ML20238B090
| ML20238B090 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1279, NUDOCS 8709090526 | |
| Download: ML20238B090 (424) | |
Text
{{#Wiki_filter:NUREG-1279 O _ Technical Specifications ! Beaver Valley Power Station, ; Unit 2 Docket No. 50-412 l Appendix "A" to i License No. NPF-73 1 I O-Issued by the , U.S. Nuclear Regulatory l Commission ! Office of Nuclear Reactor Regulation August 1987 , 1 f* * *Ut, z Ee O S F %888R 8588811a P PDR c _ __ _-_-___-__- _ _ _- _ --_-___-__________-____ -_______ - . - _ _ _ - _ _ _ _ _ _ _ - - - _ - - - - - - . - - _ -
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NUREG-1279 ( Technical SpecFcations Beaver Valley Power Station' Unit 2 Docket No. 50 412 Appendix "A" to - License No. NPF-73
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lssued by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1987 l p" ~ m,,
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w INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS i 1.1 DEFINED TERMS......................... ............................. 1 1.2. THERMAL POWER. .................................... ................ 1-1 4 1 1.3 RATED THERMAL P0WER................................................. 1-1 1.4 OPERATIONAL M0DE....,............................................... 1-1 1.5 ACTI0N......................................................... .. . 1-1 1.6 OPERABLE - OPERABILITY... .......................................... 1-1
- 1. ~? REPORTABLE EVENT.................................................... 1-1
- 1. 8 CONTAINMENT INTEGRITY.............................. ................ 1-1 1.9 CHANNEL CALIBRATION............ .................................... 1-2 1.10 CHANNEL CHECK....................................................... 1-2 p 1.11 CHANNEL FUNCTIONAL TEST........................ .................... 1-2 1.12 CORE ALTERATION............ ........................................ 1-2 1.13 SHUTDOWN MARGIN...................................................... 1-2 1.14 IDENTIFIED LEAKAGE...................... ........................... 1-3 1.15 UNIDENTIFIED LEAKAGE........... . ........ ........... ............. 1-3 1.16 PRESSURE BOUNDARY LEAKAGE... ....................... ................ 1-3 1.17 CONTROLLED LEAKAGE............ .............. ................. .... 1-3 1.18 QUADRANT POWER TILT RATIO................ ......... . ............. 1-3 1.19 DOSE EQVIVALENT I-131. ......... ............. ...................... 1-3 I
1.20 STAGGERED TEST BASIS. .. ............ ..... . ........ ............ 1-3 ; I 1.21 FREQUENCY NOTATION... ........ ......... ........................... 1-4 1.22 REACTOR TRIP RESPONSE TIME.............................. . ........ 1-4 1.23 ENGINEERED SAFETY FEATURE RESPONSE TIME...... .. .. .............. 1-4 1.24 AXIAL. FLUX DIFFERENCE.. ............ ......... ............. ....... 1-4 1.25 PHY$1CS TEST..... ........... ................ ..... ........... ... 1-4 1 1.26 E-AVERAGE DISINTEGRATION ENERGY.................................... 1-4 1.27 SOURCE CHECK....... .................................... ............ 1-5 1.28 PROCESS CONTROL PROGRAM........ . ..... ............................ 1-5
% 1.29 SOLIDIFICATION............................... .... . . ............. 1-5 1.30 0FF-SITE DOSE CALCULATION MANUAL (0DCM)......... ................... 1 BEAVER VALLEY - UNIT 2 I
INDEX DEE1!!LUO11S SECTION PAGE 1.31 GASEOUS RADWASTE TREATMENT SYSTEM =.. . ..... ...... . . .. .. 1-5 1.32 VENTILATION EXHAUST TREATMENT SYSTEM... ................ ... . 1-5 1.33 PURGE - PURGING. .... .. .... .... .. . ........ ... .. . .. 1- 5 1.34 VENTING. ..... ... ... ........ ... .................. ... ..... 1-6 1.35 MAJOR CHANGES. . .. . .. ...... . ... ... .... ..... ...... 1-6 1.36 MEMBER (S) 0F THE PUBLIC... . . ... ...... . ........ ....... 1-6 1ABLE 1.1 OPERATIONAL MODES (TABLE '.1)... .. ..... .. .. ....... . 1-7 TABLE 1.2 FREQUENCY NOTATION.. . .. ..... ...... . .. ... .... 1-8
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e-I i O BEAVER VALLEY - UNIT 2 II
g INDEX 1 i 1
}AFETY LIMITS ARD LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE. ...... .......................... ................. 2-1 2.1.2 REACTCR COOLANT SYSTEM PRESSURE................. ................ 2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1- REACTOR TRIP SYSTEM INSTRUMENTATION.SETP0INTS.................... 12-3 l
BASES SECTION PAGE i l 2.1 SAFETY LIMITS O- 2.1.1 REACTOR CORE. ................................................... B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.................... ............. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS........ ........... B 2-2 ] 1 i
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i l BEAVER VALLEY - UNIT 2 III l
INDEX L1MITING CDhalTION FOR OPER6J10tLR1Q SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...........
.... . .... .. ........ .. 3/4 0-1 3/4.1 REACTIVITY 70NTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T avg > 200% . .. . . ... .. ... .. .. 3/4 1-1 Shutdown Margin - T avg 5 200 F... . ..... ..... .... . . 3/4 1-3 Boron Dilution...... . ..... ..... .......... ... . . .... 3/4 1-4 Moderator Temperature Coetiic >nt. . ... ... . ....... 3/4 1-5 Minimum Temperature for Criticality.. ..... . .. ...... .. 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Paths - Shutdown... .. . .. . . . ... .. ... .. .. 3/4 1-7 Flow Paths - Operating......... ........... . ...... .... . 3/4 1-8 Charging Pumn Shutdown.. .. .. .... . .... ........... . 3/4 1-10 Charging Pumps - Operating... . ... . .. ..... .. .... 3/4 1-11 Boric Acid Transfer Pumps - Shutdown. . . . . . . . . .... .... . 3/4 1-12 Boric Acid Transfer Pumps - Operating. ... . ................ 3/4 1-13 Borated Water Sources - Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-14 ,
i Borated Water Sources - Operating.......... . . .... ..... .. 3/4 1-15 { Isolation of Unborated Water Sources - Shutdown. . . . . . . . . . . 3/4 1- 17 j ( 3/4.1.3 MOVABLE CONTROL ASSEMBLIES { Group Height.. ... ... . . . ... . . . ....... . .. ... 3/4 1-18 Posit!on Indication Systems - Operating......... ... .... 3/4 1-21 Position Indication Systems - Shutdown.... .... .... ........ 3/4 1-22 Rod Drop Time. . .......... ... . ... . . ... ...... .. . 3/4 1-23 Shutdown Rod Insertion Limit.... .. .... .. .. .. ...... .... 3/4 1-24 Control Rod Insertion Limits.. .. ... . . ... .......... . 3/4 1-25 i
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3/4.2 POWER DISTRIBUTION LIMITS l 1 l ! 3/4.2.1 AXIAL FLUX DIFFERENCE... ....... .............. .. .. ..... 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR -q F (Z). . . . . . . . . . . . ....... 3/4 2-5 N 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - FAH....... .... ..... . 3/4 2-9 BEAVER VALLEY - UNIT 2 IV
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INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE I i 3/4.2.4 QUAD RANT POWER TI LT RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-11 3/4.2.5 DNB PARALETERS......... ........................... ........ 3/4 2-13 l 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION. ....................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.. 3/4 3-14 l l 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring........ .................. ............ 3/4 3-39 i Movable Incore Detectors.................................... 3/4 3-45 j Seismic Instrumentation..................................... 3/4 3-46 ! Meteorological Instrumentation.. ........................... 3/4 3-49 Remote Shutdown instrumentation............................. 3/4 3-52' Chl ori ne Detec ti on Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-56 i
\s- Accident Monitoring Instrumentation......................... 3/4 3-57 l Radioactive Liquid Effluent Monitoring Instrumentation...... 3/4 3-60 Radioactive Gaseous Effluent Monitoring Instrumentation. . . . . 3/4 3-65 l
1 3/4.3.4 TURBINE OVERSPEED PROTECTION................................ 3/4 3-74 3/4.4 REACTOR COOLANT SYSTEM 3/M.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Normal Operation.... ...... ............................. 3/4 4-1
. Hot Standby ... ... ...... ..... ..... .................... 3/4 4-2 Shutdown ..... ..... . .... ............................ 3/4 4-3 Reactor Coolant Pump-Startup ....... .............. . ..... 3/4 4-7 3/4.4.2 SAFETY VALVES - SHUTDOWN. .......... ...................... 3/4 4-8 3/4.4.3 SAFETY VALVES - OPERATING........... ...................... 3/4 4-9 3/4.4.4 PRESSURIZER..... ................. .... ................... 3/4 4-10 3/4.4.5 STraM GENERATORS... .......... ................. .......... 3/4 4-11 1 ,~
l (s BEAVER VALLEY - UNIT 2 V
INDEX LIMITINGl0HDlTION FOR OPERATION AND_.511RVflLLANCE RE0t!IREBINTS SECTION PAGE 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems. ............ ... ..... .... ... 3/4 4-17 Ope rati onal Lea kage. . . . . . . . . . . . . . . . . . .... ......... ... 3/4 4-19 Pressure Isoletion Valves.... ........... .. ............ 3/4 4-21 3/4.4.7 CHEMISTRY... .. ....... ... . ........ ............ ... . 3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY.. . ...... ............. .. ........ .. 3/4 4-27 3/4.4.9 PRES $URE/ TEMPERATURE LIMITS Reactor Coolant System... . .......... ....... .......... 3/4 4-30 Pressurizer. .. ...... ... ...... .. .............. ..... S/4 4-34 Overpressure Protection Systems........ ..... ..... ....... 3/4 4-35 3/4.4.10 STRUCTURAL INTEGRITY ASME cme Class 1, 2 and 3 Components. . . . . . . . . . . . . . . . . . 3/4 4-38 3/A.4.11 RELIEF VALVES .. ...................... ........ .......... 3/4 4-39 3/4.4.12 REACTOR COOLANT SYSTEM HEAD VENTS., .. ........ . ... ..... 3/4 4-40 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS............................. ...... . . . 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T avg
> 350 F............................. 3/4 5-3 t
3/4.5.3 ECCS SUBSYSTEMS - T avg < 350 F.... ............... . . .. 3/4 5-6 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.. .. ......... ............ ......... 3/4 6-1 i l Containment Leakage........................ . .... .. . ... 3/4 6-2 i Containment Air Locks... .... .. ............. .......... .. 3/4 6-4 Int.ernal Pressure..................... . . ............... 3/4 6-6 l Air Temperature........... .......... .... .... ... . .... 3/4 6-8 I Containment Gtructural Integrity.. ...... .............. .. 3/4 6-9 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System......... .................. 3/4 6-10 O BEAVER VALLEY - UNIT 2 VI
