I ni n i nd a T oar o o r ) oa rR o - A C M C 1 CR - T ht t r N nsi a ce E ien e at M aro r . . ea U MFM A a b RW R T S N I 1 2 3 CjMn i *" { TE
f TABLE 3.3.7.1-1 (Continued) 7m l ) RADIATION MONITORING INSTRUMENTATION L./ 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 s.t'ERABLE 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 operrtion. ACTION 71 - With one of the required monitor inoperable, assure a portable ! continuous monitor with the same alarm setpoint is OPERABLE in l the vicinity of the installed monitor during any feel movement. ; If no fuel movement is being made, perform area surveys of the j (^s monitored area with portable monitoring instrumentation at least once per 24 hours. I
~
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. l 0 ( LIMERICK - UNIT 2 3/4 3-65
OA a s s LFLD e e A LE 5 m i m NSIR i ONEI , T T IOVU 3 l TIRQ l ATUE , l l RISR 2 A A ED ) PNHS , a t t O O C I. 1 ( A A CI H W
)
S N b T O ( N I R R R R E LT M EA E NR R NB I AI U HL Q CA E C R E C N A L L LA L EN I NOT E NIS M M M M V ATE 1 R HCT
- U CN 1 S U
. F 7 N
. O 3 I T
- 4. A T
E N L L E EK B M NC S S A U NE S S T R AH T HC S C N I e G g N n a g I o s r R l i r o t n O at o t c ir T ma ri t i S er ro l o ot od n l it oi
*L Na R M o e u
Di n C n o l m F mo eM r N oy y oM O ol t t o un I R p i nl R n so T p s l eo o oi A l u r a po li l t I oS o c SP ot ca D N r t i ra ni A O t r i t ti Ed R I ni n i nd a T oar o r ) oa rR A C o M C 1 CR o T ht t r N nsi a ce E i en e at M aro r . . ea U MFM A a b RW R T S _ N . . . 3 2 I 1
- 9 i c* " ga uj s l
TABLE 4.3.7.1-1 (Continued). O b ( RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
.g) .
TABLE NOTATIONS
~ *When irradiated fuel is being handled in the secondary containment.
(a) With fuel in the spent fuel storage pool. ; (b) 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 .
-i 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. ; i l i l O LIMERICK - UNIT 2 3/4 3-67
INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION l 3.3.7.2 The seismic monitoring instrumentation shown in Table 3.3.7.2-1* shall be OPE 9A9LE. APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required seismic monitoring instruments
} inoperable for morr: than 30 days, prepare and submit a Special Report I to the Commission pursuant to Specification 6.9.2 within the next j 10 days outlining the cause of the malfunction and the plans for 1-restoring the instrument (s) to OPERABLE status.
- b. The provisions of Specification 3.0.3 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 CH'.riNEL CHECK, CHANNEL FUNC-TIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.2-1. A.3 7.2.2 Each of the above requir:M seismic monitoring instruments which is accecssible during power operations and which is actuated during a seismic eve it 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 determire the magnitude of the vibratory ground motion. A Special Report shall be prepared and submi&d to the Commission pursuant to Specifica-tion 6.9.2 within Ifi days denribing the magnitude, frequency spectrum and resultant effect upor: unit teatures important to safety. Each of the above seis.r.ic monitoring instruments which is actuated during a seismic event greater than or equal to 0.01 g but is not accessible during power operation shall be restored to OPERABLE status and a CHANNEL CALIBRATION performed the next time Unit 1 enters OPERATIONAL CONDITION 4 or below. A supplemental report shall then be prepared and submitted to the Commission with 14 days pursuant ;o Specification 6.9.2 describing the addi-tional data from these instruments.
- Shared with Unit 1.
LIMERICK - UNIT 2 3/4 3-68
m s i TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION Of d MINIMUM L MEASUREMENT INSTRUMENTS INSTF.UMENTS 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 Equipment (RHR Heat. Exchanger, N
Loc. 102-R15-177)
- 6) XRSH-VA-107* Foundation of an 0 to 1 g 1 Independent Seismic Catregory I Structure (Spray Por.d # amp House, El 237')
- b. Recorders (Panel 00C693)
- 1) XR-VA-102 for XE-VA-102 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.
LIMERICK - UNIT 2 3/4 3-69
I i l
)
TABLE 3.3.7.2-1 (Continued) SEISMIC MONITORING INSTRUMENTATION MINIMUM 4 MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE i
- c. Triaxial Seismic Trigger (S/T)
]
- 1) XSH-VA-001 (Activates Items NAt 1 1.b.1) thru 5) above (Loc. Area 16, El 177')
- 2. Triaxial Peak Recording Accelerograph (P/A's) <
f a. XR-VA-151 Reactor Equipment 0-2g 1*** i (Top of reactor vessel head)
- b. XR-VA-152 Reactor Piping 0-2g 1 (Mn. Stm. Li ne ' D , ' El 313 ' ,
in containment)
- c. XR-VA-153 Reactor Equipment Outside 0 -2g 1 Containment (RHR Heat Exchanger, Loc. 203-R15-201)
- 3. Triaxial Seismic Switch l
l
- a. XSHH-VA-001 Primary Containment NAtt 1*
Foundation (Loc. 118-R16-177) Triaxial Response Spectrum Analyzer 1*, **
- 4. 1-33.5 Hz (RSA); (Loc. Control Room) tThe Triaxial Seismic Trigger setpoint is 0.005g.
ftThis switch triggers at < 0.15g horizontal and < 0.10g vertical.
*With reactor control room indication and annunciation.
** Receives signal from playback unit fed with data from the Triaxial Accelerographs, Item 1.a above.
***Not required to be OPERABLE when the Unit I reactor vessel head is removed.
l LIMERICK - UNIT 2 3/4 3-70
TABLE 4.3.7.2-1 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-R11-177)
- 4) XE-VA-105 Reactor Piping N.A. SA R 3upport (Mn. Stm. Line 'D,'
El 313', in containment) 5)' XE-VA-106 Outside Contain- N.A. SA R ment on Seismic Category I
.g)i Equipment, (RHR Heat Exchanger, Loc. 102-R15-177)
%_/
- 6) XRSH-VA-107* Foundation of N.A. SA R an Independent Seirmic Category I Structu' e (Spray Pond Pump House, f.1 237')
- 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 fur XE-VA-105 N. A. SA R
- 5) XR-VA-106 for XE-VA-106 N.A. SA R
- Includes sensor, trigger, recorder, and backup power supply.
s LIMERICK - UNIT 2 3/4 3-71
TABLE 4.3.7.2-1 (Continued) SEISMIC 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 (Activates N.A. SA R Items 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. 5tm. Line 'D,' El 313',
in containment) l c. XR-VA-153 Reactor Equipment N.A. N.A. R l Outside Containment (RHR Heat i 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 2 3/4 3-72
INSTRUMENTATION 7-,.q METEOROLOGICAL MONITORING INSTRUMENTATION
-{\_ /n . -
I LIMITING CONDITION'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 Specification 3.0.3 are not applicable.
'~'}
v SURVEILLANCE REQUIREMENTS l 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. I i
/~
i e LIHERICK - UNIT 2 3/4 3-73 _ _=
TABLE 3.3.7.3-1 j METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM 1 Tower 1 Tower 2 INSTRUMENTS i INSTRUMENT (Primary) gackup) OPERABLE I
- 1. Wind Speed
- a. Elevation 1 30 feet or 159 feet 1
)
i
- b. Elevation 2 175 feet or 304 feet 1 {
- 2. Wind Direction j
- a. Elevation 1 30 feet or 159 feet 1 l 3
- b. Elevation 2 175 feet or 304 feet 1 1 1 3. Air Temperature Diffe/ence
- a. Elevations 266 feet- 300 feet-26 feet or 26 feet 1 O
O LIMERICK - UNIT 2 3/4 3-74
y m l
, . i
' TABLE 4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
' \/
1% CHANNEL CHANNEL. ! INSTRUMENT CHECK CALIBRATION
- 1. ; Wind Speed p
; a. Elevation 1 (Tower 1 and Tower 2) D SA~ ;
- b. Elevation 2 (Tower 1 and Tower 2) D SA
.2. -Wind, Direction Elevation.1.(Tower 1 and Tower 2)
- a. . D SA
- b. . Elevation 2 (Tower 1 and Tower 2) D SA
- 3. Air. Temperature Difference
- a. Elevations 266 - 26 ft (Tower 1)~ D SA 4 .b. Elevations 300 - 26 ft (Tower 2) D SA k
l l l LIMER.1LK - UNIT 2 3/4 3-75
INS 1RUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS LIMITING CONDITION FOR OPERATION 3.3.7.4 The remote shutdown system instNmentation and controls shown in . Table 3.3.7.4-1 shall be OPERABLE. ! APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. l 1 ACTION: I
- a. With the number of OPERABLE remote sht,tdown system instrumentation I channels less than required by Table 3.3.7.4-1, restore the inoperatle 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 remote shutdown system controlc 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. O LIMERICK - UNIT 2 3/4 3-76
Q S e T v N l EE a MML v UUB / MRA 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 ITR NSE I NP MIO
)
0 0 S 1 L O - R T 0 N ( O C n o D i N t A a c N i O d I n T I n A o 1 T n i
- N o t 4 E i a
. M t c 7 U i i
) . R s d
/ \ w3E 3 T S N I e r e P
)
A o n o I n n o L M u r 8 i i B E s u 4 t t A T s s 0 a n a T S e s F c o i Y r e 2 i i d S e P r - d t a r P 1 n a R N u t 5 I c W s s e e - i h 32 e e s l g C w d g 1 v r e t r - o n i T l u r u a V L I H U a t P O h H H v a c ( e h e S r e r s r g n 3 l e g e i e n u i i E e p r t D v o s H L T , v m a a l l i s O n e e h W p e a t e e e g M o L T c m v V a r r E l i s e u e c P u r R e t r r i c P L s i t a v i e e D i s d l a h e e s t t v r k a n i r c r L o a a p r e n p I O e s u PWW m e t a y p i s r u S a T B d g m D s e e r r e P W e n e e t v e e r r d e r p i T r r a l b b u r e e e g e p r e P W a m m e t e g w e c a g i a l g V a a r a w t n o p i r n r e i n l l h h u r o a a l S v o a T B O a e e f C C s e l W h F r t h h s s e s p F c e e S c e e g c s s i n n e m e x m n S x n n n x e e l o o r e m c E e i e E i i i E V V e i i P T e i t b y t b b r R s s t v t s r c a t r r a t r r / s s l l s r a y u n s a u u e a o o y e e l l y e e S T e n e T T B e T t t t r r e e S S H g e H H N c c e p p w w C C r d C C C E a a f p p y y R R R I I e n R I I I R M e e a u u r r H H H C C m o H C C C H U R R S S S D D R R R R R E C R R R R R R % T ~~ s S . . . . . . . . . . . . . . . . . . N I 1 2 3 4 5 6 7 . 8 9 101 11 1 21 131 141 51 6 7 8 9
- Eig[ ,1 "4"
TABLE 3.3.7.4-1 (Continued) REMOTE SHUTDOWN SYSTEM CONTROLS RCIC SYSTEM HSS-49-291 Control-Transfer Switch HSS-49-292 Control-Transfer Switch HSS-49-293 Control-Transfer Switch HSS-49-295 Control-Transfer Switch HSS-49-296 Control-Transfer Switch I l HV-49-2F076 Control-Steam Line warmup bypass valve HV-49-2F060 Control-RCIC turb exhaust to suppression pool i isolation HV-50-212 Control-Turb trip throttle valve HV-50-2F045 Control-Turbine steam supply valve l HV-49-2F008 Control-Turbine steam line outboard isolation valve HV-49-2F007 Control-Turbine steam line inboard isolation valve HV-49-2F031 Control-RCIC pump suction from suppression poc1 HV-49-2F029 Control-RCIC pump suction from suppression pool
- HV-49-2F010 Control-RCIC pump suction from condensate storage
) tank HV-49-2F019 Control-Minimum f ow bypass to suppression pool HV-49-2F022 Control-Test return to condensate storage tank l HV-50-2F046 Control-RCIC turbine cooling water valve HV-49-2F012 Control-RCIC pump disch valve HV-49-2F013 Control-RCIC pump disch valve 20P220 Control-Vacuum tank condensate pump 20P219 Control-Barometric condenser vacuum pump HV-49-2F002 Control-Barometric condenser vacuum pump disch l 9 LIMERICK - UNIT 2 3/4 3-78
, 9' i
.)
1 < , Table 3.3.7.4-1 (Continued) ( ' RCIC SYSTEM (Continued)-
.HV-49-2F080L Control-Vacuum breaker outboard isolation valve ,
1 HV-49-2F08'4E Control-Vacuum breaker: inboard isolation valve -
,: .FIC-49-2R001- Controller-RCIC discharge flow control-7 E51-S45. RCIC Turbine Trip Bypass I LN_UCLEARBOILER5YSTEM~
'HSS 41-291'- Control-Transfer ~ switch PSV-41-2F013A Control-Main steam line safety / relief valve PSV-41-2F013C' Control-Main s' team line safety / relief valve PSV-41-2F013N Control-Main steam line safety / relief valve.
RHR SYSTEM HSS-51-195 Control-Transfer switch HSS-51-196 Control-Transfer switch HSS-51-292~ Control-Transfer switch HSS-51-293 Control-Transfer switch HSS-51-294 Control-Transfer switch HSS-51-295 Control-Transfer switch HSS-51-296. Centrol-Transfer switch HSS-51-297 Control-Transfer switch HSS-51-298 Control-Transfer switch HV-51-2F009 Control-RHR pump shutdown cooling suction inboard isolation HV-51-2F008 Control-RHR shutdown cooling suction outboard isolation l HV-51-2F006A Control-2A RHR loop shutdown cooling suction HV-51-2F006B Control-28 RHR loop shutdown cooling suction
'HV-51-2F004A Control-2A RHR pump suction 2AP202 Control-2A RHR pump LIEiERICK - UNIT 2 3/4 3-79
Table 3.3.7.4-1 (Continued) { RHR SYSTEM (Continued) HV-43-2F023A Control-Recirculation pump A suction valve HSS-43-291 Control-Transfer switch j HV-51-2F007A Control-2A RHR pump minimum flow bypass valve HV-51-2F048A Control-2A heat exchanger shell side bypass HV-51-2F015A Control-2A shutdown cooling injection valve ! I HV-51-2F016A Control-Reactor containment spray I HV-51-2F011A Control-2A heat exchanger flow to suppression pool ! i L HV-51-2F017A Control-2A RHR loop injection valve j l l HV-51-2F024A Control-2A RHR loop test return HV-51-2F027A Control-Suppression pool sparger isolation HV-51-2F047A Control-2A Heat exchanger shell side inlet HV-51-2F003A Control-2A Heat exchanger shell side outlet HV-51-2F026A Control-2A Heat exchanger flow to RCIC HV-51-2F049 Control-RHR Discharge to radwaste outboard isolation HV-51-225A Control-2A/2C test return line to suppression pool HV-51-2F052A Control-HPCI steam to RHR heat exchanger HV-51-253A 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 I HSS-12-016A-2 Control-Spray / bypass select 1 HSS-12-016C-2 Control-Spray / bypass select
]
l 9 LIMERICK - UNIT 2 3/4 3-80
Table 3.3.7.4-1 (Continued) [] RHR SERVICE WATER SYSTEM (Continued) HSS-12-094 Control-Transfer switch HSS-12-093- Control-Transfer switch HV-51-2F014A Control-2A RHR heat exchanger tube side inlet 0CP506 Control-RHR Service Water pump HV-51-2F068A Control-2A RHR Heat exchanger tube side outlet EMERGENCY SERVICE WATER SYSTEM 0AP548 Control-A emergency service water pump HV-11-011A Control-A emergency service water disch to RHR service water HSS-11-091 Control-Transfer switch HSS-11-092 Control-Transfer switch HSS-11-093 Control-Transfer switch
'N The following valves of the ESW and RHRSW systems are actuated 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 l 152-11509/CSR 101-D21 Safeguard SWGR fecder bkr. 152-11609/CSR 101-D22 Safeguard SWGR feeder bkr. 152-11709/CSR 101-D23 Safeguard SWGR feeder bkr. 152-11502/CSR 201-021 Safeguard SWGR feeder bkr. 152-11602/CSR 201-D22 Safeguard SWGR feeder bkr. 152-11702/CSR 201-D23 Safeguard SWGR feeder bkr.
/'
152-11505/CSR D21C Safeguard LC ):'"R breaker L LIHERICK - UNIT 2 3/4 3-81
Table 3.3.7.4-1 (Continued) STANDBY AC POWER SUPPLY (Continued) 152-11605/CSR D224 Safeguard LC XFMR breaker 152-11705/CSR D234 Safeguard LC XFMR breaker 143-115/CS Transfer switch 143-116/CS Transfer switch l 143-117/CS Transfer switch 4 9 O LIMERICK - UNIT 2 3/4 3-82 -- I
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INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION I 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. j 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 tne CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.5-1. O 1 0' LIMERICK - UNIl 2 3/4 3-84 i
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I i Table 3.3.7.5-1 (Continued) ACCIDENT MONITORING INSTRUMENTATION TABLE NOTATIONS I
*Three noble gas detectors with overlapping ranges (10 7 to 10 1, 10 4 to l 102, 10 2 to 105 pCi/cc). i
**High range noble gas monitor. )
i ACTION STATEMENTS j ACTION 80 - i
- a. With the number of OPERABLE accident monitoring instrumentation j channels less than the Required Number of Channels shown in Table ]
3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status i 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 moritor-ing the appropriate parameters within 72 hours, and
- a. Either restore the inoperable channel (s) to ?ERABLE 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.
l O LIMERICK - UNIT 2 3/4 3-86
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l I INSTRUMENTATION I i SOURCE RANGE MONITORS LIMITING CONDITION FOR OPERATION l 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:
- a. In OPERATIONAL CONDITION 2*, three.
I 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 i HOT SHUTDOWN within the next 12 hours.
I
- b. In OPERATIONAL CONDITION 3 or 4 with one or more of the above required I 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.
** Neutron detectors may be excluded from CHANNEL CALIBRATION.
***For initial fuel loading and startup the count rate may be reduced to ,
0.7 cps provided the signal-to-noise ratio is > 2. l
#During initial startup test program, SRM detectors may be partially l withdrawn prior to IRM on-scale indication provided that the SRM channels l remain on scale above 100 cps and respond to changes in the neutron flux. l LIMERICK - UNIT 2 3/4 3-88
INSTRUMENTATION ("; 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 actectors, drives and readout equipment to map the core, and
- b. Inde:ing equipment to allow all five detectors to be calibrated in a common location.
APPLIt ABILITY: When the traversing in-core probe is used for:
- 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 Specification 3.0.3 are not applicable. U 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.
O LIMERICK - UNIT 2 3/4 3-89
i INSTRUMENTATION j CHLORINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION I I 3.3.7.8.1 Two independent chlorine detection system subsystems shall be j OPERABLE with their alarm and trip setpoints adjusted to actuate at a i chlorine concentration of less than or equal to 0.5 ppm. !! APPLICABILITY: All 0PERATIONAL CONDITIONS. I l ACTION: l
)
i l a. With one chlorine detection subsystem inoperable, restore the l l 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 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.
O 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 2 3/4 3-90
l INSTRUMENTATION
/ \ T0XIC GAS DETECTION SYSTEM ,Q 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 actuate 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
[j inoperable detection system to OPERABLE status within 7 days or, within the next 6 hours, initiate and maintain operation of at least i / one control room emergency filtration system subsystem in the chlorine isolation mode of operation.
- b. With both toxic gas detection subsystems inoperable, within 1 hour initiate and maintain operation of at least one cor. trol room emer-gency filtration system subsystem in the chlorine isolation mode of operation.
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,
- b. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- c. CHANNEL CALIBRATION at least once per 18 months.
O l' LIMERICK - UNIT 2 3/4 3-91 l l
INSTRUMENTATION FIRE OETECTION 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 shall be OPERABLE. APPLICABILITY: Whenever equipment protected by 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 { 1 east once per hour, unless the instrument (s) is located inside i 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 B, or one-half or less of the Total I 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 inside an inaccessible zone, then inspect the area surrounding the inaccessible zone at least once per hour.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.9.1 Each of the above required tire 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 2 3/4 3-92
T
.z,-
TABLE 3.3.7.9-1 p FIRE DETECTION INSTRUMENTATION
-f , INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS
- i FIRE .
