ML20247K352
| ML20247K352 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 05/26/1989 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20247K345 | List: |
| References | |
| NUDOCS 8906010276 | |
| Download: ML20247K352 (151) | |
Text
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I ATTACHMENT 2 d
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LIMERICK GENERATING STATION L
Docket No. 50-352
. License No. NPF-39 PROPOSEDiTECHNICAL SPECIFICATIONS' CHANGES-GROUPL 1 GROUP 2 . GROUP -3 PageL Page Page i 13/4.4-17 3/4'3-95* 1-6,
~ii1 .3/4 5-l' 3/4 6-4 . 3/4 3-21 iv '3/4 5 3/4 6-18 3/4'3-81 '
' ,vi 3/4.5-5 3/4.6-19 3/4'3-93 L: via ~3/4.6-5 3/4.6-20 3/4 3-96
, vil 3/4 6-48 3/4 6-21 3/4 6-26
! xii. ~3/4 7-24 3/4 6-24 3/4 6-29
-xiv 3/4 11-1 3/4 6 3/4 6-31 xvi .
3/4'12-7 3/4 6-27 3/4'6-33 xviii; 3/4 12-8 3/4 6-42* 3/4 6-42*-
xx' 6-5 3/4'6-53 3/4 6-45
- xxvi B 2-5 3/4 7-22* 3/4.7-22*
3/4 3-7c B 3/4-3-2 3/4 11-11 3/4 7-27 3/4 3-13 B-3/4 3-7 B 3/4 6-1* 3/4 7-28 3/413-15 B 3/4 4-8 3/4'8-25 3/4~3-19 B 3/4 6-1* 3/4 11-3 3/4 3-20 B 3/4 6-3 5-1 3/4~3-24 B 3/4 8-3 6-7
-3/4 3-29 B 3/4 10-1 3/4 2-5 3/4 3-34 3/4 3-37*
3/4.3-78 3/4 3-55 3/4 3-104 3/4 3-95*
~3/4 3-111 3/414-10
'3/4 3-37*
- Item also shown in another group.
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M0Ek der!N:~ ION!'
SECTION~
1.0 DEFIN!TIONS ' .
PAGE
. - 1.1 A0 TION....................................................... 1-1 1.2 AVERAGE PLANAR EXP05URE...................................... 1-1
- 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 1-1 1.4 .CKANNE L C ALI B RATI 0h . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-1 .......
1.5 , CHANNEL CHECK................................................ .
1-1 1.6 CHANNEL FUNCTIONAL TE57......................................
1.-1 i 1 7 . CO RE A LTE RATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-2 ...
1.2 CRITICALPOWEKRATI0......................................... 1-2
- 1. 9 . DOS E EQUI VALE NT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-h .
1.10 E-AVET. AGE- DISINTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . .1-' ...
RL '
1.11 EMERGEN Y CORE COOLING SYSTEM (E005) RESPONSE TIME........... 1-2
- 3. 1.12 END-OF-0 OLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.... 1-2
- 1. 23 F RACTION OF LIMITING POWER DENSITY. . . . . . . . . . . . . . . . . . . l . . . . . . . 1-3 1.14 FRACTION OF RATED THERMAL P0WER.............................. 1-3 1.15 FREQUEN;Y NOTATION........................................... 1-3 1.16 I D E NTI FI ED LE AKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.17 ISO LATION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.18 LIMITING CORTROL R00 PATTERN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3
. 19 LINEAR HEAT GENERATION RATE....... .......................... 1-3 1.20 LOGIC SYSTEM FUNCTIONAL TE5T................................. 1-4
- 1. 21 MAXIMUM FRACTION OF LIMITING POWER DEN 5!TY. . . . . . . . . . . . . . . . . . . 1-4' LIMERICK - UNIT 1 i nl:* L '.M.- t
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.wDEY F. v DE :NI ION"
.c .r e..mJ DE INT IONS (Cont.inued) . P A'i!
1.20 MIMEER(5) 0F THE PUBLIC...................................... 1-4
'1. 23 MINIP JM CRITI CAL POWER RAT 2 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4
- 1. 24 0FFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4
- 1. 2 5 O P E RAE LE - 0 P E RAS I LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 j 1.25 OPERATIONAL CONDITION - CONDITI0H. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 ,
4 1
1.27 PHYSIC 5 TEST 5................................................ 1-4 i 1
1.28 PRES 5URE BOUNDARY LEAKAGE.................................... 1-5 j 1.29 PRIMARY C0h7AINM~hi INTEGRITY................................ 1-5
- 1. 3 0 P ROCESS C0hTRO L PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.31 PURGE - PUR2ING.............................................. 1-5 ,
1.32 RATED THERMAL P0 DER.......................................... 1-6 1.33 REACTOR ENCLOSURE SECONDARY C0h7AINMEhi INTEGRIN. . . . . . . . . . . . 1-6 1.34 REACTOR PROTE TION SYSTEM RESPONSE TIME...................... 1-6 1.35 REFUELING FLOOR SECONDARY C0h7AINMEhi IhTEGR 7T.............. 1-6 1.36 REPORTABLE EVEhi............................................. 1-7 1.37 R00 DEN 5!TY.................................................. 1-7 1.38 SHUT 00WN MARGIN.............................................. 1-7 .
1.39SITEB00NDARY................................................ 1-7
.1. 4 0 50,LI DI FI CATI ON. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 1.41 SOURCE CHECK................................................. 1-7 1.42 STAGGERED TEST BASI 5......................................... 1-2 1.45 THERMAL P0WER................................................ 1-8 1.44 UNIDENTIFIED. LEAKAGE......................................... 1-8 LIMERICK - UNIT I ii l-l. '. B L'O g
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$!* TION F A *. !
- 2. 2 Sari v LIM *~5 THI RWA .
- PD.*I F , L o P re s s ere or Lo. F 10. . . . . . . . . . . . . . . . . . . . . . 2-1 THERFA' . PD,*iR, Hign Pre s sure and High F1o. . . . . . . . . . . . . . . . . . . 21 q Rea: tor Coolant Syster Pressur.............................. 2-1 Rea: tor. Vessel Water Levei.................................. 2-2 2.2 LIM IN* SA~iTY SYST!w Si TINOS I
Rea: tor Prote: tion Syster Instrumentation Setpcints. . . . .. .. . '
23 Tatie 2.2.1-1 Rer:ter Prete: tion Syster Instrume .tation Set;:1nts............... 2-4 BASES 2.2 51:ETY LIM:'S e
' THIRFA'. P0dIR , Lo. Pre s s ere o r Lo. Fi o. . . . . . . . . . . . . . . . . . . . . . E2*.
THERFA' . P0d!E, Hist Pressure an: Hign F1ov.................. E 2-2 Le f t I nte nti or.al ly E 1 a n A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 2-3 Left Intentionally Eiank.... ............................... E 2-1.
Rea: tor Cc:lant System Pres s ure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 2-E 4 Rea: tor Vessel Water Leve1.................................. E 2-E 22 LIMITIN3 SArETY SYSTEM SE'i71N05 Rea: tor Prote: tion System Instrumentation Setpoints. ...... . .. B 2-6 1-LIMERICK - UNIT 2 iv Amendment h:. 7 f.U* 141!!1 f.
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-y' LIMITING CONDITIONS FCR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued)
Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB248........... 3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus:
Average Planar Exposure Initial.
Core Fuel Types P8CIB163........... 3/4 2-4 Figure 3.2.1-4 Maximum Average Planar Linear Heat
- Generation Rate (MAPLNGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB094........... 3/4 2-5' #
Figure 3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel-Types P8CIB071........... 3/4 2-6 Figure 3.2.1-6 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure For Fuel Type 8 32DA (GE8X8EB).............. 3/4.2-6a Figure 3.2.1-7 Maximum Average Planar. Linear Heat Generation Rate (MAPLHGR)
Versus Average Planar Exposure For Fuel Type BC318A (GE8X8EB)..... 3,'S 2-6b Figure 3.2.1-8 Maximum Average Planar Linear Heat Generation Rate (MAPLGHR)
Versus Average Planar Exposure For Fuel Type BC322A (GE8X8EB)..... 3/4 2-6c 3/4 2.2 APRM SETP0lNTS.......................................... 3/4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RAT 10............................ 3/4 2-8 Table 3.2.3-1 Deleted Figure 3.2.3-la Minimum Critical Power Ratio (MCPR)
Versus t (P8X8R/BP8X8R Fuel) 80C to EOC-2000 MWD /5T........................ '3/4 2-10 Figure 3.2.3-lb Hinimum Critical Power Ratio (MCPR)
Versus t (P8X8R/8P8X8R Fuel) EOC-2000 MWD /ST to E0C.......................... 3/4 2-10a LIMERICK UNIT - I vi Amendment NoN,19 l
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IN D E >'
LIMITING CONDITIONS FOR OPERATION AND-SURVEILLANCE REQUIREMENTS-SECTION g
POWER DISTRIBUTION LIMITS (Continued)
Figure 3.2.3-Ic Minimum Critical Power Ratio (MCPR)
Versus I (GE8X8EB Fuel) BOC to EOC-2000 Ma'D/ST........................ 3/4'2-10b .
I Figure 3.2.3-Id Minimum Critical Power Ratio (MCPR)
Versus t (GF8X8EB Fuel)
EOC-2000 MWD /ST to E0C................. 3/4 2-10c-Figure 3.2.3-2 K 7 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 S Response Times......ystem
................ 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance ..
Requirements...................... 3/4 3-7 1
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I LIMERICK UNIT - I via Amendment No. 19 l
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_. LIF*iING :DN' I* IONS 50F ODER *IDK AN: SURVEIL' ANCE Rii*FEu!N !
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3 . i . u>. f cA;E 3/4.3.2 ISOLATION ACTUATI0h INSTRUMENTATION. . . . . . . . . . . . . . .3/4 . . .3-s Table 3.3.2-1 Isolation Actuation Instrumentation..................... 3/4 3- n
- Taole 3.3.2-2 Isclatic'nA:tuation.
Instrumentation 5etpcints........... 3/4 3-1E.
Table 3.3.2-3 Isolation System Instrumen-tation Response Time................ 3/4 3-23 Table 4.3.2.1-1 Isolation A:tuation Instrumen-tation Surveillance Requirements...................... 3/4 3 27 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION I NSTRUME NTATI ON. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4
. . .3-32 Table 3.3.3-1 Emergency Core Cooling System Actuation Ins trumen .ation. . . . . . . . . . . 3/4 3-23
, Table 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation 5etpoints........................... 3/4 3-37 1
Table 3.3.3-3 Emergency Core Coolin Response Times.......g...............
Systee 3/4 3-25 Table 4.3.3.1-1 Emergency Core Cooling System Actuation Instrumentation -
Surveillance Requirements......... 3/4 3-40 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATV5 Re:irculation Pump Trip System Instrumentation..... 3/4 3-47 Table 3.3.4.1-1 ATW5 Recirculatieri Pump Trip System Instrumentation............ 3/4 3-43 Table 3.3.4.1-2 ATW5 Rec'Irculation Pump Trip System Instrumentation 5etpoints......................... 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 3-46 4
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'~:0N! COF 0:ER47ICA AN" SUR'.'E* ','. AN"! R-22* EMEA~5-SI
'3" c_._ A y 4, % _ _4. [G--A " ' O NI p_sm m .4.5 RESICUA. HEAT REMOVA'.
Hot Shutdown............................................ 3/4 4-25 Cold Shutdown........................................... 3/4 4-2f 3 /4. 5 EMERGEN*Y CORE C00'.IN* SY5tEMS 3/4.5.1 E C C S - O P E RAT I N 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-1 3/4.5.2 E C C S - S H UTD0h'N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-6 j 3/4.5.3 SUPPRESSION CHAW 3ER..................................... 3/4 5 3/4.6 CONTAINMENT SYSTEMS' 3/4.6.1 PRIMARY CONTAINMENT Primary Contai nme nt Inte gri ty. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-;
Priscry Contai nme nt Le a ta pe. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-2 .>
Primary Contaiteent Ai r Lodk. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5 MS IV Le a ka g e C o ntrc l Sys tert.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-7, ,
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Primary Containment Structural Integrity. . . . . . . . . . . . . . . . 3/4 6-8 Drywell and Suppression Chamber Internal Pressure. . . . . . . 3/4 6-9 ,
Daywell Average Air Temperature......................... 3/4 6-10 Drywell and Suppression Chamber Purge Syste... . . . . . . . . . . . 3/4 6-11 3/4.6.2 DEPRESSURI2ATION SYSTEMS Suppression Chamber..................................... 3/4 6-12 w
Suppression Pool r " " y - M 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 Va1ves.............................. 3/4 6-19 i
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l INDEk LIM: IN: CON: IONS FOR ODERATION AN: SURVE!LLANCE REOUIREp!N $
SE*TIDh PAGI PLANT Sv57Ew.S (Continued) 3/4.7.2 CONTROL ROOM EMERGEN:Y. FRESH AIR SUPPLY SYSTEM.......... 3/4 7-6 REACTOR CORE ISOLATION COOLING SYSTEM................... 3/4 7-9
.3/4.7.3 3/4.7.4 SNL'BBERS................................................ 3/4 7-12 Figure 4.7.4-1 Sample Plan 2) For Snutber l Functional Test.................... .
3/4 7-16 3/4.7.5 SEALED SOUR *E CONTAMINATION............................. 3/4 7-17 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fi re Suppres s i on Water System. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-15 Spray and/or Sprinkler Systems. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-22 CO 2 Systems............................................. 3/4 7-24.
s Halon Systems........................................... 3/4 7-25. ,
Fire Hos,e Stations...................................... 3/4 7-26 Table 3.7.6.5-1 Fire Hose Stations................ 3/4 7-27
. Cd Yard Fire Hydrants and Myn..-fHosej ouses. H ............. 3/4 7-29
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Table 3.7.6.6-1 Yard Fire Hydrants and i.- . ...P
[ 4. .n Mio s e4 ous e s . . . . . . . . . . . . . . . 3/4 7-30 H
3/4.7.7 FIRE RATED ASSEMBLIES............C ....................... d 3/4 7-31 3 /4. 8 ELECTRICAL POWER SYSTEMS 3/4. 8.1 A.C. SOURCES .,
A.C. Sources - Operating......~.......................... 3/4 6-1 Table 4.8.1.1.2-1 Diesel Generator Test Schedule........................ 3/4 8-8 A.C. Sources - Shutdown...................'.............. 3/4 8-9 3/4.8.2 D.C. SOURCES D.C. Sources - Operating................................ 3/4 E-10 1.IMERICK - UNIT 1 xiv E A E l'il t
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- NDEX LIM:~!NG CONDITIONS FOR OPERATION AND SURVE:LLAN E RE?U:REwEs ;
SECTIONjMr~uh/Af c e..W TicM (gun,g_) PAGE 3'74.9.10 CONTROL ROD REMOVAL Single Control Rod Remova1.............................. 3/4 9-13 Mul tiple Control Red Removal . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION Hign Water Leve1........................................ 3/4 9-17 Low Water Level..... ................................... 3/4 9 ~.!
3,4.10 SPECIAL TEST EXCEP_TIONS
.......... 3/4 10-1 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY................
3/4.10.2 ROD WORTH MINIMIZER........................... ......... 3/4 10-2, l 3/4 10-3 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS..........................
3/4 10-4 3/4.10.4 RECIRCULATION L00PS.....................................
CONCENTRATION.................................... 3/4 10-5 3/4.10.5 OXYGEN STARTUP5................... ................... 3/4 10-6 3/4.10.6 TRAINING 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 Program........................ 3/4 11-2 3/4 11-5 Dese...................................................
Liquid Radwaste Treatment 5ystem........................ 3/4 11-5 Liquid Holdup Tanks.....................................
3/4 11-7 j 3/4.11.2 GASE0US EFFLUENTS 3/4 11-8 Dose Rate...............................................
xvi Amencment No. I7 LIMERICK - UNIT 1 mat : 21989
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PA3E 5!! TION 3/40 AD D L I C AE 1 L I T Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E (*- 2 3 /4 l
3/4.3 REATIVITY C0CRD; SYSTEv.5 3/4.1.1 S HUT 00a'N MAR 3 I N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 I - 2 3/4.1.2 REA:TIVITY AN0MALIES.................................. E 3/4 2-1 3/4.1.3 CO NT RO L R00 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 1- 2 3/4.1.4 CONTRDL RDO PROGRAW. 00NTROLS. . . . . . . . . . . . . . . . . . . . . . . . . . E 3/4 1-3 3/4.1. 5 ST ANDE Y LIQ'.'ID CONTR0; 5YSTEM. . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 2-4 3 /4. 2 PD G DISTRIEU*1ON L1RITS 3/4.2.1 AVIRA3E PLANAR LINIAT. HIAi GENERATIch R ATI . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 2 - 1 LEFT I NT ENTI ON ALLY E L ANK. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 2- 3 9 3/4.2.2 A8 RS. S ET P
- I NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 2- 2 3/4.2.3 MI NI MP. C RI T I ta P ov'E F. R AT I C . . . . . . . . . . . . . . . . . . . . . . . . . . E 3 /4 2 - 4 3/4.2.4 LINE AR HE AT dENERATION R ATE. . . . . . . . . . . . . . . . . . . . . . . . . . . E 3/4 2-5 3 /4. 3 INSTRUMECA90N 3/4. 3.1 REA: TOR PROTECTION SYSTEM INSTRWINTATION. . . . . . . . . . . . . E 3/4 3-1 3/4.3.2 ISOLATION A TUATION INSTRUMENT ATION. . . . . . . . . . . . . . . . . . . E 3/4 3-2 3/4.3.3 EMERGEN Y COPi C00LIN3 SYSTEM A TUATION INSTRUMENTATION....................................... B 3/4 3-2 3/4.3.4 RE !ROULATION PUMP TRIP ACTUATION INSTRUMihTATION..... B 3/4 3-3 3/4.3.5 REA: TOR CORE ISOLATION C00 LING SYSTEM ACTUATION I N ST RW.E NT AT ! 0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3 /4 3 - 4 3/4.3. 6 CONTROL R0D SLO;K INSTRWINTATION. . . . . . . . . . . . . . . . . . . . . B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Ratistion Monitoring Instrumentation. . . . . . . . . . . . . . . . . . B 3/4 3-4 LIMIRICK - UNIT 1 xviii Arnenoment h: 7 M* i 4 HU C - ---- - - - _ - _ _ _ . . _ _ __ _ _ . _ _ _ _ _ . _ _ _ _ _
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Fk?E5 S F -* I 0 f. - ce ,E RE/.0 TOR 000' . AtI* SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIV!TY....................................... E 3/4 4-4 3/4.4.6 PRESSURE /TEMDEPATURE LIMIT 5............................. E . 3/4 4-4 Bases Table i 3/4.4.6-1 Rea: tor vessel Toughness................. E 3/4 4-7 Bases Figure E 3/4.4.6-2 Fast Neutron Fluen:e
. (E>l heV) At 1/4. T As A Fun: tion of Service.