y i -( t INDEX LIMITING CfhalIION FOR OPERATION AND SURVEILLANCE RE0VIREMENTS , a SECTION PAGE { l Containment Recirculation Spray System..................... 3/4 6-12 i Chemical Addition System..... ............................. 3/4 6-14 ; 3/4.6.3 CONTAINMENT ISOLATION VALVES............................... 3/4 6-15 l 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers......................................... 3/4 6-31 Electric Hydrogen Recombiners.............................. 3/4 6-32 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector....... .......... .................. 3/4 6-34 3/4.7 PLANT SYSTEMS 3/4 7.1 TURBINE CYCLE Safety Valves. ..................... ..................... 3/4 7-1 Auxiliary Feedwater System............................ ... 3/4 7-4 i' Primary Plar.t Demineralized Water (PPDW). . . . . . . . . . . . . . . . . . 3/4 7-6 Activity............. ......................... ......... 3/4.7-7 ; Main Steam Line Isolation Valves.......................... 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSORE/ TEMPERATURE LIMITATION........... 3/4 7-10 3/4.7.3 PRIMARY COMP 0NENT COOLING WATER SYSTEM.................... 3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM (SWS)............ ................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK - OHIO RIVER. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-13 3/4.7.6 FLOOD PROTECTION.. .................................... .. 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS............... 3/4 7-15 , 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)... 3/4 7-18 3/4.7.9 SEALED SOURCE CONTAMINATION...... ...... ....... .. ... 3/4 7-20 3/4.7.12 SNUBBERS........ ............. .................. . ..... 3/4 7-24 3/4.7.13 STANDBY SERVICE WATER SYSTEM (SWE)..... . ...... ........ 3/4 7-28 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES O Operating.... .... ....................... .. ............. 3/4 8-1
\s_ l Shutdown............ ............ ..... .......... ......... 3/4 8-6 BEAVER VALLEY - UNIT 2 VII
litill1N(LCONDITION FOR OPERATION AND $11RVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating..... ............ . ....... 3/4 8-7 A.C. Distribution - Shutdown........ ............... .. ... 3/4 8-8 D. C. Di stribution - Operati ng. . . . . . . . . . . . . . ............. 3/4 8-9 D.C. Distribution - Shutdown....... ............. . .. .. 3/4 8-12 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORCN CONCENTRATION...... . ... . ..... ..... ... ..... 3/4 9-1 3/4.9.2 INSTRUMENTATION... ... ..... ...................... ..... 3/4 9-2 3/4.9.3 DECAY TIME.. . .. ...... .... .. ... ...... .. .. .... 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.......... ........... . 3/4 9-4 3/4.9.5 COMMUNICATIONS.. ............... ... . ................. . 3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY................ .......... .. 3/4 9-6 3/4.9.7' CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING......... .. 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.............. 3/4 9-8 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM..... . ... . 3/4 9-10 3/4.9.10 WATER LEVEL-REACTOR VESSEL................................. 3/4 9-11 3/4.9.11 STORAGE POOL WATER LEVEL.... .......... ................... 3/4 9-12 3/4.9.12 FUEL E'JILDING VENTILATION SYSTEM - FUEL MOVEMENT.... ............... ..... ............. . .. . 3/4 9-13 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE....... .......... ............................... 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS I 3/4.10.1 SHUTDOWN MARGIN.... .. ...................... ............ 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.... . 3/4 10-2 3/4.10.3 PHYSICS TESTS.. .......... .. ............................ 3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS. ............. ....... .. ...... ... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM-SHUTDOWN., ... .. ... .... ... 3/4 10-5 O BEAVER VALLEY - UNIT 2 VIII
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INDEX LIMITING _ CONDITION FOR OPERATION AND SURVEILLANCE REOUIREBENTS __ SECTION PAGE
.? /4.12 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......... ................................ 3/4 11-1 l
Concentration. . ........................................ 3/4 11-1 J Dose...................................................... 3/4 11-6 Li qui d Was te T re atmect. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-7 J Li q u i d Hol dup Ta n ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11- 8 3/4.11.2 GASE0US EFFLUENTS......................................... 3/4 11-9 Dose Rate........................ ..... .................. 3/4 11-9 a s Dose - Noble Gases.............................. .......... 3/4 11-13 ! Dose cadioiodines, Radioactivity Material in Particulate l Form, and Radionuclides Other Than Noble Gases. . . . . . 3/4 11-14 Gaseous Radwaste Treatment................................. 3/4 11-15 Gaseous Waste Storage Tanks.. ............................. 3/4 11-16 Explosive Gas Mixture....... ............. ................ 3/4 11-17 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................ ... 3/4 11-18 3/4.11.4 TOTAL D0SE.......................................... ..... 3/4 11-19 l l 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING t 3/4.12.1 MONITORING PR0 CRAM................... ......... ... ....... '3/4 12-1 i 3/4.12.2 LAND USE CENSUS............................................ 3/4 12-9 l 3/4.12.3 INTERLABORATORY COMPARIS0N PROGRAM ......... .......... . 3/4 12-10 B__&SES ___ I SECTION PAGE 3/4.0 APPLICABILITY........ ........................................ B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL........... . ............................ B 3/4 1-1 .;
, s,.) 3/4.1.2 B0 RATION SYSTEMS... ..... ................................ B 3/4 1-2 l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................. B 3/4 1-4 BEAVER VALLEY - UNIT 2 IX
INDEX BASES SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE............ ........... .......... . B 3/4 2-1 3/4.2.2 AND 3/4.2.3 HEAT FLU'. AND NUCLEAR ENTHALPY HOT CHANNELFACTORSF(Z)andFfH...... q
.................. . B 3M 2-2 3/4.2.4 QUARDRANT POWER TILT RATIO....... .............. ..... ... B 3/4 2-4 3/4.2.5 DNB PARAMETERS. ....... .... . ... ......... .. ... ....... B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION... ........ ..... .... B 3/4 3-1 3/4.3.2 HGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION. ... .... ... ..................... ..... . B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION., ......................... ... B 3/4 3-4 Radiation Monitoring... .......... ..... .... .......... . B 3/4 3-4 Movable Incore Detectors... ..... ....................... B 3/4 3-4 Sei smi c Instrumentation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3- 5 Meteorological Instrumentation... .......... . ............ B 3/4 3-5 Remote Shutdown Instrumentation............... .......... . B 3/4 3-5 Chlorine Detection Systems................................. B 3/4 3-5 Accident Monitoring Instrumentation..................... .. B 3/4 ; 6 Radioactive Liquid Effluent Monitoring Instru-mentation.. .............. . . .... ...................... B 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instru- I mentation............... ..................... ..... ...... B 3/4 3-6 I
3/4.3.4 lVRBINE OVERSPEED PROTECTION.. . .... ....... ......... . . B 3/4 3-6 ) l J 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RE ACTOR COOLANT LOOPS AND COOLANT CIRCU'.1 TION. . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 AND 3/4.4.3 SAFETY VALVES.............. ..... ........... .. B 3/4 4-2 3/4.4.4 PRESSURIZER...................... ..... ........... ...... B 3/4 4-2 3/4.4.5 STEAM GENERATORS.... . .... .... ... ............ ........ B 3/4 4-2 BEAVER VALLEY - UNIT 2 X
l l. i L 1 I INDEX BASES SECTION PAGE l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............ ................ B 3/4 4-4 3/4.4~7. CHEMISTRY......... ........................................ B 3/4 4-5 3/4'.4.8 SPECIFIC ACTIVITY........ ................................. B 3/4 4-5 4 3/4.0.9 PREuSURE/ TEMPERATURE LIMITS............... ................ B 3/4 4-6 , 3/4.4.10 STRUCTURAL INTEGRITY....................................... B 3/4 4-15 3/4., il RELIEF VALVES............................................. B 3/4 4-16 3/4.4.12 REACTOR COOLANT SYSTEM HEAD VENTS.......................... B 3/4 4-16 I ( 1 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ) 3/4.5.1 ACCUMULATORS.......... .................................... B 3/4 5-1 3/4.5.2 AND 3/4.5.3 ECCS SUBSYSTEMS.................................. B 3/4 5-1 3/4.5.5 (MOVED TO BASES SECTION 3/4.1.2)
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3/4.6 CONTAINMENT SYSTEMS l 3/4.6.1 PRTMARY CONTAINMENT........................................ B 3/4 6,1-l 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.......................'B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES............. ................. B 3/4 6-2 3/4.6.4 COMBUSTIBLE GAS CONTROL.......... ........................ B 3/4 6-3 3/4.6.5 SUBATMOSPHECIC PRESSURE CONTROL SYSTEM..................... B 3/4 6-3 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE. .... ............................ ........ B 3/4 7-1 3/4.7.2 $ TEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.... ....... B 3/4 7-3 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM..................... B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM.................................. .... B 3/4 7-3 3/4.7.6 ULTIMATE HEAT SINK..... ............................... .. B 3/4 7-3 3/4.7.6 FLOOD PROTECTION.. .. . .... ............................. 8 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEM.... .......... . B 3/4 7 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS).....B 3/4 7-4
; 3/4 7.9 SEALE0 SOURCE CONTAMINATION................................ B 3/4 7-5 \w- 3/4.7.12 SNUBBERS.......... ........................................ B 3/4 7-5 BEAVER VALLEY - UNIT 2 .XI