ZONE STRUCTURE ELEV. AREA HEAT SM0KE FLAME I (x/y) (x/y) -(x/y) ! IL Control 200' Control Structure' Chillers and NA 3/0 NA i Chilled Water Pump Area 258 j 1M Control 200' Control Structure Chillers and NA 3/0 NA l Chilled Water Pump Area 263 2 Control 217' 13-kV Swit30 ear Area 336 NA 34/0 NA ;
.5 Control 217' Battery Room 10 (20) 1/0 1/0 NA 6 Control 217' Battery Room 361 (2C) 1/0 1/0 NA
)
7 Control 239' Corridor 437 NA 5/0 NA
]
10A Control 239' Battery Room 426 1/0 2/0 NA 10B Control 239' Switchgear Maintenance 2/0 1/0 NA Area 454 11 Control 239' Battery Room 427 1/0 2/0 NA
'16 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 430 17 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA
. 431 18 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA j 428 1 19 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 429 21 Control 254' Ststic Inve-ter Room Unit 2 NA 6/0 NA Area 453 ,
l l 23 Control 254' Cable Spreading Room Unit 2 NA 14/0 NA Area 450 4 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. 536 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-l Elect) 24G Control 269' Control Room Shift Supt. NA 1/0 NA 532 l LIMERICK - UNIT 2 3/4 3-93 a- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _
I TABLE 3.3.7.9-1 (Continued) FIRE DETECTION INSTRUMENTATION J.NSTRUMENTLOCATION 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 0/88 57/0 NA (PGCC (Ceiling)
Floor) 46/0 (PGCC Floor) . i 0/8 6/0 l l (Non- (Non-l PGCC PGCC l Floor) Floor) 32/0 (Terminal Cabinets) l 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) l 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) Control Control Room Fresh Air 3/0 28C 332' NA NA Intake Plenum 54 Unit 2 17/ RHR Heat Exchanger & NA 8/0 NA Reactor Pump Room 173 (A&C) l 55 Unit 2 177' RHR Heat Exchanger & NA 7/0 NA Reactor Pump Room 174 (B&D) l 56 Unit 2 177' RCIC Pump Room 179 0/3 2/0 NA Reactor l 57 Unit 2 177' HPCI Pump Room 180 0/4 3/0 NA l Reactor 58 Unit 2 177' 'B' Core Spray Pump NA 2/0 NA
. Reactor Room 181 LIMERICK - UNIT 2 3/4 3-94 l
r :n T-
- - - - - - - - ~ - - - - - - - - - - - - - - - - - -
r 4 , e
+
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, i g, . TABL'i 3.3.7.9-1 (Continued) )
FIRE DETECTION: INSTRUMENTATION'N
]
h~ JINSTRUMENT LOCATION- TOTAL' NUMBER OF INSTRUMENTS *
' FIRE: . . l
- ~ZONE STRUCTURE- ELEV.- AREA HEAT SM0KE- FLAME- i (x/y) (x/y) (x/y) 177' 'D' Core. Spray Pump
]
59! Unit ~2 NA' 2/0 NA-
' Reactor Room 184-160 -Unit 2 177'- 'C' Core Spray Pump NA 2/0 NA Reactor Room 185
- 61 Unit 2. 177' 'A' Core Spray Pump NA 2/0 NA Reactor- Room 188 62 Unit 2 177' Sump Room 186; NA 4/0 NA Reactor. Passageway 189
'63 ~ Unit.2 177' Corridor 182 NA 2/0 NA Reactor 64A Unit 2 201' RECW Equipment Area 264 0/0 S/0 ~ NA l
. Reactor
- 65 Unit 2 201' Safeguard System Access t/14 4/0 NA
, Reactor. Area 279 66 Unit 2. 217' Safeguard System Isolation NA 8/0 FA Reactor Valve Area 376
\ .
67 Unit 2 217' Safeguard System Access 0/11 28/0 NA-Reactor- Area 370 (Scutheast) 0/11 (Nort; cast) 0/14 W rthwest)
~68A Unit 2 253' CRD Hydraulic Equipment 0/10 20/0 NA Reactor Area 475 (Northsest) 0/6 (Southeast) 68B Unit 2 253' Neutron Monitoring NA 2/0 NA Reactor System Area 479
=68C Unit 2 253' CRD Repair Room 476 NA 2/0 NA
, Reactor 70A Unit 2 283' Corridor 580; General 0/15 22/0 NA Reactor Equipment Area 574 70B Unit 2 295' Isolation Valve NA 2/0 NA Reactor Compartment 593 70C 'Jnit 2 283' Fuel Pool Cooling Water NA 2/0 NA Reactor Pump and Heat Exchanger Area 585 70D Unit 2 283' Isolation Valve NA 1/0 NA 1 Reactor Compartment 58'/597 LIMERICK - UNIT 2 3/4 3-95
_ _ _ _ = _ _ _ _ - _ - _ - _ _ _ _ - - . - _ _ _
i TABLE 3. 3. 7. 9-j (Continued) FIRE DETECTION , INSTRUMENTATION INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS
- FIRE ZONE STRUCTURE ELEV. AREA HEAT SM0KE FLAME (x/y) (x/y) (x/y) 71A Unit 2 313' Laydown Areas 637 and 638; NA 18/0 NA Reactor Corrid3r and RERS Fan Area 641
- 74A Unit 2 331' RERS Filtet 2/0 NA NA l Reactor Compartment 651 (inside plenum) 74B Unit 2 331' RERS Filter 2/0 NA NA Reactor Compartment 653 (inside o plenum) 83 Diesel- 217' Diesel-Generator 1/S 4/0 1/0 Generator Cell Unit 2 84 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Gener ;or Cell Unit 2 85 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Lenerator Cell Unit 2 86 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 l Generator Cell Unit 2 122A Spray 268' ESW and RHRSW Pung Area NA 4/0 NA bnd Pump -
Structure 122E Spray 251' RHRSW Valve Cordpartment 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 . 125A Diesel- 217' Diesel-Generator Access NA 4/0 NA Generator s c*idor 317 126A Common 412' Nor; , Stack Instrument NA 2/0 NA Reactor noom 713 l (x/y): X is ?.he number of Function A (Early Warning Fire Detection and Notification Only) Instruments. Y is the number of Function B (A W vation of Fire Suppression System and Early Warning Notificatii.7) Instruments. (a) These smoke detectors are located belo.1 iN suspended ceiling in the Control Room. (b) These smoke detectors are located above the suspended ceiling in the Control Room. LIHERICK - UNIT 2 3/4 3-96
i 4 l INSTRUMENTATION ( ) LOOSE-PART DETECTION SYSTEM W l i LIMITING CON 3ITION FOR' OPERATION 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) I to OPERABLE status,
- b. The provisions of Specification 3.0.3 are notaa pplicable.
SURVEILLANCE REQUIREMENTS t ( 4.3.7.10 Each channel of the 1onse 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.
l l LIMERICK - UNIT 2 3/4 3-97
INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPER(WZN 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 ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints* of tnese channels shall be determined and adjusted in accordance with the methodology and parseeters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel clarm/ trip ratpoint less conservative than requ' red by the above specification, immediately suspend the release of radioactive l liquid effluents monitored by the affected channel, or declere 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 OPERABL.E status within the time specified in the ACTION or explain in the next Semiannual Radioactive Effleent Release Report why this inoperabilit:y was not corrected within the time specified.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.11 Each radioactive liquid effluent monitoring instrumentation ciannel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST aperations at the frequencies shown in Table 4.3.7.11-1.
- Excluding the flow rate measuring i vices which are not determined and adjusted in accordance with the ODCM.
Ol LIMERICK - UNIT 2 3/4 3-98
t, . i TABLE 3.3.7.11-1 A ( RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS
, INSTRUMENT OPERABLE- ACTION
- 1. GROSS RADI0 ACTIVITY MONITORS PROVIDING AUTOMATIC TERMINATION OF RELEACE
- a. Liquid Radweste Effluent Line 1' 100
- b. RHR Service Water Systen Effluent Line 1/ loop 101
- 2. GROSS RADI0 ACTIVITY MONITORS NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
- a. Service Water System Effluent Line 1 101-
- 3. FLOW RATE MEASUREMENT' DEVICES
- a. Liquid.Radwaste Effluent Line 1 .102
- b. Dis M rge Line 1 102.
[
\
l O i LIMERICK - UNIT 2 3/4 3-99 i
-----_---------_-------___-_-_.-_-_-_.-_n_ . _
1 l L TABLE 3.3.7 J1-1 (Continued) i J ACTION STATEMENTS f ACTION 100 - With the number of channels OPERABLE less than required by the j Minimum Channels OPERABLE requirement, effluent releases may 3 continue for up to 14 days 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 j i
- b. At least two technically qualified members of the facility l 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 tLe Minimum Channels OPERABLE requirement, effluent releases via l this pathway may continue for up to 30 days provided that, at least once per 8 hours, grab t Amples are cellected and analyzed for gross radioactivity (beta or gamma) at a limit of detet. tion of at least 10 7 microcurie /mL. ACTION 102 - With the number of channels OPERABLE less than 19quirr;d by the Minimum Channels OPERABLE requirement, effluent reieases via this pathway r'y 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 2 3/4 3-100 I
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i 1 i TABLE 4.3.7.11-1 (Continued) I TABLE NOTATIONS i (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation l of this pathway and control room alarm annunciation occur if any of the j following conditions exists: ) (
- 1. Instrument indicates measured levels above t.e alarm / trip setpoint. l
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure. ;
I (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. l
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
(3) The initial CHANNEL CALIBRATION shall be performed usirig 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 me.de at least once per 24 hours on days on which continuous, periodic, or batch releases are made. I i f I 1 O LIMERICK - UNIT 2 3/4 3-102 l O_____.___________ _ _.
INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 1 I p 3.'4.7.12 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 shall be determined in acco'rdance ! with the methodology and parameters in the ODCM. l 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 above Specification, immediately suspend the release of radioactive gaseous effluents 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 in Table 3.3.7.12-1. Restore the inoperable instrurnentation to e OPERABLE status within the time specified in the ACTION or explain 1
( why this inoperability was not corrected in a timely manner in the j next Semiannual Radioactive Effluent Release Report. ]
- c. .The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.3.7.12 Each radioactive gaseous effluent monitoring instrumentation channel I shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the l frequencies shown in Table 4.3.7.12-1.
*The alarm / trip setpoints for the Main Condenser Offgas Treatment System .,
Explosive Gas Monitoring System and the Main Condenser Offgas Retreatment Radiation Monitor are set in accordance with Specification 3.11.2.5 and 3.11.2.6, respectively. LIMERICK - UNIT 2 3/4 3-103 x
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l TABLE 3.3.7.12-1 (Continued) i TABLE NOTATIONS
*At all times.
**During operation of the main condenser steam jet air ejector and offgas treatment system.
***0uring operation of the hot maintenance shop ventilation exhaust system. I l
ACTION STATEMENTS l ACTION 110 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of main condenser l 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. ; 1 ACTION 111 - With the number of channels OPERABLE less than required by the l 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 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 and provided the mechar,ical 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 l Effluent Noble Gas Activity Monitor is OPERABLE; otherwise, be in at least HOT SHUTDOWN within 12 hours. i O LIMERICK - UNIT 2 3/4 3-106 l l
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l
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1 I' L TABLE 4.3.7.12-1 (Continued) ~f~. TABLE NOTATIONS
- At all' times. ;
** Duril:g operation of the main condenser steam' jet air ejector and offgas 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:
- l. 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.
t (2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit' calibrating the system over its intended range of energy and measurement C range. For subsequent CHANNEL CALIBRATION, sources that have been related
-1]N to the initial calibration shall be used.
(3) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- 1. 0.0 volume percent hydrogen, balance nitrogen, and
- 2. 4 volume percent hydrogen, balance nitrogen.
(4) The iodine cartridges and particulate filters will be changed at least once per 7 days. k LIMERICK - UNIT 2 3/4 3-109
INSTRUMENTATION 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM j LIMITING CONDITION FOR OPERATION 3.3.8 At least one turbine overspeed protection system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. l 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, l 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. i
- 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: l
- 1. Cycling each of the following valves through at least one complete cycle from the running position:
a) For the overspeed prctection 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.
i O LIMERICK - UNIT 2 3/4 3-110
l j 4 INSTRUMENTATION
~(
- SURVEILLANCE REQUIREMENTS (Continued) 1 L' )
b; At aast once perl31 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; j
' 1) Four.high pressure turbine control valves b) For the electrical overspeed. trip system and the. mechanical overspeed trip system; 1)- Four high pressure turbine control' valves
~
- 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 seat.s, disks and stems and verifying no unacceptable flaws or excessive corrosion. If. unacceptable flaws or excessive corrosion are found, all'other valves of that type shall be inspected.
.(
LIMERICK - UNIT 2 3/4 3-111
1 INSTRUMENTATION 3/4.3.9 FEEDWATER/ 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 j with the values shown in the Trip Setpoint column of Table 3.3.9-2. I I APPLICABILITY: As shown in Table 3.3.9-1. ACTION: ,
- a. With a feedy:9ter/ main turbine trip system actuation instrumentation !
channel trip setpoint less conservative than the value shown in the i 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. I l l l l l O l LIMERICK - UNIT 2 3/4 3-112
f 'l I L
- l I
- TABLE-3.3.9-1 FEEDWATER/ MAIN-TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE APPLICABLE
~
. . CHANNELS PEP OPERATIONAL
. TRIP FUNCTION TRIP SYSTEM CONDITIONS
- 1. Reactor. Vessel Water-Level-High, Level 8 4 1 r
k i LIMERICK - UNIT 2 3/4 3-113 4-
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
- 1 55.5 inches l *See Bases Figure B 3/4.3-1
)
1 1 I l l 1
\
l I 1 1 i l LIMERICK - UNIT 2 3/4 3-114
TABLE 4.3.9.1-l' FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS-OPERATIONAL CONDITIONS CHANNEL FOR WHICH CHANNEL FUNCTIONAL CHANNEL' SURVEILLANCE
. TRIP FUNCTION CHECK ~ TEST CALIBRATION REQUIRED
- 1. Reactor Vessel Water D M. R 1 Level-High, Level 8 I
O O LIMERICK - UNIT 2 3/4 3-115
3/4.4 REACTOR COOLANT SYSTEM l h 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS I LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation I with: i
- a. Total core flow greater than or equal to 45% of rated core flow, or i
- b. THERMAL POWER less than'or equal to the limit specified in Figure 3.4.1.1-1.
4 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,
[ 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 specifieu in Figure 3.4.1.1-1:
- 1. Determine the APRM and LPRM** noise levels (Surveillance 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 DOWER.
- 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 2 3/4 4-1
i
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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 THERMAL POWER. t I i 4.4.1.1.2 Each pump MG set scoop tube mechanical and electrical stop shall ba l demonstrated OPERABLE with overspeed setpoints less than or equal to 109% and 107%, respectively, of rated core flow, at least once per 18 months. 4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within l the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) i i 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. l 0
*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 2 3/4 4-2 l
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i REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 L'd 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 determining recirculation loop flow, total core flow and diffuser-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 recircula' ion loop flow measurements.
- c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs 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.
l Comparisons of the actual data in accordance with the criteria listed shall I commence upon the conclusion of the startup test program. O LIMERICK - UNIT 2 3/4 4-4 i t__________--- ;
REACTOR C00LtST SYSTEM
T RECIRCULATION PUMPS LJ 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. 10% 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:
- a. Restore the recirculation pump speeds to within the specified limit within 2 hours, or
- b. Declare the recirculation loop of the pump with the slower speed not in operation and take the ACTION required by Specification 3.4.1.1.
SURVEILLAN_CE REQUIREMENTS 4.4.1.3 i< circulation pump speed shall be verified to be within the limits at least once per 24 hours.
*See Special Test Exception 3.10.4.
[~h O LIMERICK - UNIT 2 3/4 4-5
i 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, ed:
- 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% tf 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. 1 O LIMERICK - UNIT 2 3/4 4-6
REACTOR COOLANT SYSTEM
'm ) 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 rvstem safety / relief valves shall be OPERABLE with the specified co6 . safety valve function lift settings:*# 4 safety / relief valves @ 1130 psig 11% 5 safety / relief valves @ 1140 psig 11% 5 safety / relief valves @ 1150 psig 11% APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. I 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 in COLD SHUTDOWN within the following 24 hours.
'b 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 leve1 by performance of a:
- a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and a
- b. CHANNEL CALIBRATION at least coce 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 rpares that have been previously set pressure tested ar.d stored in accordance with manufacturer's recommendations tested at least once per 40 months.
"The lif t setting pressere shall correspond to am'aient conditions of the valves at nominal operating temperatures and pressures
**The provisions of Specification 4.0.4 dre not applicable pewided the Surveillance is performed within 12 hours aft.sr reactor steam pressure is adequate to perform the test.
#Up to 2 inoperable valves may be replaced with spare OPERABLE valves w;th r lower setpoints until the next refueling.
## Initial setting shall be in accordance with the manufacturer's recommendation.
(k Adjustment to the valve full open noise level shall be accomplished during the startup test program. LIMERICK - UNIT 2 3/4 4-7
r i REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION l 3.4.3.1 The following reactor coolant system leakage detection systems shall I be OPERABLE:
- a. The primary containment atmosphere gaseous radioactivity monitoring l system, !
i
- b. The drywell floor drain sump and drywell equipment drain tank flow I 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 grau 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 monitoriM lyster; is inoperable; otherwise, be in at least HOT SHUTDOWN within the next 12 ..Jurs and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS 4.4.3.1 Tho reactor coolant system leakage detection systems shal' ce demonstrated OPERABLE by:
- a. Primary containment atmospnere 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.
*The primary containment atmosphere gaseous radioactivity unitor is not required to be OPERABLE until OPERATIONAL CONDITION 2.
l' LIMERICK - UNIT 2 3/4 4-8 i
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j :3.4.3.2 l Reactor coolant system leakage shall be-limited to: i a. No PRESSURE B0liN %RY LEAKAGE.
- b. . 5' gpm UNIDENTIFIFn '.EAKAGE.
- c. 30 gpm total leakage.
- d. 25 gpm total leakage averaged over any 24-hour period.
- e. I gpm leakage at a. reactor coolant system pressure of 950 i10 psig from any reactor coolant system pressure isolation valve specified.
in Table 3.4.3.2-1. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. l ACTION: o
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within I 12 hours and in COLD' SHUTDOWN within the next 24 hours.
. With any reactor coolant system leakage greater than the limits in b, c
-and/or d, above, reduce the leakage rate to within the limits within f~ 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, ur 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 statur. 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 f.ollowing 24 hours.
*Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.
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.s LIMERICK - UNIT 2 3/4 4-9
J l REACTOR COOLANT SYSTEM
-SURVEILLANCE REQUIREMENTS I
4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be I within each of the above limi+s 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 i once per 12 hours, J
- 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 (nnt 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.
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B 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 hoers 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 H0T SHUTDOWN within 12 hours and in COLD SHUTDOWN 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, resto,e the conductivity and pH to within the limit within 72 iiours, or b) Chloride concentration exceeding the limit specified in Table 3.4.4-1, restore the chloride concentration to within the limit wi'.hin 24 hours, or perform an engineering evaluation to determine the effects of the out-of-limit con 6 tion on the structural integrity of the ) reactor coolant system. Determine that the structural integrity l of the reactor coolant system remains acceptable for continued operation prior to proceeding to OPERATIONAL CONDITION ~,.
- 2. The provisions of Specification 3.0.3 are not applicable. l LIMERICK - UNIT 2 3/4 4-12
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' REACH 0RCOOLANTSYSTEM 'O4 SURVEILLANCE REQUIREMENTS L
4.4.4 The reactor coolant shall be determined to be within the specified L chemistry limit by: L
- 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 q b) 8 hours whenever conductivity is greater than the limit 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 CONDITIONI 1, 2, and 3, and
- 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.
LIMERICK - UNIT 2 3/4 4-13
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/7 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 1006 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, DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation values closed within 12 hours. The provisions of Specification 3.0.4 are not applicable.
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-(N 2. Greater than 100 6 microcurie per gram be in at least HOT SHUTDOWN with the main steam line 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 microcuries per gram DOSE EQUIVALENT I-131 or greater than 1006 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.
- c. In OPERATIONAL CONDITION 1 or 2, with:
- 1. THERMAL POWER changed by more than 15% of RATED THERMAL POWER in I 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 I is restored to within its limit.