Life...................... E 3/4 4-E 3/4.4.7 MAIN STEAM LINE ISOLATION VALVE5........................ E 3/4 4-6 3/4.4.8 STRU:TURAL INTEGR!TY.................................. . E 3/4 4-E 3/4.4.9 RESIDUAL HEAT REM 0 VAL.......'............................ E 3/t. 4-6 3 /4. 5 EMERGENOV Coc.E C00LIN3 SYSTE8.5 .
- 3/4.5.1 and 3/4.5.2 E 05 - OPERATING anc SHUTDOW. . . . . .. . . . . .E 3/4 5-1 3/4.5.3 SUPPRESSION CHAM 3ER................................ E 3/4 E-2 3/A.6 CONTAINMEN~ SYSTEw.5 "
3/4.6.1 PRIMARY CONTAINMENT .
Prima ry Contai nment Integri ty. . . . . . . . . . . . . . . . . . . . . . 2 3/4 6-1 Prima ry C ontai nme nt Le a ka ge . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 6-1 Prima y Containment Air Lo: ..................... B 3/4 6-1 MSIV Leakage Control Systas. . . . . . . . . . . . . . . . . . . . . . . . B 3/4 6-1 l Prima y Containment Structural Integrity. . . . . . . . . . . B 3/4 6-2 l
l Drywell and Suppression Chamber Internal Pressure......................................... B 3/4 6-2 Drywell Average Ai r Temper'ature. . . . . . . . . . . . . . . . . . . . B 3/4 6-2 -
Drywell and Suppression Chamber Purge System.. ... . . 8 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEK5........................... B 3/4 6-3 LIKERICK - UNIT 1 xx D iW *
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'5E**:0A PAGE 6.3 RE$*0NSI6?LITY............................................... 6-1 6.2 ORGANI2AT10N..... ........................................... 6-1 6.2.1' 0ffsite...................'...................-...... .... 6-1
~
Figure 6.2.1-1 Offsite Organization............... 6-3 6.2.2 Unit 5taff.............................'................. 6-1
. Figure 6.2.2-1 Crpenization for Condu:: of Flar.t Operations................... 6-4 Tatle 6.2.2-1 Minimur Shift Crew Cor. position......................... 6-!
6.2.3 INDEPENDEN* SAFETY ENGINEERING GRDUP (ISE4,)
Function ..............'................................ 6-6 c
Compositier................ ............................ . 6-6 -
- Re s p or.s t i11 t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 Records............................................'.... 6-6 ,
- 6. 2.' 4 SHI FT TE CHNI CAL ADC 50R'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.3 UNIT STArF 00ALIFICATIONS................................... 6-6 .
6.4 TRAINING.................................................... 6-7 6.5 REVI!v AND AUDIT 6.5.1 Plant Operations Review Committee (PORO)
Function .............................................. 6-7 Composition ........................................... 6-7 Alternates............................................. 6-7
,' Me e t i n g F re q ue n cy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 Quorum................................................. .
6-7 l
Responsibilities ...................................... E-6 Re:ords................................................ 6-9 l-
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ISOLL' ION SYS?!M INSTRUMEN*A~ ION REstDNSE TIME TCC UN ~ ION RESDONSI TIME (Setenet)#
4 HIGW PRESSURE CODLANT INJECTION SYSTEM IS0 VAT]QN
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, TABLI 3.3.7.4-1 My -
REMOTE SHUTDOWN SYSTEW CON'TROLS RCIC SYSTEM H55-49-191 Control-Trantfer Switch H55-49-192 Control-Transfer Switch Control-Transfer Switch H55-49-193 ,
'H55-49 -
195 Control-Transfer Switch H55-49-196 Control-Transfer Switch ,
!' HV-49-1F076 Control-Steam Line warms bypass valve 1
- l Control-RCIC turb exhaust to suppression poel
~
HV-49-1F060 '
isolation HV-50-11b Control-Turb trip throttle valve HV-50-1F045 Control-Turbine steam supply valve HV-49-1F008 Control-TurbinestaaElineoutboardisolationvalve HV-49-1F007 Control-Turbine steam line inboard isolation valve e-ty, !
HV-49-1F031 ,
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HV-49-1F010 Control-RCIC pump suction from condensate storage
. tank .
. HV-49-1F019 Control-Hinimum flow bypass to suppression pool HV-49-1F022 control-Test return to condensate storage tank HV-50-1F046 Control-RCIC turbine cooling water valve HV-49-1F012 Control-RCIC pump disch valve
. HV-49-1F013 control-RCIC pump disch valve 10P220 Control-Yacuum tank condensate p ap ,
! 10P219 Control-Barometric condenser vacuum pump f .
1 HV-49-1F002 Control-Barometric condenser vacuum pump disch l
l 4
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I INSTRUMENTATION 1- SURVEIL' LANCE RE0VIREMENTS (Continued)
( .~. . .
1 i -
- b. At'least once per 31 days by:
1
- 1. Cycling each of the' fo11osing valves through at least one i
j -
complete cycle from the runnimg position:
1 .
! a) - For the overspeed protection control system;
.- '1) Four high pressure turbine contro'1 valves For the electrical overspeed trip system and the mechanical b) overs. peed trip .system;
- 1) Font high pressure tu'rb'ne i control valve 5
. \ /
- c. At least once per 18 months by performance of a CHANNEL CALIBRATION of the turbine overspeed protection instrumentation.
a d. At least once per 40 months by disassembling at least one of each of i the above valves and performing a visual and surface inspection of
- all valve seats, 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.
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SURVEILLAN*E REQUIREMENTS h f;
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. 4.4.3.2.1 The reactor coolant system leakage.shall be demonstrated to be within each of the above limits by:
- a. Monitoring the primary contai' nment atmospheric gaseous radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
- c. Monitoring the drywell unit coolers condensate flow rate at least
. once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, ,
. d. Monitoring the primary containment pressure at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not a means of quantifying leakage), ,
- e. Monitoring the reactor vessel head. flange leak cietection system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,,and
- f. Monitoring the primary containment temperature at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (not a means of quantifying leakage). -
4.4.3.2.2 Each reactor coolant system pressure isolation. valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit: .
s
- a. At least once per 18 months, and ,.,
w
- b. Prior to returning the valve to service following maintenance, reoair 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. ,
4.4l3.2.3 The high/ low pressure interface valve leaksge 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 sonths. '
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-l 3/4.5 EMERGENCY CORE COOLING SYSTEMS )
(I -3/A.5.1' ECCS - OPERATING .f LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:
a.'
The core' spray system (CSS) consisting of two subsystems with each subsystem comprised of:
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and' transferring the water through the spray sparger-to the reactor vessel. -
l b. Thelowpressurecoolantin'jection(LPCi)systemoftheresicual heat removal system consisting of four subsystems with each subsystem comprised of:
) -
1 1. One 0PERABLE LPCI pump, and
- 2. An OPERABLE flev path capable of taking sG. ion from the
. suppression ch4mber and transferring the water to the reactor
$ vessel.
f/ 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. -
[. i i
- d. The automatic depressurization system (ADS) with at least five OPERABLE ADS valves.
APPLICABILITY: OPERATIONAL CONDITION 1, 2" ** #, and 3" ** N.
, pressure is less than or equal to 20.0.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. ,' '
y NTwo 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'.
l .;.
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EMERGENCY CORE COOLING SYSTEMS
' [k 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 OPERASLE, restore the inoperable ADS valve te OPERABLE status within 14 days oribe in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to i 100 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 houns and reduce reactor steam dome pressure to 1 100 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- a. With a CSS and/or LPCI hender AP instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or determine the ECCS header AP Iccally at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; otherwise, declare the associata.d CSS and/or LPCI, as applicatie,
- 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-A sitted to the Commission pursuant to Specification 6.9.2 withit.
,- 90 days describing the circums*ances of the actuation and the ' total hi* ~ ac Lumulated actuation cycles to date. The current value of the y usfage factor for each affected safety injection nszzle shall be /\
provided in this Special Report whenever its value exceecs 0.70. '"
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4 EMERGENCY CORE COOLING SYSTEMS
, , . SURVEILLANCE RE0UIREMENTS (Continued): -
( y.;..
'". 2. . For the HPCI system, verifying that:
a)' The system develops r. flow of at least' 5600 gpm against a i test line pres,sure corresponding to a rt.*ctor vessel pressure
'of g 200 psig plus head and line losses, when steam is being' <
supplied to the turbine at 200 + 15, - O psig.*" ;
i-bk P :.sctionisautomaticallytransferredfromthecondensate
<
- gage tank to the suppression chamber on a condensate l1 storage tank water level - low' signal and on a suppression -y )
charderpater leveghigh signal. , ;f
- 3. Perfoming a CHANNEL CALIBRATION of the CSS, LPCI, arid HPCI .I system discharge lire " keep filled" alarm instrumentation.
-l
- 4. Perfoming a' CHANNEL CALIBRATION of the CSS header AP instru-mentation and verifying the satpoint to be 1 the allowable value>
of 4.4 psid.
, - 5. ' Per, forming a CHANNEL CALIBRATION of the L?CI header P irstru-i . mentation and verifying the setpoint to be 1 the allowable value-i . .
of 3.0 psid.
- d. For the ADS:
'f 1. At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST' of the accumulated backup compressed gas system low pressure
- h;g..
e ,
ala m system.
. 2. At least'once per 18 months:
a) Performing a system functional test which includit simulated automatic actuation of the system throughout its.asergency .
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 4
observing that either:
, 1) The control valve or bypass valve position responds 4
accordingly, or
- 2) There is a corresponding change in the sensured steam -
flow. .
c) . Performing a CHANNEL CALIBRATION of-the accumulator backup compressed gas system low pressure alare system and verifying an alars setpoint of 90
- 2 psig on decreasing pressure.
1
(- **TheprovisionsofSpecification4.d.4arer.at'applicabieTrevidedthe *
!. surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam. pressure is adequate to perform the test.
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'CONTAINMEN SYSTEMS
(,n, , o l ' PRIMARY CONTAINMENT AIR LOCK [ l.
" LIMITING CONDITION FOR OPERATION
. 3.6.1.3 The prima'ry containment str lock shall be OPERABLE with:
- l. 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 deor shall be closed, and
- h. An overall air lock leakage rate of less than or equal to 0.05 'L, at
- P,, 44.0 psig.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2*, and 3. - ACTION: 4
\'
I -
- a. With one primary containment air lock door inoperable: ' ,
I. Maintain at least the OPERABLE air lock door closed and either
. restore the inoperable air lock door to OPERABLE status withw 24 hours or lock the OPERABLE air lock door ' closed.
th) 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 on:e per . 31 , days.
. 3. Otherwise, be in at least HOT SHUTDOWN within ti.a pext 12 hours j and in COLD SHUTDOWN within the followirg 24 hours. .
- 4. The provisions of Specification 3.0.4 are not applicable.
- b. With the primary containment air inck 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 4.ad
, in COLD SHUTDOWN within the following 24 hours.
i *
)
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38740E6670 [ G ., CONTAINMENT SYSTEMS REACTOR Ehcovadae SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES {- LIMITING CONDITION FOR OPERATION 3.6.5.2.1 The reactor enclosure secondary containment ventilation system auto-matic isolation valves shown in Table 3/6.5.2.1-1 shall be OPERABLE with isolation times less than'or equal to the times shown in Table 3.6.5.2.1-1. f APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With one or more of the reactor secondary containment ventilation system automatic isciation valves shown in Table 3. 6. 5. 2.1- 1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours either:
- a. Restore the inoperable valves to OPERAELE status, or
- b. Isolate each affected pene rt ation by.use of at least one deactivated 7
valve secured in the isolation position, or i Isolate each affected penetration by use of at least one closed manual c. valve or bline flange.
- Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT.5HUTDOWN #
within the next -12 hours -and in COLD SHUTOOWN within the following 24 hours. h 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 q OPERABLE: Prior to returning the valve to service aft'er maintenance, repair or
- a. l
' 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 iso,lation l time. )
- b. . At least once per 18 months by verifying that on a containment L
1 solation 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.
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p o l ! etANT Sv5TEss .3691002560 f I Cov SYSTEMS LIMIT]NG CONDITION FOR ODERATION l 3.7.6.3 The following low pressure CO2 system shall be OPERABLE: ,
- s. ControlRoomEntrance,HoseRackOHR601andDH0o2. .. # i APPLICABILITY: Whenever equipment protected by the CDs systeef is required to j l be OPERA 56L. p ACTION:
- a. With the above required CO2 system inoperable, within 1 hout establish a continuous fire watch witt backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hour)y fire watch patrol,
- b. The provisions of Specification 3.0.3 are not applicable. l SURVEILLANCE REQUIREMENTS 4.7.6.3.1 The above reovired low pressure CDs system shall be demonstrated OPERABLE at least once per 7 days by verifying the CO2 sturage tank level to be ,
greater than 25% and pressure to be greater than 265 psig. 4.7.6.3.2 The aoove recuired CD 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. I LIMERICK - UNIT 1 3/4 7-24 Amendment he.11 froV TE
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-2.11.1.1 The' cen:entration of radioactive material releasec in liccic effluents to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall o ' e limite: to the con =entrations specifiec in 10 CFR Part 20, Appendix E, Table II, Colume. 2 for racionu:lices einer than dissolvec or entrainec noble gases. For diss:1vec or entrainec nocle
' gases, the concentration shall be limitec to 2 x 10 4 micro:uries/mi total activity.
AoDLICAEILITY: At all times. ACTION: With the concentration of radioactive material released in licuid effiuents te UNRESTRICTED AREAS exceeding the above limits,1ernediately restore tne con:er.- tration to within the above limits. 4 SURVEILLANCE REQUIREMENTS C. 4.11.1.1.1 Radica:tive licuid wastes shall be sampled and analy:ec ac:ording to the sampling and analysis prograr. of Table 4.11.1.1.1-1. 4.11.1.1.2 Tne results of the radion:tivity analyses shall be usec ir. . i accorcan:e with the methodology and parameters in the DDOM to assure that the concentrations at the point of release are e.aintained within the licits of Spe:ification 3.11.1.1. , 1 1 i LIMERICK - UNIT 1 3/4 11-1 4# E W.P -
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i I l RAO* 0.00:0AL EWi':RONWER~A'. MONITOU NO /ANW_ ! IAE.! :.t2.2 '. (;ontinued) ) i TAEL ACTA ~;0NS "Soecific pa-ameters ed distance and cirectior, sector f ron the cen*e-line ef the *wC reactors anc soci!ior.a1 cesc'iptier. Where pertinent, shall De provicec for each and every sample location in Tacle 3.12.1-1 in a ta:1e anc figure (s) in tne ODCM. Deviations are perzittec from the reovired sampling scheoule if specimens are unobtainable due to hazardous concitions, seasonal unavailability, malf unction of automatic sampling equipment and other legitire. ate reasons. If
- specimens are unobtainable due to sampling eovipment malfunction, every effort
- shall be mace to complete corrective action prior to the end of the next saclin; period. All . deviations from the samoling schedule shall be documentet in the Annual Radiological Environmental Operating Report pursuant to Specification 6.c.1.7.
It is recognizec that, at times, it may not be possible or practicable to con-tinue to obtain sacles of the media of choice at the mort cesired location or time. In these instances sL'itatde alternative media and locations may be chosen for the particular patnway in quertion and appropriate substitutions made within 30 days in the radiological environmental monitoring pectrar.. Pursuant to Specification 6.9.1.8, identify t'ie cause of the unavailability of samples for that pathway and identify tne new locat on(s) for obtai ing replacement samples in the next Semiannual Radioactive Effluent Release Report
'and also incluce in the repo-t a revised figure (s) and table for the 00?'
reflecting the new locationfs), b One or more instruments, such as a pressurized ion chamber, for measuring and C recording dose rate continuously may be used in place of, or ir. aedition ic. integrating dosimeters. For the purposes of this *.able, a thermolurinescent i dosimeter (TLD) is considered te be one phosphor; two or more phospnoes in a packet are consicered as twc or more cosimeters. Film bacges snail not ce used as cosimeters ,for measuring direct radiation.