INDEX BASES SECTION PAGE I 3/4.7.13 STANDBY SERVICE WATER SYSTEM (SWE). . .. . .. . . B 3/4 7-6 3/4.8 ELECTRICAL POWER SYSTEMS l 3/4.8.1 A.C. SOURCES........... ........ ... .... .. . . ..... . B 3/4 8-1 f l 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS. . ...... .. .......... B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION... .. . ... .. .. . ............ ... B 3/4 9-1 3/4.9.2 INSTRUMENTATION............... ... ..... . ........ .... .. B 3/4 9-1 3/4.9.3 DECAY TIME... ........ .. ..... ..... ................ .. B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS... .. .. . ... .. . B 3/4 9-1 3/4.9.5 COMMUNICATIONS. .. . . ..... ... ...... ..... ... B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE OPERABILITY. .. .. ..... ...... . .. B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING......... .. ... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION...... .. . . B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM...... .... . B 3/4 9-2 3/4.9.10 AND 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND STORAGE POOL., ... .. ......... .. ... . .......... ...... . B 3/4 ^-3 3/4.9.12 AND 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM.. ..... ... . B 3/4 ' 3/4.10 SPECIAL TEST EXCEPTIONS i 3/4.10.1 SHUTDOWN MARGIN. . ..... . ... . . .... ..... ...... ... . B 3/4 10-1 1 1 i 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS. . .......... .......................... ......... B 3/4 10-1 3/4.10.3 PHYSICS TESTS. .... . ......... .. . .... ......... . ... B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS.. . .... ..... .. ......... .... . . B 3/4 10-1 3/4.10.5 POSITION IWDICAfl0N SYSTEM-SHUTDOWN. . ... ... .. .. .. B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS...... .. .. ......... . . ....... .... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS.. . ... . . .... . .......... ....... B 3/4 11-2 BEAVER VALLEY - UNIT 2 XII
t. INDEX-BASES SECTION PAGE 3/4.11.3 SOLID RADI0 ACTIVE WASTE.................................. B 3/4 11-5 3/4.11.4 TOTAL DOSE, ........................... ................ B 3/4 11-5
'3/4.12 RADIOLOGICAL ENV A0NMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM........ ................. .. ........ B 3/4 12-1 3/4.12.2 LAND USE CENSUS.......................................... B 3/4 12-1 j 3/4.12.3 INTERLABORATORY COMPARIS0N PROGRAM....................... B 3/4 12-1 q i
O i 9 BEAVER VALLEY . UNIT'2 XIII l
INDEX RESIGNF%T.EES 5.1 SITE Site Boundary for Gaseous Effluents... .. .. . . ....... ..... 5-1 Site Boundary for Liquio Ei fluents. . ... . . . ... . . .... .. 5-1 Exclusion Area. . . . . . . . ..... ............. ..... . .... ..... 5-1 Low Population Zone. ... .. ... . .. .. ................ .. 5-1
)
Flood Control.. ..... . . .. ....... .. ....... ........ .. 5-1 l l i I 5.2 CONTAINMENT Configuration. . ..... . ..... ........ . ..... .... .... .. 5-1 Design Pressure and Temperature. ............... ..... . ....... 5-6 Penetrations. ...... ........ ...... ..... ....... ....... 5-6 5.3 REACTOR CORE Fuel Assemblies. . ......... .. .... . ..... ....... .......... 5-6 Control Rod Assemblies... . ....................... ........ 5-6 5.4 REACTOR COOLANT SYSTEM l Design Pressure and Temperature..... . .......... ...... . .... 5-6 l Volume. . .. ...... .... ..... .... ... ... ....... ... ..... 5-6 5.5 EMERGENCY CORE COOLING SYSTEM........... ..... ... ......... .... 5-7 I l -5.6 FUEL STORAGE Criticality... . ..... .... ........... .... .. ....... .. .. .. 5-7 Drainage.. ................... ....... ... ..... . .............. 5-7
)
Capacity. ........ ..... ............. . ..... .. ............. 5-7 5.7 SE'SMIC CLASSIFICATION..... .. ........ ............ ......... .. 5-7 l 2 5.8 METEOROLOGICAL TOWER L0 CATION.................................... 5-7 i O BEAVER VALLEY - UNIT 2 XIV l
INDEX AQ111FISTRATLVLf0NIROLS 6.1 RESPONSIBILITY....... ........ ...... . ........ ..... .... ........ 6-1 6.2 ORGANIZATION Offsite.. ....... .... . .... ..... ..... .. . ........ .... . 6-1 Facility Staff. .. ......... .. . . .. .. ........ ........ . 6-1 6.3 FACILITY STAFF QUALIFICATIONS. ...... ...... . .......... . .... 6-6 6.4 TRAINING... . . . .. ....... ....... .. .. ... .... ....... ...... 6-6 6.5 REVIEW AND AUDIT 6.5.1 ONSITE SAFETY COMMITTEE (OSC) Function.. ..
.. .. .. ........ .. . .. ..... ........ . 6-6 Compo,ition... ....
6-6 Alternates.. 9 . Meeting Frequency.. ... ... 6-7 6-7 Quorum.... . .. .. . . ..
.......... . .. . .. ..... . 6-7 Responsibilities.... ........ ......
..... ... ........... 6- 7 Authority.... ..... . ...... ... ... ...... .. ......... .. 6-8 Records.......... .... ..... .. ... .. .. ...... ... ...... .. 6-8 6.5.2 0FFSITE REVIEW COMMITTEE (ORC)
Function. . . 6-8 Composition. .... . . ... . ... ... .... .. . . 6-9 Alternates. . . . . . . . .. .. . ... . .. . . ..... 6-9 Consultants. ... . . . . .. . .. . ...... ....... .. . 6-9 Meeting Frequency.. . . . . .. .
...... .. . . .... . . 6-9 Quorum. . . ..... ... .... . .. . ........ . . . 6-9 Review. ... .. . ... ..... . .. . . ... . . . . . . .. 6-9 Audits.... . .... . ...... . ....... ..... .. . . .... .. .. 6-10 Authority. . . . ... . .... . .. .. .. ...... .. .... 6-11 Records. .. .... .... . . ...... . . .. .. .... . .. 6-11 O
BEAVER VALLEY - UNIT 2 XV
INDEX ADMINISTRATIVE CONIRO15 6.6 REPORTABLE EVENT ACTION.... .. . ...... .......... . ........ . 6-11 6.7 SAFETY LIMIT VIOLATION. ........... ... .... ........... . . .. 6-12 6.8 PROCEDURES.. . ... ... ......... . . .. ..... . ...... ......... 6-12 6.9 REPORTING REQUIREMENTS.. .... .. . .. ........ ... ... ........ 6-13 6.9.1 ROUTINE REPORTS. . ..... . ... ........ ......... .... . 6-13 Startup Reports... . ... .. .... ....... ................ 6-14 l Annual Reports. ..... . ...... ... ............ ......... 6-14 Monthly Operating Report. ..... ..... ... ..... ...... . 6-15 Annual Radiological Environmental Report. . ..... .. .. .. 6-15 Semi-Annual Radioactive Effluent Release Report.. . .... .. 6-17 Radial Peaking Factor Report. . . . . . . ..... .. ... . .. ...... 6-18 I 6.9.2 SPECIAL REPORTS... ... ............ ..... ...... ............ 6-18 O 6.10 RECORD RETENTION.. ..... ..... ..... ...... . ......... ..... . 6-19
)
6.11 RADIATION PROTECTION PROGRAM.... .. ... ...... ... ....... . 6-20 i i 1 6.12 HIGH RADIATION AREA. .......... .. . .. .............. .. .. 6-20 1 i 6.13 (DELETED) i 6.14 PROCESS CONTROL PROGRAM (PCP). . ........ . ................ 6-22 l i 6.15 0FFSITE DOSE CALCULATION MANUAL.(ODCM).. .. . ....... ........ 6-22 1 i O BEAVER VALLEY - UNIT 2 XVI i
l O INDEX-ADMINISTRATIVE CONTROLS ,, 6.16 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS............ 6-22 6.17 RADIOLOGICAL ENVIRONMENTAL MONITORING PR0 GRAM................... 6-24 O O BEAVER VALLEY - UNIT 2 XVII
1.0 DEFINITIONS - ( DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. RATED THERMAL POWER l 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2652 MWt. OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1. l ACTION
- 1. 5 ACTION shall be those additional requirements specified as corollary statements to each principal specification and shall be part of the specifications.
OPERABLE - OPERABILITY l 1.6 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instru-mentation, controls, normal and emergency electric power sources, cooling or-seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related safety function (s). REPORTABLE EVENT
- 1. 7 A REPORTABLE EVENT shall be any of those conditions specified in Sec-tion 50.73 to 10 CFR Part 50.
CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when: 1.8.1 All penetrations required to be closed during accident conditions are either:
- a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or BEAVER VALLEY - UNIT 2 1-1
DEEJNITIONS s CONTAINMENT INTEGRITY (ContiNed)
- b. Closed by manual valves, blind flanges, or deactivated auto- i matic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.
1.8.2 All equipment hatches are closed and sealed. 1.8.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3. , and 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2. 1.8.5 The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0 rings) is OPERABLE. l C;ANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel l output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated. CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indi-cations and/or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST j i i l 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions. CORE ALTERATION i 1.12 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the j vessel. Suspension of CORE ALTERATIONS shall not preclude ccmpletion of move- I ment of a component to a safe conservative position. SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster asser.bly of highest reactivity worth which is assumed to be fully withdrawn. BEAVEP, VALLEY - UNIT 2 1-2
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l l l-l l DfflNITIONS IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the container.nt atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the secondary system.
UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. PRESSURE B0UNDARY LEAKAGE 1.16 PRESSURE B0UNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. i CONTROLLED LEAKAGE 1 1.17 CONTRDLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. l QUADRANT POWER TILT RATIO 1.18 QUADRAT 4T POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one (1) excore detector inoperable, the remaining three (3) detectors shall be used for computing the average. DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (pCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I '31, I-132, I-133, 1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation, shall be those listed in Regulatory Guide 1.109, 1977 or TID 14844. STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of: BEAVER VALLEY - UNIT 2 1-3
l DEEIRIIl0NS STAGGERED TEST BASIS (continued)
- a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals;
- b. The testing of one (1) system, subsystem, train or other designated component at the beginning of each subinterval.
FREQUENCY NOTATION 1.21 The FREOUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2. REACTOR TRIP SYSTEM RESPONSE TIME 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall ne the time interval from when the monitored parameter exceeds its trip setpaint at the channel sensor until loss of stationary gripper coil voltage.
. ENGINEERED SAFETY FEATURE RESPONSE TIME 1.2S The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that +,ime interval from when the monitored parameter exceeds its ESF actuation setpcint at the channel sensor until the FSF equipment is capable of performing its sr.fety function (i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in nori.alized flux signals between the top and bottom halves of a two-section excore neutron detector. PHYSn'S TESTS 1.25 PHYSICS TESTS chall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and
- 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.
E - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average sum (weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling) of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. O BEAVER VALLEY - UNIT 2 1-4
j I gyrganas SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. PROCESS CONTROL PROGRAM 1.28 A PROCESS CONTROL PROGRAM (PCP) shall be the manual or set of operating parameters detailing the program of sampling, anclysis, and evaluation by which SOLIDIFICATION of wet radioactive wastes is assured. Requirements of the PCP are provided in Specification 6.14. '
-{ <
SOLIDIFICATION 1,29 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements. OFFSITE DOSE CALCULATION MANUAL (00CM) 1.30 An 0FFSITE DOSE CALCULATION MANUAL (ODCM) shall be a' manual containing the methodology and parameters to be used in the calculation of offsite~ doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. Requirenats e of the ODCM are provided in Specification 6.15. GASEOUS RADWASTE TREATMENT SYSlEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effsuents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. VENTILATION EXHAUST TREATMENT SYSTEM 1.32 VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gasecus radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorber3 .' and/or HEPA filters for the purpose of removing iodines or particulate from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety ! Feature (ESf) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions, in such a manner that replacement air or gas is required to purify the confinement. D BEAVER VALLEY - GNIT 2 1-5
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DUINIT10115 VENTING 1.34 VENTING is the controlled process of discharging air or gas from a con-finement to maintain temperature, pressure, humidity, concentre. tion or other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. f MAJOR CHANGES 1.35 MAJOR CHANGES to radioactive waste systems, as addressed in Para-graph 6.16.2, (liquid, gaseous and solid) shall include the following:
- 1) Major changes in process equipment, components, structures, and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evaluation Report (SER)
(e.g., deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems);
- 2) Major changes in the design of radwaste treatment systems (liquid, gaseous, and solid) that could significantly increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement);
- 3) Changes in system design which may invalidate the accident analysis as described in the SER (e.g. , changes in tank capacity that would alter the curies released); and
- 4) Changes in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g. , use of temporary equipment without adequate shielding provisions).
EMBER (S)0FTHEPUBLIC l 1.36 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally l associated with the plant. This category does not include employees of the utility, its contractors, or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries and persons who traverse portions of the site as the consequence of a public highway, railway, or waterway located wi? 5in the confines of the site bounc ary. This category does include persons who use porLions of the site far re:reational, occupational, or other purposes not associated with the plant. O BEAVER VALLEY - UNIT 2 1-6
r i TABLE 1.1 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER
- TEMPERATURE eff 1
- 1. POWER OPERATION >0.99 >5% >350 F
{
- 2. STARTUP >0.99 15% >350 F
- 3. HOT STANDBY <0.99 0 >350 F
- 4. HOT SHUTDOWN <0.99 0 350 F >T avg
>200 F
- 5. COLD SHUTDOWN <0.99 0 <200 F 1
- 6. REFUELING ** 10 95 0 1140 F 1
l 1 O i
- Excluding decay heat.
** Reactor vessel head unbolted or removed and fuel in the vessel.
BEAV.1 VALLEY - UNIT 2 1-7
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I TABLE 1.2 FREQUENCY NOTATION NOTATION 'REQUENCY S At least once per 12 hours. l 1 0 At least once per 24 hours. ! W At least once per 7 days. M At least once per 31 days. ; Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. P Completed prior to each release. N.A. Not applicable. O 4 1 i I O\ BEAVER VALLEY - UNIT 2 1-8
; f0 SAFET_Y LIMllS AND LIMITING SAFETY SYSTEM SETTIMS _
'2.1 SAFETY. LIMITS REACTOR. CORE I
2.1.'1 The combination of THERMAL POWER, pressurizer pressure, and the highest L operating . loop coolant temperature (Tavg) shall not exceed the limits shown in l Figure 2.1-1 for 3-loop operation. APPLICABILITY: MODES 1 and 2. ACTION:'
.Whenever the' point defined by the combination of the highest. operating loop' average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2,.3, 4, and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour. MODES 3, 4, and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, i
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I 4 I (8 BEAVER VALLEY - UNIT 2 2-1
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FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION BEAVEFi' VALLEY - UNIT 2 2-2
b L l p 2.2 LIMITING SAFETY SYSTEM SETTINGS J A[N REACTOR TRIP SYSTEM INSTP.UMENTAlION SETPOINTS 2.2.1 The Reactor Trip Systen Instrumentation and Interlock Setpoints shall be consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION:
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less con-servative than the value shown in the Trip Setpoint column but more conser-vative than the value shown in the Allowable Value column of Table 2.2-1 adjust the Setpoint consistent with the Trip Setpoint value.
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 was ;
satisfied for the affected channel or
- 2. Declare the channel inoperable and apply the applicable ACTION state- )
ment requirement of Specification 3.3.1 until the channel is restored j to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value. EQUATION 2.2-1 Z + R + S < TA where: Z = The value for column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor . error, or the value of Column S (Sensor Error) of Table 2.2-1 for ) the affected channel, and 1 TA = The value from Column TA (Total Allowance in % of span) of Table 2.2-1 for the affected channel. O BEAVER VALLEY - UNIT 2 2-3
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A S A T Y T S ( P E I C R N T A W R O . . . . __ O L A A. A A. A A T L . . . . . . C A N N N N N N A E R k o c rt o e l bt _ r o mu _ e n n nt ap _ t e o o o h n n g6 r r rt CI I n - t t t u ( _ d n t m e R aP N u e N u e up er NI i e s3 l 1 a s ex ( u-y t u e e e pP p S al g g g0 m i iF n8 n9 n1 I , _ r p d a - a- a- e T i en RP RP RP er T rs mo rr r r , r , nu I c Tk , i s N i c et ex ex ex) b s) U t a ro ol t u ne wu ol wu ol wu7 ol - r e7 ur - L mc t r I N PF PF PFP TPP A oi ce N t g at O uo en . . . . . I AL RI a b c d e T C N . . U 1 2 F 2 2 m 5pm7 c2H m 5 Nnc j
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n l T a l i i A r i P 2 T a t o t R e t K R s u u n u I n r t e R - y e s o s a g T b p t s t t t y m n a n R K T o a T n n O s a A o a T { A c t n t i t C g s; e s d ; t s A d ns p n e F c n E e a o m o t n o R A r l c3 o c a / u c u - c c ; 7 f 5 s a d a e= m g e i 8 1 e
) e e m d 2 0 e m i 2 a i n . . h i 3 M l T 1 L T I 1 0 T T 3
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= = = = = = = = = =
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p u f e t o g r n a % a t 7 r s 1
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w o al A M R E l a h s p p R H h l t E T s l 0 e g H a h n T l t n h s
= t i a r D t i 6 f u E o o t 1 o d T t p n A t i b
, s e R s e o r r s i S d p t o o n t n t t t o n 0 p a e i a c p e 9 i S e s c r ; .
s S n t e r
- T R E
p R t e e r e i E T N p d p q t T W r W S I m t A O T O m n e P P O o e r d e T i P c n h L A L T ; t m a A T E g ) t u t A S a R o r b M e M l E b t q , , R h R d P W s y % E t E I g O ; d n d l 3 H H R v P ) n i n e 3 T , T ) T a e a a v - % T L r d i D 9 D d p e tt s E + E e N) A u q c T u Od d M s o r d T n I e e R s t u e e A s A r E e s e p e R d R i T u n a r s c e t An u H r n Ti s T p e e e c x t e t i o Nt ; a e m h r e a c a g e D g ; w w x C E n n e b e e e ( Mo v m E n 1 t r q u u UC a T i - e o , bl 1 R( T e A t s b 0 o l T h R a d c - a 9 a 2 SN d t ; r , e e = v v NO e t g e r c s e t ~ r n i p o n a ) h ; q s s 2 II tR t tt l ; u i s o t e b I E MA F s g p a r A E f i q i L ET a d v S r e d ( fW o B TO e e a , C e f e y oO f e f A SN , m z T e R p f t f P e o d o T Y e i r o i c s d u S r n l l u l d e , eL u % t % a s a m % vA t 2 i 5 u o i s n r d e l 9 l M i 5 n 7 F t t n + aR n g I a r u i e i o e s g 2 a 1 R r o m r m f t hE e t t o P o s a e d H a m T n N n c b n mT m y y p a N r a a o b e b R m s a ( ( i e h O e n t e r d o tD t d t e s F z g t n t % tE h d T C p n ; i i i 3 oT t e e e m o 2 2 r s e s 3 bA c t c A u p c e n R t u a u E g o c . 8 - a d h d R a c 6 0 s a h i d e r e 7 0 s 5 l t a n nf h e t e g m 5 0 e 3 p g e ao t r r r 2 a f e t v a i . L o h w pt t y n y A L T < 0 P 2 t t on n l e l n i e t e e l c l o w b c c a r a i er r c e c
= t ; bh e e i p i
= = = = = = = p t t c s q t p a h a n r c m u e - nn h m f b ii c o a o m: ( a t e t a at q RR e u u h a EE a r a i
s ch t ro WW OO r o e F o e 6 n F PP F b b 1 ) oh I ic l + ( a rs u T 6 3 I 1 'T K P 'P S y a
)
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= 0 s S a g 6 y e v T t r a
+
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) T e s r d
i p o t n u c e N) a m e t u Od S t o p a g n I e 6 n c m r a i T u ' e ; o ; e l t An m T g c R p - n Ti + u A a E m e o Nt r l g W e t C E n 1 t d - a O t a ( Mo s e d P r O l UC n r a T e 1 R( I u e A e L g e
- T s l h A a h 2 SN d a d t M r t NO b l e n e R e 2 II 7 o m i r n E v y T 5 1 f u i H a b E MA i n d s T L ET n o e a d g d B TO a z e e D n e A SN 1 M r i m z E i t T Y ( o l i T s a S D t i n l A a r 5 T a t o i R e e P K R s u t r n I n r u t c e R - y e s o a n g T b p t t t i R
4 m n a n T n K T o a s a A r o O T { A c t n t o i s e s d f t C d g n p n e c A e a o m o t ; F n E r l c o c a 1 u R u - c c 8 / f
$ s d e e i 7 2 a a m g m d 0 0 e
) e e i a i n . h 3
3 M L T L T I 1 0. T 1
= = = = = = = = =
y f T T A A R E W S b b O : 1 2 p 3 7 P e T T I T S 1 R r 1 7 E e + + ,
+ g T +
V T h y 3 T 4 5 O A W 1 1 T 1 1 A K K 1 O 3 E T O N M3 9 <RE
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= o o r f m p 6
1 e y m r b a
, u e r t t t o a n s t r i a e o r s p p o n m t f S
T e e e . N p t S e d I m r n O o n p u a P c o i s T i r s , E g t T e % S a r 0 1 r d p . P b e 1 I g ; i t r R v l u o s ) T T a P C a p m f i d ( o n w e N) $ c a o u Od d p n I e e T R l i T u r E d s s f t A n u r W n t r n Ti s o O a i f o e o Nt ; a f P d t C E n 9 e ; a ( Mo V m 0 L 1 e % UC " A - e 3 w 1 R( T e M s c 7 d T h R x . e 2 SN d t 6 E ; , e 0 e
. NO e K H ) r f 2 I I r n T F o t d r
T u i d ; t o n E MA s n F D 2 a n a o L ET a d a E . r f B TO e e T 6 e l % , m z , A 7 p l 2 A SN e o a 7 % T Y i "T R 5 0 S n l r 5 h . o i > u t m s 1 . P t t a , r 1 I r u T a n o t s R o r go f n i s T t t r e vi s . i i a n o p at n I o e R s a f m T a a A p r w O n t e t r t u o T e s F T d n t l e t l C p n e e l S a f A m o / e t m e a r E o c 2 g a u c p e m R c 1 a c r a r i p a e 0 r i t i o r m e g m 0 e d s p f T e t a i . v n n a t s L T 0 A I i L 0 m r u r
. m o o i f f x
= = = = = = a . r r
=
mna o r o r s p r r
' s e e l
eT r r na n o s o s . S af n nn 6 ) h o e ea I c s s p t s 1 A % e
+ ( e6 e 6 6 " 2 h . h hf 1 1 K T T S f T2 T T o 4 5 6 E E E T T T O O O N N N E E 9 <E E<'_z $
p I 2.1 SAFETY LIMITS v BASES I 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission pro-ducts to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could re-i sult in excessive cladding temperatures because of the onset of departure from ) nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coef-ficient. DNB is not a directly measurable parameter during operation and there-fore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 R-Grid correlation. The W-3 R-Grid DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uri-f orm and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal opera-tional transients, and anticipated transients is limited to 1.30, ihis value q corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curve of Figure 2.1-1 shows the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. The curves are based on an enthalpy hot channel factor, F H, f 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F H at reduced power based on the expression: F N = 1.55 [1 + 0.3 (1-P)] AH where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(ol) function of the Overtemperature AT trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT trip will reduce the setpoint to provide protection consistent with c core safety limits. k i