O *Not applicable during the startup test program. LINEWICK - UNIT 2 3/4 4-15
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l REACTOR COOLANT SYSTEM 1 SURVEILLANCE REQUIREMENTS Oj) I 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
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t-REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION i 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 70 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 SHUTDOWN 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 above 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. 1 O LIMERICK - UNIT 2 3/4 4-18
F REACTOR COOLANT SYSTEM i 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 vt;sel 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 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 70 F:
- a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
- 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 2 3/4 4-19
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/ / CURVES A,B,C ARE VAUD
/ FOR 2.5 EFPY OF OPERATION O i i I O
100 200 300 400 500 600 l MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 . i LIMERICK - UNIT 2 3/4 4-20 l
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- O 3 R
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. E I 6 C
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- 4 A
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E E L V B R A U T S L A I R E N T LO A EI M ST 0 0 0 SA 3 2 0 L EC 1 3 E VO S L S E V R O T C A E R 3 2 1 _ 0 0 0 0 0 0 G G G 7 7 7 T 1 1 1 E 7 7 7 4 LR 7 7 7 / UE 1 SB C C C _ PM 1 1 1 t AU 3 3 3 A _ CN 1 1 1
- C h ,_ E , C 5 [ wbi5 l ll
REACTOR COOLANT SYSTEM REACTOR STEAM DrfjE LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1020 psig. APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2^. l ACTION: With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure i to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within l 12 hours. l SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig at least once per 12 hours. l
*Not applicable during anticipated transients.
O LIMERICK - UNIT 2 3/4 4-22
;. s REACTOR COOLANT SYSTEM 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 th6n or equal to 5 seconds.
; 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 mtin steam line by !se of a deactivated MSIV in the closed position.
- 2. Otherwise, be in at least H0T SHUTDOWN within.the next 12 hours (n 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. l LIMERICK - UNIT 2 3/4 4 23 l
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. l ACTION: 1
- a. With the structural integrity of any ASME Code' Class I component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate i
the affected component (s) prior to increasing the reactor coolant system temperature more than 50 F above the minimum temperature required by NDT considerations. I
- b. With the structural integrity of any ASME Code Class 2 component (s) l 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.
SURVEILLANCE REQUIREMENTS 4.4.8 No requirements other than Specification 4.0.5. O : LIMERICK - UNIT 2 3/4 4-24 I
m< , REACTOR COOLANT SYSTEM' 1 3/4.4.9 . RESIDUAL HEAT REMOVAL vs HOT SHUTDOWN LIMITING CONDITION FOR OPERATION. 3.4.9.1 o Two* shutdown cooling mode loops of the residual heat removal (RHR)
. system shall be OPERABLE and, unless at bast one recirculation pump is in operation, at least one shutdown. cooling mob loop shall be in operation ** ***
with each loop _ consisting of at least:
- a. One OPERABLE RHR pump, and
- b. One OPERABLE RHR heat exchanger.
i 3 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 L
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 SHUTDOWN within 24 hours.****
'N b.- With-ao RHR shutdown cooling mode loop in operation, immediately initiate corrective action to rtturn 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.
LIMERICK - UNIT 2 3/4 4-25
REACTOR COOLANT SYSTEM COLD SHUTDOWN
. LIMITING CONDITION FOR OPERATION 3.4.9.2 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.
l APPLICABILITY: OPERATIONAL CONDITION 4. 1 ACTION:
- a. With less than the above required RHR shutdown cooling mode loops i OPERABLE, within 1 hour and at least once per 24 hours thereafter, l 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 I
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 periot provided the other loop is OPERABLE.
l
***The shutdown cooling mode loop may be removed from operation during hydrostatic testing.
I LIMERICK - UNIT 2 3/4 4-26
g, e, R me . .g g ! 3/4.5L-EMERGENCY CORE COOLING SYSTEMS- [3- 3/4.5.1 ECCS --OPERATING. i 2 F LIMITING CONDITION FOR OPERATION
'3.'5.1 .The emergency core cooling systems shall be OPERABLE with: !
- a. The core spray system (CSS) co.nsisting of two subsystems with each subsystem comprised of:
- 1. Two OPERABLE CSS pumps, and 2.- An OPERABLE flow path capable of taking suction from the l: su'ppression ~ chamber and transferring the water through the spray .
sparger to the reactor vessel.
- b. The low pressure coolant infection' (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 suppression chamber and transferring the water to the reactor 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 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 in the shutdown cooling mode when reactor vessel pressure is less than the RHR Shutdown cooling permissive setpoint.
LIMERICK - UNIT 2 3/4 5-1 L _ __ - - _ _ _ _ _ _ _ _
E
. EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
- a. For the core spray ystem:
- 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-51-282 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 l within 72 hours,'or be in at least H0T SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
l
- 3. With no RHR cross-tie valves (HV-51-282 A, B) 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 SHUT 00WN 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 SHUTDOWN 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 5 200 psig within the following 24 hours.
QWhenever 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 2 3/4 5-2
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 H0T 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 1 100 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,
- f. In the event an ECCS 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 O' 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
O LIMERICK - UNIT 2 3/4 5-3
i I EMERGENCY CORE COOLING SYSTEMS i 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 j system piping from the pump discharge valve to the system j isolation valve is filled with water. I b) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or l otherwise secured in position, is in its correct
- position. J
- 2. For the LPCI system, verifying that both LPCI system subsystem cross-tie valves (HV-51-282 A, B) are closed with power removed l from the valve operators.
- 3. For the HPCI system, verifying 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 > 105 psid plus head and line losses.
- 2. Each LPCI pump in cach 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
> 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 includes 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 en 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. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than 200 psig within the following 72-hours.
LIMERICK - UNIT 2 3/4 5-4 J i
- - - - - _ _ - - _ l
+ 1I i
EMERGENCYCOREC0bLINGSYSTEMS . L D SURVEILLANCE REQUIREMENTS'(Continued) dj L , 2. For the HPCI system, verifying that: L 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 l supiilied to the turbine' at 200 + 15, - O psig.** b) .Th'e 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 $ 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 allovrable value of 3.0 psid.
'd. For the ADS:
- 1. .At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST 7 of the accumulator backup compressed gas system low pressure
. alarm system.
(A
- 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 back compressed gas system low pressure alarm system and veritying l-an alarm setpoint of 90 t 2 psig on decreasing pressure.
l
**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. If ADS or HPCI OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than 100 psig or 200 psig, respectfully, within the following 72 hours.
LIMERICK - UNIT 2 3/4 5-5 L-_-----__-
1 EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN 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 l 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 l lj b) When the suppression chamber water level is less than the limit or is drained, from the 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 2 3/4 5-6
EMERGENCY CORE COOLING SYSTEMS
'xs _, St!RVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.
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). l
? ,O\
() LIMERICK - UNIT 2 3/4 5-7 [
i 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 4 condensate sto %ge tank and transferring the water through the spray sparger to the reactor vessel.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5*. i 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 SHUTDcWN within the following 24 hours.
- b. In OPERATIONAL CONDITION 3 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.
l
*The suppression chamber is not requireci t o be OPERABLE provided that the i reactor vessel head is removed, the cavity is flooded or being flooded from I 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 f
l Specifications 3.9.8 and 3.9.9. LIMERICK - UNIT 2 3/4 5-8 l L_______. 1
EMERGENCY CORE COOLING SYSTEMS . yy M ) i\s' SURVE1LLANCE REQUIREMENTS 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 i 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 !
r.atisfied, or j
- b. Verify footnote conditions
- to be satisfied.
l lOJ t, N. s I l 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 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 ,h Specifications 3.9.8 and 3.9.9. '%.)
LIMERICK - UNIT 2 3/4 5-9
3/4.6 CONTAINMENT SYSTEMS
; ) 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT IIITEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained. ,
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3. ACTION: Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within i hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE REQUIREMENTS I 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 Pa, 44.0 psig, l and verifying that when the measured leakage rate for these seals is
!O) added to the leakage rates determined pursuant to Surveillance i
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,. i
- b. At least once per 31 days by verifying that all primary containment l penetrations ** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, exr.ept as provided in Table 3.6.3-1 of Specification 3.6.3.
- c. By verifying the primary containment air lock is in compliance with j the requirements of Specification 3.6.1.3.
- d. By verifying the suppression chamber is in compliance with the requirements of Specification 3.6.2.1.
l
*See Special Test Exception 3.10.1 l **Except valves, blind flanges, and deactivated f.omatic valves which are located inside the containment, and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD l l
SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often
, than once per 92 days.
{- i LIMERICK - UNIT 2 3/4 6-1
1 i l l CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 1
- 3. 6.1. 2 Primary containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of less than or equal to La , 0.500 percent by weight of the containment air per 24 hours at Pa , 44.0 psig.
- b. A combined leakage rate of less than or equal to 0.60 L, 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 1 tested per Table 3.6.3-1, subject to Type B and C tests when pressurized to P,, 44.0 psig.
j 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 ,t22.0 psig.
- d. A combined leakage rate of less than or equal to 1 gpm times the total mimber of containment isolation valves in hydrostatically testeJ 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 primary containment leakage rate exceeding 0.75 La , 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 L,, or
- 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
- Exemption to Appendix J of 10 CFR Part 50.
LIMERICK - UNIT 2 3/4 6-2
CONTAINMENT SYSTEMS
'A l LIMITING CONDITION FOR OPERATION (Continued)
- Q)
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 valves which are hydrostatically tested per Table 3.6.3-1, subject to Type B and C tests to less than or equal to 0.60 L,, and
- c. The leakage rate to less than or equal to 11.5 scf per hour for any 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 shali l' ) be conducted at 40 1 10 month intervals during shutdown at P , 44.0 psig, during each 10 year service period. The third test of each et shall be conducted during the shutdown for the 10 year plant inservice inspection. ;
- b. If any periodic Type A test fails to meet 0.75 L,, the test schedule for subsequent Type A tests shal? be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L,,
a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L,, at which time the above test l schedule may be resumed.
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,. The formula to be used is: [L g + L am - 0.25 L,] 5 L c
=
5 [L g+ L,, + 0.25 L ]awhere L c= supplemental test result; L g superimposed leakage; L,, = measured Type A leakage.
- 2. Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
- 3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be between 0.75 La and 1.25 L,.
(/
- Exemption to Appendix "J" to 10 CFR Part 50.
LIMERICK - UNIT 2 3/4 6-3
1 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- d. Type B and C tests shall be conducted with gas at ?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 l 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. ,
O
*Unless ; hydrostatic test is required per Table 3.6.3-1.
LIMERICK - UNIT 2 3/4 6-4
CONTAINMENT SYSTEMS h i
\s / PRIMARY CONTAP1ENT AIR LOCK 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 La at P,, 44.0 psig, APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3.
ACTION:
- a. With one primary containment air lock door inoperable:
\
~
- 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed.
(O)_, 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. l
- 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 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
*See Special Test Exception 3.10.1.
O v LIMERICK - UNIT 2 3/4 (-5
CONTAINMENT SYSTEMS < SURVEILLANCE REQUIREMENTS
- 4. 6.1. 3 The primary containment air lock shall be demonstrated OPERABLE: I
- 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 l 10 psig: ,
- 1. within 72 hours after each closing, ecept when the air lock is !
being used for multiple entries, then at least once per 72 hours; I 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, j 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 E.pplicable.
** 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 containmen.t is inerted, provided that the airlock doors' interlock is tested within 8 hours af ter the primary containment has been deinerted and provided Lne shield door to the airlock is maintained locked closed.
LIMERICK - UNIT 2 3/4 6-6 _ _ _ _ _ _ - - _ _ . - _ _ _ _ J
CONTAINMENT SYSTEMS MSIV LEAKAGE CONTROL SYSTEM LIMITING CONDITION 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 ieast HOT SHUTDOWN within the next 12 hcurs 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-O 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 O at 100 scfm. b) Outboard valves, 15" H2O at 200 scfm.
- d. By verifying the operating instrumentation to be OPERACLE 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.
( C LIMERICK - UNIT 2 3/4 6-7 l 1
-__.__m____________-______m.____-. - _ . _ _ _ _ _ _ _ _ _ _ _ .
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTEGRITY l
, LIMITING CONDITION FOR,0PERATION i 1
1 1 3.6.1.5 The structural integrity of the primary containment shall be l maintained at a level consistent with the acceptance criteria in Specification l 4.6.1.5. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
!fTION:
With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits witt;n 24 hours or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. i l ( ' 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 i 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 o the Type A containment leakage rate test to 'ierify no apparent changes in appearance or other abnormal degradation. 4.5.1.5.2 Reports Any abnormal degradation of the primary containment 4 structure detected during the above required inspections shall be reported in I 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. l
)
O i LIMERICK - UNIT 2 3/4 6-8 l w--________________- . _ _ _ _ _ _ _
CONTAINMENT SYSTEMS ( )
' (_/ 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 % following 24 hours. I
,--~
(_) 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. ( l n LIMERICK - UNIT 2 3/4 d-9
}
l l l I CONTAINMENT SYSTEMS i DRYWELL AVERAGE AIR TEMPERATURE 1 LIMITING CONDITION FOR OPERATION l
- 3. 6.1. 7 Drywell average air temperature shall not exceed 135 F. {
l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ; t 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 ! vst ) HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following I 24 hours. l l SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average aii temperature shall be the volumetric average 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' 125 , 225 , 345
- c. 260' 50 , 165 , 300
- d. 248' 11 , 74 , 150 , 182 , 253 , 337
*At least one reading from each elevation is required for a volumetric j average calculation.
Ol LIMERICK - UNIT 2 3/4 6-10
l CONTAINMENT SYSTEMS
~
j
/3 D_RWELL AND SUPPRESSION CHAMBER PURGE SYSTEM LIMITING CONDITION FOR OPERATION l
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 iine 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 witt;in 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 ("' for more than 90 hours in the previous 365 days.* (
- Valves open for pressure control are not subject to the 90 hour 9er 365 day A limit provided the 1-inch /2-inch bypass line is being utilized.
LIMERICK - UNIT 2 3/4 6-11
1 L 1 l l CONTAINMENT SYSTEMS i 3/4.6.2 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 122,120 fts and 134,6003ft , 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. fCTION:
- 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 % F within 24 hours or be in at least HOT SHUTDOWN within the iaxt 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 ce in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWh' 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. j b) 110 F, place the reactor mode switch in the Shutdown I position and operate at least one residual heat removal loop l in the suppression pool cooling mode. '
- Includes the volume inside the pedestal.
LIMERICK - UNIT 2 3/4 6-2?
CONTAINMENT SYSTEMS ( LIMITING CONDITION FOR OPERATION (Continued) k j)' 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 drywe' * ~ 9-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor ()olant temperature above 200 F.
SURVEILLANCE REQUIREMENTS i 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 during testing which adds heat to the suppression chamber, by verifying the suppression chamber average water temperature less than 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 core 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, l
by verifying suppression chamber average water temperature less than or equal to 120 F. LIMERICK - UNIT 2 3/4 6-13 1
I CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) I
- 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: l l
- 1. High water level 124'll" s
- 2. High water temperature:
a) First setpoint i 95 F b) Second setpoint i 105*F c) Third setpoint i 110 F d) Fourth setpoint i 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/# calculated from the measured leakage is within the specified limit. If any drywell-to-suppression chamber bypass leak test fails 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.
l ! o! LIMERICK - UNIT 2 3/4 6-14 l - - . - . . _ _ j
CONTAINMENT SYSTEMS'
- SUPP!1ESSION' POOL SPRAY
((f 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). 1 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ;
)
ACTION:
]
l
- a. With one suppression pool spray loop inoperable, restore the inoperable 1 loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN I 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 i following 24 hours. I
,f ! k 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 SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
LIMEP.ICK - UNIT 2 3/4 6-15
CONTAINMENT SYSTEMS , SUPPRESSION POOL COOLING I LIMITING CONDITION FOR OPERATION I 1 1 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 j
- b. An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger. l l
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ' ACTION:
- a. With 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 I
HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN
- within the next l 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 puh.ps 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.
*Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN j as required by this ACTION, maintain reactor coolant temperature as low as '
practical by use of alternate heat removal methods. O 1 l LIMERICK - UNIT 2 3/4 6-16 i 1
m _
\
CONTAINMENT SYSTEMS. 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES N/ LIMITING CONDITION FOR OPERATION i i 3.6.3 The primary containment isolation valves and the reactor instrumentation 4 line excess flow check valves shown in Table 3.6.3-1 shall be OPERABLE with : inolation times less than or equal to those shown in Table 3.6.3-1. L APPLICABILITY: 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 or.e de-activated automatic valve secured in the isolated position,* or
- 3. Isolate each affected penetration by use of at least one closed manual valve'or blind flange.*
- 4. The provisions of Specification 3.0.4 are not applicable provided that within 4 hours the affected penetration is isolated in accordance with ACTION a.2. or a.3. above, and provided that b 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 SHUTDOWN 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 4 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 I
- 2. The instrument line is isolated and the associated instrument is declared inoperable. 1 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.
l lO ; LIMERICK - UNIT 2 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 OPERABLt 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 isolation 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 ?, cification 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 i 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 l 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.
l 1 O LIMERICK - UNIT 2 3/4 6-18 L______-._ _ _ _ _ _ _ _ _ ._
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u TABLE 3.6.3-1 )
- PRIMARY CONTAINMENT ISOLATION VALVES l {
NOTATION
\
NOTES
- 1. Instrumentation line isolation provisions consist of an orifice and excess 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.
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.
- 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 l provided by Type A leakage rate test. Type C test is not required.
l 13. The motor operators on HV-13-209 and HV-13-210 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.
\
LIMERICK - UNIT 2 3/4 6-41
TABLE 3.6.3-1 PRIMARY CONTAINMENT ISCLATION VALVES NOTATION NOTES (Continued)
- 15. Check valve used instead of flow orifice.
- 16. Penetration is se, led by a flange with double 0-ring seals. These seals are leakage rate fested by pressurizing between the 0-rings. Both the TIP Purge Supply (Penetration 35B) 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-241A-E) are Type C tested. It is not practicable to leak test the shear valves (XV-240A-E) because squib firing is required for closure.
Shear valves (XV-240A-E) are normally open.
- 17. Instrument line isolation provisions consist of an excess flow check valve.
Becau w the instrument line is connected to a closed cooling water system insidt c s,tainment, no flow orifice is provided. The line does not isolate ; during a LOCA and can leak only if the line or instrument should rupture. Leaktir htness of the line is verified during the integrated leak Nte test (Type . 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 are flanged to facilitate removal and are equipped with double 0 ring seal assemblies on the flange closest to primary containment. 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 the PCRVICS isolation signal (s) that initiate closure of each automatic isolation valve.
In addition, the following nor-PCRVICS isolation signals also initiate closure of selected valves: EA Main steam line high pressure, high steam line leakage flow, low l MSIV-LCS dilution air flow LFHP With HPCI pumps running, opens on low flow in associated pipe, closes i when flow is above setpoint l l 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, clo es when flow is above setpoint l LFCC Steam supply valve fully closed or RCIC turbine stop valve fully c h ea All power operated isolation valves may be opened or closed remote manually. O LIMERICK - LNIT 2 3/4 6-42
}
i TABLE 3.6.3-1
'O PRIMARY CONTAINMENT ISOLATION VALVES }s NOTATION NOTES (Continued)
- 21. Automatic isolation signal causes TIP to retract; ball v 5 e 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 Type A test. Type C test not required.
- 24. Both isolation signals required for valve closure.
- 25. Deleted
- 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 deteccion 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-g- 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. (DELETED)
- 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 41-2016 is an outboard isolation barrier for penetrations X-9A, B and X-44. Leakage through valve 41-2016 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 syrtems during plant startup.
- b. Reactor pressure vessel hydrostatic testing, which is conducted follow-ing each refueling outage prior to commencing plant startup.
Therefore, valve stroke timing in accordance wi.h Specification 4.0.5 is not required.
- 33. Valve also constitutes a Unit 1 Reactor Enclosure Secondary Containment Automatic Isolation Valve and a Refueling Area Secondary Containment Automatic Isolation Valve as shown in Table 3.6.5.2.1-1 and Table 3.6.5.2.2-1, respectively.
n 34. Isolation signal causes recombiner to trip; valve closes when recombiner 1 A is not operating, j LIMERICK - UNIT 2 3/4 6-43
h CONTAINMENT SYSTEMS 3/4._6.4 VACUUM RELIEF SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Each pair of suppression chamber - drywell vacuum breakers shall be OPERABLE and closed. l l _ APPLICABILITY: OPERATIONAL CONDITIONS 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 32 hours and in COLD SHUTDOWN within the following 24 hours.
- b. With one suppression chamber - drywell vacuum breaker open, verify i 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 er be in at least HOT SHUTDOWN within the next 12 hours and in COLD fHUTDOWN 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 msintained 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 hcurs and in COLD SHUTDOWN within the following 24 hours. O LIMERICK - UNIT 2 3/4 6-44 1 i
CONTAINMENT SYSTEMS r .I
\
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, i 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 optning setpoint, from the closed position, to be 0.5 psid i 5%, and b) Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.
c) Verifying that each outboard valve's position indi ator is capable of detecting disk displacement >0.050", and each inboard valve's position indicator is capable of detecting disk displacement >0.120". ( LIMERICK - UNIT 2 3/4 6-45
l CONTAINMENT SYSTEMS I 3/4.6.5 SECONDARY CONTAINMENT REACTOR'ENCt050RE SECONDARY CONTAINMENT INTEGRITY l LIMITING CONDITION FOR OPERATION 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained. I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. 1 ACTION: Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY within 4 hours or be in at least HOT SHUT 00WN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SLTvtILLANCE REQUIREMENTS 4.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon-itrated by:
- a. Verifying at least once per 24 hours that the pressure within the reactor enclosure secondary containment is greater than or equal I to 0.25 inch of vacuum water gauge.
l
- 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 i;alation dampersbralves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers 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 with wind speeds of 5 7.0 mph as measured on the wind instrument on Tower I elevation 30' or, if that instrument is unavailable, Tower 2, elevation 159'.