" Methodology for' recovery of radiciodine shall be described in the .00 M.
d Airborne particulate sample filters shall be analy:ed for gross beta radio-activity 24 hours or more after sampling to allow for racon and theren 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 cuantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. -
f The " upstream sample" shall be taken at a distance beyond significant influence of the dischaege. The
- downstream" sample shall be taken in an area beyond but near the mixing zone. -
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R ADICi.0GI C A' EWIRONW.IR~AL M0t!!TORINO /we I TAELE L 'l.2 . . (;ontinued) o i 4 iAE I NOTAT*0N! EA composite sam;1e is one in whict.'the cuan*ity (alicuot) of 'licuid sampiec is proportional to the quantity cf ficaing licuid anc in which the methoc of sampling employed results in a specimen that is representative of the licuit flow. In this program composite sample alicuots shall be collected at time intervals that are very short (e.g. , hourly) relative to the compositing period (e.g. , acnthly) in order to assure cetaining a representative sample, h Groundwater samples shall be taken when this source is, tappec for erinking or irrigation purposes in areas where the hycraulic gradient or recharge properties are suitable for contamination.
's
'The dose shell be calculated for the maximum organ and age group, using the methodology and parameters in the ODOM.
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VITHUNIik2[INCON0!TIDh1,2,OF.3 POSITION NUM3!3 Of IN0!V70VALS REOL' IRE TD r:Li PO!!* 0h CON 0! TION 1. 2. or 3 00N *!Oh a e- 5 55 l' l' SRC l' l' RO 2" ;
~AL - NLO 2" 1 STA l' hene TABLE NOTATIONS "Indivicual may fill the same position on Unit 2.
**0ne of the two recuired incivicuals may fill the same pcsition on Unit 2.
55 - Shift Superintendent or Shift. Supervisor with a Senice Ope ster license on Unit 1. SRO - Incivicual with a Senior Operator license on Unit 1. RD - Incividual with an Operator license on Unit 1. HLO - Non-licensed operator properly Qualifisc to su; port the unit te which assignec. STA - Shift Technical Advisor Except for Shift Supervision (55), the shift crew composition may be one less than the sinimum requirements of Tacle 6.2.2-1 for a perioc of time not te exceed 2 hours in creer to at:ommodate unexpectec absence of on-duty snift crew members provided immediate action is taken to restore the shift crew como:- sition to within the minimum requirements of Table 6.2.2-1. This provision does not perir.it any shif t crew position to be unmannec upon shif t change cue to an oncoming shift crewman being late or absent. During any absence of Shift Supervision (55) from the control room while the uni is in OPERATIONAL CONDITION 1, 2, or 3, an individual (other than the Shif t Technical Advisor) with a valid Senior Operator license shall be oesignatec to assume the control room command function. During any absence of Snift 1 Supervision from the control room wnile the unit is in OPERAi!ONA; 00N117 0N
- i or 5, an individual with a valic Senior Operater license or Operator license i shall be designatec to assume the control room commanc functica.
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i I i J SAFE *Y LIMITS f..:. h
. BASES i 2.1. 3 ' REACTOR COOLANT SYSTEM PRESSURE The Saf'ety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code 1958 Edition, including Addenda through Summer 1969, which permits a maximum pres-sure transient of 110%, 1375 psig, of design pressure 1250 psig. The Safety Limit of 1325 psig, as sensured Dy the reactor vessel steam dome pressure indicator, is equivalent .to 1375 psig at the lowest elevation of the reactor cociant system. The reactor colant system is. designed to the ASME Boiler and Pressure Vessel Code, Edition, including Addenda through Summer 1977 for the reactor recirculatief 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. 'he pressure Safety Limit is, selected to be the lowest transient overpressure allowed by the ASME Boiler and Pressure Vessel Code
; Section III Class I. (, 79 7p 7p ,
2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is
,JE . __ shutdown, _ consideration must be given to water level requirements due to the effect of decay heat.' If the water level should drop below the top of the
?,,f active irradiated fuel during 'this period, the ability to remove decay heat is reduced. This reduction in cooling capability'could lead to elevated cladding temperatures and clad perforation in the event that the water level became l'ess 'l than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.
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BASE! 3 /4. 3. 2 ISO.Ai:DA A*iUA*10A IN$tRUWEN"A*]OA This specificatier. ensures the effectiveness of the instrumentation usec to mitigate the consequences of accioents by prescribing the OPERABILITY trip setpoints anc response times for isolation of the reactor systems. When necesse*y, one channel may be inoperable for brief intervals to concuct recuired surveillance.
, Some of the trip settings may have tolerances explicitly stated where both the high
*anc low values are critical and may have a substantial effect on safety. ine set-
' points of other ins rumentation, where only the high or lo, and of the setting have a direct bearing or safety, are established at a , level away from the normal rperating range to prevent. inadvertent actuation of the systems involved.
. Except for the P.57Vs, the safety analysis does not accress individual sense
- response times or the response times of the logic systems to which the sensers are connectec. For D.C. operated valves, a 3 seconc celay is assumec before
.the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency ciesel generators. In this event, a time of 13 seconds is assumec before.the valve starts to move. In addition to the pipe break, the failure of the D.C. operatec
' valve is assumed; thus the signal delay (sensor resporise) is concurrent witr the 10-second diesel startup and the 3 seconc load center loading celay. Tne safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second celay. It follows #
that checking the valve speeds and the 13-second time for emergency power estac-lishment will establish the response time for the isolation functions. Operation wit'h a trip set less conservative than its Trip 5etpoint but within its specifiec Allowable Value is acceptable on the basis that the difference between each Trip 5etpoint and the Allowable Value is an allowance for instrument drif t specifically allocated for each trip in the saf ety analyses. 3/4.3.3 EMERGEN0Y CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accicents that are beyond the ability of the operator to control. This specification provices the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be usec to send the actuation signal to more than one system at the same time.
' Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Yalue is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance h[ for instrisnent drift specifica11/y allocated for each trip in the safety.antlyses. .
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Ju?~RuvECA~;Dh EASES , T: s. . r, a . :~~ y-v~r~ m -- ' ~ce r-3/4.3.7.12 EA:~0A0~1VE GASIOUS EFFLUEN' ON:TOF.:NO ]NSTRUMECAT:Dh The radioactive gaseous effivent ins.runentation is provice: to monitor and control -as applicable, the releases of radioactive materials ir gaseous effluents curing actual or potential releases of gaseous effluents. The alarr./ trip setpcints for these instruments sna11 be calculatec in accordan:e witt. the proceoutes in the ODOM te ensure that the alarr./ trip will oc:ur prior te
-exceeding the limits of 10 CFR Part 20. This~ instrumentation also in:1uces provisions for monitoring the concentrations of potentia 11) explosive gas mix-tures in the off gas system. The OPERABILITY anc use of this instrw.mntation-is consistent witn the requirements of General Design Criteria 60, 63, and e4 of Appendix A to 10 CFR Part 50. '
3/4.3.8 TUREINE OVEr.5PEEC PRDTECTION SYS Ev. This specification is penviced to ensure that the turbine overspeed prote: tion system instrumentation and the turbine speed control valves are OPERABLE anc will protect the turcine from excessive overspee:. Protectic-from turbine excess.ive overspeed is required since excessive overspee: of the turbine could generate potentially dar, aging missiles which could impa:t anc damage safety related components, eovipment or structures. 3 /4. 3. 9 FEEDWATER/ MAIN TUREINE TRIF SYSTEM ACTUATION INSTRUMENTATION C- Tne feedwater/ main turbine trip systee actuation instrumentation is proviced to initiate action of the feedwater system /mair. turbine -i; syster in the event of failure of feeowater controller under maximum deman:..
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. BASES FIGURE B 3/4.4.5-1
- At 90% of RATED THERMA!. POWER.and 90% avai1 ability.
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3 /t . E C0CA!NMEC SYSTEMS . BASES , 3 /4. E . 2 cRIMARY CONTAINMENT 3 /4. 6. 2. 2 PRIMARY CONTAINMENT ICESRITY PRIMARY CONTAINMENT Ih7EGRITY ensures that the release of radioactive mate-rien, from the containment atmosphere will be restrictec to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BO'JNDARY.
. radiation doses to within the limits of 10' CFR Part 100 during accident concitions.
3 /4. 6.1. 2 PRIMARv CONTAINMENT LEAKAGE ' i The limitations'en primary containment leakag'e rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accioent pressure of f.4.02 psig, P As an addec conserva-tism, the measured overall integrated leakage rate is f6r.ther.limitec to less-during performance of the periodic tests to acceunt f or than possible or equal to 0.75 L*the degradation of containment leakage barriers between leakage tests. Operating experience .vith the majn staan line isolation valves har indicated that degradation.has occasionally occurred in the ' leak tightness of the valves; therefore the special requirement for testing these valves. The surveillance testing for measuring leakage rates is consistent with 8 the requirements of Appendix J of 10 CFR Part 50 with the exception of - . exemptions granted for leak testing of the main steam isolation valve:, the airlock $TIP shear valves.'
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3/4.6.1.3 PRIMARY CONTAINMENT AIR LO0rJ . MT e limitations on closure and leak rate for the primary containment air lockfare required to meet the restrictions on PRIMARY CONTAINMENT INTEGF.ITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2. The specification makes all ces for the fact that there may be long periods of time when the air loc ill be in a closed and secured porition during reactor operation. O y one closed door in esek air lock y. is required to maintain the integrity of the containment. h /\ 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postuisted LOCA situations would be a small fraction of the 10 CFR Part 200 guidelines, provided the main steam line sys+.em from the isolation valves up to and including the turbine rondenser remains intact. Operating experience has indicated that degradati, a has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage
- requirements have not always been maintained continuously. The requirement for the leakage control system will reduce the untreated leakage from the MSIVs ,
when isolation of the primary system and containment is required.
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BASES 3/4.6.2. DEPRESSURIIATION SYSTEMS The specifications of this section ensure that the primary containment pressure-will not exceed the design pressure of 55 psig during primary system blowdown from full operatin; pressure. 1
- The suppression enamber water provides the heat sink for,the reactor l coolant system energy release following a postulatec rupture of the system. '
The suppression chamber water volume must absort the associated decay and structural sensible heat released during reactor coolant syster blowcown from 1040 psig. Since all of the gases in the d ywell are purged into the suopres-sion chamter air space during a less-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant M " M : M d s discharged to the , suppression chamoet and that the drywell volume is purged tothe suppression chamber. p . Using the a:inimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is approximately 30 psig wnich is below the design pressure of 55 psig. ' Maximur water volume M- of 134,600 ft8 results in a downcomer submergence of 12'3" and the minieur volume.cf 122,120 ft 8 results ir. a submergence approximately 2'3" less. The
) sajority of the Bodega tests were run with a submerged length of 4 feet and with complete condensaticti.. Thus, with respect to the downcomtr submergence, ~
this specification is adeauste. The maximum temperature at the end of the , blowdown' tested during the Humbolet Bay and Bodega Bay tests was 1*C'F and this is conservatively taken to De the limit for complete condensation of the reactor
. coolant, although condensation would occur for temperatures above 170*F.
Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3. Under full power operating conditions, blowdown through safety / relief valves assuming an initial suppression chamber water temperature of 95'F results in a bulk water temperature of approximately 136*F immediately following blowdown l' ' which is below the 190'F bulk. temperature limit used for complete condensation via T-quencher eevices. At this temperature and atmospheric pressure, the avail-able NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LDCA operations. Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained e below 200*F during any period of relief valve operation for T-quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially hign suppre;sion chamber loadings. E.~ I Sr. LIMERICK - UNIT 1 B 3/* 6-3 l e
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1 ELECTRICAL F0WER SYSTEw.S I BASES 3/4.6.4 ELECTRICAL EOUIDMENT PROTEtiTVE DEVICES Primary containment electrical penetrations and penetration concacters . are protected by either de-energizing circuits not rer,iuired during reactor operation er demonstrating the OPERABILITY of primary and backup overcurrent
, protection circuit br:akers by periodic surveillance.
. The surveillance reQuiremen'.s applicable to lower voltage circuit breake-s provides assurance of breaker reliability by testing at least one representative sample of each manufacturers brand of circuit breaker. Esch manufacturer's molded case circuit breakers are grouped into representative samples which are th/ntestedonarotatingbasistoensurethatallbreakersaretestec.
(C. [ De bypassing of the motor operated valves themal overiced protection continuously by integral bypass devices ensures that the themal overicad pro-tection vill not prevent safety related valves from perfoming their function. The Surveillance Requirements for demonstrating the bypassing of the themel overload protection continuously are met by functionally testing the automatic operation of the motor ope.ated valva and ensuring that the motor themt1
- overload protection design does not change and is in accorcance with Pegulatory
- Guide 1.106 Themal DvMead Protection for Electric Motors on Motor Operatec g Valves", Revision 1. March 1977.
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3/4.10' SPECIAL TEST EXCEDTIONS / BASES
- 3/4.10.1 oRIMARY CONTAINMENT INTEGR:TY ine requirement- for PRIMARY C0HTAINMENT INTEGRITY is not applicable during .
the period when open vessel tests are being performed curing the low power
. PHYSICS TESTS.
3/4.10.2 ROD WORTH MINIMIZER , l In order to perform the te .s required in the technical specifications it is necessary to bypass the quence restraints on control. rod movement. The-additional surveillance recui ents ensure that the specifications on heat generation rates and. shutdown margin requirements are not exceeded during the
-k period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.
3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed-requires additional restrictions in order to ensure that criticality coes not occur. These additional restrictions are specified in this LCO. 3/4.10.4 RECIRCULATION LOOPS This special test exceotion permits reactor criticality under no flow conditions and is required to J.erform certain startup and PHYSICS TESTS while at low THERMAL POWER levels. 3/4.10.5 OXYGEN CONCENTRATION Relief from the oxygen concentration specifications .is necessary in orcer to. provide access to the primary containment curing the initial startup and testing phase of operation. Without this access the startup and test program could be restricted and delayed. 3/4.10.6 TRAINING STARTUPS l l This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the rapicactive waste disposal ',ystem. l LIMERICK - UNIT 1 2 3/4 10-1 Amenoment NO. 17 MAE 2 219I! 1
TAELY 3.3.7.9-1 (Continued)
,- FIRE DEE: TION INSTRUv!NTATICN I
INS RUMENT LO"ATION TOTAL NUu!Eo Or INSTRUu!C !' FIRE 20NE STRUCTURE ELEV. AREA HEAT SMOKE FLAME (x/y) (x/y) IEE 36 Unit 1 177' 'C' Core Spray Pump NA 2/0 NA Reactor Room 113 37 Ur.it 1 177' 'D' Core Spray Pur.: NA 2/C NA
. Reactor Roo: 114 35 Unit i 177' 'B' Core Spray Pump NA 2/0 NA Reactor Room 117 39 Unit 1 .177' Su :. Roo .115; NA 4/0 NA Rea: tor Passageway 118 .
4* Unit 1 177' Corricor .5t*~ NA 2/0 NA
- Reactor j,j, j,,
41 Unit 1 20;' Pl*V E uipment Area 2C7 0/10 3/0 NA Rea: tor 42A Unit 1 20 Safeguard Syster A::ess 0/12 3/C NA Rea:ter Area 200 43 Unit 1 217' Safeguard Syster Isolation NA 8/C NA Rea: tor Valve Area 305
@ ,. 44 Unit 1 217' Safeguar: Syste- A::ess 0/5 27/0 NA
! Reacter Area 304 (Southwest) 0/14 (hartheast) 45A Unit 1 253' CRC byeraulic E::vipment 0/',6 20/0- NA Rea: tor Area 4C2 455 Unit 1 253' Neutron Monitoring 0/2 2/; NA .
Reactor System Area 405 45: Unit 1 253' CR" Repair Room 403 NA 1/0 NA Reactor 47A Unit 1 253' Corridor 506; General 0/18 21/0 NA Reactor Equipment Area 500 478 Unit 1 295' Isolation Valve NA 2/0 NA Reactor Compartment 523 47; Unit 1 283' Fuel Pool Coolin'g Water NA 2/0 NA Reactor Pump and Heat Exchanger Area 511 47D Unit 1 283' Isolation Valve NA 1/0 NA Reactor Compartment 510/522 y, , - ., ...:. <n. so.... .. . . . . . , s .. s 1: n _ . . . n
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l CONTA!NMEN" SYSTEMS SURVEILLANCE RE0VIREME'iT5 (Continued) 1 a
- d. Typ6 E_an C tests shall be condu:ted with gas at P , 44.c psig , ,
at in .e-vals ne greater tnan 24 m:ntnsd:ept for tests involvin;: l
~
- 1. ' Air lo:ks , - -
- 2. Main stea line isolation valves,- ]
-Containment isolation valves in hydrostatically teste lines !
- 3. l wni:n penetrate the primary containettnt, anc'
- e. Air lo:ks sna11 be te' stet and cer.onstrated OPERABLE'per Surveillan:e Requirement 4.6.1.2.
. Main steam line isola-icn valves shall be leak tested at least on:e per 18 montns.
Containment isclation valves in hydrostatically tected lines whi:n g. penetrate the prima y containment shall be lear tested a- least on:s per 16 montns. 4 l[
- h. The provisions c' 5:e:ificatien 4.0.2 are not a:plicable ::-Spe:ifi: -
ti ons 4. 6.1. 2a. , 4. 6.1. 2: . , 4. 6.1. 2:. , 4. 6.1. 2 . , and 4. 6.1. 2 e.
=
'"Unless. a hydrostatic test'is required per Table 3.6.3-1.
*A Type c test interval extension to May 26,1986-is permissw ee Tor primaryD )
.[ containment isolation valves identified by an asterisk in the inboard and outboard isolation barrier columns of Table 3.6.3-1,'Part A, as discussed in L lf Application for Amendment of Facility Operating License dated December IS, IC. .' ." .
liti. 3 "' 3/4 6-4 Amendment N:. 2 LIMERICK - UNIT 1
.. . .. v . .- -
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- P *I #
e p a w' o + . - ' _.
- - _ ~ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _________ _ ___ _
CONTAINMENT SYSTEMS SURVE!LLANCE REQU1REwfNTS
-- 1 4.6.3.2 .iach primary containment isolation valve shown in Table 3.6.3-2 shall be demonstrated CPERABLE prior te returning ?he valve to service after mainte-nance, repair or replacement work is performed en the valve or its associated actuator, control or power circuit by cycling the valve through at least ene complete cycle of full travel and verifying the specified isolation time.