)
BEAVER VALLEY - UNIT 2 B 2-1
{ SAFETY LIMITS L me 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radio-nuclides contained in the reactor coolant from reaching the containment atmosphere. i The reactor pressure vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plants l which permits a maximum transient pressure of 110% (2735 psig) of design pres- 1 sure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation. 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation j and design basis anticipated operational occurrences and to assist the Engineered . Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for coibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allow-able Values for the reactor trip setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative thaa the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its trip setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertain-ties in calibrating the instrumentation. In Equation 2.2-1, Z + R + $ < TA, the interactive effects of the errors in the rack and the sensor, and tiie "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, Detween the trip setpoint and the value used in the analysis for reactor trip. R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified trip setpoint. S or Sensor Drift is either the "as measured" deviation of the sensor from its calibration point or the value O BEAVER VALLEY - UNIT 2 B 2-2 i __-______A
O LIMITING SAFETY SYSTEM SETTINGS MSES specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. i The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the
, trip setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. Power Range, Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core. l l protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low setpoint provides redund-ant protection in the power range for a power excursion beginning from low power. l The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 10 percent of RATED THERMAL POWER). Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above 1.30 for control rod drop accidents. At high power a multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power . by action of the automatic rod control system, could cause an unconservative l local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which
/ DNBRs will be greater than 1.30.
BEAVER VALLEY - UNIT 2 B 2-3 L_-_____-_-_-_-_.
i 2.2 LIMITING SAFETY SYSTEM SETTINGS MSES Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condi-tion. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels. The Source Range Channels will initiate a reactor trip at about 10+5 counts per second unless manually blocked when P-6 becomes active. The intermediate range channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. Although no explicit credit was taken for operation of the Source Range Channels in the accident analyses, operability requirements in the Technical Specifications will ensure that the Source Range Channels are available to mitigate the consequences of an inadvertent control bank withdrawai in MODES 3, 4 and 5. Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power dis-tribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low pressure reactor trips. This set-point includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown on Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1. Overpower AT The Overpower AT reactor trip provides assurance of fuel integrity, e.g., f no melting, under all possible overpower conditions, limits the required range for Overtemperature AT protection, and provides a backup to the riigh Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. The Overpower AT trip provides J protection to mitigate the consequences of various size steam line breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Release." O\ BEAVER VALLEY - UNIT 2 B 2-4 !
F LIMITING SAFETY SYSTEM SETTINGS
- t ,
BASES Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pres- l sure range in which reactor opcration is permitted. The High Pressure trip is ' backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig) The Low Pressure trip protects against low pressure which could lead to DNB ts tripping the reactor in the event of a loss of reactor coolant pressure. On decreasing power the Low Pressure trip is automatically blocked by P-7 q (a power level of approximately 10% of RATED THERMAL POWER or turbine impulse chamber pressure at approximately 10% of full power equivalent); and an increas-ing power, automatically reinstated by P.7. On-decreasing power, the low setpoint trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with turbine impulse' chamber pressure at approximately 10 percent of full power equivalent); and on increasing power, automatically reinstated by P-7. Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor O Coolant System overpressurization by limiting the water level to a volume suf-ficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. On-decreasing power, the pressurizer high water level trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10 percent of full power equivalent); and on increasing power, automatically reinstated by P-7. No credit was taken for operation of this trip in the acci-dent analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System. Loss of Flow l i The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps. Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90 percent of nominal full loop flow. Above 30 percent (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90 percent of nominal full loop flow. Steam Generator Water Level , 4 The Steam Generator Water Level Low-Low trip provides core protection by O preventing operation with the steam generator water level below the minimum d volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system. BEAVER VALLEY - UNIT 2 B 2-5
LIMITING SAFETY SYSTEM SETTINGS MSES Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Pro-tection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by > 1.55 x 106 lbs/ hour in any loop. The Steam Generator low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is idnimized. Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or under-frequency to more than one reactor coolant pump. The specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached. Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfre-quency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is reached shall not exceed 0.6 seconds. On decreasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10 percent of full power equivalent); and on increasing power, reinstated automatically by P-7. Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9. Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System. BEAVER VALLEY - UNil 2 B 2-6
LIMITING SAFETY SYSTEM SETTINGS a BASIS Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor pratective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels which initiate a safety injection signal are shown in. Table 3.3-3. Reactor Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB resulting from the opening of two or more pump breakers above P-7. These t:ips are blocked below P-7. The open/close position trips assure a reactor tr y signal is generated before the low flow trip setpoint is reached. No credit was taken in the accident analyses for operation of these trips. .Their functional capability at the open/close position settings is. required to enhance the overall reliability of' the Reactor Protection System. Reactor Trip System Interlocks The Reactor Trip System interlocks perform the followin0 functions: P-6. Above the setpoint P-6 allows the manual block of the Source Range reactor trip and de-energizing of the high voltage to the detectors. Below the setpoint source range level trips are automatically reactivated and high voltage restored. P-7 Above the setpoint P-7 automatically enables reactor trips on low flow or coolant pump breaker open in more than one primary coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. Below the setpoint the above listed trips are automatically blocked. P-8 Above the setpoint P-8 automatically enables reactor trip on low flow in one or more primary coolant loops. Below the setpoint P-8 auto-matically blocks the above listed trip. P-9 Above the setpoint P-9 automatically enables a reactor trip on turbine trip. Below the setpoint P-9 automatically blocks a reactor trip on turbine trip. P-10 Above the setpoint P-10 allows the manual block of the Intermediate Range reactor trip and the low setpoint Power Range reactor trip; and automatically blocks the Source Range reactor trip and de-energizes the Source Range high voltage power. Below the setpoint the Intermediate Range reactor trip are automatically reactivated. Provides input to P-7. ( P-13 Provides input to P-7. BEAVER VALLEY - UNIT 2 B 2-7
1 4 I 1 1 1 SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l
ft 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l 3/4.0 APPLICABILITY LIMITINGC301 TION FOR OEERATIO'N 3.0.I' Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. ' 3.0.2. Noncompliance with a specification shall exist when the requirements of the Limiting. Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is. restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When'a limiting Condition for Operation is not met except as'provided in' the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
- 1. At least HOT STANDBY.within the next 6 hours,
- 2. At least H0T SHUTDOWN within the following 6 hours, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours.
- Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements requirements. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.
)
3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable limiting Con-dition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied within 2 hours, action shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply, by placing it, as applicable, in:
- 1. At least HOT STANDBY within the next 6 hours,
- 2. At least HOT SHUTDOWN within the following 6 hours, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours.
This specification is not applicable in MODES 5 or 6. BEAVER VALLEY - UNIT 2 3/4 0-1
APPLICABILITY SURVEILLANCE RE0VIREMENTS 4.0.1 Surveillance Requirements shall be met durfng the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveillance interval, and
- b. The combined time interval for any 3 consecutive surveillance irtervals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condi-tion for Operation have been performed within the stated surveillance interval or as otherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Ccdc Class 1, 2 and 3 components and inservice testing af ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applic C.. Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted Ly the Commission pursuant to 10 CFR 50, Sec-tion 50.55a(g)(6)(i).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as l follows in these Technical Specifications:
l 1 0' BEAVER VALLEY - UNIT 2 3/4 0-2 1
f APPLICABILITY l: - i SEVEILLANCE_8E@lREMENTS i ASHE Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per'92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days.
- c. The provision of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection ;
and testing activities.
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
/' e. Nothing in tne ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
l j 1 O BEAVER VALLEY - UNIT 2 3/4 0-3
f 3/4.1 flEACTIVITY CONTROL SYSTEMS U 3/4.1.1,_,BORATION CONTROL SHUTDOWN MARGIN - T >200 F avo l I LIMIT _ING. CONDITION FOR OPERATION 1 3.1.1.1 The SHUTDOWN MARGIN shall be >1.77% ak/k. APPLICABILITY: MODES 1, 2*, 3, and 4.
- ACTION:
With the SHUTDOWN MARGIN <1.77% ak/k, immediately initiate and continue boration at >30 gpm of > 7000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIR is restored. MRYELLjgE RE0VIREMENTS _ 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be >1.77% ak/k:
- a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If. the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).