LIMERICK - UNIT 2 3/4 6-46
I CONTAINMENT SYSTEMS ( 3/4.6.5 SECONDARY CONTAINMENT l
}
REFUELING AREA SECONDARY CONTAINMEN1 INTEGRITY LIMITING CONDITION FOR OPERATION _ _ _ 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained. APPLICABIQTY: OPERATIONAL CONDITION *. ACTION: Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, susnend handling of irradiated fuel in the secondary containment, CORE AllERAuCNS 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 b'J:
- a. Verifying at least once per 24 hours that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.
I b. Verifying at least once per 31 days that: U
- 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 OPERABLE secondary containment automatic iso-lation dampers / valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers / valves secured in ,
position.
- c. At least once per 18 months:
Operat.ing one standby gas treatment subsysic 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.
*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.
O LIMERICK - UNIT 2 3/4 6-47
CONTAINMENT SYSTEMS REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISGLATION 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 mere of the reactor secondary containment ventilation system automatic isolation valves shown in Table 3.6.5.2.1-1 inoperable, maintain at , least one isolation valve OPERA'BLE 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, blind flange or slide gate damper.
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 2 3/4 6-48
)
l TABLE 3.6.5.2.1-1 (3 REACTOR ENCLOSURE SECONDARY CONTAINMENT VENTILATION SYSTEM i V) AUTOMATIC ISOLATION VALVES REACTOR ENCLOSURE (ZONE II) MAXIMUM ISOLATION TIHE ISOLATIgg) VALVE FUNCTION (Seconds) SIGNALS
- 1. Reactor Enclosure Ventilation Supply Valve HV-76-207 5 B,H,5,U
- 2. Reactor Enclosure Ventilation Supply Valve HV-7'.i-208 5 B,H,5,U
- 3. Reactor Enclosure Ventilation Exhaust Valve HV-76-257 5 B,H,5,U
- 4. Reactor Enclosure Ventilation Exhaust Valve HV-76-258 5 B,H,S,0
- 5. Reactor Enclosure Equipment Compartment Exhaust Valve HV-76-241 5 B,H,S,0
- 6. Reactor Enclosure Equipment Compartment Exhaust Valve HV-76-242 5 B,H,5,U
- 7. Drywell Purge Exhaust Valve HV-76-030 5 B,H,5,U,R,T l 8. Drywell Purge Exhaust Valve HV-76-031 5 B,H,5,0,R,T
- 9. Drywell Purge Exhaust Inboard Valve 5 B,H,5,U,W,R,T o HV-57-114 (Unit 1)
_ 10. Drywell Purge Exhaust Outboard Valve 6 B,H,5,U,W,R,T HV-57-115 (Unit 1)
- 11. Suppression Pool Purge Exhaust Inboard Valve 5 B,H,5,U,W,R,T HV-57-104 (Unit 1)
- 12. Suppression Pool Purge Exhaust Outboard 6 B,H,5,U,W,R,T Valve HV-57-112 (Unit 1) 1 l
l (a)See Specification 3.3.2, Table 3.3.2-1, for isolation signals that operate each automatic valve. LIMERICK - UNIT 2 3/4 6-49 l _ _ _ _ _ _ _ -
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 !.o 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 1
I
- 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, blind flange or slide gate damper.
Otherwise, in Operational Condition *, suspend handling of irradiated l 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: i a. Prior to returning the valve to service af ter 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. I
- b. At least once per 18 months by verifying that on a containment I
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-ment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
LIMERICK - UNIT 2 3/4 6-50
TABLE 3.6.5.2.2-1 f ^g REFUELING AREA SECONDARY CONTAINMENT VENTILATION SYSTEM -( ) AUTOMATIC ISOLATION VALVES v REFUELING AREA (ZONE III) MAXIMUM ISOLATION TIME ISOLATIgg) 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 1
- 4. Refueling Area Ventilation Exhaust Vahe 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 Ventilation Exhaust Valve HV-76-267 (Unit 2) 5 R,T
- 8. Refueling Area Ventilation Exhaust Valve HV-76-268 (Unit 2) 5 R,T
- 9. Drywell Purge Exhaust Valve HV-76-030 5 B,H,5,U,R,T
- 10. Drywell Purge Exhaust Valve HV-76-031 5 B,H,5,U,R,T
- 11. Drywell Purge Exhaust Inboard 5 B,H,5,0,W,R,T Valve HV-57-114 (Unit 1)
- 12. Drywell Purge Exhaust Outboard 6 B,H,5,U,W,R,T l Valve HV-57-115 (Unit 1)
- 13. Suppression Pool Purge Exhaust Inboard 5 B,H,S,U,W,R,T i i
Valve HV-57-104 (Unit 1)
- 14. Suppression Pool Purge Exhaust Outboard 6 B,H,S,U,W,R,T Valve HV-57-112 (Unit 1)
O LIMERICK - UNIT 2 3/4 6-51
t, TABLE 3.6.5.2.2-1 (Continued) ,f) REFUELING AREA SECONDARY CONTAINMENT VENTILATION SYSTEM .() AUTOMATIC ISOLATION VALVES REFUELING AREA (ZONE III) MAXIMUM ISOLATION TIME .ISOLATIgg)
-VALVE FUNCTION (Seconds) SIGNALS
. 15. Drywell Purge Exhaust Inboard 5 B,H,5,U,W,R,T y Valve HV-57-214 (Unit 2)
~
- 16. - Drywe.il Purge Exhaust Outboard 6 B,H,S,U,W,R,T Valve HV-57-215 (Unit 2)
- 17. Suppression Pool Purge Exhaust Inboard 5 B,H,5,U,W,R,T Valve HV-57-204 (Unit 2),
- 18. Suppression Pool Purge Exhaust Outboard 6 B,H,S,U,W,R,T Valve HV-57-212 (Unit 2) 4 (a) See Specification 3.3.2, Thole 3.3.2-1, for isolation signals that
( operate each automatic isolation valve. ! LIMERICK - UNIT.1 3/4 6-51a
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEM - COMMON 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 *. AITION:
- a. In OPERATIONAL CONDITION 1, 2, or 3:
- 1. With the Unit 1 diesel generator for one standby gas treatment subsystem inoperable for more than 30 days, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. The provisions of Specification 3.0.4 are not applicable. !
- 2. With one standby gas treatment 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 f9 10 wing 24 hours. I
- 3. With one standby gas treatment subsystem inoperable and the other standby gas treatment subsystem with an inoperable Unit I diesel generator, restore the inoperable subsystem to OPERABLE status or restore the inoperable Unit 1 diesel generator 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.
- 4. With the Unit I diesel generators for both standby gas treatment system subsystems inoperable for more than 72 hours, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION *:
- 1. With one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or 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.
I
- 2. With both standby gas treatment subsystems inoperable, 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.
1
*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 2 3/4 6-52
CONTAINMENT SYSTEMS p ), ' t j 'w/ SURVEILLANCE REQUIREMENTS l 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 i subsystem operates with the heaters OPERABLE.
i l l
.v i i
O V LIMERICK - UNIT 2 3/4 6-52a
CONTAINMENT SYSTEMS .j L' SURVEILLANCE REQUIREMENTS (Continued)
- 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 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.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. Verify that when the fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III) when tested in accordance with ANSI N510-1980.
Ix
- 4. Verify that the pressure drop across the refueling area to SGTS prefilter is less than 0.25 inches water gage whi'e operating at a flow rate of 2400 cfm i 10%.
- c. After every 720 hours of charcoal adsorber operation by verifying within 31 days af ter 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 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 9.1 inches i water gauge while operating the filter train at a flow rate of '
8400 cfm i 10%. I LIMERICK - UNIT 2 3/4 6-53
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 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 temperature differential across each heater is > 15 F when tested in accordance with ANSI N510-1980.
- e. After 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 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%.
- g. Prior to initial criticality of Unit 2 or after any major system alteration:
- 1. Verify that when the SGTS fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III).
- 2. Verify that one standby gas treatment subsystem will drawdown reactor enclosure Zone 11 secondary containment to greater than or equal ;o 0.25 inch of vacuum water gauge in less than or equal to 121 seconds with the reactor enclosure recirculation system in operation and the adjacent reactor enclosure and refueling area zones are in their isolation modes.
O LIMERICK - UNIT 2 3/4 6-54
'l t
CONTAINMENT SYSTEMS i REAC10R ENCLCSURE 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 HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE: 1
- 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.5.a, C.S.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 Guiae 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 operation when tested in accordance with ANSI N510-1980.
LIMERICK - UNIT 2 3/4 6-55
)
l 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%.
l
- d. At least once per 18 months by:
1 l 1. Verifying that the pressure drop across the combined prefilter, l upstream and downstream HEPA filters, and charccal 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. ,
- e. After 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 critiria 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 ficw rate of 60,000 cfm i 10%.
O l LIMERICK - UNIT 2 3/4 6-56 i l 1
i l CONTAINMENT SYSTEMS 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL . f] C/ PRIMARY CONTAINMENT HYDROGEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION I 3.6.6.1 Two independent primary containment hydrogen recombiner systems shall ! be'0PERABLE. 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.
- 3. A Heater Coil Check. )
- 4. A verification of valve operation by stroking all the valves to their proper positions. j
- b. At least once per 18 months by:
- 1. Performing a CHANNEL CALIBRATION of all control room recocbiner 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 one (1) megohm. ,
- 3. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure; i.e., loose l wiring or structural connections, deposits of foreign r:aterials, etc. ]
- 4. Verifying during a recombiner system functional test that the minimum heater outlet gas temperature increases to greater than or equal to '
1150 F within 120 minutes and maintained for at least one hour.
- c. By measuring the system leakap 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 3, 44.0 psig, on the schedule required by Specification 4.6.1.2, and including the measured leakage as a part of the leakage determined in accordance with Specification 4.6.1.2.
LIMERICK - UNIT 2 3/4 6-57
il CONTAINMENT SYSTEMS DRYWELL HYDR 0 GEN MIXING SYSTEM
, LIMITING CONDITION FOR OPERATION l 3.6.6.2 Four independent drywell unit cooler hydrogen mixing subsystems
- (2AV212, 2BV212, 2GV212, 2HV212) shall be OPERABLE with each subsystem consist-l ing of one unit cooler fan.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With one drywell unit cooler hydrogen mixing subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS _ 4.6.6.2 Each drywell unit cooler hydrogen mixing subsystem shall be demonstrated j 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.
l
)
O LIMERICK - UNIT 2 3/4 6-58 l
CONTAINMENT SYSTEMS , p) {
\s_/ DRYWELL AND-SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION ,
3.6.6.3 The drywell and suppression chamber atmosphere oxygen concentration shall be less than 4% by volume. APPLICABILITY: OPERATIONAL CONDITION 1*, during the time period: '
- a. Within 24 hours ** after THERMAL POWER is greater than 15% of RATED l THERMAL POWER, following startup, to ll
- b. Within 24 hours ** prior to reducing THERMAL POWER to less than 15% of j RATED THERMAL POWER, preliminary to a scheduled reactor shutdown.
ACTION: With the drywell and/or suppression chamber oxygen concentration exceeding the limit, restore the oxygen concentration to within the limit within 24 hours or be in at least STARTUP within the next 8 hours. ( (
)
SURVEILLANCE REQUIREMENTS I 4.6.6.3 The dryvell 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. ]
'I i
i
'i
*See Special Test Exception 3.10.5.
** Specification 3.6.1.8 is applicable during this 24 hour period.
i l LIMERICK - UNIT 2 3/4 6-59 "
- .t l 3/4.7 PLANT SYSTEMS i -3/4.7.1 SERVICE WATER SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM - COMMON SYSTEM i I
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 Unit 2 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 3.4.9.2, 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 one RHRSW pump inoperable, 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 one RHRSW pump in each subsystem inoperable, restore at laast 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 SHUTDOWN within the following 24 hours.
?. With one RHRSW subsystem otherwise inoperable, restore the inoperable subsystem to OPERABLE status with at least one OPERABLE RHRSW pump within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 4. With both RHRSW subsystems otherwise 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.
*Whenever 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.
LIMERICK - UNIT 2 3/4 7-1
PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- 5. With two RHRSW pump / diesel generator pairs
- inoperable, restore at least one RHRSW pump / diesel generator pair
- to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours.
The provisions of Specification 3.0.4 are not applicable.
- 6. With three RHRSW pump / diesel generator pairs
- inoperable, ;
restore at least one RHRSW pump / diesel generator pair
- to j i
OPERABLE status within 7 days or be in at least HOT SHUTDOWN j i within 12 hours and in COLD SHUTDOWN within the following l 24 hours. I 7. With four RHRSW pump / diesel generator pairs
- inoperable, l l restore at least one RHRSW pump / diesel generator pair
- to )
OPERABLE status within 8 hours cr be in at least HOT SHUTDOWN within 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, inoperable, declare the associated RHR loop inoperable and take the ACTION required by Specification 3.4.9.1 or 1 5.4.9.2, as applicable. ,
i
- c. In OPERATIONAL CONDITION 5 with the RHRSW subsystem (s), which is associated with an RHR loop required OPERABLE by Specification l 3.9.11.1 or 3.9.11.2, inoperable, declare the associated RHR system l inoperable and take the ACTION required by Specification 3.9.11.1 or 3.9.11.2, as applicable.
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.
*A RHRSW pump / diesel generator pair consists of a RHRSW pump and its associated diesel generator. If either an RHRSW pump or its associated diesel generator becomes inoperable, then the RHRSW pump / diesel generator pair is inoperable.
1 O LIMERICK - UNIT 2 3/4 7-2
e i i l l
, PLANT SYSTEMS
(( EMERGENCY SERVICE WATER SYSTEM - COMMON' SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops., ~ l 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 suppliea from the spray pond or j the cooling-tower basin and transferring the water to the associated ' Unit 2 and common safety-related equipment, I 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:
- 1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 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 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 otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable **,
restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
*When' handling irradiated fuel in the secondary containment. )
**The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed. Those diesel generators not aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.
l l LIMERICK - UNIT 2 3/4 7-3
L j PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- 4. With three ESW pump / diesel generator pairst inoperable, restore at least one ESW pump / diesel generator pairt to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the ecxt 12 hours and in COLD SHUTDOWN within the following 24 hou: s,
- 5. With four ESW pump / diesel generator pairst inoperable, restore at least one ESW pump / diesel generator pairt 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 4 or 5:
- 1. 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 declare i the associated safety related equipment inoperable and take the I ACTION required by Specifications 3.5.2 and 3.8.1.2. i i
- c. In OPERATIONAL CONDITION *
- 1. 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 verify adequate cooling remains available for the diesel generators required to be GPERABLE or declare the associated diesel genera-tor (s) inoperable and take the ACTION required by Specifica-tion 3.8.1.2. The provisions of Specification 3.0.3 are not applicable.
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 per 18 months 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.
*When handling irradiated fuel in the secondary containment.
tan ESW pump / diesel generator pair consists of an ESW pump and its associated diesel generator. If either an ESW pump or its associated diesel generator becomes inoperable, then the ESW pump / diesel generator pair is inoperable. LIMERICK - UNIT 2 3/4 7-4
i
, PLANT SYSTEMS ULTIMATE HEAT SINK 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'-10" Mean Sea Level, and
- b. A pond water temperature of less than or equal to 88 F.
, APPLICABILITY: OPERATIONAL CONDITIONS 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 eroergency 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 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 hmit at least once per 24 hours.
- b. By verifying the water surface temperature (within the opper 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.
*When handling irradiated fuel in the secondary containment.
LIMERICK - UNIT 2 3/4 7-5
I i PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMFRGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM l LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems chall be OPERABLE. l APPLICABILITY: All OPERATIONAL CONDITIONS and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2, or 3:
- 1. With the Unit 1 diesel generator for one control room emergency fresh air supply subsystem inoperable for more than 30 days, be in at least HOT SHUTDOWN within the next 12 hours and in COLD ;
SHUTDOWN within the following 24 hours. The provisions of I Specification 3.0.4 are not applicable. j i
- 2. With one control room emergency fresh air supply subsystem l
inoperable, restore !Se inoperable subsystem to OPERABLE status l within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. l l 3. With one control room emergency fresh air supply subsystem l inoperable and the other control room emergency fresh air supply subsystem with an inoperable Unit 1 diesel generator, restore the inoperable subsystem to OPERABLE status or restore the Unit 1 diesel generator 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.
- 4. With the Unit 1 diesel generators for both control room emergency fresh air supply subsystems inoperable for more than 72 hours, I
be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.
- b. In OPERATIONAL CONDITION 4, 5 or *:
- 1. With one control room emergent., fresh air supply subsysttm inoperable, restore the inoperatde subsystem to OPERABLE :, Latus within 7 days, or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.
- 2. With both both control room emergency fresh air supply subsystem 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 *.
*When irradiated fuel is being handled in the Secondary Containment.
l J LIMERICK - UNIT 2 3/4 7-6
PLANT SYSTEMS
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l\ m 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-l ture to be less than or equal to 85 F effective temperature.
- b. At least once per 31 days on a STA'aGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.
- c. 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 communicating with the subsystem by:
- 1. Verifying that the subsystem satisfies the in place penetration and bypass leakage testing acceptance criteria cf less than 0.05% and uses the test procedure guidance in Regulatory Positions C.6.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, i March 1973, and the system flow rate is 3000 cfm i 10%.
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LIMERICK - UNIT 2 3/4 7-6a
I l PLANT SYSTEMS I
\ l 1
'(,) 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 latroratory 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 I000 cfm i 10% during subsystem operation when tested in accordance with ANSI N510-1980.
- d. After every 720 hoes of charcoal adsorber operation by verifying within 31 days aftet removal that a laboratory analysis of a repre-sentative carbon sanple obtained in accordance with Regulatory Position C.6.b of Ragulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.i,?. Revision 2, March 1978, for a methyl iodide penetration of less Ban 1%.
- e. 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 (9 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. I ij *An allowable outdoor airflow rate of less than or equal to 2100 cfm is per-missible until the issuance of the Unit 2 full power operating license. LIMERICK - UNIT 2 3/4 7-7
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) Ol
- 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 N510-1SB0 while operating the system at a flow rate of 3000 cfm i 10%.
- g. After each complete or partial replacement of a charcoal adsorber i bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakt.ge 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%.
l l l l O O! LIMERICK - UNIT 2 3/4 7-8
PLANT SYSTEMS ( 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 1 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLL . flow path capable of automatically taking suction from the suppressiori 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 tne 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.
/"
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 ccntroller is in the correct position.
- b. At least once per 92 days by verifying that the P.CIC 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 p adequate to perform the test. If OPERABILITY is not successfully demonstrated Q within the 12-hour period, reduce reactor steam pressure to less than 150 psig within the following 72 hours.
LIMERICK - UNIT 2 3/4 7-9
1
\
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4 W
- c. At least once per 18 months by: 1
- 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 l vessel may be excluded. d
- 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.* 1
- 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 leveElow signal.
- 4. Performing a CHANNEL CALIBRATION of the RCIC sjstem discharge line " keep filled" level alarm instrumentation.
O 1 l I l
*heprovisicasofSpecification4.0.4arenotapplicableprovidedthe siirveillar.ce is performed within 12 hours af ter reactor steam pressure is adequate to perform the tests. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce, reactor steam dome pressure to less than 150 psig within the following 72 hours.
LIMERICK - UNIT 2 3/4 7-10
PLANT SYSTEMS gx 3/4.7.4 SNUBBERS ( LIMITING CONDITION FOR OPERATION 3.7.4 All snubbers shall be OPERABLE. l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE in those OPERATIONAL CONDITIONS. ACTION: l l 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 j i attached system inoperable and follow the appropriate ACTION statement for that I system. SURVEILLANCE RE0VIREMENTS 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of l Specification 4.0.5.