4.6.3.2 Each primary containment automatic isoistion valve shown in Table 3.6.3-1 shall be demonstrated OPERA!LE during COLD SHUTDOWN er REFUELING at least once per 18 months by verifying that on a containment isolation test signal asen automatic isclation 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 3.6.31 shall be determinee to be within its limit when tested pursuant to specification 4.0.5. " 4.6.3.4 Each reacter instrumentation.1ine excess flow che:k valve shown in I I Table 3.6.3-1 shall be demonstrated OPERABLE at least once per 18 .sn hs by - l verifying that the valve checks flow. 4.6.3.5 Ea:h traversing in-core probe system explosive isolation valve shall be demonstrated CPERAELE:
- a. . At least once per 31 days by . verifying the continuity of the' explosive charge. .
s-
- b. At least once per 18 senths by removing the ex;1esive scuit fro- the (
- explosive valve, such that eaen explosive squib in each explosive valve will be tested at least on=e per 90 eenths, and initiating the explcsive s:cib. The recla:esant charge fer the ex;1cce: squib sna11 be frot the sase s.anufa:tured batch as the ons fire: cr frc . : .cther batch which has been certifier' by having at least one of that batch successfully fired. Ne squib shall rer.ain in use beycad the expiraticn of its shelf-life and/ot operating life, as applicable.- ,
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TABLE 3.6.3-1 PRIMARY CONTAIN6NT ISCLATION VALVES t0TATION NOTES (Continued)
- 15. Check valve used instead of flow orifice.
- 16. Penetration is sealed oy a flange with double 0-ring seals. These seals -
are leakage rate tested by pressurizing between the 0-rings. Both the TIP Purge Supply (Penetration 35A) and the TIP Drive Tubes (Penetration ~ 35C-G) are welded to their respective flanges. Leakage througn these seals is included in the' Type C. leakage rate total for this-penetration. The ball valves (XV-141A-E) are Type C tested. It is not practicable to leak test the shear valves (XV-140A-E) because squib firing is required for closure. Shear valves (XV-140A-g) are normally open. 17. Instrument line isolatkn provisions consist of an excess flow check valve. g Because the instrument line .is connected to a closed cooling water systerr. - inside containment, no flow orifice is provided. The line does. net 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).
- l8' . In addition to double "0" ring seals, this penetration is tested by pres-surizing volum between doors per Specification 4.6.1.3.
- 19. The RHR system safety pressure relief valves c'" is ._.vt:f *$= +ha
\
iriti:,1 LL"T. 'h;'r li;f ; & c:: '- th::: :: c'" M e;-f rf + r ---+ r 4-- r: . ; n :: = t r' ; th; '-iti;.1 IL",7 ...
'1 I' 11 ..i.. m m, I .' . - . . I , .. . i t':n. ::fi'inti :: ; " i: g rf;r; d ;.; ;;,, 7-: .. ,.f 1;... .. f..: , ...
in:1 t= ting r -::: :1 :nd i:::t t;: tin;':1 th; nii:' C;n #n; : 2 n;r-t LL"- "r:0 - ' i r' n' r: which are flanged to facilitate removal will be equipped with double 0-ring seal assemblies on the flange closest
- to primary containment,5y " e-f r* *?? # d -~ -- ' " " C ^ " *
- C - These 1 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 non-PCRVICS isolation signals also initiate closure of selected valves:
. EA Main steam line high pressure, high steam line leakage flow, low MSIV-LCS dilution air flow LFHP With HPCI pumps running, opens on low flow in associated pipe, closes when flow is above setpoint LFRC With RCIC pump running, opens on low flow in associated pipe, closes when flow is above setpoint
. LFCH With CSS pump running, opens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve fully closed or RCIC turbine stop valve fully -
closed L All power operated isolation valves may be opened or closed remote manually. LIMERICK - UNIT 1 3/4 6-42 IE i
tog A!w.EW~ SYSTPS
'5UPVEILLANCE: REQUIREMENTS (Continued',
t. At least on:e pe- IE mentns or (1) af ter any stru:tural mainte an:e on the HEPA filtee or enarcoal acseroer housings, or:(2) follo.ing painting, fire, or enerrical release in any ventilation':ene communicating with the suosystem by:
- 1. Ve-ifyin; that the sutsyster satisfies the ir-place penetration and Dysass leakage testing a:ceptan:e criteria of less tnan 0.051-and uses the test procedure guidance in Regulatory Positions C.E.a, C.E.c 'anc C.S.c of Regulatory Guide 1.52, Revision 2, Mar:n 1975, an: tne system flow rate is 3000 cfm : 10% .-
( Ve ifying witnin 31 days after removal that'a laboratory analysis 2. cf a re:resentative careen sam:ie ottsinec in a:corean:e vitt Regulate y Position C.E.t of Reguistory Guice 1.52, Revision 2, Mar:n 197E, meets the laborate y testin; :riteria e' Regulate y Position C.E.a of Regulato*y Guice 1.52, Revision 2, Mar:h 1978,
'for a metnyl iocide penetration cf less than 0.175%; an:
Ve i'y the; wnen the f an is running the subsyster fiewrate is
- 3. U 28% cfr minimum f rom ea:n reactor en:losuae a (Zone (Zones I:I) when I en: I:) f anc 220C-cfr minimum free the refueling teste: i* a::ercan:e .itn ANI! N510-1980
- 4 Verify that the pressure crop across the refueling area te-5G~5 p-efilter is less than 0.25 inches water gage whine operating at a fio. rate cf 2400 cfm : 101.
- c. Afte- every 720 heu-s cf char::a1 a:scrber operatict by ve-i'ying witnin 31 cays af ter removal inat a laceratory analysis cf a re:re-sentative carcen sar:ie obtained in s.::crean:e with Regdate y Position C.E.b of Regulatory Guice 1.52, Revision 2 Mar:n 157E, meets the laboratory testing critern of Regulatory Positic . C.E.a of Regulatory Guice 1.52, Revision 2, Mar:n 1978, for a metnyi iodice penetration of less than 0.175%.
- d. At least on:e per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA .
. filters and charcoal assorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rat.e of 8400 efa
- 10%.. .. . w-
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./
T ["5pecifisc suosystem flow rate is for a two unit operation. During the Unit 2 \
.,.' .' construction phase, the Unit I subsystem flow rate will be 2800 cfr minimum
'"" .;? fro: the' rea: tor en:losure and 2200 cfm minimum from the refueling area
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[' h pcAN~ SYSTEM 5 3g; g SDRAY AN:/05 58tihe,ER SYSTEMS LIM:i:N: CONDITION FOR OPERs'10N 3.7.f.2 ine following spray 'an sprinkler systems shall et 0:ERAE.!: L Fire Zone Description Rea: tor En:losure Hatchway heter Curtains:
- 1. EL 253'
{ 2. EL 2E3'
- 3. EL 313'
. Fire Area Separation Water Curtains: .
4EA 1. Area 602, EL 313' 45A 2. Area 402 EL 253' W9 44 3. Area 304, EL 217' (2 cu-t ins) 22 CatieSprescingRoom,Rooe4W,EL254[ j 2/ Control Structure Fan Room. EL 304' 27 CREFAS System Filters, EL 304' M EE SGT5 Filters , Compartment 624, n: :~~:
~..... .. .... EL 332' 33 RCIC Pump Room, Room 108. EL 177' HDCI Pum; Fre 3 com 105, EL 177' 34 4 41 RECW Areat EL 20; 5 t o ?
42A 4 Safe;uarc Syster Access Area 200, EL 201' \ 44 Safeguarc Systen A::ess Area 304, - EL 2'.7' (Partiai) (-2 systies) 45A CR; Hyerauli: Ecuipment Area 402, Rea:ter En:lesu e,
. EL 253' (Fattial) .
45E hectron M:nitoring Systee Area 406 El 2E3' (Pa-tial) < 47A General Equipment Area.50C anc Corricer SDE, Rea:te-Enclosure, EL 283' (Partial) E;A & E Rea:ter En:lesure Recirculation Systee Filters, EL 33;' 75,60,61,82 Diesel Generator cells (4 Cells) , 1 ADDL1:A!Iti'Y: Whenever ecuipment prote:ted by the spray av/or sprinkler systems is recuired te be OPERAELE. ACTION:
- a. With one or more of the above required spray and/or sprinkler systems inoperable, within I hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or colnponents could be damaged; for other areas, establish an hourly fire watch patrol,
- b. The provisions of Specification 3.0.3 are not applicable. l
*::.^. . . wired to De urt.ne .d ' C . 6 .%.<e... "' d R AT F n "' 8"I' * *1'~ ~ A
[ LIMERICK - UNIT 1 3/4 7-22 Amen: ment he. :1 gy 't ge!
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'O TA!.! 4 11.2.1.2 1 (Continvec)
' TAE;E cTA'!cs; a
t San.: ling an: ana1yses snail.41sc De pe feare: fcliowi,;g shutcow , sta-tu , c- a THERMAL PowEk change er ee:ing 15% cf the RATED THERMA. PO=Et within a .
- hcu- pe-io:. This re:Licene.t coes net a :1y if (1) analysis sho s that tne DOSE EOL'IVALEhi I 13; cence.testien ia the prima y cociant hat n:t in:-ease: me-e *. nan a f a:ter of 3; anc (2) the mair. concense- of f gas pre-treatmert racioa:tivity monito- shows that etfluent activity has not increasec mere than a fa:ter cf 3.
c Sam:les shall be changee at least once per 7 cays and analyses sna11 ee com lete: within 48 heu s af te changing, or af ter removal f ree sam:le-Sam: ling snail aise et pe f orinec at least on:e per 24 hou s fc at least 7 cays fc11owing es:h shutcown, sta tu , or THERKAL POWER change exceecin; 15% of RATIC THERMAL P0d!R in 1 hout anc analyses comcleted wit.hin 48 hou s cf changing. Vnen sam:les celle:teg fo 24 hours a*e analy:ec, the corresponding LLDs may De in:-ease: by a factor of 1*. Tnis reavireme.i coes not ap:1y if (1) analysis shows that the DOSE E0.' VALENT I-131 con:er.tration in the prima y coelant has net increasec mere than a fa: tor cf 3: and (2) the nc:le gas monitor shows that effluent a:tivity has n:t increasec more than a f a: tor of 3. c The ratic cf the sample fio. rate te the sampled strear fio= rate shall be know*. for the time pe*ic: cove-ed by ta:n c:se o- ccse rate calc.*ation mace in accorcan:e witt. Specifications 3.11.2.1, 3.1'. 2.2, an: 3.1". 2.3. e The p*incipal game.a eritte's for which the LLO see:ification a: lies in:1uce the fello.in; racionu:lices: Ar-67, Kr-SS , Xe-133, . xe ;3 3e , Xe-135. Xe *.35e a.: Ae-138 fc gasecas e-issions anc Mr-St. Fe-is, Cc-55, Cc-6~, 2:-65, Mc-Ei, * '.r., ts-134, Cs-137, Ce-121 anc Ce *44 f c- ;a~ ticulate erissions. This list cces net mean that cnly these nu:lices a-e te De consicerec. Other gama pear.s which are icentifia:1e, togetne w:tn these of the a:cve nu:lices, shall aisc be analy:e: an: rep: te: in the Seeta .nual Racica:tive Ef fluent Release Report, pursuant to Spe:ifica-tion 6.5.1.E. f Under .he previsions cf foctnote e, above, only ncble gases nee: te te consic,erec. g O-tnEc.*.'- e w .. - , . .. r : -- --< : " = . a. ,- . a. ~ . n . a., k, voe4.e4.amat.: ht Recuired for the het maintenance shop ventilation exhaust only curing opera-tien of.the hot maintenance shop ventilation exhaust system. A""- S!!!i LIMERICK - UNIT 1 3/4 11-11
- .--~:-- - - - - . . ~ . ~ ~~- . = v w
l 1 CE-*CTIONI RATED THE:HAL 00WE:.
)
1 1.32 RATE: THERMAL POWER shall be a total rea: tor core heat transfer rate to
- the rea:.or coolant of 3253 Wt. .
REACTORENOLOSURESE*0NDARYCONTANMEriTINTEGRITY 1.33 REACTOR ENOLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when: (
- a. All reactor enclosure seconda y containment penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE seconcary c:ntainment 2.
automatic isolation system, or Closed by at least one manual valve, blind flange, g g g) automatic valve secured in its closed position, ex: gor eptdeactivated as providec in Table 3.6.5.2.1-1 of Specification 3.6.5.2.1.
- c. All rea: tor en:losure secondary containment hat:hes and blowout panels are closed and sealed.
- c. The standby gas treatment system is in compliance with the requirements or Spe:ification 3.E.5.3.
- d. The reactor enclosure re:irculation system is in compliance with the requirements of Specification 3.6.5.4.
4 e. At least one door in each a :ess to the rea: tor enclosure se:encar~v # containment is closed. .
- f. The sealing mechar.is associated with each rea:ter enclos"ub seconca containment penetration, e.g. , welds, bellows, or 0-rings, is OcERASLE.
- g. The pressure within the rea: tor enclosure secondary containment is less than or equal to the value required by Spe:ification 4.E.5.1.;a.
REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PRCTECTION SYSTEr1 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. 1 REFUELING FLOOR SECONDARY CONTAINMENT IKTEGRITY 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:
- a. All refueling floor. secondary co'ntainment penetrations required to ,
be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or gg h , ~ 9 *
- 2. Closed by. at least one manual valve, blind flange,Aer deactivated automatic valve secured in its closed position, except as provided in Table 3.6.5.2.2-1 of Specification 3.6.5.2.2.
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, Ta:1e 3.3.7 4-1 (Continued)
~
RHF $!RV :E WATER SYSTEM (Continued) H55-12-094 Centrol-Transfer switch ! H55-12-052 Control-TransfAr switch - l j 51-1F014A Control-1A-RHR heat excharger tube side inlet . CAP 506 , . Control-RHR Service Water pump
,HV-51-1F068A' Control-1A RHR Heat exchanger tube side outlet
. EMERGENCY SERVICE WATER SYSTEM OAP548 Cont'ol-r emergency service' water pump -
[ H n-011A Control- emergency service water disch to RHR' service water i H55-n-091 - Contrel-Transfer switch g . H55-11-092 Control-Transfer switch H55-D-093 Contrel-Transfer switch Ac
' ."". - .The fellowing valves of the ESW and RHR5W systects are actuatec by signals fro-the transfer switches:
ESW and RHR5W put s wetwell intertie gate HV 12-005 H5'-n-015A E5W loop A ditcharge to RHRSV loop B i ' ( HV-12-017A ESW and RHR5W cooling tower return cross-tie STANDBY AC POWER SU8 PLY i 152- n509/CSR 101-011 Safeguard SWGR feeder bkr. 152-n609/CSR 101-D12 Safeguard SWGR feeder bkt. 152-n709/CSR 101-013 Safeguard SWGR feeder bkr. i 152-11502/CSR 201-Dn Safeguard SWGR feeder bkr. , 152-11602/CSR 201-D12 Safeguard SWGR feeder bkr. _.1 - 9 152-n702/CSR 201-013 Safeguard SWGR feeder bkr. 152-n505/CSR D U4 Safeguard LC XFMR breaker "
- .: 5 113 b.' LIMERICK - UNIT 1 3/4 3-81 . . . . . . _g
~ ~ ~ ~
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I 4 l IAE!l 3.3.7.9-1 TIREDEi! ION INSTRUMENTATION TOTAL NUp!E: 0* IN3~:lgvECS* h_. INSTRUMENT LOCA* ION FIRE ZONE STRUCTURE ELEV. AREA HEAT SMOKE F.Aw!- (x/y) (x/y) (x/y, IL Control 200' Control Structure Chillers and NA 3/0 NA Chilled Water Pump Area 258 1.M Control 200' Control Structure Chillers and NA , 3/0 NA
- Chilled Water Pump Area 263 2 Control 217' 13-kV Swit hgear Area 336 NA 34/0 hA 3 Control 217' Battery Room 323 (1D) 1/0 1/0 NA
;4 Control 217' Battery Room 324 (1C) 1/0 1/0 NA 7 Control 239' Corridor 437 'A N 5/0 NA B Control 239' Battery Room 425 (IE U 1E2) 1/0 ,
2/C NA
'S Control 239' Battery Room 435 (1A1/1A2) 1/0 2/0 NA
~
12 -Control 239' 4-kV Switchgear Compartment 2/0 2/C NA 434 (013) . 13 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 435 (011) 14 Control 239' 4-kV Swit:hgear Compartment 2/0 2/0 NA 4 432 (014) f.. c. 15 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 433 (012) , 20 Control 254' Static Inver'ter Room Unit 1, NA 4/0 NA Area 452 Control 254' Cable Screading Room Unit 1, NA 14/0 NA 52 Area 449 24A Control 269' Control Room 533 NA 23(a)/0 NA 11(b)/0 Control Control Room Utility Room 529 NA 1/0 NA 24B 2.69' 24C Control 269' Control Room Office 531 g NA 1/0 NA Control 269' Control Room Shift Supt. E r NA 1/0 NA 24D , Control 269' Control Room Shop 534 NA 1/0 NA 2'4E (Photo-Elect) Control 269' Control Room Instrument NA 1/0 NA 24F Lab 535 *(Photo-4 Elect) 269' Control Room Shift Supt. NA 1/0 NA 24G Control - 53 3/4 3-93 C- E%
- l. '. LIMERICK - UNIT I a een. g
TAS'.E 3.3L1 4-1 (Continued) .. FIRE DE~ECTION INSTRUMENTATION ( INSTRUMENT LOCATION - TOTAL NUMEEP Or INSTRUME C 5' FIFE. 20NE STRUCTURE ELEV. . AREA HEAT ~ SM0KE FLAu! (X/y)' (x/y) (X/y) - 48A Unit 1 313' Laydo'wn Areas 601 and 602; NA 8/0 NA Reactor , Corricer and RERS Fan Area 605 51A Unit 1 331'- RERS Filter 2/0 NA . NA Reactor Compartment 618 (inside - plenum) SIE Unit 1 331' RERS Filter 2/0 NA NA Reactor Compartment $12 (inside plenum) 79 Diesel- 217' Diesel-Generater 1/5 4/0 1/0 . Generator Cell Unit 1 80 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1
~
81 Diesel- 217' Diesel-Generat:r 1/5 4/0 1/C Generator . Cell Unit 1 82 Diesel- 217' Diesei-Generator 1/5 4/C 1/C Generator Cell Unit 1 , p 122A Spray 2 6 8.'. .. ESW and RHRSW Pume Area NA 4/0 NA Pond Pum; - ( Structure
. 122E Spray ' 251' RHR5W Valve Compartment NA 2/0 NA -
Pond Pump . Structure.