- b. When in MODES 1 or 2,# at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specifica-tion 3.1.3.6.
- c. When in MODE 2,## at least once during control rod withdrawal and at least once per hour thereafter until the reactor is critical.
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.
- e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:
*See Special Test Exception 3.10.1
#With Keff>1.0
##With Keff(1.0 BEAVER VALLEY - UNil 2 3/4 1-1
REliCTIVITY CONTROL SYSTEMS SURVEILLANCE RE0VlREMENTS (Continued)
- 1. Reactor Coolant System boron concentration,
- 2. Control rod position,
- 3. Reactor Coolant System average temperature,
- 4. Fuel burnup based on gross thermal energy generation, j
- 5. Xenon concentration, and
- 6. Samarium concentration,
- f. The Reactnr Coolant System shall be borated to at least the COLD SHUTDOWN boron concentration prior to manually blocking the Low Pressurizer Pressure Safety injection Signal and shall remain at this boron concentration or greater at all times during which this sgnal is blocked.
4 1.1.1.2 The overall core reacLivity balance shall be compared to predicted values to demonstrate agreement within 1% Ak/k at least once per 31 Ef fective Full Power Days (EFPO). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel barnup of 60 Effective Full Power Days after each fuel ',oading. a i I i i O BEAVER VALLEY UNIT 2 3/4 1-2
p REACTIVITY CONTROL-SYSTEMS kI SHUTDOWN MARGIN - T ava
< 200 F l'
. LIMITING CONDITIQN FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be 1 1.0% ak/k.
APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN < 1.0% ak/k, immediately initiate and continue boratioa at > 30 gpm of > 7000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined te be 1 1.0% ak/k-
- a. Within 1 hour after detection of an inoperable control rod (s) 7_
and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTOOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).
- b. At least once per 24 hours by consideration of the following factors:
- 1. Reactor Coolant System boron concentration,
- 2. Control rod position,
- 3. Reactor Coolant System average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
O BEAVER VALLEY - UNIT 2 3/4 1-3
REACTIV11Y . CONTROL SYSTEMS BORON DlLUTION LIMITING _LDEDll10N F0fLQEERATION 3.1.1.3 The flow rate of reactor coolant through the core shall be 1 3000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made. APPLICABILITY: All MODES. ACTION: With the flow rate of reactor coolant through the core < 3000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System. SURVE1LLMICE RE0_VIREMENT.S 4.1.1.3 The flow rate of reactor coolant through the core shall be determined to be > 3000 gpm prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:
- a. Verifying at least one reactor coolant pump is in operation, or
- b. Veritying that at least one RHR pump is in operation and supplying 1 3000 gpm through the core.
I O\ l BEAVER VALLEY - UNIT 2 3/4 1-4 ___________a
.p REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT (MTC)
LIMITING CONDI]I0H l0R OPERATION 3.1.1.4 The Moderator Temperature Coefficient (MTC) shall be:
- a. Less positive than 0 x 10 4 Ak/k/ F,
- b. Less negative than -5.0 x 10 4 ok/k/ F at RATED THERMAL POWER.
APPLICABILITY *# MODES 1 and 2 ACTION: With the Moderator Temperature Coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours. 51!RVflLLANCE RE0VIREMENTS. 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits. 4.1.1.4.2 The MTC shall be determined at the following frequencies'and THERMAL POWER conditions during each fuel cycle:
- a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
- b. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.
.l i
I
*With K > 1. 0. l
{ eff
#See Special Test Exception 3.10.3.
I BEAVER VALLEY - UNIT 2 3/4 1-5 , t
REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITIEGl0ND1Il0N FQJLOPERAUSN . _ _ , 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (Tavg) < shall be 1541 F when the reactor is critical. APPLICABILITY: MODES 1 and 2. ACTION: With a Reactor Coolant System operating loop temperature (Tavg) < 541 F, restore (T avg) to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLAt{CE RE0VIREMENJ1 4.1.1.5 The Reactor Ccalant System temperature (Tavg) shall be determined to be > 541 F.
- a. Within 15 minutes prior to achieving reactor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T avg is less than 551 F with the (Tavg) deviation alarm not reset.
*See Special Test Exception 3.10.3. #With K eH 1 1.0.
BEAVER VALLEY - UNIT 2 3/4 1-6
j a REACTIVITY CONTROL SYSTEMS
\V) 3/4.1.2 B0 RATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION m 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:
- a. A flow path from the boric acid storage system via a boric acid transfer pump to a charging pump to the Reactor Coolant System if only the boric acid storage tank is OPERABLE as given in Specifi-cation 3.1.2.7.a for MODES 5 and 6 or as given in Specifica-tion 3.1.2.8.a for MODE 4; or
- b. The flow path from the refueling water storage tank via a charging pump or a low head safety injection pump (with an open RCS vent of greater than or equal to 3.14 square inches) to the Reactor Coolant System if the refueling water storage tank is OPERABLE as given in Specification 3.1.2.7.b for MODES 5 and 6 or as given in Specifica-cion 3.1.2.8.b for MODE 4.
(~ APPLICABILITY: MODES 4, 5 and 6
\
ACTION With none of the above flow paths OPERABLE, suspen: i operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.
.5118YULLANCE RE0VIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
- 2. Verifying that the temperature of the heat traced portion of the flow path is > 65 F when a flow path from the boric acid tanks is used and the ambient air temperature of the Auxiliary .
Building is < 65 F.
- b. At least once per 31 days by verifying that each valve (manual, O power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
BEAVER VALLEY - UNIT 2 3/4 1-7
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LiliLIING_C0j@lTION FOR APERATION 3.1.2.2 Each of the following baron injection flow paths shall be OPERABLE: , a. The flow path from the boric acid tanks via a boric acid transfer pump and one charging pump to the Reactor Coolant System, and
- b. The flow path from the refueling water storage tank via one charging pump to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2 and 3*. ACTION:
- a. With the flow path from the boric acid tanks inoperable, restore the inoperable flow path to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivaleni, to at least 1% ok/k at 200 F within the next 6 hours; restore the flow path to OPERABLE status within the next ' days or be in HOT SHU100WN within the next 6 hours.
- b. With the flow path from the refueling water storage tank inoperable, restore the flow path to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SBElLULNCLREMIEfMNIL 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.
- 2. Verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is > 65 F when the ambient air temperature of the Auxiliary Building is < 65 F.
- b. At least once per 31 days by verifying that each valve (manual, i power operated or automatic) in the flow path that is not 1ccked, f sealed, or otherwise secured in position, is in its correct position. {
- (
*The provisions of Specifications 3.C.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to '
Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature cf one or more of the RCS cold legs exceeding 375 F, whichever comes first. 1 3 BEAVER VALLEY - UNIT 2 3/4 1-8 1 __-____-_ a
% REACTIVITY CONTROL SYSTEMS d
SURVEILlattCE_ REQUIREMENTS (Cnotinued)
- c. At least once per 18 months during shutdown by cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.
O 4
)
1 l 1 BEAVER VALLEY - UNIT 2 3/4 1-9
REACTIVITY CONTROL SYSTEMS CHARGING PUMP-SHUTDOWN LIMITING C0llDil10lLLOR_0PERAT10N 3.1.2.3 One charging pump in the boron injectiLn flow path required by Specification 3.1.2.1 or Low Head Safety injection Pump (with an open Reactor Coolant System vent of greater than or equal to 3.14 square inches) shall be OPERABLE and capable of being powered from an OPERABLE emergency bus. APPLICABILITY: MODES 4, 5 and 6 ACTION: With none of the above pumps OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump or Low Head Safety Injection pump is restored to OPERABLE status.
.SWNEILLMiCEJEQUISMENIL _
4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a differential pres-sure of > 2437 psid when tested pursuant to Specification 4.0.5. 4.1.2.3.2 All charging pumps, except the above required charging pump, shall be demonstrated inoperable
- by verifying that the control switches are placed in the PULL-TO-LOCK position and tagged within 4 hours af ter entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325 F, whichever comes first, and at least once per 12 hours thereafter.
4.1.2.3.3 When the Low Head Safety Injection pump is used in lieu of a charg-ing pump, the Low Head Safety Injection pump shall be demonstrated OPERABLE by: I a. Verification of an OPERABLE RWST pursuant to 4.1.2.7 and 4.. 2.8
- b. Verification of an OPERABLE Low Head Safety Injection Pump pursuant to Specification 4.5.2.b.2,
- c. Verification of an OPERABLE Low Head Safety Injection flow path from the RWST to the Reactor Coolant System once per shift, and
- d. Verification that the vent is open at least once per 12 hours.^^
l
*An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position. **Except when the vent path is provided with a valve which is locked or provided with remote position indication, or sealed, or otherwise secured in the open position, then verify these valves open at least once per 7 days.
BEAVER VALLEY - UNIT 2 3/4 1-10
REACTIVITY CONTROL SYSTEMS j l CHARGING PUMPS-0PERATING
)
i LIMITING CONDITION FOR OPERATION j
)
i 3.1.2.4 At least two charging pumps shall be OPERABLE I APPLICABILITY: MODES 1, 2 and 3*. ACTION: With only.one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a StiUTDOWN MARGIN equivalent to at least 1% Ak/k at 200F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours. 1 SURVEILLANCE RE0VIREMENTS l 4.1.2.4.1 At least-two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a differential j
-( pressure of > 2437 psid when tested pursuant to Specification 4.0.5. j 1
)
l
*The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry
)j into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or nare of the RCS cold legs exceeding 375 F, whichever comes first BEAVER VALLEY - UNIT 2 3/4 1-11
_ _ - _ _ _ _ - _ _ _ - __ _ -- - l
REACTIVITY CONTROL SYSTEMS ' BORIC ACID TRANSFER PUMPS - SHUTDOWN LIM 111[G_CONDillQ. NFOR OPJiR&UON ___ . 3.1.2.5 One boric acid transfer pump shall be O h RABLE and capable of being powered from an OPERABLE emergency bus if only the flow path thru the boric acid transfer pump of Specification 3.1.2.1.a, is OPERABLE. APPLICABILITY: MODES 4, 5 and 6. ACTION: With no boric acid transfer pump OPERABLE as required to complete the flow path of Specification 3.1.2.1.a. suspend all operations involving CORE ALTERA-TIONS or positive reactivity changes until at least one boric acid transfer pump is restored to OPERABLE status. eke 1EILLANCLREQUIRfEENTS 4.1.z.5 The above required boric acid transfer pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a dif-ferential pressure of > 102 psid when tested pursuant to Specification 4.0.5. l O BEAVER VALLEY - UNIT 2 3/4 1-12
L l l 1 1: \ l
/'] REACTIVITY CONTROL SYSTEMS
() BORIC ACID TRANSFER PUMPS - OPERATING h Llfilllh0EMITION FOR OPERATION 3.1.2.6 At least one boric acid transfer pump in the boron injection flow path required by Specification 3.1.2.2.a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump in Specification 3.1.2.2.a is OPERABLE. APPLICABILITY: MODES 1, 2 and 3. 1 j ACTION: With no boric acid transfer pump OPERABLE, restore at 10ast one boric acid transier pump to OPERABLE STATUS within 72 hours or be in at least HOT STANDBY j within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to 1% Ak/k ( at 200 F; restore at least one boric acid transfer pump to OPERABLE status ' within the next 7 days or be in HOT SHUTDOWN within the next 6 hours. SURVEILLANCE RE0VIREMENTS t' ( 4.1.2.6 The above required boric acid pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a differential pres-sure of > 102 psid when tested pursuant to Specification 4.0.5. 1 I l E L./ j BEAVER VALLEY - UNIT 2 3/4 1-13 I
l REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCES - SHUTDOWN LIMll1NG__C.0MQU.IDILE0lLOP1.fLAUDN 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE- !