- a. Inspection Types l(,
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 l after 4 months but within 10 months of commencing POWER OPERATION and l shall include all snubbers. If all snubbers of each type on any system l 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 i schedule:
LIMERICK - UNIT 2 3/4 7-11
PLANT SY TEMS SURVEILLANCE REQUIREMENTS (Continued) No. Inoperable Snubbers of Each Type on Any System Subsequent Visual per Inspection Period Inspection Period *# 0 18 months i 25% 1 12 months i 25% 2 6 months i 25% . I 3,4 124 days i 25% 5,6,7 62 days i 25% 8 or more 31 days i 25%
- c. Visual Ir.. 0ection 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 J 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 3' for that particular siubber and for o+.her snubbers irrespective of type on that system thet may be generically susceptible; and/or (2) the , affected snubber is functionally tested in the as found condition and - l determined OPERABLE per 5 specifications 4.7.4f. For those snubbers l common to more than one <ystem, the 0PERABILITY 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 thereafter if no inoperable snubbers of that type are found on that system. #The provisions of Specification 4.0.2 are not applicable.
LIMERICK - UNIT 2 3/4 7-12 l
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- i '- PLANT SYSTEMS tj SURVEILLANCE REQUIREMENTS (Continued)
- e. Functional Tests During t' first refueling shutdown and at least once per 18 months thereafter,-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 Administrator shall be 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:
'~
- 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 sampit of each type of snubber shall be functionally tested in accordance with Figure 4.7.4-1. "C" is the total number of snubt ers of a type found not meeting the acceptance requirements m b cification 4.7.4f. The cumulative number of snubbers of a type tc'ad 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 tested. If at any time the point plotted falls on or below the
" Accept" line, testing of snubbers 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 " Accept" region or the " Reject" region, or all the snubbers of that type have been tested. Testing equipment failure 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
- 3) An initial representative sample of 55 snubbers of each type shall 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 point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept" line, testing of that type of snubber may be terminated. If the point plotted falls above the " Accept" line, testing must continue until the point falls on or below the " Accept" line or all the snubbers of that type have been tested.
4 LIMERICK - UNIT 2 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 l the sample plan, and failure of this functional test shall not be the l 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 l functional testing. l f. Functional Test Acceptance Criteria l The snubber functional test shall verify that: l 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 tre specified range in both directions of travel; and
- 4) For snubbers specifically required not to displace under continuous load, the ability of the snu)ber 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. Functi,onal 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 2 3/4 7-14
PLANT SYSTEMS' '%/ f' '} _ 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 l stated in Specification 4.7.4e. for snubbers not meeting the j' functional test acceptance criteria.
- h. Functional Testing of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test l acceptance criteria shall be repaired or replaced. Replacement L 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 l The service life of all snubbers shall be monitored to ensure that l the service life is not exceeded between surveillance inspections.
l The maximum expected service life for various seals, springs, and i other critical parts sha'll 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.0PERABLE. The , parts replacements shall be documented and the documentation shall I be retained in accordance with Specification 6.10.3. l l l l LIMERICK - UNIT 2 3/4 7-15 i
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l l FIGURE 4.7.4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST O LIMERICK - UNIT 2 3/4 7-16
PLANT SYSTEMS
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3/4.7.5 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.5 Each sealed source containing radioactive material either in excess of j 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha l emitting material shall be free of greater than or equal to 0.005 microcurie l of removable contamination. l 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.
- b. The provisions of Specification 3.0.3 are not applicable.
( 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
- 2. In any form other than gas.
I LIMERICK - UNIT 2 3/4 7-17
f 1 l !- .i 1 1 PLANT 5YSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. l Stored sources not in use - Each sealed source and fission detector i shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources anc fission f
{ detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.
- c. Startup sources and fission detectors - Each sealed startup source
- and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair k or maintenance to the source. {
l 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.
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*Except the Cf-252 startup sources which shall be tested within 6 months prior to being subjected to core flux or installed in the core and following repair i or maintenance to the source. I i
LIMERICK - UNIT 2 3/4 7-18 l
L PLAN 1 sVSTEMS 3/4.7.6 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 The tire 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 capabis of taking suction from the Unit 1 Cooling Tower Basin and tht: Unit 2 Cooling Tower Basin and transfer-ring the water through di c ribution 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 p e-action sprinkler system and the last valve ahead of the fire hose :tations' required to be OPERABLE per Specifica-tions 3.7.6.2, 3.7.6.5, ed 3.7.6.6.
. APPLICABILITY: At all times.
T' 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 Specification 3.0.3 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 mirates on recirculation flow.
l l c. At least once per 31 days by verifying that each valve (manual, power-l operated, or automatic) in the flow path is in its correct position. l 1 LIMERICK - UNIT 2 3/4 7-19
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 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:
l 1. Verifying that each fire suppressic,n pomo develops at least l 2500 gpm at a system head of 125 psig, l l
- 2. Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, 1 and i
- 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.
l
- 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 Tabir 1 of ASTM D975-77 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 o' service.
O LIMERICK - UNIT 2 3/4 7-20 f 1
i PLANT SYSTEMS
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1 U SURVEILLANCE REQUIREMENTS (Continued) 4.7.6.1.3 The diesel-driven fire pump starting 24-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1. The electrolyte level of each cell is above the plates,
- 2. The pilot cell specific gravit1, corrected to 77 F and full electrolyte level, is greater than or equal to 1.260, and
- 3. The overall battery voltage is grtater 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.
- 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
,m 2. Battery-to-battery and terminal connections are clean, tight,
( free of corrosion, and coated with anticorrosion material. N LIMERICK - UNIT 2 3/4 7-21
I PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS i$1TINGCONDITIONFOROPERATION 3.7.6.2 The follow;ng spray and sprinkler systems shall be OPERABLE: Fire Zone Description l Reactor Enclosure Hatchway Water Curtains:
- 1. EL 253'
- 2. El 283' 1
- 3. EL 313' i Fire Area Separation Water Curtains: l 71A 1. Area 638, EL 313' 68A 2. Area 475, EL 253' 67 3. Area 370, EL 217' 23 Cable Spreading Room, Room 450, EL 254',
27 Control Structure Fan Room, Room 619, EL 304' I 27 CREFAS System Filters, EL 304' i 28A SGTS Access Area 625, EL 332' 28B SGTS Filters, Compartment 624, EL 332' 56 RCIC Pump Room, Room 179, EL 177' 57 HPCI Pump Room, Room 180, EL 177' 64A RECW Area 284, EL 201' 65 Safeguard System Access Area 279, EL 201' 67 Safeguard System Access Area 370, El 217' (Partial) (3 systems) 68A CRD Hydraulic Equipment Area 475, Reactor Enclosure, EL 253' (Partial) (3 systems) 70A General Equipment Area 574 and Corridor 580, Reactor Enclosure, EL 283' (Partial) 74A & B Reactor Enclosure Recirculation System Filters Areas 651 & 653, EL 331' 83,84,85,86 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 are not applicable.
O LIMERICK - UNIT 2 3/4 7-22 l
\,
r PLANT SYSTEMS r 'i 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 1er 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 sigtal, and b) Cycling each valve in the flow path that b not testable during plant operation through at least one complete cycle of full travel.
- 2. By a visual inspection of the dry pipe spray and sprinkler O
3. headers to verify their integrity, and 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 vceifying each open head spray nozzle and sprinkler heacler 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.
1. 1 O LIMERICK - UNIT 2 3/4 7-23
PLANT SYSTEMS CO2 SYSTEM _ 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 OHR602.
APPLICABILITY: Whenever equipment protected by the CO2 system 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 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 CO2 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. l l 9 LIMERICK - UNIT 2 3/4 7-24
I i PLANT SYSTEMS I (('h1 HALON SYSTEMS p (Q/ L 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 I 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 are not applicable.
SURVEILLANCE REQUIREMENTS
- ('
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 l 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.
m LIMERICK - UNIT 2 3/4 7-25
PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 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 one or more of the fire hose stations shown in Table 3.7.6.5-1 inoperable, provide 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 connected to a length of hose sufficient to provide coverage for the area left unprotected by the inopereble hose station.
Where it can be demonstrated that the physical routing of the fire i hose would result in a recognizable hazard to operating technicians, l plant equipment, or the hose itself, the fira 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 primary means of fire suppression; otherwise route the additional hose within 24 hours.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.6.5 Each of 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 equipment is at the station.
- b. At least once per 18 months by:
- 1. Visual inspection 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 alve 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. l l
)
LIMERICK - UNIT 2 3/4 7-26 I
il J TABLE 3.7.6.5-1 ) l [~'N , FIRE HOSE STATIONS l t
) 1 H0SE RACK LOCATION ELEVATION 10ENTIFICATION
- 1. , Control
Enclosure:
Stairwell 350' 1HR-141 Stairwell, Outside SGTS Room 332' 1HR-140 Stairwell, Outside Fan Room 304' 1HR-103 Outside 13kV Switchgear Room 217' 2HR-116 Stairwell, Outside Aux Equip Rm 289' 1HR-130 Inverter Room Number 2-453 254' 2HR-25t' Wall, Outside 4kV Switchgear & Battery Rooms 239' 2HR-251 Corridor 466, South Side of 4kV Switchgear & Battery Rooms 239' 2HR-122 Wall, Corridor 277 200' 2HR-120 Wall, Corridor 166 180' 2HR-121
- 2. Refueling Area:
I' ( SE Corner Refuel Floor 352' 2HR-201 NE Corner Refuel Floor 352' 2HR-202 North Wall-Center 352' 2HR-203 South Wall-Center 352' 2HR-204
- 3. Reactor Enclosure Unit 2:
SE Corner Reactor Enclosure 331' 2HR-205 SE Corner Reactor Enclosure (RERS Fan Area) 313' 2HR-207 NE Corner Reactor Enclosure (Laydown Area 638) 313' 2HR-208 SW Corner Reactor Enclosure (Near Refuel Floor Exh. Fans) 313' 2HR-209 NW Corner Reactor Enclosure (Near Load Center) 313' 2HR-210 SE Corner Reactor Enclosure (Corridor 580) 283' 2HR-215 NE Corner Reactor Enclosure (Corridor 580) 283' 2HR-216 t LIMERICK - UNIT 2 3/4 7-27
TABLE 3.7.6.5-1 (Continued) FIRE HOSE STATIONS HOSE RACK ! LOCATION ELEVATION IDENTIFICATION
- 3. Reactor Enclosure Unit 2: (Continued)
SW Corner Reactor Enclosure (SLC Pumps Area 574) 283' 2HR-217 NW Corner Reactor Enclosure 283' 2HR-218 SE Corner Reactor Enclosure l (Area 475, Near CR0 Maintenance i Room) 253' 2HR-223 i NE Corner Reactor Enclosure l (Near Elev. & Stair No. 6) 253' 2HR-224 West Wall Reactor Enclosure (Near Unit 1/ Unit 2 Airlock) 253' 2HR-225 West Wall Reactor Enclosure i (Near Stair No. 2) 253' 2HR-226 l SE Corner Reactor Enclosure ) (Near RCIC Equip Hatch) 217' 2HR-232 NE Corner Reactor Enclosure (Near Supp Pool Access Hatch) 217' 2HR-233 West Wall Reactor Enclosure (Near Personnel Airlock 366) 217' 2HR-234 NW Corner Reactor Enclosure ( % r 5tair No. 2) 217' 2HR-235 SE Corner Reactor Enclosure (Near Stair No. 5) 201' 2HR-240 NE Corner Reactor Enclosure (Near Elev. & Stair No. 6) 201' 2HR-241 West Wall Reactor Enclosure (Near RECW Heat Exchangers) 201' 2HR-242 NW Corner Reactor Enclosure ; (Near RECW Pumps) 201' 2HR-243 SE Corner Reactor Enclosure 177' 2HR-252 NE Corner Reactor Enclosure 177' 2HR-253 NW Corner Reactor Enclosure 177' 2HR-236 O LIMERICK - UNIT 2 3/4 7-28
PLANT SYSTEMS YARD FIRE HYDRANTS AND HOSE CART HOUSES LIMITING CONDITI0'4 FOR OPERATION 3.7.6.6 The yard fire hydrants and hose cart houses shown in Table _3.7.6.6-1 i shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE. i ACTION: i
- 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 service 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 Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.6.6 Each of the yard fire hydrants and hose cart houses shown in i Table 3.7.6.6-1 shall be demonstrated OPERABLE: j j
.a . At least once per 31 days by visual inspection of the hose cart j house to assure all required equipment is at the hose house. i
- 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 and verifying that the hydrant barrel is dry and that the hydrant is not damaped.
i
- c. At least once per 12 inor- . 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.
LIMERICK - UNIT 2 3/4 7-29
TABLE 3.7.6.6-1 YG0 FIRE HYDRANTS AND HOSE CART HOUSES LOCATION HYORANT NUMBER East of Diesel Generator Enclosure FH #9 South of Diesel Generator Enclosure FH #8 LOCATION HOSE CART HOUSE NUMBER East of Diesel Generator Ere h r..are HCH #5 O l O:f l LIMERICK - UNIT 2 3/4 7-30
-l l
4 i PLANT SYSTEMS' 3/4.7.7 FIRE RATED ASSEMBLIES 1 LIMITING CONDITION F R 0 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.
, 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.
- l. b. The provh.~ons of Specification 3.0.3 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 inspec+ ion of an additional 10% of each type of sealed penetration shall be ma6e. This inspection process shall !
continite 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. LIMERICK - UNIT 2 3/4 N 1
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 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 j latches at least once per 6 months, and by verifying: - That each locked-closed fire door is closed at least once per ( a. 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.
l b. By verifying the OPERABILITY of the fire door supervision system l for each electrically supervised fire door by performing a CHANNEL l FUNCTIONAL TEST at least once per 31 days. 1
- c. By inspecting the closing mechanism and latches at least once per 6 months.
I i O LIMERICK - UNIT 2 3/4 7-32
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.-1 'A.C. SOURCES A~ A.C. SOURCES - OPERATING
= LIMITING CONDITION FOR OPERATION J
3.8.1.1 As a minimum, the following A.C. electrical power sources shal1~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 L 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 O at least once per 7 days thereafter; restore the inoperable diesel generator to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUT 00WN within the following 24 hours. See also ACTION e.
- 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 v least once per 8 hours the eafter; restore at least one of the inoper-able diesel generators to Or'ERABLE status within 721:ours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. See also ACTION e,
- c. With three diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of tha remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a. and 4.8.1.1.2a.4. , for one diesel pnerator 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. See also ACTION e,
- d. With one offsite circuit and one diesel geaerator of the above required A.C. electrical power sc,urces 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 I hour and at least once per 8 hours thereafter. Restore at least two offsite circuits and at LIMERICK - UNIT 2 3/4 6-1 m- _ - - - - - - - - - - - - - , - - _ _ - - - - - - - - _ _ _ . . _ . _ _ _ . , _ , _ , . _ _ . _ _ _ _ , , _ _ _ . _ _ _ _ _ _ _ , _ _ _ _ _
~
i l 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. See also ACTION e.
- e. In addition to the ACTIONS above:
- 1. For two train systems, with one or more diesel generators of the above required A.C. electrical power sources inoperable, verify within 2 hours and at least once per 12 hours thereafter that at least one of the required two train system subsystem, train, components, and devices is OPERABLE and its associated
! diesel generator is OPERABLE. Otherwise, restore either the j inoperable diesel generator or the inoperable system subsystem i to an 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.
- 2. For the LPCI systems, with two or more diesel generators of the above required A.C. electrical power sources inoperable, verify l within 2 hours and at least once per 12 hours thereafter that at least two of the rsquired LPCI system subsystems, trains, components and devices are OPERABLE and its associated diesel
! penerator is OPERABLE. Otherwise, be in at least HOT SHUTDOWN l thin the next 12 hours and in COLD SHUTDOWN within the following
/f hours.
Th.s ACTION does not apply for those systems covered in Specifications 3.7.1.1 and 3.7.1.2.
- 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 4.8.1.1.2a.4., for one diesel generator at a time, within 3 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 hours.
- g. With 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 I 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 wi+n 72 hours from time of initial loss or be in at least HOT SHUTD0kA . thin the next 12 hours ll and in COLD SHUTDOWN within the following 24 hours.
1 LIMERICK - UNIT 2 3/4 8-2 l
s ELECTRICAL POWER SYSTEMS
;m N
U
)'
LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- h. With one offsite circuit and two diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY j 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. See also ACTION e. I Na
/O lV LIMERICK - UNIT 2 3/4 8-2a
ELECTRICAL POWER SYSTEMS L[ } SURVEILLANCE REQUIREMENTS J 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class IE distribution system shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct i breaker alignments and indicated power availability, and j
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by j transferring, manually and automatically, unit power supply from the I normal circuit to the alternate circuit.
4.8.1.1.2 Each of the above required diesel generators shall be demonstrated OPERABLE:
- a. In accordance whh the frequency specified in Table 4.8.1.1.2-1 on a STAGGERED TEST BASIS by:
- 1. Verifying the fuel level in the day fuel tank.
- 2. Verifying the fuel level in the fuel storage tank.
- 3. Verifying the fuel transfer pump starts and transfers fuel from thr. 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.
The generator voltage and frequency shall reach 4285 1 420 volts and 60 1 1.2 Hz within 10 secondr after the start signal. The [h 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 a
'ctuation 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 200 seconds, and operates with this load for at least 60 minutes.
- 6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
- 7. Verifying the pressure in all diesei 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 stress and wear on the diesel engine is minimized. ,C **If diesel generator started manually from the control room,10 seconds af ter
( the automatic prelube period. LIMERICK - UNIT 2 3/4 8-3
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. By removing accumulatt d water:
- 1) From the day tank at least once per 31 days and after cach occa-sion when the diesel is operated for greater than 1 hour, and j
- 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 {
0975-81 prior to addition to the storage tanks that the sample has: 1 a) An API Gravity of within 0.3 degrees at 66 F or a specific gravity of within 0.0016 at 60/60 F, when compared to the l supplier's certificate or an absolute specific gravity l at 60/60 F of greater than or equal to 0.83 but less than or I equal to 0.89 or an API gravity at 60 F of greater than or equal to 27 degrees but less than or equal W 39 degrees. I b) A kinematic viscosity at 40 C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification. c) A flash point equal to or greater than 125 F, and d) A clear and bright appearance with prop e color when tested in accordance with ASTM D4176-82.
- 2) By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM 0975-81 exceot that the analysis for sulfur may be performed in accordance with
- ASTM D1552-79 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, except that the I filters specified in ASTM D2276-78, Sections 5,1.6 and 5.1.7, may I have a nominal pore size of up to three (3) microns,
- 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 manuf acturer'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 1 420 volts and frequency at 60 1.2 bz.
LIMERICK - UNIT 2 3/4 8-4
;;W s -
'~
~ l ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Coritinued)
, 3. Verifying the ' diesel = generator capability to reject a. loa'd 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 buses and load shedding from the emergency buses. I b) Verifying the diese' generator starts on the auto start signal, energizes the emergency buses within 10 seconds, energizes the auto-connected loads throtTh the individual-load timers and operates for greater than or equel to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency buses shall be insintained at 4285 1 420 volts and 60 1 1.2 Hz during this test.
- 5. Verifying that on an ECCS actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall reach O 4285 1 420 volts and 60 1 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 ettuation test signal, and:
a) Verifying deenergization of the emergency buses and load sheddin0 from the emergency buses. b) Verifying the diesel generator starts on the auto-start signal, energizes the emergency buses within 10 seconds, energizes the auto-connected shutdown loads through the individual load timers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads, After energization, the steady-state voltage and frequency of the emergency buses shall be maintained at 4285 1 420 volts and 60 i L2 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.
LIMERICK - UNIT 2 3/4 8-5
( i r ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 8. Verifying the diesel generator operates for at least 24 hours.
During the first 2 hours of this test, the diesel generator shall i be loaded to greater than or equal 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 cap hility 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 status.
- 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. **If diesel generator started manually from the control room, 10 seconds after the automatic prelube period.
I O 1 1 LIMERICK - UNIT 2 3/4 8-6 j a
i i ELECTRICAL POWER SYSTEMS l [ \ SURVEILLANCE REQUIREMENTS (Continued)
- 13. Verifying that the following diesel generator lock 6ut features i prevent diesel generator starting only when required: I a) Control Room Switch In Pull-To-Lock (With Local / Remote Switch in Remote) b) local / Remote Switch in Local. -i 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 diesel 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 t 2. Performing a pressure test of those portions of the diesel fuel
( 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. c* LIMERICK - UNIT 2 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 l
2 At least once per 14 days 3 At least once per 7 days
>4 At least once per 3 days l
l
- 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 v& lid tests." Entry into this test schedule shall be made at the 31-day test frequency.
i O LIMERICK - UNIT 2 3/4 8-8
1 I ELECTRICAL POWER SYSTEMS 1s A.C. SOURCES - SHUTDOWN 1 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 *. ACTION:
- 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.
SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electricc1 power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2, and 4.8.1.1.3, except for the requirement of Specification 4.8.1.1.2a.5.
*When handling irradiated fuel in the secondary containment.
LIMERICK - UNIT 2 3/4 8-9
ELECTRICAL POWER SYSTEMS 3/4.8.2 D.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 2A1 (2 AID 101).