. 123A Spray 26E' ESW and'RHR5W Pump Area NA 4/0 NA Pone Pump 4
~
Structure 123E Spray 251' RHRSW Valve Compartment HA 2/0 NA
- Pond Pump Structure 124A Diesel- 217' Diesel-Generator Access NA 4/0 NA Generator Corridor 313 Vr 126A Common 412' North Stack Instrument NA 2/0 NA Reactor Room7/342~3
" (x/y): X is the number of Function A (Early Warning Fire Detection and -
Notification Dnly) Instruments. ' ' 1 Y is the number of Function B (Activation of Fire Suppression 7 System and Early Warning Notification) Instruments. 1 (a) These smoke detectors are located below the sus;iended ceiling in the Control Room.
~
(b) These smoke detectors are located above the suspended ceiling in the g. ' Control Room. LIMERICK - UNIT 1 3/4 3-96 W !'0 t
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- Crywell va*uut Dreater sha'* 0,
- 4. ke'ifie: :lest: at least on:e per 7 Cays.
2 De*:-st ate: OCIRA!'!: . A; leas
- Cnce pe* 21 Cays en: within 2 hcuas afte* 2"y c'.s:ca ;t 1 a'.e3 cf stear : the su;;-ession cnar:er free the safety / relief ty Cy:Iin; e a*h vaggue Ortaker thrcugn a* leas
- cne Cc"; .t'.e cy:le cf 1L*1 travel.
2 en:e pe* j; cays by ve ifyin; be*h cesition in:i:at:-s j v-:ng:.: 0, c:sta.in; exce:te: valve .e.:vemer.; : .:n; tne gy:::r; at..le.ast. test.
- 3. A *. l e a s *. C *: e De
- 16 E-* ths by ;
a) Ve-ifyin; ea:n valve's c:enin; set:: int, fr:- : e :', se:
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a.: ea:- 's k[ ca:a:'e cf ce:e: tin; cist dis;ia:eee.; ' in::t : valve's ::s tice in icater is it::: .e :' ce:e::ir; c'st cis:la:e er.: [*.*2*". . e d n (l'J i t u..r r. .t e. g . UN *. *> i. 3/4 6-48. A'.,,: 8 b...e:
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7--=_----__-. 762iI 28 7. 6. 5- 2 ' i
,. FIRE H05? STATIONS
' HOSE RA;L i
LO:4 ~ ! O'. ELEVA 10h 7 DEN ~~r::s~:0N
- 1. Coetaci En:1osure:
Stairwell 350' IHR-241 Stairwell, Outside SGTS Room 332' 1HR-240 i
. Stairwell, Outsice Fan Room 304' 1Fh-103 j
~ '
Dutsice 13kV Svitchgear Root 217' 1HR-116 Stairwell, Outside Aur Equip Rm 289' 1HR-130 Stairwell, Outside Cable Spreading Rn; 254' 1HR 250 Wall, Outsice.4kV Svit:hgea? &'Ea*te y
}
Rooms 23S' 1HR-251 Corricer L4E, Soutt Sice cf 4kV
.5vit:n;eer & Eatte y Reors 235' 1HR-124 Wall, Cerricor 2E! 200' IHP-12:
Well, Corridor 164 , 180' IHR-12;
- 2. Refueline Area:
C. SW Corner Refuel Floor 252' 1HR-201
; NW Corner Refuel Floer 357' 1HR-20 North Wall
- Center- 352' 1HR-2 3 .
So. n Wall-Center 352' 1HR-204
- 3. Reaet:- En:1esu-e: .
SW Corner Rea: tor Entiosure 331' 1HR-205 SW Corner Reactor En:losure (RERS Fan Area) 313' 1HR-207 NW Corner Reactor En:losure (Laydown Area 601) 313'. 1MR-20E SE Corner Reactor En:1osure 313' 1HR-209 (Near Refuel Floor Exb. Fans) NE Corner Reactor Enclosure 313' 1HR-210 L
$ d W(Near D124 Load. Center)
^ ' Reactor Enclosure 283' 1HR-215 f
, (Corridor 506) .
NW Corner Reactor Enclosure 283' 1HR-216 (Corricor 506) I- I i '- LIMERICK - UNIT 1 3/4 7-27 t e e.s me, _gg p sog , esio y , e.pe e 4 . e ,p , gg, p $ig9 g ee gg s g
- 9 se oe S tse e s. ace g. e mee e y e tag, 9) P 'rf96M8'* .
***W**We &G %
/
i TAE'. E ? . 7. E. 5-1 (Continuet) FIRE HOSE STA*!DN! H:5! RA:t E '.E V A * ! O N 10Ex ;r;;s ;os LCOATION .
- 3. Rea:to* En:1rtv e:.(Continued)
SE. Corner Renctor Enclosure (SLC Pumps Area 500) 2E3' IHF-217 NE Corner Reactor Enclosure 283' 1HR*.:1F D fw r . 41 Wm Reactor Enclosure (Ares 402A, Near CR0 Repa'ir) 2E3' 1HR-203 NW Corner Rea: tor Enclosure . (hear Depe11 Ecuip Hatch) 2E3' 1HR-224 SE Corner Reactor En:1osure (hear Drpell Personnel Lo:k) 2E3' 1HR-005 East Wall Reactor En:losure Near TIP Machines) 2E3' 1HR-225 9"}" W" Reactor En:losure f'- (hear JPe* Equip Hater.) 217' 1HR-232
/
e it MC.
-A W Corner Reactor Enclosure f (hear Supp Pool A::ess Het:n) 217' 1HR-233 East Well Reactor En:1osure (hem
- Eouipment Atricck 300) 217' 1HC-234 NE Corner Rea: tor En:iosure. .
0124- R-G) 217' 1HR-23E N(hear W-.'.1MO: p k. Reactor En:losure (Near M C D134-R-H) 201' 1HR-240 . Nw' Corner Reactor Enclosure 201' 1HR-241 (Hear MCC C134-R-H1) East Wall Reactor Enclosure (Near RECW Heat Exchangers) 201' 1HR-242 NE Corner Reactor Enclosure ' (Naar RECW P. umps) 201' 1HR-243 SW Cor,ner Reactor Enclosure 177' 1HR-252 NW Coraer Reactor Enclosure 1 71' 1HR-253 NE C W er Reactor Enclosure 177' 1HR-142 k.3 C s'r - -.--- " LIMERICK - UNIT 1 3/4 7-25 i
* ~~
,.e =. ., ..
e ,. , , , g.. a . . . . , *,.. , , , . , , , ,.
. . . . . . . . . . .. . , . _ m . ..... . .. ,. ,,...,,, ,... ...v.,.
t c r TABLE 2.E.L.P ; (Continued) c F.:Mt. v CON siwn!K o!NETpt ION 000U TO: , l , DVER URREC oROTEC !VE DEV::E!
; 2. 4EO-V0;" MOLDE: CASE EF,EAKERS (Continue (1 CIR UIT SYSTEMS OR BREAKER NO, LOCATION TYFES EQUIPMENT POWERED 52-2241C D14- E- E IM H~E100 112 Drywell Area Unit-TM HFE100 Cooler 1E2V212 52-22411 Clu-R-E IM HFB100 1H2 Drywell Area Unit TM HFE100 Cooler 1H2V212 52-2241E DIM-R-E IM HFE50 HP I Mn Str Su;;1y Int-:
TM HFE150 Isol Viv HV-55-1F002 52-22516 114E-R-C IM HFE25 1A Rea: Re:ir: Pum: TM HFE10; Su: tion VLV HV-43-1F023A 52-2251E M45-R-0 IM HFE25 1A Reac Re:ir: Pum; TM HFE100 Discharge VLV HV-43-1FC21A 52-2252: n45-E-C IM H'E25 Rea:ter Bottom Hea: Drain V;V TM HF540 HV-4-1F10; . .
,C gwc.g .Tvje f ffcM du AWC. loofi 52-22E26 n45-R-C IM HFE25 etm . . .. ; . . . . .. . . ; . . . * 'l^-
TM H:E4; HV 1F105 52-22554 n45-R-0 IM HFE25 Res tor Vessel Head Vent TM HFB40 HV-41-1F001 52-22535 1145-R-C IM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-1F005 52-22537 M 45-R- TM HFE15 Disposal Cask Removal Cart TM HFE20 Heist 10H236 i 52-22538 n45-R-C TM HFB15 Control Red Drive Platfore TM HFB20 Heist 10H229 52-22608 1245-R-C TM Hi'515 CRD Equipment Handling TM HFB20 Platfors 10N22608 52-22618 1245-R-C IM HFB25 1B Rese. Recire. Pump TM HFE100 , Discharge VLV HV-43-1F031B
'j52-22622 124f-R-C TM HFB125 Permanent Plant In-Containment Welding System 10NW201 l
LIMirsICK - UNIT 1 3/4 8-25 # "'I I Mi .
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TAB LE 4.1'.. L 2. El (Continued) 8 TAB EASTATIONS "The' LLC is cefined, for purpores of these specifications, as the smallest concentration of radioactive caterial in a sample that will yield a net count, above syster background, that will be detected with 95% probability with nnly 5% prot,adility of falsely concluding that a blank observation represents a
real" signal.
For a particular measurement system, which may include radiochemical separation: 4.66s D
'LLD = -
E V ,2.22 x 108 [Y er.p (- ut) Where: LLD is the e priori lower limit of detection' as defined above (as microcuries per unit mass or volume), s is the standard' deviation of the background counting rate or of the cbunting rate of a blank sample as appropriate (as counts per minute), E is the counting efficiency. as counts per disintegration, p V_ is the sample size, in units of mass or volume,
) 2.22 x 105 is the number of disintegrations per minute per ricrocurie, Y is the fractional radiochemical yield, when applicatie '
A is the radioactive decay constant for the particular radionuclides, anc At for the plant effluents is the elapsed time between the siepoint of sample collection and time of counting. . Typical values of E, V, Y, and at sh;,.lld be used in the calculation. It should be recognized that the LLD is defined as an a prie*i (be*cre the
' fact.) limit representing the capability of a measuremeiit systee at not as an a posteriori (after the fact) limit for a particular measurement.
1
) LIMERICK - Uh'IT 1 3/4 11-3 i
.'~
~
. . . . . . .: . . ..... . . . ~. ~-~-
5.C DESIGN 8'E ATUF.E!
/
[ E.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. LOV POPULATION ZONE
^
' 5.1. 2 The low population zone shall be as shown in Figure 5.1.2-1.
MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE CASEOUS AN* L]OUIO E % UENTS 5.1. 3 Inferr.ation regarding radioactive gaseous and licuid effluents, which will allow icer.tification of structures and release points as well as defini-tion cf UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMEER OF THE PUBLIC, shall.be as shovn in Figures 5.1.3-la and 5.1.3-It. METEOROLOGICAL TOWER LOCATION - 5.1. 4 The metecrologica'l towers shall be located as shown on Figure 5.1.4-1. b 5.2 C0KTAINu.EN~
~ ~ ~ ' '-
)
l' . CONFIGURATION - 5.2.1 The primary containment is a steel lined reinforced concrete structu*e consisting of a drywell and suppression chacer. The drywell is a steel-lined reinforced concrete vassel in the shape et a truncated cone on top of a water . filled suppression chamber and is separated by a diaphragm slab and connected to the suppression chacer through a series of downcomer vents. The crywell l has a maximum free air volume of 243,580 cubic feet at a minimum suppression
- pool level of 22 feet. The suppression chamber has a maximum air region of l 155,540 cubic feet'and a minimum water region of 122,1 0 cubic feet.
DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:
- a. Maximum internal pressure 55 psig.
- b. Maximum internal temperature: drywell 340*F.
, 4 ,. suppression pool 210'F.
- c. Maximum externa ressure5psib.d -
dMer w / l d. Maximum floor dif.ferential pressure: 30 psid, downward. 20 psid, upward. 1 LIMERICK ~ UNIT I 31 L.: ; ,z.. t
r e 3'"'9:002560 l A0wN:STRATIVE C0C ROLS (a TRt!h:N3 f.a A retrainin; an: re:la:eme : tt aining cre;-ae 'er :*e uM sta " sna. :e main.ainec unce the cirettico cf the site Training c ;ad:atieny.: sna M mee; er ex:eec the requirements of' Ah!!/ANS 3.1-157E anc i;;r:- I
- 10 CFR Fa-; !!
anc the sue:1emental reevirements sce:ifie:.in Se:tiens A an: C cf En:1csu e 1. /' - cf :ne Mar:n 28,1980 NR* 1ette te a*1 licensees, anc snail in:1uce f acilia-i:t-tion witn relevant inCustry C;erational ex;trien:t. E.! REV!!w ANC AVO!' E. 5.'. DLAN OcERA :oNS REVIE. ::ws:- EE .POR:) FUNO :0h E.5.1.1 The POR: shall fun::'en te a: vise the Plan Manage- c til me::e s l reia*.e: :: nu: lear safety. C0we05**:0N E.5.1.2 Tne P R: small be : m:esec cf the: Cnairman: Suce-intence.:-Ope-ations Mee.ce : Su:e-intencea. Teennical Mem:e r: Su:erintencent-kaintenan:e/Instrumentatice.an: Conti:1s Mem:er: Sweetintencent-Plant Services b Memet : Assistant Superintendent-Operations i Meete : Repulatory Engineer Mem:e : Te:nnt:ai En;tnee- : Pe.:e-: Snift Su:e-intencer Wet *t
. : Mai.'.e.a*.:e Enginet*
e
- '. !:s:~Es 6.5.1.3 All alternate memte-s snail te a:::intec in w-iting'ty the POR*
C na i r.t.a n :: seave en a tem;cre y easis; however, ne me e tnan tw alte-nates snail participate as vcting mem.:ers in POR: a::ivities at any one time, w!E*:NG FREDUENCY E . 5.1. 4 The POR; shall meet at least on:e per calencar menin and as ccnvene: by the POR: Chaire.an er his cesignatec alternate. OUDRUM_ E.5.1.5 The cuerum of the POR: necessa y for the performance of the POR: responsibility and authority provisiens of these Te:nnical Specifications sna11 consist of the Chairman er his cesignatec alternate anc four semeets inclucing alternates. L w.!RI:K - UNIT 1 E-7 A~4n:~2** N#
- 10
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r.. TABLE /5.3.5-2 REACTOR CORE ISOLATION C00LINL.STSTEM' ACTUATION INSTRUMENTATION' $ETPOINTS ALLOW /.BLE -_ - FUNCTIONAL UNITS' TRIP SETPOINT VALUE
- a. ReactorYessel.Wateri.evel-Low Low,, Level 2 1-38 inches
- 1-45 inches
- t. Reactor Vessel Water Lovel -
- Hign, Level 8 f,, 54 inches 5, 60 inches .
- c. Condensate. Storage Tank Level -
Low 1 135.8** inches 1 132.3 inches
- d. Manual Initiation- N. A. N.A.
"See Bases ~ Figure B 3/4.3-1. .
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4 4 INDEX- ll-p , DEFINITIONS SECTION~ m.. l'. 0 DEFINITIONS PAGE 1.1' 'ACTI0N.......................................................- 1 1,2' AVERAGE PLANARIEXP0SURE............... ...................... ;1 1
~1.? ' AVERAGE PLANAR. LINEAR HEAT GENERATION RATE................... 1-1
.1.4 CHANNEL CALIBRATION........................................... 1-1 1.5 CHANNEL CHECK................................................- 'l-1 1.61 CHANNEL FUNCTIONAL TEST...................................... 1-1 1.7 CORE ALTERATION.............................................. 1-2.
1.8 CRITICAL' POWER RATI0......................................... 1.-2 : 1.9 DOSE EQUIVALENT I-131........................................ l'- 2 1.10 ~E-AVERAGE DISINTEGRATION ENERGY.............................. 1-2' 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME........... 1 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.... 1-2. 1.13 FRACTION OF LIMITING POWER DENSITY........................... 'l-3 1.14 FRACTION OF RATED' THERMAL POWER.............................. 1 1.15< FREQUENCY N0TATION............................................ 1-3' 1.16- IDENTIFIED LEAKAGE............................... . ........... 3 1.171 ISOLATION SYSTEM RESPONSE TIME................................ 1-3 1.18 LIMITING CONTROL R00 PATTERN................................. 1-3 1.19 -LINEAR HEAT GENERATION RATE.................................. 1-3 1 1.20 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-4 l'.21 MAXIMUM FRACTION OF LIMITING POWER DENSITY................... 1-4 l
- LIMERICK - UNIT 1 i !