- a. A boric acid storage system with:
- 1. A minimum contained volume of 2315 gallons,
- 2. Between 7000 and 7700 ppm of boron, and
- 3. A minimum solution temperature of 65 F.
- b. The refueling water storage tank with:
- 1. A minimum contained volume of 217,000 gallons,
- 2. A minimum boron concentration of 2000 ppm, and
- 3. A minimum solution temperature of 45 F.
APPLICABILITY: MODES 5 and 6. O ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one barated water source is restored to OPERABLE status. SufWEl.LLANCE_REqEEMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Verifying the buron concentration of the water,
- 2. Verifying the water level of the tank, and
- 3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside ambient air temperature is < 45 F BEAVER VALLEY - UNIT 2 3/4 1-14
l l-l f REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIE 111NG CONDITION 10R OPERATION 3.1.2.8 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2.
- a. A boric acid storage system with:
- 1. A minimum contained volume of 13,390 gallons,
- 2. Between 7000 and 7700 ppm of boron, and
- 3. A minimum solution temperature of 65 F.
- b. The refueling water storage tank with:
- 1. A minimum contained volume of 859,248 gallons,
- 2. A boron concentration between 200d and 2100 ppm, and 7 3. A solution temperature of > 45 F and < 50 F.
APPLICABILITY: MODES 1, 2, 3 & 4. ACTION: !
- a. With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1%
ak/k at 200 F within the next 6 hours; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours,
- b. With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. '
SURVEILLANCE 3EQlIREMENIS , 4.1.2.8 Each borated water source shall be demonstrated OPERABLL: b v BEAVER VALLEY - UNIT 2 3/4 1-15
REACTIVITY CONTROL SYSTEMS EURVE1LLMiCE3EDIREMENTS(Continued) ;
- a. At least once per 7 days by:
- 1. Verifying the boron concentration in each water source,
- 2. Verifying the water level in each water source, and
- 3. Verifying the boric acid storage system solution temperature.
- b. At least once per 24 hours by verifying the RWST temperature when the RWST ambient air temperature is > 50 F or < 45 F.
l 9 l O BEAVER VALLEY - UNIT 2 3/4 1-16
I i O REACTIVITY CONTROL SYSTEMS U l ISOLATION OF UNB0 RATED WATER SOURCES - SHUTDOWN I l LIMITING CONDITION FOR OPERATION 3.1.2.9 Provisions to limit flow capability frum unbarated water sources to the reactor coolant system to less than or equal to 85 gpm by means of a flow limiting orifice shall be OPERABLE. APPLICABILITY MODES 4 and 5. 1 ACTION l With the requirements of the above specification not satisfied immediately , suspend all operations involving positive reactivity changes and, if within 1 hour the required SHUTDOWN MARGIN is not verified, initiate and continue boration as specified in Specification 3.1.1.1 for MODE 4 and 3.1.1.2 for MODE 5 until the SHUTDOWN MARGIN is restored. I S M ELLLANCE RE0VIREMENTS 4.1.2.9 The provisions to limit flow capability to less than or equal to 85 gpm from unborated water sources to the reactor ccolant system shall be determined to be OPERABLE at lenst once per 31 days by verifying that valve 2CHS-37 is closed. l BEAVER VALLEY - UNIT 2 3/4 1-17
REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LJMITING_C0flDll10B FOR OPERATI0B 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within i 12 steps (indicated position, as determined in accordance ! with Specification 3.1.3.2) corresponding to their respective group demand l counter position. APPLICABILITY: MODES 1* and 2* ACTION:
- a. With one or more full length rods inoperable due to being immovable I as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT i
STANDBY within 6 hours.
- b. With more than one full length rod inoperable or misaligned from the group demand cuunter position by more than 1 12 steps (indicated position determined in accordance with Specification 3.1.3.2), be in HOT STANDBY within 6 hours.
- c. With one full length rod trippable but inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group demand counter position by more than i 12 steps (indicated position determined in accordance with Specification 3.1.3.2), POWER OPERATION may continue provided that within one hour either:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and in- 1 sertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or
- 3. Tte rod is declared inoperable and the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
a) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the hour and, within the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal to 85% of i'ATED THERMAL POWER.
*See Special Test Exceptions 3.10.2 and 3.10.3 BEAVER VALLEY - UNIT 2 3/4 1-18
. ____________-_ A
l-I q REACTIVITY CONTROL SYSTEMS I [V" LIMIIING_C0HDITIQH.FOR DEERATION (Continued) b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours, c) A power distribution map is obtained from the movable incore detectors and Fq(Z) and F H are verified to be within their limits within 72 hours, d) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions. SURVEILLANCE RE0VIREMENTS 4.1.3.1.1 Each shutdown and control rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one
,A direction at least once per 31 days. ;
i 4.1.3.1.2 Each full length rod position shall be determined to be within
.i 12 steps of the associated group demand counter by verifying the individual i rod position at least once per 12 hours except during intervals when the Rod Position Deviation monitor is inoperable, then verify the group position at l i
least once per 4 hours. ) l I l l l O G BEAVER VALLEY - UNIT 2 3/4 1-19
l TABLE 3.1-1 ; 1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF l AN INOPERABLE FULL LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment loss of Reactor Coolant From Small Ruptured Pipes Or From Cracked Large Pipes Which Actuates The Energency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) Major Secondary Systems Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) O O BEAVER VALLEY - UNIT 2 3/4 1-20
. (~] REACTIVITi rPMROL SYSTEMS N-POSITION INDICATION SYSTEMS - OPERATING LIMITING _ CONDITION FQB_0PERATION 3.1.3.2 The Digital Rod. Position Indication System and the Demand Position ; Indication System shall be OPERABLE and capable of determining the control rod I positions within i 12 steps. APPLICABILITY: MODES 1 and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable either:
- 1. Determine the position of the nonirdicating rod (s) int "ectly by the movable incore detectors at least once per 8 hc 3 and immediately after any motion of the nonindicating rod which 4 exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER
,A j , within 8 hours.
s b
- b. With a maximum of one demand position indicator per bank inoperable either:
1
- 1. Verify that all digital rod position indicators for the affected '
bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours, j
SURVEILLARCLIEQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation monitor is in-operable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours. b ( BEAVER VALLEY - UNIT 2 3/4 1-21
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTDOWN i LIMlIlRG CONDlI10N F0Jl OPERATION 3.1.3.3 One digital rod position indicator (excluding demand position indica-tion) shall be OPERABLE and capable of determining the control rod position within 12 steps for each shutdown or control rod not fully inserted. APPLICABILITY: MODES 3*# , 4^# and 5*# ACTION: With less than the above required group demand position indicator (s) OPERABLE, open the reactor trip system breakers. SURVllLLANCE REQUlREMEHIS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 18 months.
*With the reactor trip system breakers in the closed position. #See Special Test Exceptions Specification 3.10.5.
BEAVER VALLEY - UNIT 2 3/4 1-22
f: l a l ( i (3 1 REACTIVITY CONTROL SYSTEMS I J ROD DROP, TIME LIMITING CONDITION FOR OPERATION
.3.1.3.4 The individual full length (shutdown and control) rod drop time from 4 the fully withdrawn position shall be < 2.2 seconds from beginning of decay '
of stationary gripper coil voltage to Hashpot entry with
- a. T avg
> 541 F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODE 3. ACTION: i
- a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
i SURVEILLANCE RE0VIREMENTS { O 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality: 1
- a. For all rods following each removal of the reactor vessel head. I i
- b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could I affect the drop time of those specific rods, and I
- c. At least once per 18 months.
1 i n ) U BEAVER VALLEY - UNIT 2 3/4 1-23
REACTIVITY CONTROL SYSTEM SHUTDOWN R0D INSERTION LIMIT LIM 111tLCtl0HDITION FOR OEEWLTION , _ . . 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2*# ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveil-lance testing pursuant to Specification (4.1.3.1.1), within one hour either:
- a. Fully withdraw the rod, or
- b. Declare the rod to be inoperable and apply Specification (3.1.3.1).
SURVEILLANCE RE0VIREttENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:
- a. Within 15 minutes prior to withdrawal of any rods in control banks A.