- 2. 125-Volt Battery 2A2 (2A2D101).
- 3. 125-Volt Battery Charger 2BCA1 (2 AID 103).
- 4. 125-Volt Battery Charger 2BCA2 (2A2D103). J
- b. Division 2, Consisting of:
- 1. 125-Volt Battery 2B1 (2B10101).
- 2. 125-Volt Battery 2B2 (2B20101).
- 3. 125-Volt Battery Charger 2BCB1 (2810103).
- 4. 125-Volt Battery Charger 2BCB2 (2B2D103).
- c. Division 3, Consistiag of:
- 1. 125-Volt Battery ?C (2CD101). '
- 2. 125-Volt Battery Charger 2BCC (2CD103).
- d. Division 4, Consisting of:
- 1. 125-Volt Battery 2D (2DD101).
- 2. 125-Volt Battery Charger 2BCD (20D103).
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 B 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 i
- 2. Total battery terminal voltage for each 125-volt battery is i greater than or equal to 131 volts on float charge.
LIMERICK - UNIT 2 3/4 8-10
i l ELECTRICAL POWER SYSTEMS e. l SURVEILLANCE REQUIREMENTS (Continued) 3
- b. At least once per 9? days and within 7 days after a battery discharge with battery termir,a1 voltage below 105 volts or battery overcharge l
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, or the connection resistance of these items is less than 150 x 10 6 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 less than or equal to 150 x 10 6 ohm excluding cable intercell connections, and 7
k 4. The battery chargers will supply the currents listed below at a minimum of 132 volts for at least 8 hours: Charger Current (Amperes) 2BCA1 300 2BCA2 300 2BCB1 300 2BCB2 300 2BCC 75 2BCD 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:
LIMERICK - UNIT 2 3/4 B-11
[ h: ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) LOAD CYCLE (amps) Division Battery 0-1 Min. 1-239 Min. 239-240 Min. I 2A1 546 168 187 2A2 449 129 147 II 2B1 889 158 321 2B2 823 119 282 III 2C 193 31 31 IV 20 169 21 21 Each 125/250-volt battery is rated at 1500 ampere-hours at an 8-hour oischarge 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 i;.ast 30% of the manufacturer's rating when subjected to a performance discharge test. At this once per 60 month interval, this performance discharge test ;nay be performed in lieu of the battery service test (Specification 4.8.2.1.d).
- f. At least once per 18 months during shutdcwn 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 2 3/4 8-12
a l 7- , I TABLE 4.8.2.1-1 .[ BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2)
\
Parameter Limits for each Limits for each A110wable(3) designated pilot connected cell value for each cell connected cell Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and < \" above and < k" above and not maxiEum level maxiium level overflowing-indication mark indication mark Float Voltage 1 2.13 volts 3 2.13 volts (4) > 2.07 volts Not more than 0.020 below the-Specific average of all Gravity (5)- 2 1.195(6) 1 1.190 connected cells Average of all Average of all connected cells connected cells
> 1.200 1.190(6)
(1)For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the 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.
(2)For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within their allowable values and provided the Category B parameter (s) are restored to within limits w' thin 7 days. (3)Any Category B parameter not within its allowable value indicates an inoperable battery. (4)May be corrected for average electrolyte temperature. ( ) Corrected for electrolyte temperature of 77 F and full level. (6)0r battery charging current is less than 1 amperes when on float charge. ( l LIMERICK - UNIT 2 3/4 8-13
i I 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-l {
- a. Division 1, Consisting of:
- 1. 125-Volt Battery 2A1 (2 AID 101). {
l
- 2. 125-Volt Battery 2A2 (2A2D101). I
- 3. 125-Volt Battery Charger 2BCA1 (2 AID 103). l
- 4. 125-Volt Battery Charger 2BCA2 (2A2D103). {
j
- b. Division 2, Consisting of:
! 1. 125-Volt Battery 2B1 (2B10101). l 2. 125-Volt Battery 2B2 (2B2D101).
- 3. 125-Volt Battery Charger 2BCB1 (2B1D103).
- 4. 125-Volt Battery Charger 2BCB2 (2B20103).
- c. Division 3, Consisting of:
- 1. 125-Volt Battery 2C (2C0101).
- 2. 125-Volt Battery Charger 2BCC (200103).
- d. Division 4, Consisting of:
- 1. 125-Volt Battery 2D (20D101).
- 2. 125-Volt Battery Charger 2BCD (2DD103).
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 i potential for draining the reactor vessel. !
l
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS i 4.8.2.2 At least the above required batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.
*When handling irradiated fuel in the secondary containment.
LIMERICK - UNIT 2 3/4 8-14
ELECTRICAL POWER SYSTEMS
,c-s ,
3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS l 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. Unit 2 Division 1, Consisting of:
a) 4160-VAC Bus: D21 (20A115) b) 480-VAC Load Center: D214-(20B201) c) 480-VAC Motor Control Centers: D214-R-C (20B213) 0214-R-G (20B211) D214-R-G1 (20B215) D214-D-G (20B515) d) 120-VAC Distribution Panels: 20Y101 20Y206
- 2. Unit 2 Division 2, Consisting of:
a) 4160-VAC Bus: D22 (20A116) b) 480-VAC Load Center: D224 (20B202) c) 480-VAC Motor Control Centers: 0224-R-C (20B214) / D224-R-G (20B212)
- t. D224-R-G1 (20B216) 0224-D-G (20B516) d) 120-VAC Distribution Panels: 20Y102 20Y207
- 3. Unit 2 Division 3, Consisting of:
a) 4160-VAC Bus: D23 (20A117) b) 480-VAC Load Center: D234 (208203) c) 480-VAC Motor Control Centers: D234-R-H1 (20B221) D234-R-H (20B217) 0234-R-E (20B223) D234-D-G (20B517) d) 120-VAC Distribution Panels: 20Y103 20Y163
- 4. Unit 2 Division 4, Consisting of:
a) 4160-VAC Bus: D24 (20A118) b) 480-VAC Load Center: D244 (20B204) c) 480-VAC Motor Control Centers: D244-R-H1 (20B222) 0244-R-H (20B218) D244- R- E (20B224) 0244-D-G (20B518) d) 120-VAC Distribution Panels: 20YiO4 20Y'.' 64 O LIMERICK - UNIT 2 3/4 8-15
i ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
)
- 5. Unit 1 and Common Division 1, Consisting of:
a) 4160-VAC Bus: D11 (10A115) b) 480-VAC Load Center: D114 (10B201) l c) 480-VAC Motor Control Centers: 0114-R-C (108213) J D114-R-C1 (10B219) ; D114-D-G (108515) l D114-5-L (00B519) d) 120-VAC Distribution Panels: 10Y101 1 10Y206 01Y501 ]
- 6. I' nit 1 and Common Division 2, Consisting of: i l a) 4160-VAC Bus: D12 (10A116) b) 480-VAC Load Center: 0124 (108202) c) 480-VAC Motor Control Centers: D124-R-C (10B214) 0124-R-C1 (10B220) 0124-D-G (108516)
D124-S-L (00B520) d) 120-VAC Distribution Panels: 10Y102 10Y207 02Y501
- 7. Unit I and Common Division 3, Consisting of:
a) 4160-VAC Bus: D13 (10A117) b) 480-VAC Load Center: 0134 (10B203) c) 480-VAC Motor Control Centers: D134-R-E (10B223) D134-C-B (00B131) D134-D-G (10B517) D234-S-L (008521) d) 120-VAC Distribution: 10Y103 10Y163 03Y501
- 8. Unit 1 and Common Division 4, Consisting of:
a) 4160-VAC Bus: D14 (10A118) b) 480-VAC Load Center: 0144 (10B204) c) 480-VAC Motor Control Centers: 0144-R-E (108224) D144-C-B (00B132) D144'D-G (10B518) D244-5-L (00B522) d) 120-VAC Distribution: 10Y104 10Y164 04Y501 O LIMERICK - UNIT 2 3/4 8-16
)
v
+
ELECTRIr.AL POWER SYSTEMS A l kj LIMITING CONDITION FOR OPERATION (Continued)
- b. D.C. Power Distribution Panels
- 1. Unit 2 Division 1, Consisting of:
a) 250-V DC Fuse Box: 2FA (2AD105) b) 250-V DC Motor Control Centers: 2DA (20D201) c) 125-V DC Distribution Panels: 2 PPA 1 (2AD102) 2 PPA 2 (2AD501) 2 PPA 3 (2AD162)
- 2. Unit 2 Division 2, Consisting of:
a) 250-V DC Fuse Box: 2FB (2BD105) b) 250-V DC Motor Control Centers: 2DB-1 (20D?02) 2D8-2 (200203) c) 125-V DC Distribution Panels: 2 PPB 1 (2BD102) 2 PPB 2 (2BD501) 2 PPB 3 (2BD162)
- 3. Unit 2 Division 3, Consisting of:
a) 125-V DC Fuse Box: 2FC (20D105) b) 125-V DC Distribution Panels: 2PPC1 (200102) 2PPC2 (200501) 2PPC3 (200162)
, 4. Unit 2 Division 4, Consisting 6f:
1 V) a) b) 125-V DC Fuse Box: 125-V DC Distribution Panels: 2FD 2 PPD 1 (20D105) (200102) 2 PPD 2 (2DD501) 2 PPD 3 (2DD162)
- 5. Unit 1 and Common Division 1, Consisting of:
a) 250-V DC Fuse Box: 1FA (1AD105) b) 125-V DC Distribution Panels: 1 PPA 1 (1AD102) IPPA2 (1AD501)
- 6. Unit 1 and Common Division 2, Consisting of:
a) 250-V DC Fuse Box: IFB (1BD105) b) 125-V DC Distribution Panels: IPPB1 (18D102) IPPB2 (1BD501)
- 7. Unit I and Common Division 3, Consisting of:
a) 250-V DC Fuse Box: 1FC (ICD 105) b) 125-V DC Distribution Panels: IPPCI (ICD 102) IPPC2 (ICD 501)
- 8. Unit 1 and Common Division 4, Consisting of:
a) 250-V DC Fuse Box: IFD (IDD105) b) 125-V DC Distribution Panels: IPPD1 (1DDIN) IPPD2 (1DD501) LIMERICK - UNIT 2 3/4 8-16a
i I l 4
~
ELECTRICAL POWER SYSTEMS rs ,
\s / LIMITING' CONDITION FOR OPERATION (Continued) i
- l l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
-ACTION:
L a. With one' of the above required Unit 2 A.C. distribution system l l 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. l i
- b. -Mith one of the above required Unit 2 D.C. distribution' system !
civisions 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. I
- c. With any of the above required Unit 1 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.
I SURVEILLANCE REQUIREMENTS 1/ . it 4.8.3.1 . Each of the above required power distribution system divisions.shall ls\ be determined energized at least once per 7 days by verifying correct breaker i alignment and voltage on the busses /MCCs/ panels. i l ! I 1 LIMERICK - UNIT 2 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. Unit 2 Division 1, Consisting of:
a) 4160-VAC Bus: D21 (20A115) b) 480-VAC Load Center: D214 (20B201) c) 480-VAC Motor Control Centers: D214-R-C (20B213) D214-R-G (20B211) D214-R-G1 (20B215) D214-D-G (20B515) d) 110-VAC Distribution Panels: 20Y101 20Y206
- 2. Unit 2 Division 2, Consisting of:
1 a) 4160-VAC Bus: D22 (20A116) b) 480-VAC Load Center: D224 (20B202) c) 480-VAC Motor Control Centers: D224-R-C (20B214) D224-R-G (20B212) D224-R-G1 (20B216) 0224-D-G (20B516) d) 120-VAC Distribution Panels: 20Y102 20Y207
- 3. Unit 2 Division 3, Consisting of:
a) 4160-VAC Bus: D23 (20A117) b) 480-VAC Load Center: D234 (20B203) c) 480-VAC Motor Control Centers: 0234-R-H1 (20B221) D234-R-H (20B217) D234-R-E (208223) D234-D-G (20B517) d) 120-VAC Distribution Panels: 20Y103 20Y163
- 4. Unit 2 Divis-ion 4, Consisting of:
a) 4160-VAC Bus: D24 (20A118) b) 480-VAC Load Center: D244 (20B204) c) 480-VAC Motor Control Centers: D244-R-H1 (20B222) D244-R-H (20B218) D244-R-E (20B224) 0244-D-G (20B518) d) 120-VAC Distribution Panels: 20Y104 20Y164
- 5. Unit 1 and Common Division 1, Consisting of:
a) 4160-VAC Bus: D11 (10A115) b) 480-VAC Load Center: 0114 (10B201) LIMERICK - UNIT 2 3/4 8-18
t i ELECTRICAL POWER SYSTEMS g). x - LIMITING CONDITION FOR OPERATION (Continued) ; c) 480-VAC Motor Control Centers: D114-R-C (108213) D114-R-C1 (10B219) D114-D-G (10B515) D114-S-L (00B519) d) 120-VAC Distribution Panels: 10Y101 10Y206 01Y501
- 6. Unit 1 and Common Division 2, Consisting of:
a) 4160-VAC Bus: D12 (10A116) b) 480-VAC Load Center: D124 (10B202) c) 480-VAC Motor Control Centers: D124-R-C (108214) D124-R-C1 (108220) D124-D-G (108516) 0124-S-L (00B520) d) 120-VAC Distribution Panels: 10Y102 10Y207 02Y501
- 7. Unit 1 and Common Division 3, Consisting of:
a) 4160-VAC Bus: 013 (10A117) b) 480-VAC Load Center: D134 (10B203) f'w) t
v c) 480-VAC Motor Control Centers: D134-R-E (10B223)
D134-C-B (00B131) D134-D-G (10B517) D234-S-L (008521) d) 120-VAC Distribution: 10Y103 10Y163 03Y501 ,
- 8. Unit 1 and Common Division 4, Consisting of:
a) 4160-VAC Bus: 014 (10A118) b) 480-VAC Load Center: D144 (108204) c) 480-VAC Motor Control Centers: D144-R-E (10B224) D144-C-B (00B132) D144-D-G (108518) D244-S-L (00B522) d) 120-VAC Distribution: 10Y1C4 10Y164 04Y501
- b. D.C. power distribution:
l
- 1. Unit 2 Division 1, Consisting of: i a) 250-V DC Fuse Box: 2FA (2AD105) b) 250-V DC Motor Control Center: 2DA (200201) 4 LIMERICK - UNIT 2 3/4 8-18a
4 ELECTRICAL POWER SYSTEMS i ,m
) LIMITING CONDITION FOR OPERATION (Continued) _ ,
,v ' c) 125-V DC Distribution Panels: 2 PPA 1 (2AD102) 2 PPA 2 (2AD501) < 2 PPA 3 (2AD162) )
- 2. Unit 2 Division 2, Consisting of: .l a) 250-V DC Fuse Box: 2FB (2BD105) b) 250-V DC Motor Control Centers: 2DB-1 (20D202) i 2DB-2 (20D203) c) 125-V DC Distribution Panels: 2 PPB 1 (2BD102) 2 PPB 2 (2BD501) 2 PPB 3 (2BD162)
- 3. Unit 2 Division 3, Consisting of:
a) 125-V DC Fuse Box: 2FC (200105) b) 125-V DC Distribution Panels: 2PPCI (2CD102) 2PPC2 (200501) 2PPC3 (2CD162)
- 4. Unit 2 Divicion 4, Consisting of:
a) 125-V DC Fuse Box: 2FD (20D105) b) 125-V DC Distribution Panels: 2 PPD 1 (200102) 2 PPD 2 (200501) 2 PPD 3 (2DD162)
- 5. Unit 1 and Common Division 1, Consisting of:
/ a) 250-V DC Fuse Box: IFA (1AD105)
(9) b) 125-V DC Distribution Panels: 1 PPA 1 1 PPA 2 (1AD102) (1AD501)
- 6. Unit 1 and Common Division 2, Consisting of:
a) 250-V DC Fuse Box: 1FB (18D105) b) 125-V DC Distribution Panels: IPPB1 (IBD102) IPPB2 (1BD501)
- 7. Unit 1 and Common Division 3, Consisting of:
a) 250-V DC Fuse Box: 1FC (1C0105) b) 125-V DC Distribution Panels: IPPC1 (ICD 102) IPPC2 (ICD 501)
- 8. Unit 1 and Common Division 4, Consisting of:
a) 250-V DC Fuse Box: IFD (1D0105) b) 125-V DC Distribution Panels: 1PPC. (1DD102) IPPD2 (1DD501) APPLICABILITY: G ERATIONAL CONDITIONS 4, 5, and *. ACTION:
- a. With less than two divisions of the above required Unit 2 A.C.
distribution systems energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
*When handling irradiated fuel in the secondary containment.
Lli4ERICK - UNIT 2 3/4 8-19
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- b. With less than two divisions of the above required Unit 2 D.C.
distribution systems energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the r5 actor vessel.
- c. With any of the above required Unit 1 and common AC and/or DC distribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for that system.