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R l __ ___ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ - _ _ - _ - - _ - _ _ _ __ - _ _ _ A
INDEX l s DEFINITIONS i l 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 MA NUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.25 OPERABLE - OPERABILITY........................................... 1-4 1 1.26 OPERATIONAL CONDITION - CONDITION................................ 1-4 l 1 1.27 PHYSIC 5 TESTS.................................................... 1-4 1.28 PRESSURE BOUNDARY LEAKAGE........................................ 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY.................................... 1-5 . I 1.30 PROCESS CONTROL PR0 GRAM.......................................... 1-5 1.31 PURGE - PURGING.................................................. 1-5 1.32 RATED THERMAL P0WER.............................................. 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTf: .NT INTEGRITY................ 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME.......................... 1-6 1.35 REFUELING FLOOK SECONDARY CONTAINMENT INTEGRITY.................. 1-6 1.36 REPORTABLE EVENT................................................. 1-7 1.37 R0D DENSITY...................................................... 1-7
- 1. 3 8 S HUT DOW N MARG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 7
- 1. 3 9 S I T E 8 0 V H D A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 7
- 1. 40 SO L I D I F I CAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 7
- 1. 41 S O U R C E C H E C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 7 1.42 STAGGERED TEST BASIS............................................. 1-8 1.43 THERMAL P0WER.................................................... 1-8 1.44 UNIDENTIFIED LEAKAGE............................................. 1-8 LIMERICK - UNIT 1 11 I
? i INDEX
]
SAFETY LIMITS,AND LIMITING SAFETY. SYSTEM SETTINGS t
'SECTION PAGE-2.1 SAFETY LIMITS
' THERMAL POWER, Low Pressure or Low Flow................. 2-1 THERMAL POWER, High Pressure and High Flow.............. 2-1 Reactor Coolant System Pressure......................... 2-1 Reactor Vessel Water Leve1.............................. 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 ' BASES 2.1 SAFETY ___ LIMITS THERMAL POWER, Low Pressure or Low Flow................. B 2-1 THERMAL POWER, High Pressure and High Flow.............. B 2-2 Left Intentionally Blank ............................... B 2-3 Left Intentionally-Blank ............................... B 2-4
. Reactor Coolant System Pressure......................... B 2-5 Reactor Vessel Water Leve1.............................. B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints..... B 2-6 LIMERICK - UNIT 1 iv 4
l! _ m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
INDEX 1 q, '
' LIMIT 2NG COND1TIONS_FOR OPERATION AND SURVEILLANCE REQUIREMENTS iSECTION PAGE- l POWER DISTRIBUTION LIMITS'(Continued) m figure 3.2.1-2 Maximum Average Plar. Linear Heat. l Generation Rate (MA' ttGR) Versus Average Planar Expr.JreLInitial I Core Fuel Tv .. . W IB248 '3/4 2 ~
Figure.3.2.1 Maximum Average Planar' Linear Heat. .l Generation Rate-(MAPLHGR) Versus l Average Planar Exposure-Initial. ' Core Fuel Types P8CIB163 3/4 2-4
' Figure 3.2.1-4 Maximum Average Planar Linear Heat !
Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial o- Core Fuel Types P8CIB094 3/4 2-5' { Figure 3.2.1-5 Maximum Average Planar: Linear Heat ; Generation Rete (MAPLHGR)-Versus l Average Planar Exposure initial Core. Fuel Types P8C18071 3/4 2-6 ! Figure 3.2.1-6 Maximum Average Planar Linear Heat- ; Generation Rate (MAPLHGR).Versus
' Average Planar Exposure For Fuel Type BC320A (GE8X8EB).. 3/4 2-6a Figure 3.2.1-7 Maximum Average Planar Linear i
~ Heat Generation Rate:(MAPLHGR)
.Versus Average Pl.inar Exposure ,
H For Fue* Type BC318A (GE8X8EB)' 3/4 2-6b ! Figure 3.2.1-8 Maximum Average. Planar Linear 'I Heat Generation Rate (MAPLHGR). Versus Average Planar Exposure . For Fuel. Type BC322A (GE8X8EB) 3/4 2-6c
' 3/4 2.2 APRM.SETPOINTS- 3/4 2-7 3/4 2.3 MINIMUM CRITICAL P0WER RATIO 3/4 2-8 Table 3.2.3-1 Deleted --
l Figure 3.2.3-la Minimum Critical Power Ratio (MCfh) i Versus T (P8X8R/BP8X8R Fuel) BOC to ; EOC-2000 MWD /ST 3/4 2-10 J Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR) Versus t (P8X8R/BP8X8RFuel) E0C-2000 MWD /ST to E0C 3/4 2-10a LIMERICK UNIT 1- vi Amendment No. 19 i
-_ _ _ _ _.___ _ __ u
i l INDEX LIMITING CONDIT, IONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS g TION PAGE P'JWER DISTRIBUTION LIMITS (Continued) Figure 3.2.3-1c Minimum Critical Power Ratio (MCPR) Versus T (GE0X8EB Fuel) B0C to E0C-2000 MWD /ST 3/4 2-10b Figure 3.2.3-1d Minimum Critical Power Ratio (MCPR) Versus T (GE8X8EB Fuel) E0C-2000 MWD /ST to E0C 3/4 2-10c Figure 3.2.3-2 Kf 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 iable 3.3.1-1 Reactor Protection System 3/4 3-2 Instrumentation i Table 3.3.1-2 Reactor Protection System 3/4 3-6 Response Times Table 4.3.1.1-1 Reactor Protection 3/4 3-7 Instrumentation Surveillance Requirements i ( , LIMERICK UNIT 1 via Amendment No. 19 l' l
INDEX L
-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
'SECTION PAGE INSTRUMENTATION (continued)
{ 3/4.3.2 ISOLATION ACTUATION INSTRUMENTS ..................... 3/4 3-9 Table 3.3.2-1 Isolation-Actuation Instrumentation ...................... 3/4 3-11 a 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' Table 4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements .. .................... 3/4 3-2? 3/4.3.3 EF.ERGENCY 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 Couling System Response. Times ....................... 3/4 3-39
~ Table 4.3.3.1-1 Emergency Core 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 l Instrumentation Surveillance "
Requirements ....................... 3/4 3-45 i End-of-Cycle Recirculation Pump Trip System i Instrumentation ........................................... 3/4 3-46 LIMERICK - UNIT 1 vii i p
INDEX < LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I 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 3/4.5.2 ECCS - SHUTD0WN............................................... 3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER........................................... 3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity................................. 3/4 6-1 Primary Containment Leakage................................... 3/4 6-2 Primary Containment Air Lock.................................. 3/4 6-5 MSIV Leakage Control System................................... 3/4 6-7 Primary Containment Structural Integrity...................... 3/4 6-8 Drywell and Suppression Chamber Internal Pressure............. 3/4 6-9 Drywell Average Air Temperature............................... 3/4 6-10 Drywell and Suppression Chamber Purge System.................. 3/4 6-11 3/4.5.2 DEPRESSURIZATION SYSTEMS Suppression Chamber.......................................... 3/4 6-12 S.ppression 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 Va1ves...................................... 3/4 6-19 LIMERICK - UNIT 1 xii
INDEX
'l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS-I SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.2 . CONTROL ROOM EMERGENCY FRESH AIR S'lPPLY SYSTEM. . . . . . . . . . . . . 3/4 7-6 3/4.7.3 SYSTEM......................
~
REACTOR CORE ISOLATION COOLING 3/4 7-9 3/4.7.4 SNUBBER.................................................... 3/4 7-11 Figure 4.7.4-1 Sample Plan 2) For Snubber Functional Test........................ 3/4 7-16 3/4.7.5 SEALED SOURCE CONTAMINATION................................ 3/4 7-17 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System.............................. 3/4 7-19" Spray and/or Sprinkler Systems............................. 3/4 7-22' C02 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 3/4.8.2 D. C. SOURCES D.'C. Sources - Operating.................................. 3/4 8-10 LIMERICK - UNIT 1 yiv
- 4 I
\; l t - - - - . - . - . - . - - _ - - . - _ _ - - - - _ - _ - - - _ - . - - - _ - - - - _ _ _ _ _
-r l
INDEX l . . LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REFUELING OPIRATIONS (Continued) 3/4.9.10 CONTROL ROD REMOVAL
. Single Control Rod Remova1................ ................. 3/4 9-13 4 Wi tipl e Control Rod 1temoval . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT-CIRCULATION High Water teve1............................................ 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 ROD WORTH MINIMIZER.......................................... 3/4 10-2 3/4.10.3 SHUTDOWN 14ARG IN DEMDNSTRAT I ONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-3 3/4.10.4 RECIRCULATION L00PS......................................... 3/4 40-4 3/4.10.5 DXYGEN C0NCENTRATlDR........................................ 3/4 10-5.
3/4.10.6 TRAINING STARTUPS........................................... 3/4 10-6 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1. 1.IQUID EFFLUENTS Concentration.............................. _ ..... . . 3/4 11-1 Table 4.11.1.1.1-1 Radioactive Liquid Maste Stapling and Analysis Progras........... ..... .... .. 3/4 11-2 0ose........................................................ 'J/4 11-5 liquid Radwaste Treatment System. . . . . . . .. . . .. .... . . . . ..- . . 3/4 11-6 Li qu i d Hol dup ' Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 3/4 11-7 3/4.11.2 SAtFMt EFF1 DENTS Dose Rate................................................... 3/4 11-8 r LIMERICK - UNIT 1 xvi Amendment No. 17
INDEX 'BAerg SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTRO: SYSTEMS 3/4.1.1 SHUTDOWN MARGIN............................................. B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES......................................... B 3/4 1-1 ' 3/4.1.3 CONTROL R0DS.. ............................................. B 3/4 1-2 i 3/4.1.4 CONTROL R00 PROGRAM CONTR0LS................................ B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL 5YSTEM............................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT CCNERATION i RATE........................................................ B 3/4 2-1 LEFT INTENTIONALLY BLANK............................................... 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 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION........................... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation........................ B 3/4 3-4 i LIMERICK - UNIT 1 xviii _________-_w
1
+
INDEX
. BASES SECTION , PAGE REACTOR COOLANT SYSTEM (Continued) ,
l 3/4.4.5 SPECIFIC ACTIVITY........................................... B 3/4 4-4 ( 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.................................. B 3/4 4-4 { l Bases Table B 3/4.4.6-1 Reettor Vessel J Toughness................... B 3/4 4-7 i 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 , 1 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................ B 3/4 4-6 l 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 I 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 CONTA!NMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment 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 l Drywell Average Air Temperature...................... B 3/4 6-2 Drywell and Suppression Chamber Purge System......... B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS............................. B 3/4 6-3 LIMERICK - UNIT 1 xx
hp, INDEX A-
- ADMIN'ISTRATIVE CONTROLS
- SECTIOND 0
-PAGE 6'.1= RESPONSIBILITY........e.......................................... 6-1 6.2 ~ ORGANIZATION..................................................... 6-1 6.2.1 0ffsite.a.................................................... 6-1 pq . Figure.6.2.1-1 DELETED.................................' 6-3 6.2.2. Unit Itaff.................................................... l6-1 Figure 6.2.2-1 DELETED................................. 6-4~
Table 6.2.2-1 Minimum Shift Crew Composition............................. 6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)- Function..................................................... 6-6 Composition.................................................. . .6-6 Responsibilities............................................. 6-6 Records...................................................... 6-6 16.~ 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 Meeting Frequency............................................ 6-7 L
e Quorum....................................................... 6-7 Responsibilities............................................. 6-8 Records...................................................... 6-9 l' l l LIMERICK'- UNIT 1 xxvi g i 1_... L_ ]
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TABLE 3.3.2-3 (Continued) ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION' RESPONSE TIME (Seconds)#
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
- a. HPCI Steam Line A Pressure - High 5 13(a)
- b. HPCI Steam Supply Pressure - Low 5 13(a)
- c. HPCI Turbine Exhaust Diaphragm Pressure - High N.A.
- d. HPCI Equipment Room Pressure - High N.A. I
- e. HPCI Equipment Room A Temperature - High N.A.
- f. HPCI Pipe Routing Area Temperature - High h.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. RC": Equirment Room A Temperature - High N.A.
- f. RCIC Pipe Routing Area Temperature - High N.A. I
- g. Manual Initiation H.A.
LIMERICK - UNIT 1 3/4 3-24 1
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, REMOTE SHUTDOWN SYSTEM CONTROLS
.RCIC SYSTEM l 'HSS-49-191 Control-Transfer Switch HSS-49-192- Control-Transfer Switch fU ~HSS-49-193 Control-Transfer Switch HSS-49-195 Control-Transfer Switch HSS-49-196 Control-Transfer Switch HV-49-1F076 Control-Steam Line warmup bypass valve HV-49-1F060 Control-RCIC turb exhaust to suppression pool isolation HV-50-112 Control-Turb trip throttle valve HV-50-1F045 . Control-Turbine steam supply valve HV-49-1F008 Control-Turbine steam line outboard isolation valve HV-49-1F007 Control-Turbine steam line inboard isolation valve HV-49-1F031 Control-RCIC pump suction from suppression pool
, HV-49-1F029 Control-RCIC pump suction from suppression pool HV+49-1F010' Control-RCIC pump suction from condensate storage tank HV-49-1F019 Control-Minimum flow' bypass to suppression pool HV-49-1F022 Control-Test return to condensate storage tank HV-50-1F046 Control-RCIC turbine. cooling water valve HV-49-1F012 Control-RCIC pump disch valve HV-49-1F013 Control-RCIC pump disch valve 10P220 Control-Vacuum tank condensate pump 10P219 Control-Barometric condenser vacuum pump HV-49-1F002 Control-Barometric condenser vacuum pump disch LIMERICK - UNIT 1 3/4 3-78 -_=-_____--_________________.-________-___________-_____-_____-_- _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - - _ _ _ _ _ _ _ _ _ - - _ _ _ _ -
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t I INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued)
~!
- b. IAt least once per 31 days by:
- 1. Cycling each of the following valves through at least one' complete cycle from the running position:
a) For the overspeed protection control system;
- 1) Four high pressure turbine control valves b) .For the electrical overspeed trip system and the mechanical.
overspeed trip system;
- 1) Four high pressure turbine control valves ll
- c. At least once per 18 months by performance of a CHANNEL CALIBRATION of the turbine overspeed protection instrumentation.
- d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of all valve seats, disks and stems and verifying no unacceptable flaws or excessive corrosion.. If unacceptable flaws or excessive corrosion are-found, all other valves of that type ~shall be inspected.
l 1 LIMERICK - UNIT 1 3/4 3-111 l
REACTOR' COOLANT SYSTEM , l SURVEILLANCE REQUIREMENTS -l i 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by: a.- Monitoring the primary containment atmospheric gaseous radioactivity at least once per 12 hours (not a means of quantifying-leakage),
- b. Monitoring the drywell floor drain sump and drywell equipment drain .
tank flow rate at least once per 12 hours,
- c. Monitoring the drywell unit coolers condensate flow rate at least once per 12 hours,
- d. Monitoring the primary containment pressure at least orece per 12 hours (not a means of quantifying leakage), ;
q
- e. Monitoring the reactor vessel head flange leak detection system at I least once per 24 hours, and
- f. Monitoring the primary containment temperature at least once per 24 hours (not a means of quantifying leakage).
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified ', 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: l
- a. At least once per 18 months, and
- b. Prior to returning the valve to service following maintenance, repair j 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 l 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.
l LIMERICK - UNIT 1 3/4 4-10
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( 3/4.5 EMERGENCY. CORE' COOLING SYSTEMS .j 3/4.5.1 ECCS - OPERATING , LIMITING ~ CONDITION FOR OPERATION L 3.5.1 :The emergency core cooling systems shall be OPERABLE with:
- a. The core spray system (CSS) consisting of two subsystems'with each subsystem comprised of:
- 1. Two OPERABLE CSS pumps, and l l
- 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the. spray sparger to the reactor vessel.
- b. The low pressure coolant injection (LPCI)' system of the residual heat removal system consisting of four subsystems with each subsystem
; comprised of:
1 One OPERABLE LPCI pump, and
- 2. An OPERABLE flow path capable of taking suction form 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 lease 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.
ifTwo 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 1 3/4 5-1
f EMERGENCY CORE C00LfNG SYSTEMS ; LIMITING CONDITION.FOR OPERATION (Continued) o. ACTION: (Continued)
- d. For the ADS:
- 1. -With one of-the above required ADS valves-inoperable. provided the HPCI system, . ; CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 !. curs and reduce reactor steam dome pressure to < 100 psig within the next' 24 hours.
- 2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours and reduce reactor steam dome pressure to _5 100 psig within the next 24 hours,
- e. With a CSS and/or LPCI header AP instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours or. determine the ECCS header AP locally at least once per 12 hours; otherwise, declare the associated CSS end/or LPCI, as applicable, inoperable.
'f. In the event en ECCS system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage l.-
factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70. l l l . l LIMERICK - UNIT 1 3/4 5-3
i
, EMERGENCY CORE: COOLING SYSTEMS !
i
' SURVEILLANCE REQUIREMENTS (Continued)
]
l
- 2. ,For the HPCI system, verifying that:
a) The system develops a flow of at least'5600 gpm against a test line pressure corresponding to a reactor vessel pressure of
> 200 psig plus head and line losses, when steam is being supplied to the turbine at 200 + 15, - O psig.** i b) The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chmber water level - high signal. l
- 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 instrumentation and verifying the setpoint to be 1 the allowable value of a 4.4 psid.
- 5. Performing a CHANNEL CALIBRATION of the LPCI header AP instrumentation and verifying the setpoint to be 1 the allowable value of 3.0 psid.
- d. For the ADS:
- 1. At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the accumulator backup compressed gas system low pressure alarm system.
- 2. At least once per 18 months:
a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation, b) Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig** and observing that either:
- 1) The control valve or bypass valve postion responds accordingly, or
- 2) There is a corresponding change in the measured steam flow.
c) Performing a CHANNEL CALIBRATION of the accumulator backup compressed gas system low pressure alarm system and verifying an alarm setpoint of 90 1 2 psig on decreasing pressure.
**The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
LIMERICK - UNIT 1 3/4 5-5 [
- CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK- l LIMITING CONDITION FOR OPERATION-25 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 shall1be closed, and
- b. 'An overall. air lock leakage rate of.less than or. equal to 0.05 La at Pt , 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 clo,ed.
- 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, he 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.
- b. With the primary containment air 'ock inoperable, except as a result of an inoperable air lock door, ma, tain at least one air lock door closed; o restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT CHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
' *See Special Test Exception 3.10.1.
LIMERICK - UNIT 1 3/4 6-5
a L CONTAINMENT SYSTEMS REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITED CONDITION FOR OPERATION 3.6.5.2.1 The reactor enclosure secondary. containment ventilation system automatic' isolation valves shown in Tab':t. 3.6.5.2.1-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.5.2.1-1. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: With one or more of the reactor secondary containment ventilation system autons'.ic isolation valves shown in Table 3.6.5.2.1-1 inoperable, maintain at least one isolation valve OPERABLE in each affected -inetration that is open and within 8 hours either:
- a. Restore the inoperable valves to OPERABLE status, or
- b. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or
- c. Isolate each affected penetration by use of at least one closed manual valve or blind flange.
Otherwise, in OPERATIONAL CONDIT.10N 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN withf.. the following 24 hours. SURVE1LLANCE 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 1 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 wonths 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.
LIMERICK - UNIT 1 3/4 6-48
l PLANT SYSTEMS \: l C09 SYSTEMS LIMITING CONDITION FOR OPERATION 1 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 C02 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 automatic) in the flow path is in its correct position. LIMERICK - UNIT 1 3/4 7-24 Amendment No. 11 _ _ _ - - - _ - _ _ _ - - _ - - _ _ . - - - - _ _ - - - _ _ - - -- )
L.
~3/4.11 RADIOACTIVE EFFLUENTS l
3/4.11.1 LIQUID EFFLUENTS CONCENTRATION
)
LIMITING CONDITIONS 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 nobles 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 ! concentration to within the above limits. SURVEILLANCE REQUIREMENTS 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. LIMERICK - UNIT 1 3/4 11-1 - _ - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ 1
I i TABLE 3.12.1-1 (Continued) 4 RADIOLOGICAL ENVIRONMENTAL MONITORING 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 ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant ;o Specification 6.9.1.7. It is recognized that, at times, it may not be possible or practicable to continue 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). b One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may oe used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. methodology for recovery of radioiodine shall be described in the ODCM. d Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after scmpling 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, e Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. I The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" samples shall be taken in an area beyond but near the mixing zone. LIMERICK - UNIT 1 3/4 12-7
, TABLE 3.12.1-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATIONS 9 A 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 collect 0d 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 ODCf;. e l l LIMERICK - UNIT 1 3/4 12-8
. ~ -l
-TABLE 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMMON CONTROL ROOM WITH UNIT 2 IN CONDITION 4 OR 5 OR DEFUELED l'
-POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SS 1* 1*
SRO 1 1*- R0 2 1 NLO 2 2**- STA 1 None 1 WITH UNIT 2 IN CONDITION 1, 2, OR 3 l POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITION 1, 2. or 3 CONDITION 4 or 5 4 SS 1* 1*
.SR0 1* 1*
R0 2** 1 NLO 2** 1 STA 1* None TABIE NOTATIONS
- Individual'may fill the same position on Unit'2.
**0ne of the two required individuals may fill the same position on Unit 2.
SS - Shift Superintendent or Shift Supervisor with a Senior Operator License on Unit 1. SRO - Individual with a Senior Operator license _on Unit 1. RO - Individual with an Operator license on Unit 1. NLO - Non-licensed operator properly qualified to s , port the unit to which assigned. STA - Shift Technical Advisor Except for Shift Supervision (SS), the shift crew composition may be one less than the minimum requirements of Table 6.2.2-1 for a period of time not to exceed 2 hours-in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. l During any absence of Shift Supervision (SS) from the control room while the unit is in OPERATIONAL CONDITION 1, 2, or 3, an individual (other than the Shift Technical Advisor) I with a valid Senior Operator license shall be designated to assume the control room
- command function. During any absence of Shift Supervision from the control room while the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command j function. !
l L LIMERICK - UNIT 1 6-5 i l i_ __ _--_ _ . _ - - - __
- _ l
.SAFETV LIMfTS BASES
.q
)
2.1.3 REACTOR COOLANT SYSTEM PRESSURE j Tne Safety Limit for the reactor coolant system pressure has been selected I such that it is at a pressure below which it can be shown that the integrity of l the system is not endangered. The reactor pressure vessel is designed to Section ; III of the ASME Boiler and Pressure Vessel Code 1968 Edition including Addenda ' through Summer 1969, which permits a maximum pressure transient of 110%, 1375 4 psig, of design pressure 1250 psig. The Safety Limit of 1325 psig, as measured by l 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 j
' designed to the ASME Boiler and Pressure Vessel Code, 1977 Edition, including l.
Addenda through Summer 1977 for the reactor recirculation piping, which permits a maximum pressure transient of 110%, 1375 psig of design pressure, 1250 psig for suction piping and 1500 psig for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the ASME Boiler and: Pressure Vessel Code Section III, Class I. i 2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shutdown, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide cdequate margin for effective action. I LIMERICK - UNIT 1 B 2-5
I INSTRUMENTATION 1 BASES I l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to , mitigate the. consequences of accidents by prescribing the OPERABILITY trip i setpoints and response times for isolation of the reactor systems. When I necessary, one channel may be inoperable for brief intervals to conduct required ) surveillance. Some of the trip settings may have tolerances explicitly stated j where both the high and low values are critical and may have a substantial i effect on safety. The setpoints of other instrumentation, where only the high 1 or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved. Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost aid is restored by startup of the emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts t move. In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second celay. It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION , The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. l LIMERICK - UNIT 1 B 3/4 3-2
1 l INSTRUMENTATION BASES l 3/4.3.7.12 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation.is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm /tiip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the off-gas system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3/4 3.8 TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures. 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand. LIMERICK - UNIT 1 8 3/4 3-7
.i w.
I 1.2 4 f 8 1.0 ' x Y I >
- 0.8-A.
E 0.6 p t I E 0.4 1 Z
/
/ '
0.2 0 10 20 30 40 Service LWe (Years') FAST NEUTRON FLUENCE (E>l MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE
- BASES FIGURE B 3/4.4.6-1
*At 90% of RATED THERMAL POWER and 90% availability.
LIMERICK - UNIT 1 B 3/4 4-8
i 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure of 44.02 psig, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests. Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves. The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for leak testing of the main steam isolation valves, the airlock and TIP shear valves. ( 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCK The limitations on closure and leak rate for the primary containment air lock are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.5.1.1 and 3.6.1.2. The 4 specification makes allowances for the fact that thare may be long periods of time when the air lock will be in a closed and secured position during reactor operation. Only one closed door in the air lock is required to maintain the ! integrity of the containment. l 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR Part 100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously. The requirements for , the leakage control system will reduce the untreated leakage from the MSIVs when l isolation of the primary system and containment is required. LIMERICK - UNIT 1 8 3/4 6-1
CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary system blowdown from full operating pressure. The suppression chamber water provides the heat sink for the reactor coolant system, energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1040 psig. Since all of the gases in the drywell are purged into the suppression chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber. l Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is approximately 30 psig which3is below the design pressure of 55 psig. Maximum water volume of 134,600 ft rgsultsinadowncomersubmergenceof12'3"andtheminimumvolume of 122,120 ft results in a submergence approximately 2'3" less. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. Themaximumtemperatureatgheendoftheblowdowntestedduringtne Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken tobethelimitforcompletecondensationoftheregctorcoolant,although condensation would occur for temperatures above 170 F. Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3. Underfullpoweroperatingconditions,blowdownthroughsafgty/reliefvalves assuminganinitialsuppressionchamberwatgrtemperatureof95Fresultsina bulk water temperature of approximately 136 F immediately following blowdown which is below the 1900F bulk temperature limit used for complete condensation via T-quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations. Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 2000 F during any period of relief valve operation for T-quencher devices. Specification.s have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings. LIMERICK - UNIT 1 B 3/4 6-3
ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary containment electrical penetrations and penetration conductors are protected by either de-energizing circuits not required during reactor operation or demonstrating the OPERABILITY of n*4 mary and backup overcurrent protection circuit breakers by periodic surveillance. The surveillance requirements applicable to lower voltage circuit breakers provides assurance of breaker reliability by testing at least one representative sample of each manufacturers brand of circuit breaker. Each manufacturer's molded case circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. l The bypassing of the motor operated valves thermal overload protection continuously by integral bypass devices ensures that the thermal overload protection will not prevent safety related valves from performing their function. The Surveillance Requirements for demonstrating the bypassing of the thermal overload protection continuously are met by functionally testing the automatic operation of the motor operated valve and ensuring that the motor thermal overload protection design does not change and is in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves", Revision 1, March 1977. 1 l LIMERICK - UNIT 1 B 3/4 8-3
3/4.10 .SPECIAL TEST EXCEPTIONS BASES i 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement of PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS-TESTS.. 3/4.10.2 R00 WORTH MINIMIZER In order to perform the tests required in the technical specifications it is-necessary to bypass the sequence restraints on control rod movement. The additional surveillance requirements ensure that the specifications on heat l generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis. 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO. 3/4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels. 3/4.10.5. OXYGEN CONCENTRATION Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the ini.tial startup and testing phase of op3 ration. Without this access the startup and test program could be restricted and delayed. 3/4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system. LIMERICK - UNIT 1 B 3/4 10-1 Amendment No. 17
r m j 1 w TABLE-3.3.7.9-1-(Continued) FIRE DETECTION INSTRUMENTATION
.INSTRUMENI LOCATJON TOTAL NUMBER OF INSTRUMENT J FIRE ~ ..
' ZONE STRUCTURE ELEV. AREA HEAT SMOKE. F' LAME (x/y) (x/y) -(x/y);
136 Unit.1 177' 'C' Core Spray Pump NA 2/0 NA Reactor Room 113 4 o 37 Unit:1' 177' 'D' Core Spray Pump NA 2/0 NA Reactor Room 114 38 Unit 1 177' 'B' Core Spray Pump NA 2/0 NA Reactor Room 117 39 Unit 1 177' Sump Room'115; NA 4/0 NA Reactor Passageway 118 40 Unit.1- 177' Corridor 111 NA 2/0 NA: l Reactor 41 Unit 1 -201' RECW Equipment Area 207 0/10 3/0 NA. l Reactor 42A Unit 1 201' Safeguard System Access 0/12 '3/0 NA l' Recctor- -Area 200.
> 43 Unit 1 217' Safeguard System' Isolation- NA 8/0 NA Reactor Valve Area 309 44 Unit 1 217'- Safeguard System Access 0/8 27/0 NA-Reactor Area 304 (Southwest) 0/14 (Northeast)-
' 5A 4 Unit 1 253' CRD Hydraulic Equipment 0/16 20/0 NA Reactor Area 402 458- Unit 1 253' Neutron Monitoring 0/2 2/0 NA Reactor System Area 406 45C Unit 1 253' CR0 Repair Room 403 NA 1/0 NA Reactor 47A Unit 1- 283' Corridor 506; General 0/18 21/0 NA l Reactor Equipment Area 500 478 Unit 1 295' Isolation Valve ,, NA 2/O' NA Reactor . Compartment 523 47C Unit 1 283' Fuel Pool Cooling Water NA 2/0. NA Reactor Pump and Heat Exchanger Area 511 47D Unit 1 283' Isolation Valve NA 1/0 NA' l Reactor Compartment 510/522 LIMERICK'- UNIT 1 3/4 3-95
' CONTAINMENT SYSTEMS SUR'VEILLANCE REQUIREMENTS (Continued)
I
- d. Type B and C tests shall be conducted with gas at Pa , 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 per1 18 months.
- h. The provisions of Specification 4.0.2 are not applicable to Specifications 4.6.1.2a., 4.6.1.2b., 4.6.1.2c, 4.6.1.2d, and 4.6.1.2e.
*Unless a hydrostatic test is required per Table 3.6.3-1.
i l LIMERICK - UNIT 1 3/4 6-4 1 _. __-____-__ _ _ a
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS i 4.6.3.1 Each primary containment isolation valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE prior to returning the valve to service after 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 l 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 Specification 4.0.5. 4.6.3.4 Each reactor instrumentation line excess flow check valve shown in Table 3.6.3-1 shall be demonstrated OPERABLE at least once per 18 months by verifying that the valve checks flow. 4.6.3.5 Each traversing in-core probe sl, stem explosive isolation valve shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying the continuity of the explosive charge,
- b. At least once per 18 months by removing the explosive squib from the explosive valve, such that each explosive squib in each explosive valve will be tested at least once per 90 months, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the -
same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life and/or operating life, as applicable. LIMERICK - UNIT 1 3/4 6-18
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1 5 I f TABLE 3.6.3-1 l i PRIMARY CONTAINMENT ISOLATION VALVES l NOTATION
]1 I
NOTES (Continbed) I l
- 15. Check valve used instead of flow orifice.
i
- 16. Penetration is sealed by a flange with double 0-ring seals. These seals L are leakage rate tested by pressurizing between the 0-rings. Both the TIP '
Purge Supply'(Penetration 35A) and the TIP Drive Tubes (Penetration 35C-G) l_ are welded to their respective flanges. Leakage through these seals is included in the Type C leakage rate total for this penetration. The ba' valves (XV-141A-E) are Type C tested. It is not practicable to leak test the shear valves (XV-140A-E) because squib firing is required for closure. Shear valves (XV-140A-E) are normally open. l
- 17. Instrument line isolation provisions consist of an excess flow check valve.
Because the instrument line is connected to a closed cooling water system inside containment, no flow orifice is provided. The line does not isolate during a LOCA and can' leak only if the line or instrument should rupture. Leaktightness of the line is verified during the integrated leak rate test (Type A test).
- 18. In addition to double "0" ring seals, this penetration is tested by pressurizing volume between doors per Specification 4.6.1.3.
- 19. The RHR system safety pressure relief valves which are flanged to facilitate removal will be 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 non-PCRVICS isolation signals also initiate closure of selected valves:
EA Main steam line high pressure, high steam line leakage flow, low MSIV-LCS dilution air flow. LFHP With HPCI pumps running, opens on low flow in associated pipe, closes when flow is above setpoint. LFRC With RCIC pump running, opens on low flow in associated pipe, closes when flow is above setpoint. LFCH With CSS pump running, opens on low flow in associated pipe, closes when flow is above setpoint. LFCC Steam supply valve fully closed or RCIC turbine stop valve fully closed. All power operated isolation valves may be opened or closed remote manually. LIMERICK - UNIT 1 3/4 6-42
G. .
- bONTAINMENTSYSTEMS lSURVEILLA'CE-REQUIREMENTS N 1(Continued).
t
.b. At.leastione 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'O.05% and uses the test procedure guidance in Regulatory Positions C.5.a. C.S.c and C.5.d'of Regulatory Guide 1.52, Revision 2. March.1978,'and the system-flow rate is 3000 cfm i 10%.
s
- 2. . Verifying within'31 days after removal that.a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Pos.ition C.6.b of: Regulatory Guide 1.52, Revision 2, March 1978, meet's
'the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.5.2, 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 flow rate 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..
- 4. Verify that the pressure drop across the refueling; area to SGTS prefilter is less than 0.25 inches water gage while operating at a. flow rate of 2400 cfm i 10%.
.c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2 March 1978, meets the laboratory testing' criteria of.
Regulatory Position C.6.a of Regulatory Guide 1.52,. Revision 2. March 1978, for a methyl iodide penetration of less than 0.175%. , d. At least once per 18 months by: l 1. Ver ifying that the pressure drop across the combined HEPA filters and . charcoal adsorber banks is less than 9.1 inches water gauge while operating'the filter train at a flow rate of 8400 cfm i 10%. 1 LIMERICK - UNIT 1 3/4 6-53
l I
- i. !
y PLANT SYSTEF SPRAY AND/0R SPRINKLER SYSTEMS l i LIMITING CONDITION FOR OPERATION - 3.7.6.2 The following spray and sprinkler systems shall be OPERABLE: Fire Zone Descriptio_n Reactor Enclosure Hatchway Water Curtains:
- 1. EL 253'
- 2. EL 283'
- 3. EL 313' Fire Area Separation Water Curtains:
48A 1. Area 602, EL 313' 45A 2. Area 402, EL 253' 44 3. Area 304, EL 217' (2 curtains) 22 Cable Spreading Room, Room 449, EL 254' l 27 Control Structure Fan Room. EL 304' 27 CREFAS System Filters EL 304' 28A SGTS Access Area 625. EL 332' l 288 SGTS Filters, Compartment 624 EL 322' 33 RCIC Pump Room, Room 108. EL 177 34 HPCI Pump Room, Room 109. EL 177' 41- RECW Area 207, EL 201' l 42A . Safeguard System Access Area 200, EL 201' I 44 Safeguard System Access Area 304, EL 217' (Partial) (2 systems) 45A CRD Hydraulic Equipment Area 402, Reactor Enclosure, EL 253' (Partial) 45B Neutron Monitoring System Area 406, EL 253' (Partial) 47A General Equipment Area 500 and Cperidor 506, Reactor Enclosure, EL 283' (Partial) 51A & B Reactor Enclosure Recirculation System Filters, EL 331' 79,80,81,82 Diesel Generator cells (4 Cells) APPLICABILITY: Whenever equipment protected by the spray and/or sprinkler systems is required to be OPERABLE. ACTION:
- a. With one or more of the above required spray and/or sprinkler systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specification 3.0.3 are not applicable.