- d. 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. O l 1 \ l i l i O LIMERICK - UNIT 2 3/4 8-20
i i ELECTRICAL POWER SYSTEMS i, V 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES 4 l PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES I \
\
LIMITING CONDITION FOR OPERATION 3.8.4.1 All primary containment t 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 2 3/4 8-21
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQylREMENTS (Continued) l
- 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 functi.onal 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-l 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 l 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. l
- b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures j prepared in conjunction with its manufacturer's recommendations. 1 O
LIMERICK - UNIT 2 3/4 8-22
TABLE 3.8.4.1-1 ty PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
- 1. 4160-VOLT CIRCUIT BREAKERS CIRCUIT SYSTEMS OR BREAKER NO. LOCATION EQUIPMENT POWERED 152-20101 20A201 2A Reactor Recirc Pump
'A' RPT Breaker 152-20102 20A201 2A Reactor Recirc Pump
'B' RPT Breaker 152-20201 20A202 28 Reactor Recirc Pump '
'A' RPT Breaker 152-20202 20A202 2B Reactor Recirc Pump
'B' RPT Breaker
- 2. 480-V0LT MOLDED 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 52-21108 D214-R-G IM HFB100 2A1 Drywell Area Unit TM HFB100 Cooler 2A1V212 52-21109 D214-R-G IM HFB100 2E1 Drywell Area Unit TM HFB100 Cooler 2E1V212 52-21110 D214-R-G IM HFB100 2C1 Drywell Area Unit TM HFB100 Cooler 2C1V212 52-21111 D214-R-G IM HFB100 2G1 Drywell Area Unit TM HFB100 Cooler 2G1V212 52-21124 D214-R-G IM HFB25 RHR S/D Clg. Suction Inbrd TM HFB100 Isol Viv HV-51-2F009 52-21126 0214-R-G IM HFB50 RWCU Inbrd TM HFB100 Isol Vlv HV-44-2F001 i 52-21138 0214-R-G IM HFB25 Mn Stm Line Drain Inbrd TM HFB40 Isol V1v HV-41-2F016 52-21141 0214-R-G IM HFB25 Inst Gas Compr Suct Line TM HfB40 Inbrd Isol V1v HV-59-201 O O 1 t LIMERICK - UNIT 2 3/4 8-23
TABLE 3.8.4.1-1 (Continued) PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
- 2. 480-VOLT M0LDED CASE BREAKERS (Continuea)
CIRCUIT SYSTEMS OR BREAKER NO. LOCATION TYPES EQUIPMENT POWERED 1 52-21208 D224-R-G IM HFB100 2B1 Drywell Arca Unit TM HFB100 Cooler 2B1V212 52-21209 D224-R-G IM HFB100 2F1 Drywell Area Unit TM HFB100 Cooler 2F1V212 52-21210 D224-R-G IM HFB100 2D1 Drywell Area Unit TM HFB100 Cooler 201V212 52-21211 0224-R<G IM HFB100 2H1 Drywell Area Unit TM HFB100 Cooler 2H1V212 52-21216 D224-R-G IM HFB25 2B Reactor Recirc Pump TM HFB100 Suction V1y HV-43-2F023B 52-21331 D214-R-C IM HFB25 RCIC Mn Stm Supply Inbrd TM HFB40 Isol V1v HV-49-2F007 Emergency Power l 52-21309 D214-R-C IM HFB50 Feedwater Line 'A' Inbrd TM HFB150 Maint V1v HV-41-2F011A 52-21707 D234-R-H IM HFB100 2C2 Drywell Area Unit TM HFB100 Cooler 2C2V212 52-2170B D234-R-H IM HFB100 2G2 Drywell Area Unit ! TM HFB100 Cooler 2G2V212 52-21807 D244-R-H IM HFB100 202 Drywell Area Unit TM HFB100 Cooler 202V212 52-2180S D244-R-H IM HFB100 2F2 Drywel! Area Unit TM HFB100 Cooler 2F2V212 52-22310 0234-R-E IM HFB100 2A2 Drywell Area Unit TM HFB100 Cooler 2A2V212 52-22311 D234-R-E IM HFB100 2E2 Drywell Area Unit TM HFB100 Cooler 2E2V212 52-22313 0234-R-E IM HFB25 RCIC Mn Stm Supply Inbrd iM HFB40 Isol Vlv HV-49-2F007 52-22314 0234-R-E IM HFB50 Feedwater Line 'B' Inbrd Maint V1v HV-41-2F011B TM HFB100 LIMERICK - UNIT 2 3/4 8-24
p I
- g. TABLE 3.8.4.1-1 (Continued). '
PRIMARY CONTAINMENT PENETRATION CONDUCTOR l OVERCURRENT PROTECTIVE DEVICES l
' 2. 480-VOLT MOLDED CASE BREAKERS (Continued)-
CIRCUIT . SYSTEMS OR i BREAKER NO. 10 CATION TYPES EQUIPMENT POWERED 52-22410 D244-R-E IM HFB100 282 Drywell Area Unit TM HFB100 ' Cooler 2B2V212 - 52-22411 D244-R-E IM HFB100 2H2 Drywell Area Unit j TM HFB100 Cooler 2H2V212' i l 52-22418 D244-R-E IM HFB50 HPCI Mn Stm Supply Inbrd j TM HFB150 Isol Viv HV-55-2F002 1 52-22516 214B-R-C IM HFB25 2A Reac Recirc Pump TM HFB100 Suction Viv HV-43-2F023A 52-22518 214B-R-C IM HFB25 2A Reac Recirc Pump TM HFB100 Discharge Viv HV-43-2F031A f 52-22520 214B-R-C IM HFB25 TM HFB40 Reactor Bottom Head Drain Viv HV-44-2F100 52-22536 214B-R-C IM HFB25 RWCU Flow Control Viv i TM HFB40 HV-44-2F105 l 52-22534 214B-R-C IM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-2F001 52-22535 2148-R-C IM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-2F005 > 52-22537 214B-R-C TM HFB15 Disposal Cask Removal Cart l TM HFB20 Hoist 20H236 l l 52-22538 214B-R-C TM liFB15 Control Rod Drive Platform . TM HFB20 Hoist 20H229 l 52-22608 2248-R-C TM HF815 CRD Equipmen+ Handling l TM HFB20 Platform 20N226t8 52-22618 2248-R-C IM HFB25 2B Reac. Recirc Pump , TM HFB100 Discharge V1v HV-43-2F031B j
*52-22622 2248-R-C TM HFB125 Permanent Plant In-Containment Welding System 20NW201 1~
O LIFEJICK - UNIT 2 3/4 8-25 l
I 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 N0. LOCATION TYPES EQUIPMENT POWERED
*52-22626 2248-R-C TM HFB50 Unit 2 Reactor Enclosure 2L36 (Main Breaker) 2L36 EB3090** Lighting XFMR 2X28
*52-22630 224B-R-C TM HFB20 2A Reac. Recirc. Pump TM HFB20 Motor Hoist 2AH203
*52-22631 224B-R-C TM HFB20 2B Reac. Recirc. Pump TM HFB20 dotor Holst 28H203 52-22634 2248-R-C IM HFB25 Feactor Vessel Head Vent TM HFB40 iV-41-2F002
*52-22707 214C-R-A TM HFB15 fin Stm Relief Vlv Removal TM HFB15 loist 20H232
*52-22708 214C-R-A TM HFB15 Mn Stm Relief Vlv Removal TM HFB15 Hoist 20H230
*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 ABBREVIATIONS:
TM Thermal Magnetic IM Instantaneous Magnetic l l 1 O LIMERICK - UNIT 2 3/4 8-26
ELECTRICAL POWER SYSTEMS K/ MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION I LIM.1 TING CONDITION FOR OPERATION ! l 3.8.4.2 The thermal overload protection of all Class 1E motor operated valves ' shall be either:
- a. Continuously bypassed for all valves with maintained position control switches; or,
- b. Bypassed only under accident conditions 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 oc; 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) for the affected system (s). 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 whera 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 letst 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 performed to verify that the thermal overload protection will be bypassed under accident conditions. %/ LIMERICK - UNIT 2 3/4 8-27
j l I ELECTRICAL POWER SYSTEMS I l REACTOR PROTECTION SYSTEM ELECTRICAL POWER MONITORING 9I LIMITING CONDITION FOR OPERATION l I 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 thannel to OPERABLE status within 72 hours or remove the associated RPS Inverter or alternate power supply from service.
l b. With both RPS electric power monitoring channels for an inservice RPS l Inverter or alternate power supply inoperable, restore at least one i electric power monitoring channel to OPERABLE status within 24 hours l or remove the associated RPS Inverter or alternate power supply from ! service. l 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 underfrequency protective instruments-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 3 57 Hz.
i l O\ l LIMERICK - UNIT 2 3/4 8-28 1 l
L w i l t J 3/4.9 REFUELING OPERATIONS j ( 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 l OPERABLE:
- 1. All rods in.
- 2. Refuel platform position.
- 3. Refuel platform hoists fuel-loaded.
- 4. Service platform hoist fuel-loaded.
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 I 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 CONDITION 5 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensicned or with the head removed.
LIMERICK - UNIT 2 3/4 9-1
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS O 4.9.1.1 The reactor mode switch shall be verified to be locked in the l Shutdown or Refuel position as specified:
- a. Within 2 hours prior to:
- 1. Beginning CORE ALTERATIONS, and
- 2. Resuming LtJRE ALTERATIONS when the reactor mode switch has been unlocked.
- b. At least once per 12 hours. l 4.9.1.2 Each of the above required reactor mode switt.h 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.
1 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 l CHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or CORE l 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 '
i 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. I O LIMERICK - UNIT 2 3/4 9-2 l l
{ , f , REFUELING OPERATIONS '
.r
- i 3/4.9.2, INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 'At least two source range monitor.(SRM) channels
- shall be OPERABLE l and' inserted to the normal operating level with: 9
- 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 suspend all operations involving CORE ALTERATIONS and insert all insertable t ' 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 adjacentquadrant. '
*These channels are not required when sixteen or fewer fuel assemblies, ad-jacent to the SRMs, are in the core. The use of-special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is per-missible as long as these special detectors are connected to the normal SRM circuits. j .
**Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
***See Special Test Exception, Specification 3/4.10.7.
LIMERICK - UNIT 2 3/4 9-3
1 REFUELING OPERATIONS O SURVEILLANCE REQUIREMENTS (Continued) ,
- b. Performance of a CHANNEL FUNCTIONAL TEST:
- 1. Within 24 hours prior to the start of CORE ALTERATIONS, and 1
- 2. At least once per 7 days.
I
- c. Verifying that the channel count rate is at least 3.0 cps:* l
- 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 t. ce any control rod is withdrawn,** or
- 2. Shutdown margin demonstrations.
L l l l
*For initial fuel loading and startup the count rate may be reduced to 0.7 cps provided the signal-to-noise ratio is > 2. These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core. **Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 2 3/4 9-4
,'l bi , '
REFUELING OPERATIONAL Y
!<'t d' - 3/4.9.3 CONTROL ROD POSITION LIMITIhd CONDITION FOR'0PERATION o
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 l
4.9.3. All control rods shall be verified to be inserted, except as above
'\ specified;
- a. Within 2 hours prior to:
- 1. ?The start of CORE ALTERATIONS.
- 2. The withdrawal of one control rod under the control of the reactor mode switch Refuel position one-rod-out inter 1cck.
- 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 LIMERICK - UNIT 2 3/4 9-5 l l
REFUELING OPERATIONS O 3/4.9.4 DECAY TIME LIMITING CONDITION FOR OPERATION 4 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 subtritical for less than 24 hours, 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 2 3/4 9-6
.. ' REFUELING OPERATIONS r
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. i
)
LIMERICK - UNIT 2 3/4 9-7
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 reacter pressure vessel 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 50 pounds.
- b. Demonstrating operation of the hoist loaded control rod block interlock I on the main hoist when the load exceeds 485 50 pounds. I
- 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.
i O LIMERICK - UNIT 2 3/4 9-8
I REFUELING OPERATIONS
-~
( w SURVEILLANCE REQUIREMENTS (Continued) 4.9.6.2 The f'sfueling 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 2000 1 50 pounds.
- b. Demonstrating operation of the uptravel mechanical stop on the frame l mounted hoist when uptravel brings the top of active i 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. j O 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.
l 4 ( I LIMERICK - UNIT 2 3/4 9-9
REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOL O 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: 1 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. 1 SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks which prevent crane travel over fuel assemblies in the spent fuel storagt 9001 racks shall be demonstrated OPERABLE within 7 days prior to and at least once per 7 days during crane operation. I O LIMERICK - UNIT 2 3/4 9-10 C____________________. _ _ J
REF UELING_0PERATIONS
/^\
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 cont;ol 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 I SURVEILLANCE REQUIREMENTS , I 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. l b LIMERICK - UNIT 2 3/4 9-11
I REFUELING OPERATIONS 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL O 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 r.rane 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 kast at its minimum required depth at least once per 7 days. i LIMERICK - UNIT 2 3/4 9-12 l L _-_-- --
REFUELING OPERATIONS
/'M 3/4.9.10 CONTROL ROD REMOVAL SINGLE CONTROL RCD REMOVAL LIMITING CONDITION FOR OPERATION 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 cequired to
,m be 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 re:noved 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. I
\
LIMERICK - UNIT 2 3/4 9-13 w___-___--_______ _ _ _ _ _ _ _ i
W.- I 1 I 1 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS O 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- l 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. T se SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied r er Specification 3.9.10.lc.
- d. All other control rods in a five-by-five array centered on the control rod being removed are inserted and electricuily 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.
I O LIMERICK - UNIT 2 3/4 9-14 w__ _ _ - _
p ! h i 4 REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL D 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
( four fuel assemblies removed from the core cell. r
- 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 pressure vessel and initiate action to satisfy the above requirements. I
\
l LIMERICK - UNIT 2 3/4 9-15
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS O 4.9.10.2.1 Withir, 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 Pours thereafter until all control rods and control rod drive mechanisms are r? installed and all control rods are inserted in the core, vertfy that:
- a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1 or 4.9.1.2, ais applicable, and locked in the Shutdown position or in :
the Refuel p(sition per Specification 3.9.1. I
- 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 j 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 2 3/4 9-16
i REFUELING OPERATIONS n 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION i 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 OPE DBLE 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 pr*,ssure 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 of at least One W ernate method capable of decay heat removal.
Otherwise, suspen6 all operations involving an increase in the (, reactor decay heat, load aad establish SECONDARY CONTAINMENT INTEGRITY
' within 4 hours.
- b. With no RHR shutdown cooling mede loop in operation, within 1 hour establish reactor coolant circulation by an alternate method and monitor reactor .oolant 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.
m LIMERICK - UNIT 2 3/4 9-17
REFUELING OPERATIONS LOW WATER LEVEL i LIMITING CONDITION FOR OPERATION l 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 0PERABILITY 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 i 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 cooliry 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.
O LIMERICK - UNIT 2 3/4 9-16
j f 3/4.10 SPECIAL TEST EXCEPTIONS
. ~/
\ 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY l
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 rLwer-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
- r. lace the rerr. tor mode switch in the Shutdown position.
1 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. J I LIMERICK - UNIT 2 3/4 10-1
SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD WORTH MINIMIZER O 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 may be suspended for the following i 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.
- APPLICABILITY
- OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less than l & equal to 10% of RATED THERMAL POWER.
l ACTION: With the requirements of the above specifications not satisfied, verify that l the RWM is OPERABLE per Specification 3.1.4.1. SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RWM are bypassed, verify:
- a. That movement of control rods is blocked or limited to the approved I control rod withdrawal 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.
i 1 I t h , LIMERICK - UNIT 2 3/4 10-2 i _ _ _ _ _ _ _ _ _ - - _ _ _ _ - _ . i
p -1 4 4-SPECIAL TEST EXCEPTIONS 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 more 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-se:quence movement of the control rods.
i
- d. No other CORE ' ALTERATIONS are in progress.
APPLICABILITY: OPERATIONAL CONDITION 5, during shutdown margin demonstrations. 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.
LIMERICK - UNIT 2 3/4 10-3
I SPECIAL TEST EXCEPTIONS
-3/4.10.4 RECIRCULATION LOOPS Oi LIMITING CONDITION FOR OPERATION j l
l 3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 that j
- 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 l RATED THEltMAL POWER, or
]
- b. The Startup Test Program.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, during PHYSICS TESTS and the l Startup Test Program. ACTION: o With the above specified time limit exceeded, insert all control rods,
- b. With the above specified THERMAL POWER limit exceeded during PHYSICS w TESTS, immedi.'tely place the reactor mode switch in the Shutdown position.
SURVEILLANCE REQUIREMENTS l 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. i O i LIMERICK - UNIT 2 3/4 10-4
[ i SPECIAL TEST EXCEPTIONS r ( k 3/4.10.5 OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION J 3.10.5 The provisions of Specification 3.6.6.3 may be suspended during the performance of the Startup Test Program until 6 months after initial criticality. APPLICABILITY: OPERATIONAL CONDITION 1. ACTION With the requirements of the above specification not satisfied, be in at least ' STARTUP within 6 hours. i n (s SURVEILLANCE REQUIREMENTS 4.10.5 The number of months since initial criticality .shall be verified to be less than or equal to 6 months at least once per 31 days during the Startup Test Program. 4 I O LIMERICK - UNIT 2 3/4 ~.'0-5
SPECIAL TEST EXCEPTIONS 3/4.10.6 TRAINING STARTUPS O LIMITING CONDITION FOR OPERATION 3.10.6 The provisions of Specification 3.5.1 may be suspended to permit one l 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: I l With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position. SURVEILLANCE REQUIREMENTS 4.10.6 The reactor vessel shall be verified to be 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 2 3/4 10-6 _____,m__--_m___.i_-
l SPECIAL TEST EXCEPTIONS ! 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING
% i I
l j LIMITING CONDITION FOR OPERATION 3.10.7 During initial core loading within the Startup Test Program the provisions of Specification 3.9.2 may be suspended provided that at least two source range monitor (SRM) channels with detectors inserted to the normal operating level are OPERABLE with:
- a. One of the required SRM channels continuously indicating
- in the control room, j
- b. 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,**
- c. Tne RPS " shorting links" shall be removed prior to and during fuel loading,
- d. The reactor mode switch is OPERABLE and locked in the REFUEL position.
APPLICABILITY: OPERATIONAL CONDITION 5 ACTION: With the requirements of the above specifications not satisfied, immediately U suspend all operations involving CORE ALTERATIONS and insert all insertable control rods. SURVEILLANCE REQUIREMENTS 4.10.7 Each of the above required SRM channels shall be demonstrated OPERABLE by:
- a. Within 1 hour prior to and at least once per 12 hours during CORE ALTERATIONS:
- 1. Performance of a CHANNEL CHECK ***
- 2. Confirming that the above required SRM detectors are at the normal operating level and located in the quadrants required by Specification 3.10.7. ,
*Up to 16 fuel bundles may be loaded without a visual indication of count rate.
**The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.
, *** Check may be performed by use of movable neutron source. Movement of the i
movable neutron source is not a CORE ALTERATION. LIMERICK - UNIT 2 3/4 10-7
SPECIAL TEST EXCEPTIONS SURVEILLANCE REQUIREMENTS (Continued) 4.10.7 (Continued)
- 3. The RPS " shorting links" are removed.
- 4. The reactor mode switch is locked in the REFUEL position.
- b. Performance of a CHANNEL FUNCTIONAL TEST withir, 24 hcars prior tc the start and at least once per 7 days during CORE ALTERATIONS.
- c. Verifying for at least one SRM channel that the count rate is at least 0.7 cps *:
- 1. Immediately following the loading of the first 16 fuel bundles.
l 2. At least once per 12 hours thereafter during CORE ALTERATIONS. i 1 O. l i i i i
*Provided signal-to-noise is > 2 (for initial startup only). Otherwise, 3 cps.
LIMERICK - UNIT 2 3/4 10-8
3/4.11 RADI0 ACTIVE EFFLUENTS (D)- w/ 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 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/ml total activity. APPLICABILITY: At all times. ACTION: 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. SURVEILLANCE REQUIREMENTS D 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. 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. O LIMERICK - UNIT 2 3/4 11-1
I j 1 TABLE 4.11.1.1.1-1 l RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 3 l I l LOWER LIMIT MINIMUM TYPE 3F 0FDETEC{ ION LIQUID RELEASE SAMPLING ANALYSIS ACTIVITY (LLD) TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/mL) A. Batch Write P P Principa} Gamma 5x10 7 Releve Each Batch Each Batch Emitters b Tanks I-131 1x10 6
- 1. Floor Drain P M Dissolved and 1x10 5 Sample Tank One Batch /M Entrained Gases No. 2 (Gamma Emitters)
- 2. Laundry P M H-3 1x10 5 Drain Sample d Tank Each Batch Composite Gross Alpha 1x10 7 P Q Sr-89, Sr-90 5x10 8 d
Each Batch Composite Fe-55 1x10 6 B. Continuogs W W Principa} Gamma 5x10 7 Release Grab Sample Emitters I-131 1x10 6
- 1. RHR Servi:e W W Dissolv and 1x10 5 Water System Grab Sample Entrai,ned Gases Effluent Line f (Gam *a Emitters)
- 2. Service Water W M H-f, 1x10 5 d
System Grab Sample Composite _, Effluent Gross A*.pha 1x10 7 Line O Sr-89, Sr-90 5x10 8 d Grab Sample Composite Fe-55 1x10 6 i l O LIMERICK - UNIT 2 3/4 11-2 l
l I TABLE 4.11.1.1.1-1 (Continued) { - TABLE NOTATIONS
.t x
a The LLD is defined, for purposes of these specifications, as the smallest l 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), 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 x 106 is the number of disintegrations per minut.e per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclides, and at for the plant effluents 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 measuremeiit system and not as an a posteriori (after the fact) limit for a particular measurement. LIMERICK - UNIT 2 3/4 11-3
TABLE 4.11.1.1.1-1 (Continued) TABLE NOTATIONS f 1 l l b A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling. c l The principal gamma emitters for which the LLD specification applies include I the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, l 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 analy2ed 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 l sampling employed results in a specimen that is representative of the liquids released. l 'A continuous release is the discharge of liquid wastes of a nondiscrete l volume, e.g., from a volume of a system that has an input flow during the continuous release. Whenever effluent releases are in excess of the monitor's setpoint. I O' t 1ERICK - UNIT 2 3/4 11-4 m
t ! n , RADI0 ACTIVE EFFLUENTS I 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 the site to UNRESTRICTED AREAS (See Figure 5.1.3-1) shall be limited:
- a. During any calendar quarter to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ, and
- b. During any calendar year to less than or equal to 6 mrems to the total body and to less than or equal to 20 mrems to any organ.
APPLICABILITY: At'all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit _to the Commission within 30 days, pursuant to Specifica-tion 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 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.
- b. The provisions of Specif kation 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1.2 Cumulativo 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 onco per 31 days. LIMERICK - UNIT 2 3/4 11-5
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- i 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 I effluent, from the site, 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. l l- APPLICABILITY: At all times. ACTION: l \ I l a. With radioactive liquid waste being discharged without treatment and ) in excess of the above limits, prepare and submit to the Commission ' witbin 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information: j
- 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 l
status, and 1
- 3. Summary description of action (s) taken to prevent a recurrence. l l
- b. The provisions of Specification 3.0.3 are not applicable. ;
SURVEILLANCE REQUIREMENTS i 4.11.1.3.1 Doses due to liquid releases from the site 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 - UNIT 2 3/4 11-6
RADIOACTIVE EFFLUENTS i
- LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 1 I
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 Specification 3.0.3 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. t LIMERICK - UNIT 2 3/4 11-7
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASE0US EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION l 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 B0UNDARY (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; J 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 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 obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11.2.1.2-1. O LIMERICK - UNIT 2 3/4 11-8
v T( I MN IO) 4 0 4 0 8 0 2 0 1 1 00 1 1 0 1 1 0 1 1 0 6 d3
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I s s s r r r e e e t t t t t t M i i i a A m m m m R S E E E m G I a O S a a a) G R (
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,ii
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 concluaing 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), sbis 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 100 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 radionuclides, and At is the elapsed time between midpoint of sample collection and time of ; counting (for plant effluents, not environmental samples) The value of s used in the calculation of the LLD for a detection system shall be basedbon 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 !LD 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. l O LIMERICK - UNIT 2 3/4 11-10
L I TABLE 4.11.2.1.2-1 (Continued) , O g TABLE NOTATIONS I' 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 colupleted within 48 hours ef ter changing, or after removal' from sampler. Sampling shall also be performed at least once per 24 hours for at least l 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in I hour and analyses completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the L corresponding LLDs may be increased by a factor of 10. This requirement I 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 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 made.in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. 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, In-65, Mo-99, I-131, Cs-li4, 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 Deleted.