LIMERICK - UNIT 1 3/4 7-22 Amendment No. 11
- _ _ _ _ _ _ - _ __ _. _ _ _ _ _ _ _ _ _ _ _____-___-__-______---__-______-_______-__-__-__-_-__-______-_-D
TABLE 4.11.2.1.2-1 (Continued) l TABLE NOTATIONS b Sampling and analyses shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1- i hour period. This requirement does not apply if (1) analysis shows that the j 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 I radioactivity monitor shows that effluent activity has not increased more than l a factor of 3. c Samples shall be changed at least once per 7 days and analyses shall be ! completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at least 7 days following eac'. shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour and analyses completed within 48 hours of changing. When samples collected for 24 hours are analy2ed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not 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 LLO 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, 2n-65, Mo-99, I-131, Cs-134 Cs-137, Ce-141 and Ce-144 for particulate 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 l Report, pursuant to Specification 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 operation of the hot maintenance shop ventilation exhaust system. LIMERICK - UNIT 1 3/4 11-11
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:
- 1. All reactor enclosure secondary containment penetrations required to be closed during accident condition 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 deactivated automatic valve secured in its closed position, except as provided in Table 3.6.5.2.1-1 of Specification 3.6.5.2.1.
- b. All reactor enclosure secondary containment hatches and blowout panels are closed and sealed.
- c. The standby gas treatment system is in compliance with the requirements of Specif1caticn 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 teattor enclosure secondary Containment is CIDsed.
- 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 PROTICTIDW SYSTEM RESPONSE TIME 1.34 REACTOR PRUTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-enen312ation of the scram pi1 Lot valve solenoids. The response time may be measured by any seH ? sequential, overlapping or total steps such that the entire response time is measured. REFUELING FLOOR SECONDARY CONTAINMENT INTEGRTTY 1.35 REFUELING FLOOR SECONDARY CDHTAINMENT INTEGRITY shall exist when:
- a. All 1 refueling floor secondary containment penetrations required to be closed during accident conditions are either:
l 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, j or deactivated automatic valve secured in its closed position, except es provided in Table 3.6.5.2.2-1 of Specification 3.6.5.2.2.
LIMERICK - UNIT 1 1-6
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i s-1 1 TABLE 3.3.7.4-1 (Continued) RHR SERVICE WATER SYSTEM (Continued) HSS-12-094 Control-Transfer switch HSS-12-093 Control-Transfer switch y HV-51-1F014A Control-1A RHR heat exchanger tube side inlet ~ 0AP506 Control-RHR Service Water pump HV-51-1F068A Control-1A RHR Heat exchanger tube side outlet EMERGENCY SERVICE WATER SYSTEM OAP548 Control-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 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 152-11509/CSk' 101-D11 Safeguard SWGR feeder bkr. 152-11609/CSR 101-D12 Safeguard SWGR feeder bkr. 152-11709/CSR 101-D13 Safeguard SWGR feeder bkr. 152-11502/CSR 201-D11 Safeguard SWGR feeder bkr. 152-11602/CSR 201-012 Safeguard SWGR feeder bkr. 152-11702/CSR 201-D13 Safeguard SWGR feeder bkr. 152-11505/CSR D114 Safeguard LC XFMR breaker LIMERICK - UNIT 1- 3/4 3-81
- - - - - - - - - - - _ - - - --. - - - - - - . x- - - - -- ----_----.-_.---<--.-~------------_.___------_-a---------_a.__ - - -__-__--_-__-____--- --_ __,------_---_a_._----- - - . - - - - , - - -- A
TABLE 3.3.7.9-1 FIRE DFTECTION INSTRUMENTATION
' INSTRUMENT LOCATION TOTAL NUMBER'0F INSTRUMENT'
. FIRE-ZONE- STRUCTURE ELEV. AREA HEAT l
SM0KE FLAMER (x/y) (x/y) Ti/yJq
)
IL Control 200' Control Structure Chillers and NA 3/0 NA-Chilled Water Pump Area 258 IM Control 200' Control Structure Chillers and NA '3/0 NA Chilled Water Pump Area 263 2 Control 217' 13-kV Switchgear Area 336 NA 34/0 NA 3 Control 217' Battery Room 323(10)- 1/0 1/0 .NA 4 Control 217' Battery Room 324 (1C) 1/0 1/0 NA 7 Control 239' Corridor 437. NA 5/0 NA-8 Control 239' Battery Room 425 (1B1/182) 1/0 2/0 NA 9 Control 239' Battery Room 436 (1A1/1A2) 1/0 2/0 NA 12 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA 434 (D13) 13' Control 239' 4-kV Switchgear. Compartment 2/0 2/0 NA' 435 (D11) 14 Control- 239' 4-kV Switchgear Compartment 2/0 2/0 NA 432 (014) 15 Control 239' 4-kV Switchgear Compartment 2/0 2/0 NA. 433 (012) 20 Control 254' Static Inverter Room Unit 1 NA 4/0 NA' Area 452 22 Control 254' Cable Spreading Room Unit 1,- NA 14/0 NA Area 449 24A Control 269' Control Room 533 NA 23(a)/0 NA 11(b)/0 24B Control 269' Control Room Utility Room 529 NA 1/0 NA 24C Control 269' Control Room Office 531 NA 1/0 NA 24D Control 269' Control Room Shift Supt. 536 NA 1/0 NA l 24E Control 269' Control Room Shop 534 NA 1/0 NA (Photo-Elect) 24F Control 269' Control Room Instrument NA 1/0 NA Lab 535 (Photo-Elect) 24G Control 269' Control Room Shift Supt. NA 1/0 NA 532 l LIMERICK-UNIT 1 3/4 3-93 i _ _ _ _ _ - . _ _ - . - - - . _ . _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' - ~ ^ - ^ ^ -~ ~ ' ' - ^ ' ^ ^ ~ ~ ~ ~ ^ ~ ~ ~ ~ ~
l I-TABLE'3.3.7.9-1 (Continued) FIRE DETECTION INSTRUMENTATION
, INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENT FIRE
' ZONE STRUCTURE ELEV. AREA HEAT SMOKE FLAME (x/y) (x/yJ (x/y)
.48A' Unit 1 313' Laydown Areas 601 and 602; NA 8/0 NA.
Reactor Corridor and RERS Fan Area 605 51A. Unit 1 331' RERS Filter 2/0 NA NA Reactor Compartment 618 (inside plenum) 51B Unit 1 331' RERS Filter -2/0 NA NA Reactor Compartment 612 (inside plenum) 79 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1 80 Diesel- 217' Diesel-Generator 1/5 4/0- 1/0 Generator Cell Unit 1 81 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 , Generator Cell Unit 1 82 Diesel- 217' Diesel-Generator 1/5 4/0 1/0 Generator Cell Unit 1 122A Spray 260' ESW and RHRSW Pump Area NA 4/0 NA Pond Pump Structure 122E Spray 251' RHRSW Valve Compartment' NA 2/0 NA Pond Pump Structure 123A Spray 268' ESW and RHRSW Pump Area NA 4/0 NA Pond Pump Structure 123E Spray 251' RHRSW Valve Compartment NA 2/0 NA Pond Pump Structure 124A Diesel- 217' Diesel-Generator Access NA 4/0 NA Generator Corridor 313 126A Common 412' North Stack Instrument NA 2/0 NA Reactor Room 713 1
* (x/y): X is the number of Function A (Early Warning Fire Detection and Notification Only) Instruments.
Y is the number of Function B (Activation of Fire Suppression ~; System and Early Warning Notification) Instruments. l (a) These smoke detectors are located below the suspended ceiling in the Control Room. (b) These smoke detectors are located above the suspended ceiling in the Control Room. LIMERICK - UNIT 1 3/4 3-96
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I CONTAINMENT SYSTEMS l
' SURVEILLANCE REQUIREMENTS l
4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be:
- a. Verified closed at least once per 7 days.
- b. Demonstrated OPERABLE:
- 1. At least once per 31 days and within 2 hours after any discharge of steam to the suppression chamber from the safety / relief valves, by cycling each vacuum breaker through at least one complete cycle of full travel.
- 2. At least once per 31 days by verifying both position indicators OPERABLE by observing expected valve movement during the cycling test. .
- 3. At least once per 18 months by; a) Verifying each valve's opening setpoint, from the closed position, to be 0.5 psid i 5%, and b) Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.
c) Verifying that each outboard valves's position indicator is capable of detecting disk displacement 10.050", and each inboard valve's position indicator is capable of detecting l disk displacement 10.120". i I LIMERICK - UNIT 1 3/4 6-45 _ _ _ _ _ _ _ . - _ _ - - - _ - i
- - _ _ _ _ = _ _ _ _ . . - _
- l. .
FI S iiONS l HOSE RACK LOCATION ELEVATION IDENTIFICATION l 1.- Control
Enclosure:
Stairwell 350' 1HR-141 Stairwell, Outside SGTS Room 332' 1HR-140 Stairwell, Outside Fan Room 304' 1HR-103 l .Outside 13kV.Switchgear Room 217' - 1HR-116 Stairwell, Outside Aux Equip Rm 289' 1HR-130 Stairwell, Outside Cable Spreading Rm 254' 1HR-250 Wall .Outside 4kV Switchgear & Battery Rooms 239' 1HR-251 Corridor 448, South Side of 4kV Switchgear & Battery Rooms 239' 1HR-124 Wall, Corridor 265. 200' 1HR-120. Wall, Corridor 164 180' 1HR-121
- 2. Refueling Area:
-SW Corner Refuel Floor 352' 1HR-201 NW Corner Refuel Floor 352' 1HR-202 North Wall-Center 352' 1HR-203 South Wall-Center 352' 1HR-204
- 3. Reactor
Enclosure:
SW Corner Reactor Enclosure 331' 1HR-205 SW Corner Reactor Enclosure (RERS Fan Area) 313' 1HR-207 NW Corner Reactor Enclosure (Laydown Area 601) 313' 1HR-208 SE Corner Reactor Enclosure (Near Refuel Floor Exh. Fans) 313' 1HR-209 NE Corner Reactor Enclosure (Near D124 Load Center) 313' 1HR-210 SW Corner Reactor Enclosure (Corridor 506) 283' 1HR-215 i NW Corner Reactor Enclosure (Corridor 506) 283' 1HR-216 LIMERICK - UNIT 1 3/4 7-27
TABLE 3.7.6.5-1 (Continued) FIRE HOSE STATIONS-HOSE RACK LOCATION ELEVATION IDENTIFICATION
- 3. Reactor
Enclosure:
(Continued) SE Corner Reactor Enclosure (SLC Pumps Area 500) 283' 1HR-217 NE Corner Reactor Enclosure 283' 1HR-218 SW Corner Reactor Enclosure (Area 402A, Near CRD Repair) 253' 1HR-223
-NW Corner Reactor Enclosure (Near Drywell Equip Hatch) 253' 1HR-224 SE Corner Reactor Enclosure (Near Drywell Personnel Lock) 253' 1HR-225 East Wall Reactor Enclosure (Near TIP Machines) 253' 1HR-226 SW. Corner Reactor Enclosure (Near RCIC Equip Hatch) 217' 1HR-232 NW Corner Reactor Enclosure (Near Supp Pool Access Hatch) 217' 1HR-233 East Wall Reactor Enclosure (Near Equipment Airlock 300) 217' 1HR-234 NE Corner Reactor Enclosure (Near NCC D124-R-G) 217' 1HR-235 SW Corner Reactor Enclosure (Near MCC 0134-R-H) 201' 1HR-240 NW Corner Reactor Enclosure (Near MCC D134-R-H1) 201' 1HR-241 East Wall Reactor Enclosure I (Near RECW Heat Exchangers) 201' 1HR-242 !
NE Corner Reactor Enclosure (Near RECW Pumps) 201' 1HR-243 SW Corner Reactor Enclosure 177' IHR-252 NW Corner Reactor Enclosure 177' 1HR-253 NE Corner Reactor Enclosure 177' 1HR-142 LIMERICK - UNIT 1 3/4 7-28 i l
a L
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TABLE 3.8.4.1-1 (Continued) PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
-2. 480-V0LT MOLDED CASE BREAKERS (Continued)
CIRCUIT SYSTEMS OR BREAKER NO. LOCATION TYPES EQUIPMENT POWERED 52-22410 0144-R-E IM HFB100 1B2 Drywell Area Unit TM HFB100 Cooler 182V212 52-22411- D144-R-E IM HFB100 1H2 Drywell Area Unit TM HFB100 Cooler 1H2V212 52-22418 D144-R-E IM HFB50 HPCI Mn Stm Supply Inbrd TM HFB150 Isol Viv HV-55-1F002 52-22516 114B-R-C IM HFB25 1A Reac Recirc Pump TM HFB100 Suction Viv HV-43-1F023A 52-22518 1148-R-C IM HFB25 1A Reac Recirc Pump TM HFB100 Discharge Viv HV-43-1F931A 52-22520 1148-R-C IM HFB25 Reactor Bottom Head Drain Viv TM HFB40 HV-44-1F100 52-22536 114B-R-C IM HFB25 P.WCU Inlet from Rx Recirc loop TM HFB40 HV-44-1F105 52-22534 1148-R-C IM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-1F001 52-22535 114B-R-C TM HFB25 Reactor Vessel Head Vent TM HFB40 HV-41-1F005 52-22537 114B-R-C TM-HFB15 Disposal Cask Removal Cart TM HFB20 Hoist 10H236 52-22538 114B-R-C li4 HFB15 Control Rod Drive Platform TM HFB20 Hoist 10H229 52-22608 124B-R-C TM-HFB15 CRD Equipment Handling TM HFB20 Platform 10N22608 52-22618 124B-R-C IM HFB25 IB Reac. Recirc. Pump TM HFB100 Discharge Viv HV-43-1F031B
*52-22622 124B-R-C TM HFB125 Permanent Plant In-Containment Welding System 10NW201 LIMERICK - UNIT 1 3/4 8-25
TABLE 4.11.1.1.1-1 (Continued) > I TABLE NOTATIONS j aThe LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66sb LLD = E V
- 2.22 x 106 . y . exp (-A At)
Where: LLD is the a priori lower limit of detection as defined above (as microcuries per unit mass or volume), sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute). E is the counting efficiency, as counts per disintegration, V is the sample size, in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclides, and A t for the plant effluents is the elapsed time between the midpoint of sample collection and time of counting. 1 Typical values of E, V, Y, and at should be used in calculation. I 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. LIMERICK - UNIT 1 3/4 11-3 1
5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in figure 5.1.2-1. MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND EIOUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE B0UNDARY that are accessible to MEMBER OF THE PUBLIC, shall be shown in Figures 5.1.3-la and 5.1.3-1b. METEOROLOGICAL TOWER LOCATION 5.1.4 The meteorological towers shall be located as shown on Figure 5.1.4-1. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel lined reinforced concrete structure consisting of a drywell and suppression chamber. The drywell is a steel-lined reinforced concrete vessel in a shape of a truncated cone on top of a water filled suppression chamber and is separated by a diaphragm slab and connected to the suppression chamber through a series of downcomer vents. The drywell has a maximum free air volume of 243,580 cubic feet at a minimum suppression pool level of 22 feet. The suppression chamber has a maximum air region of 159.540 cubic feet and a minimum water region of 122,120 cubic feet. DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:
- a. Maximum internal pressure 55 psig.
- b. Maximum internal temperature: drywell 3dOOF.
suppression pool 2200F.
- c. Maximum external to internal differential pressure 5 psid.
- d. Maximum floor differential pressure: 30 psid, downward. q 20 psid, upward. j l
l J i LIMERICK - UNIT 1 5-1
1 ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 A retraining and rep'acement training program for the unit staff shall be maintained under the direction ef the site Training organization and shall meet or. l exceed the requirements of ANSI /ANS 3.1.-1978 and 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter'to all licensees, and'shall include familiarization with relevant industry operational experience. 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) FUNCTION 6.5.1.1 The PORC shall function to advise the Plant Manager on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The PORC shall be composed of the: Chairman: Superintendent-Operations Member: Superintendent-Technical , Member. Superintendent-Maintenance / Instruments. tion and Contivls Member: Superintendent-Plant Services Member: Assistant Superintendent-Operations Member: Regulatory Engineer Member: Technical Engineer Member: 5hift Superintendent Member: Maintenance Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC fMirman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities at any one time. MEETING FREQUENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as corwened by the PORC Chairman or his designated alternate. l QUORUM l 6.5.1.5 The quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these Technical Specifications shall
- consist of the Chairman or his designated alternate and four members including alternates.
i
.* ...i LIMERICK - UNIT 1 6-7 Amendment No. 10
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s MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE' : INITIAL CORE FUEL TYPES P8CIB094 FIGURE 3.2.1-4 9 .1 LIMERICK - UNIT 1 3/4 2-5 )
- p. .
l ' ~ ___x--_________-___-___-___-________ -_
TkBLE-3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS TRIP ALLOWABLE FUNCTIONAL UNITS SETPOINT VALUE
- a. Reactor Vessel Water Level - - .
Low Low, Level 2 > -38. inches *- > -45 inches t, b.- Reactor Vessel Water Level - -< 54 inches -< 60 inches High, Level-8
- c. . Condensate Storage Tank Level - > 135.8** inches
> 132.3 inches Low
- d. Manual Initiation N.A. -N.A.
i
*See Bases Figure B 3/4.3-1
** Corresponds to 2.3 fcet indicated.
LIMERICK - UNIT 1 3/4 3-55 _ _ - - _ - - - - - - - - - - - - - - - - - - - - - - _ - _ - - _ - - - - - - - - - . - - - - - - - - - - - - - - - - - - - _ - - - _ - _ _ - - _ - _ _ _ _ _ _ - - -}}