- h. Required for the hot maintenance shop ventilation exhaust only during opera-tion of the hot maintenance shop ventilation exhaust system.
LIMERICK - UNIT 2 3/4 11-11
RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION l l l 3.11.2.2 The air dose due to noble gases released in gaseous effluents from i the site to areas at anj beyond the SITE B0UNDARY (see Figure 5.1.3-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation, and
- b. During any calendar year: Less than or equal to 20 mrads for gamma radiation and less than or equal to 40 mrads for beta radiation.
APPLICABILITY: At all times. ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, 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 releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
l l b. The provisions of Specification 3.0.3 are not applicable. 1 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 2 3/4 11-12
RADIOACTIVE EFFLUENTS n ( DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF lHC 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 the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the followin(.
- a. During any calendar quarter: Less than or equal to 15 mrems to any organ and,
- b. During any calendar year: Less than or equal to 30 mrems to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above f-m limits, prepare and submit to the Commission within 30 days, pursuant I 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 that subsequent releases will be in compliance with the above limits.
- b. The provisions of Specification 3.0.3 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 parametars in the ODCM at least once per 31 days. i LIMERICK - UNIT 2 3/4 11-13 l l
RADI0 ACTIVE EFFLUENTS VENTILATION EXHAUST TREATMEN~. SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM shall 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 the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1.3-1) when averaged over 31 days would exceed 0.6 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 incluces 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 descr*ption of action (s) taken to prevent a recurrence.
i b. The provisions of Specification 3.0.3 are not applicable. l SURVEILLANCE RF.QUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site to areas at and beyond i- the SITE BOUNDARY 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 EXHAUS7 TREATMEki 3 STEM shall be demonstrated OPERABLE by meeting Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. O l LIMERICK - UNIT 2 3/4 11-14 l
U RADI0 ACTIVE EFFLUENTS 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. 1 APPLICABIl.ITY: 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 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 the hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specifica-(N) ( v tion 3.3.7.12. i O LIMERICK - UNIT 2 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 Kr-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*. i ACTION: l l With the rate of the sum of the activities of the specified noble gases at the recombiner after condenser discharge exceeding 330 millicuries /seconi, J restore the gross radioactivity rate to within its limit within 72 hours or be j 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 continuously 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 af ter-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 Retreatment Radioactivity Monitor, of greater than 50%, after factoring out increases due to changes in THERMAL POWER level or air in-leakage, in the nominal stea h*-state fission gas release from the primary coolant.
- c. The provisons of Specification 4.0.4 are not applicable.
j
*When the Inin condenser air ejector is in operation. 1 O
LIMERICK - UNI'. 2 3/4 11-16 l
.e l L'
p RADI0 ACTIVE EFFLUENTS ft VENTING OR PURGING LIMITING CONDITION FOR'0PERATION 3.11.2.I VENTING or PURGING'of the Mark 11 containment shal1~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. 1he provisions of Specification 3.0.3 are not applicable. l l
SURVEILLANCE REQUIREMENTS 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 O 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 , system assure that: l
- a. Both standby gas treatment system trains are OPERABLE whenever the i purge system is in use, and {
- b. Whenever the purge system is in use during OPERATIONAL CONDITION 1 or 2 or 3, only one of the standby gas treatment system trains may be used.
l i
*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. ! O ; LIMERICK - UNIT 2 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 received at the disposal site. APPLICABILITY: At all times. ACTION:
- a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the solid waste system as necessary to prevent recurrence.
- b. With the SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1) test the improperly processed waste l in each container to ensure that it meets burial ground and shipping l
requirements and (2) take appropriate administrative action to pra 'ent recurrence. t
- c. The provisions of Specification 3.0.3 are not applicable.
i SURVEILLANCE REQUIREMENTS 4.11.3.1 The PROCESS CONTROL PROGRAM shall be used to verify that the properties of the pachiged 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. l O\ LIMERICK - UNIT 2 3/4 11-18
RADIOACTIVE EFFLUENTS i l p) ( V SURVEILLANCE REQUIREMENTS (Continued) l i 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 I specimen from at least every tenth batch of each type of wet radioactive waste i (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 least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided fm in Specification 6.13, to assure SOLIDIFICATION of subsequent batches
( of waste. L L
/
LIMERICK - UNIT 2 3/4 11-19
RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and 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 L equal to 75 mrems. APPLICABILITY: At all times. ACTION: l
- a. With the calculated doses from the release of radioactive materials
, in liquid or gaseous effluents exceeding twice the limits of Specifi-l 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 l 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 lose (s) exceeds 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 Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from licuid and gaseous effluents i 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 l 3.11.4, ACTION a, Cumulative dose contributions from direct radiation from 4 unit operation shall be determined in accordance with the methodology and I parameters in the ODCM. LIMERICK - UNIT 2 3/4 11-20 ! I i
- v. ;
i 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONIT0f!ING PROGRAM L , LIMITING CONDITION FOR OPERATION l i 3.12.1 The radiological environmental monitoring program shall be conducted I as specified in Table 3.12.1-1. APPLICABILITY: At all times. ACTION:
- a. With the radiological environment 1' 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 reasens 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 )
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 taken to reduce radioactive effluents so that the potential annual O 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: concent ation (1) concentration (2) reporting1.0level (2) + '" reporti1g level (1) 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. LIMERICK - UNIT 2 3/4 12-1
i RADIOLOGICAL ENVIRONMENTAL MONITORING { LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- c. With inilk or fresh leafy vegetable samples unavailable from one or i more of the sample locations required by Table 3.12.1-1, identify j locations for obtaining replacement samples and add them to the l 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 i identify the new location (s) for obtaining replacement samplec in the <
next Semiannual Radioactive Effluent Release Report and also include I in the report a revised figure (s) and table for the ODCM reflecting 4 the new location (s). ;
- d. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS l 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 3.12.1-1, the detection capabilities required by Table 4.12.1-1. I i O LIMERICK - UNIT 2 3/4 12-2
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TABLE 3.12.1-1 (Continued) RADIOLOGICAL ENVIRONMENTAL 110NITORING PROGRAM TABLE NOTATIONS a specific parameters of distance and direction sector from the centerline of
~the two reactors and additional, description where pertinent, shall be provided. ;
for each and every sample location in Table 3.12.1-1 in a table and figure (s)
~
in the ODCH. Deviations are permitted from the required sampling schedule if l E specimens are unobtainable due to hazardous conditions, seasonal unavailability, i malfunction of automatic sampling equipment and other legitimate reasons. If ; specimenstare unobtainable due to sampling equipment malfunction, every effort 1 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 i Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.7 2 It is recognized that, at times, it may not be possible or practicable to con-a tinue to obtain samples 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 monitoring program. Pursuant to Specification 6.9.1.8, identify the cause of the unavailability of samples for 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). j b 0ne or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in aCdition to, ,
' integrating dosimeters. For the purposes of this table, a thermolaminescent J 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-j' emitting radionuclides that may be attributable to the effluents from the facility. i I
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 f but near the mixing zone. . 4 LIMERICK - UNIT 2 1/4 12-7
TABLE 3.12.1-1 (Continued) . RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 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 maximum organ and age group, using the methodology and parameters in the ODCM. O , I O LIMERICK - UNIT 2 3/4 12-8
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w TABLE 4.12.1-1 (Continued) TABLE NOTATTONS
'(
! (a)This' list does'not mean that only these nuclides are to be coaJidered. Other j l 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- 1 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 radioactiv'e' 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 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 c5unting 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 picocurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclides, and 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. l
\ ,
LIMERICK - UNIT 2 3/4 12-11 !
i eb'. e 4.12.1-1. (Continued) 105.L.E,,,NO,I3JJ ON,S, 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. Analyses shall to 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 l i O j LIMERICK - UNIT 2 3/4 12-12 1 l
i e i s'. ' 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 m2 (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-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 Specification 3.0.3 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 i to Specification 6.9.1.7.
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE B0UNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 3.12.1-1 item 4.c. shall be followed, including analysis of control samples.
LIMERICK - UNIT 2 3/4 12-13
l RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM O. I LIMITIfiqCONDITIONFOROPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. i APPLICABILITY: At all times. )
.A;;TLDt
- a. With analyses noc beitig ;:,arformeo . s required eture, repcrt 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 Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS l 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. j O' LIMERICK - UNIT 2 3/4 12-14
--m-f D
BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
.O 1
O _a---- ----. - - - - - . -- - - - - - - - - - - - - -- -'---- ----'- - - '
[ e 6 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.
l n . O
3/4.0 APPLICABILITY , n I l (C) BASES Specifications 3.0.1 through 3.0.4 establish the general requirements applicable to Limiting Conditions for Operation. These requirements are based on the requirements for Limiting Conditions for Operation stated in the Code of Federal Regulations, 10 CFR 50.36(c)(2):
" Limiting Conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any ,
remedial action permitted by the technical specification until the i condition can be met." Specification 3.0.1 establishes the Applicability statement within each individual specification. as the requirement for when (i.e. , in which OPERATIONAL CONDITIONS or other specified conditions) conformance to the Limiting Conditions for Operation is required for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limiting l Condition for Operation are not met. It is not intended that the shutdown ACTION requirement be used as an operation convenience which permits (routine) voluntary removal of a system (s) or component (s) from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. 2 There are two basic types of ACTION requirements. The first specifies the (d remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. In this case, conformance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTION requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for Operation must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed within the allowable outage time limits, a shutdown is required to place the facility in an OPERATIONAL CONDITION or other specified condition in which the specification no longer applies. The specified time limits of the ACTION requirements are applicable from the point of time it is identified that a Limiting Condition for Operation is not met. The time limits of the ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems. Individual specifications may include i a specified time limit for the completion of a Surveillance Requirement when l equipment is removed from service. In this case, the allowable outage time limits of the ACTION requirements are applicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered an OPERATIONAL CONDITION in which a new specification becomes applicable. In this case, the time limits of the ACTION requirements would apply from the point in time that i the new specification becomes applicable if the requirements of the Limiting
\ Condition for Operation are not met.
LIMERICK - UNIT 2 B 3/4 0-1
APPLICABILITY BASES Specification 3.0.2 establishes that noncompliance with a specification exists w1en the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the specified time interval. The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with a Limiting Condition of Operation is restored within the time interval specified in the associated ACTION requirements. " Specification 3.0.3 establishes the shutdown ACTION requirements that must be I implemented when a Limiting Condition for Operation is not met and the l condition is not specifically addressed by the associated ACTION requirements. l The purpose of this specification is to delineate the time limits for placing I the unit in a safe shutdown CONDITION when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This time permits the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower CONDITIONS of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and l within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the primary coolant system and the potential for a plant upset that could challenge safety systems under conditions for which this specification applies. If remedial measures permitting limited continued operation of the facility under the provisions of the ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for Operation. Therefore, the shutdown may be terminated if the ACTION requirements have been met or time limits of the ACTION requirements have not expired, thus prov wing an allowance for the completion of the required actions. The time limits of Specification 3.0.3 allow 37 hours for the plant to be in COLD SHUTDOWN when a shutdown is required during POWER operation. If the plant is in a lower CONDITION of operation when a shutdown is required, the time limit for reaching the next lower CONDITION of operation applies. However, if a lower CONDITION of operation is rAached in less time than allowed, the total allowable time to reach COLD fHUTDOWN, or other OPERATIONAL CONDITION, is not reduced. For example, if STARTUP Is reached in 2 hours, the time allowed to reach HOT SHUTDOWN is the next 11 hours because the total time to reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours. Therefore, if remedial measures are completed that would permit a return to POWER operation, a penalty is not incurred by having to reach a lower CONDITION of operation in less than the total time allowed. LIMERICK - UNIT 2 B 3/4 0-2 _ - - - _ _ ____-____-______a
APPLICABILITY ! BASES l The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requirements for one specification results in entry into an OPERATIONAL CONDITION or condition of operation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes applicable in less time than specified, the difference may be added to the allowable outage time limits of the second specification. However, the allowable outage time of ACTION requirements for a higher CONDITION of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower CONDITION of operation. The ahutdown requirements of Specification 3.0.3 do not apply in CONDITIONS 4 and 5, because the ACTION requirements of individual specifications define the remedial measures to be taken. Specification 3.0.4 establishes limitations on a change in OPERATIONAL CONDITIONS when a Limiting Condition for Operation is not met. It precludes placing the facility in a higher CONDITION of operation when the requirements for a Limiting Condition for Operation are not met and continued noncompliance to these conditions would result in a shutdown to comply with the ACTION requirements if a change in CONDITIONS were permitted. The purpose of this s specification is to ensure that facility operation is not initiated or that
; higher CONDITIONS of operation are not entered when corrective action is being
\ taken to obtain compliance with a specification by restoring equipment to OPERABLE status or parameters to specified limits. Compliance with ACTION requirements that permit continued operation of the facility for an unlimited period of time provides an acceptable level of safety for continued operation without regard to the status of the plant before or af ter a change in OPERATIONAL CONDITIONS. Therefore, in this case, entry into an OPERATIONAL CONDITION or other specified condition may be made in accordance with the provisions of the ACTION requirements. The provisions of this specification should not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERt.BLE status before plant startup.
When a shutdown is required to comply with ACTION requirements, the provisions of Specification 3.0.4 do not apply because they would delay placing the facility in a lower CONDITION of operation. Specification 4.0.1 through 4.0.5 establish the general requirements applicable to Sur'eillance v Requirements. These requirements are based on the Surveillance Requirements stated in the Code of federal Regulations 10 CFR 50.36(c)(3):
" Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met."
Specification 4.0.1 establishes the requirement that surveillance must be performed during the OPERATIONAL CONDITIONS or other conditions for which the requirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirement. The purpose of this specification is to ensure that surveillance are performed to verify the LIMERICK - UNIT 2 B 3/4 0-3
l l 1 APPLICABILITY BASES operational status of systems and components and that parameters are within i specified limits to ensure safe operation of the facility when the plant is in an OPERATIONAL CONDITION or other specified condition for which the individual Limiting Conditions for Operation are applicable. Surveillance Requirements . do not have to be performed when the facility is in an OPERATIONAL CONDITION for which the requirements of the associated Limiting Condition for Operation do not apply unless otherwise specified. The Surveillance Requirements asso-ciated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception to the requirements of a specification. Specification 4.0.2 establishes the conditions under which the specified time interval for Surveillance Requirements may be extended. Item a. permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operation conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. Item b. limits the use of the provisions of item a. to ensure that it is not used repeatedly to extend the surveillance interval beyond that specified. The limits of Specification 4.0.2 are based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. These o mvisions are sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval. Specification 4.0.3 establishes the failure to perform a Surveillance l Requirement within the allowed surveillance interval, defined by the
- provisions of Specification 4.0.2, as a condition that constitutes a failure l to meet the OPERABILITY requirements for a Limiting Condition for Operation.
Under the provisions of this specification, systems and components are assumed to be OPERABLE when Surveillance Requirements have been satisfactorily performed within the specified time interval. However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are found or known to be inoperable although still meeting , the Surveillance Requirements. This specification also clarifies that the ! ACTION requirements are applicable when Surveillance Requirements have not been completed within the allowed surveillance interval and that the time limits of the ACTION requirements apply from the point in time identified that ; a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the Surveillance Requirements within the allowable outage time limits of the ACTION requirements restores compliance with the requirements of Specification 4.0.3. However, this does not negate the fact that the failure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specifica-tion 4.0.2, was violation of the OPERABILITY requirements of a Limiting Condi-tion for Operation that is subject to enforcement action. Further, the failure to perform a surveillance within the provisions of Specification 4.0.2 consti-tutes a failure to meet the OPERABILITY requirements for a Limiting Condition for Opention and any reports required by 10 CFR 50.73 shall be determined based on tiie length of time the surveillance interval has been exceeded, and the corresponding Limiting Conditions for Operation ACTION time requirements. LIMERICK - UNIT 2 B 3/4 0-4
L F APPLICABILITY' BASES If the allowable outage time limits of the ACTION requirements are less than 24 hours or a shutdown is required to comply with ACTION requirements, e.g., Specification ~3.0.3,~a 24-hour allowance is provided to permit a delay in implementing the ACTION requirements. This provides an adequate time limit to complete Surveillance Requirements that have not been performed. The purpose of this allowance is to permit the completion of a surveillance before a shutdown would be required to comply with ACTION requirements or before other remedial measures would be required that may preclude the completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability cf personnel, the time required to perform'the surveillance, and the safety significant of the delay in completing the required surveillance. This provision also provides a time limit fer the completion for Surveillance Requirements that become applicable as a consequence of CONDITION changes imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to the requirements of Specification 4.0.4 is allowed. If a surveillance is not completed within 24-hour allowance, the time limits of the ACTION requirements ere applicable at that time. When a surveillance is performed within the
'4-hour
- allowance and the Surveillance Requirements are not met, the time iimits of the ACTION requirements are-applicable at the time that the surveillance is terminated.
Surveillance Requirements do not have to be performed on inoperable equipment because the ACTION requirements define the remedial measures that apply. However, the Surveillance Requirements have to be met to demonstrate that inoperable equipment has been restored to OPERABLE status. Specification 4.0.4 establishes the requirement that all applicable surveillance must be met before entry into an OPERATIONAL CONDITION or other condition of operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into an OPERATIONAL CONDITION or other specified condition for which these systems and components ensure safe operation of the facility. This provision applies to changes i' OPERATIONAL CONDITIONS or other specified conditions associated with plant shutdown as well as startup. Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage. When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower CONDITION of operation. Specification 4.0.5 establishes the requirement that inservice inspection of ASME Code Class 1,2 r.nd 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall 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 50.55a. These requirements apply except when relief has been provided in writing by the Commission. L LIMERICK - UNIT 2 B 3/4 0-5
APPLICABIL. TTY ; BASES Gl This specification includes a clarification of the frequencies for performing l the inservice inspection and testing activities required by Section XI of the ! ASME Boiler and Pressure Vessel Code and applicable Addenda. This i clarification is provided to ensure consistency in surveillance intervals l throughout the Technical Specifications and to remove any ambiguities relative I to the frequencies for performing the required inservice inspection and I testing activities. 1 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. The requirements of Specification 4.0.4 ) to perform surveillance activities before entry into an OPERATIONAL CONDITION !' or other specified condition takes precedence over the ASME Boiler and Pressure Vessel Code provision that allows pumps and valves to be tested up to one week after return to normal operation. The Technical Specification definition of OPERABLE does not allow a grace period before a component, which is not capable of performing its specified function, is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision that allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. 9 l 9 l LIMERICK - UNIT 2 B 3/4 0-6
p, v4 3/4.1 REACTIVIT.Y.
- - - - - C.ONTROL
.. SYSTEMS I
BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHitTDOWN MARGIN ensures that (1) the reactor can be made subtritica' 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 I in units of % a k/k is the difference between the calculated value of minimum j shutdown margin during the operating cycle and the calculated shutdown margin i at the time of the shutdown margin test at the beginning of cycle. The value i 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 Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN. The highest worth rod may be determined analytically or by test. The SHUTDOWN O' 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 l withdrawn. This reactivity characteristic has been a basic assumption in the analysis I of plant performance and can be best demonstrated at the time of fuel leading, but the margin must also be determined anytime a control rod is incapable of insertion. 3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is staall, a careful check on actual conditions to the predicted conditions 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 l change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients. O LIMERICK - UNIT 2 8 3/4 1-1
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. l Damage within the control rod drive mechanism could be a generic problem, ! therefore with a control rod immovable because of excessive friction or l 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 specification, but the occurrence of eight inoperable rods could be indicative of 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 rasulting 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 times 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 even under the most unfavorable depressurization of the reactor. LIMERICK - UNIT 2 8 3/4 1-2
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REACTIVITY CONTROL SYSTEMS 1 BASES CONTROL RODS-(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 i- and therefore this check must be performed prior to achieving 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-stration. 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. . 1 The control rod housing support restricts the outward movement of a control i 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 pressure to act as a driving force to rapidly eject a drive housing. The required surveillance intervals are adequate to determine that the I c rods are OPERABLE and not so frequent as to cause excessive wear on the system components. 3/4.1.4 CONTROL R0D PROGRAM CONTROLS 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 1 result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control j rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% 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 RWM to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate cantrol. The RWM provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted. The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3. Additional pertinent analysis is also contained in Amendment 17 to the Reference 4 Topical Report.
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The RBM is designed to automatically prevent fuel damage in the event of l erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block
\ erroneous rod withdrawal to prevent fuel damage. This system b n.ks up the written sequence used by the operator for withdrawal of control rods.
l I LIMERICK - UNIT 2 B 3/4 1-3
l l 1 3 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 con-}}