B, C, or D during an approach to reactor criticality, and
- b. At least once per 24 hours thereafter.
f
*See Special Test Exception 3.10.2 and 3.10.3 #With Keff > 1.0 0
BEAVER VALLEY - UNIT 2 3/4 1-24
r' REACTIVITY CONTROL SYSTEMS (' CONTROL ROD INSERTION LIMITS LIMITING CORDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1. APPLICABILITY: MODES 1* and 2*# ACTION: With the control banks inserted beyond the above insartion limits, except for surveillance testing pursuant to Specification 4.1.3.1.1, either:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the above figure, or
- c. Be in at least H0T STANDBY within 6 hours, k SURVEILLANCE _REQUIREMERIS 4.1.3.6 The position of each control bank snall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours.
i l 1 l 4 l
*See Special Test Exception 3.10.2 and 3.10.3 O #with Keff > 1.0 BEAVER VALLEY - UNIT 2 3/4 1-25 i
O (Fully ~ ::thdrawn)
, i ) i i !
l ! 228 . . ._ , _ _ . ._. _____ l t [ l , BANK C i j l 200
\l g I
i i 2 l t 9 l l s M l O l # 0 150-g . .' W , j F . . _ _ . __ -- D 100 ! O l i
% [
e l - i ! o , 4 7 i sn
~~
- s t i i i t
0- , , l 0 .2 .4 .6 .8 1.0 (Fully inserted) FR ACTION OF RATED THERMAL POWER k l FIGURE 3.1- 1 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER THREE LOOP OPERATION 9 f BEAVER VALLEY - UNIT 2 3/4 1-26
l [] - 3/4.2 POWER DISTRIBUTION LIMITS ; ij-
' 1 AXIAL FLUX DIFFERENCE (AFD) i LIM 1IltfE C0BDITION 0 FOR OPERATION _
3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within a 17 percent target band (flux difference units) about the target flux difference. APPLICABILITY: . MODE 1 above 50 Percent RATED THERMAL POWER
- j ACTION:
- a. With the indicated AXIAL FLUX DIFFERENCE outside of the 17 percent target band about the target flux difference and with THERMAL POWER:
- 1. Above 90 percent of RATED THERMAL POWER, within 15 minutes:
a) Either restore the indicated AFD to within the target band j limits, or b) Reduce THERMAL POWER to less than 90 percent of RATED
. THERMAL POWER.
- 2. Between 50 percent and 90 percent of RATED THERMAL POWER: i a) POWER OPERATION may continue provided:
- 1) The indicated AFD has not been outside of the 7 percent target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and
- 2) The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within 30 minutes and reduce the Pcwer Range Neutron Flux-High Trip Setpoints to < 55 percent of RATED THERMAL
~
POWER within the next 4 hours.
]
I b) Surveillance testing of the Power Range Neutron Flux Chan-nels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.
\ '
*See Special Test Exception 3.10.2
( I BEAVER VALLEY - UNIT 2 3/4 2-1
POWER DISTRIBUTION LIMITS LIMLIIL4G C0t{QIJ10B FOR OPERATION (Continued) ACTION: (Continued)
- b. THERMAL POWER shall not be increased above 90 percent of RATED THERMAL POWER unless the indicated AFD is within the i 7 percent target band and ACTION a.2.a) 1), above has been satisfied.
- c. THERMAL POWER shall not be increased above 50 percent of RATED ,
THERMAL POWER unless the indicated AFD has not been outside of ) the 17 percent target band for more than 1 hour penalty deviation , cumulative during the previous 24 hours. SURVEILLANCE RE0VIREMENTS _ 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15 percent of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- 2. At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status,
- b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its 1 7 percent target band when at least 2 of 4 or 2 of 3 OPERABLE excore thannels are indicating the AFD to be outside the target band. POWER OPERATION outside of the i 7 percent target band shall be accumulated on a time basis of: I a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50 percent of RATED THERMAL POWER, and ,
- b. One-half-minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels below 50% of RATED THERMAL POWER.
O BEAVER VALLEY - UNIT 2 3/4 2-2
POWER DISTRIBUTION' LIMITS SURVEILLANCE RE0VIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of-Specification 4.0.4 are not applicable. 4.2.1.4 The target flux difference shall be updated at least once per 31 Effec-tive Full Power Days by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently measured value and 0 percent at the end of the cycle life. The provisions of Specifica-tion 4.0.4 are not applicable. b("'N i I l l i i BEAVER VALLEY - UNIT 2 3/4 2-3
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Fd!LDIFFERENCE (AI)%
\. ' ' i , FIGURE 3.2-1 AXIAL FLUX OIFFERENCE LIMI15 A5 A FOHCTION OF RATED <
i s THERMAL. POWER A w w > BEAVERVALLLY-UNITh 3/C 2-4 m . , .
.[ .p-
4 1 1 POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F (Z) n l LIBITING CONDITION FOR OPERATION' l 3.2.2 F q(Z) shail be limited by the following relationships: F0 (Z) 5 [2.32] [K(Z)] for P > 0.5 P Fq (Z) s [4.64] [K(Z)] for P $ 0.5 THERMAL POWER where P = RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: MODE 1 ACTION: With Fq (Z) exceeding its limit: '
- a. Reduce THERMAL POWER at least 1 percent for each 1 percent F (Z) exceeds the limit within 15 minutes and similarly reduce the Power 0 Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subse-quent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1 percent for each 1 percent Fn (Z) exceeds the limit. The Overpower AT Trip Setpoint reduction snall be performed with the reactor subcritical.
- b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER: THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.0
( C BEAVER VALLEY - UNIT 2 3/4 2-5
POWER DISTRIBUTION LIMITS JURVEILLA!1CEREQUIRMBIS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 limit by: F*Y shall be evaluated to determine0 if F (Z) is within its
- a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5 percent of RATED THERMAL POWER.
- b. Increasing the measured F xy component of the power distribution map by 3 percent to account for manufacturing tolerances and further increasing the value by 5 percent to account for measurement uncertainties,
- c. Comparing the F xy computed (F x ) obtained in b, above to: l
- 1. The F xy limits for RATED THERMAL POWER (F P) for the appropriate measured core planes given in e and f below, and
- 2. The relationship:
F'xy
=F RTP xy [140. 2(1-P)]
where F x is the limit for fractional THERMAL POWER P operation expressed as a function of F and P is the fraction of RATED THERMAL POWER at which F xy was measured,
- d. Remeasuring F according to the following schedule:
xy
- 1. When F is greater than the F P
limit for the appropriate x measuredcoreplanebutlessthantheFhrelationship, x additional power distribution maps shall be taken and P F compared to F and F x x a) Either within 24 hours after exceeding by 20 percent of RATED THERMAL POWER or greater, the THERMAL POWER at which F was last determined, or b) At least once per 31 EFPD, whichever occurs first. I BEAVER VALLEY - UNIT 2 3/4 2-6 i ( _ _ _ _ _ _ _ _ _
I i (3 POWER' DISTRIBUTION LIMITS u SUMEllMNCE RE0VIREMD1I5_1Cpatinuedi m
- 2. When the F x
is less than or equal to the F limit for the appropriate measured core plane, addition:1 power distribution l maps shall be taken and*Ye C compared to once per 31 EFPD. Y F*RP and F*Y at least
- e. The F limit for Rated Thermal Power (FRTP) shall be provided for x
all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.14.
- f. The F xy limits f e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:
- 1. Lower core region from 0 to 15 percent, inclusive.
(G
%J 3 2. Upper core region from 85 to 100 percent inclusive.
- 3. Grid plane regions of core height (i 2.88 inches) measured from grid centerline.
- 4. Core plane regions within i 2 percent of core height (1 2.88 inches) about the bank demand position of the bank "D" control rods.
- g. WithFf.exceedingF xy on F9 (Z) shall be
, the effects of F x x evaluated to determine ifqF (Z) is within its limit.
4.2.2.3 When Fq (Z) is measured pursuant to Specification 4.10.2.2, an overall measured nF (Z) shall be obtained from a power distribution map and increased by 3 percent to account for manufacturing tolerances and further increased by 5 percent to account for measurement uncertainty, i O BEAVER VALLEY - UblT 2 3/4 2-7
K(Z) - NORMALIZED Fq (Z) AS A FUNCTION OF CORE HEIGHT 3-LOOP BEAVER VALLEY - UNIT 2 1.2 6.0, 1.0 i ---- 10.8, .94 __ 1.0 m 0.8
~ ' 12.0, .647
\ .
I 0.4 0.2 - 0.0 0 2 4 6 8 to 12 CORE EIGHT (FEET FROM BOTTOM) FIGURE 3.2-2 BEAVER VALLEY'- UNIT 2 3/4 2-8
Ie (*'} POWER DISTRIBUTION LIMI S NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F H llMllING CONDITION FOR OPERATION l I N 3.2.3 F AH shall be limited by the following relationship: N F g i 1.55 [1 + 0.3 (1-P)] where P _ THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION: With F H exceeding its limit:
- a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to 1 55% of RATED THERMAL POWER within the next 4 hours.
- b. Demonstratethroughin-coremappingthatFfg is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 2 hours, and
- c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER, subsequent POWER OFERATION may proceed provided that F is demonstrated through in-core mapping to be H
within its limit at a nominal 50 percent of RATED THERMAL POWER prior to exceeding this THERMAL power, at a nominal 75 percent of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours after attaining 95 percent or greater RATED THERMAL POWER. I l
/-
[
\
BEAVER VALLEY - UNIT 2 3/4 2-9
POWER DISTRIBUTION LIMITS SEVEILLA!iCLREQldEEMEfils _ N 4.2.3.1 F g shall be determined to be within its limit by using movabla incore detectors to obtain a power distribution map: i i,
- a. Prior to operation above 75 percent of RATED THERMAL POWER e 'er each fuel loading, and I
- b. At least once per 31 Effective Full Power Days. I 4.2.3.2 The measured F f 4.2.3.1 above, shall be increated by 4% for measurement uncertainty.H O
e!, BEAVER VALLEY - UNIT 2 3/4 2-10
l l' ) r"N POWER DISTRIBUTION LIMITS ( ;
\ OVADRANT POWER TILT RATIO LIMITING CONDJIl0N _ FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY: MODE I above 50 Percent 0F RATED THERMAL POWER
- ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but i 1.09:
- 1. Within 2 hours:
a) Either reduce the QUADRANT POWER TILT RATIO to within its 1init, or b) Reduce THERMAL POWER at least 3 percent for each 1 percent of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High ' Trip Setpoints within the next 4 hours. l n 2. Verify that the QUADRANT POWER TILT RATIO is within its limit ( ) within 24 hours after exceeding the limit or reduce THERMAL
's / POWER to less than 50 percent of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours.
- 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50 percent of RATED THERMAL power may proceed provided j that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95% or j greater RATED THERMAL POWER. '
- b. With the QUADRANT POWER TILT RA110 uetermined to exceed 1.09 due to misalignment of either a shutdown or control rod:
q
- 1. Reduce THERMAL POWER at least 3 percent for each 1 percent of l indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.
- 2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER 1 to less than 50 percent-of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Set- j
'~
points to i 55 percent of PsATED THERMAL POWER within the next 1 4 hours.
*See Special Test Exception 3.10.2 BEAVER VALLEY - UNIT 2 3/4 2-11
POWER DISTRIBUTION LIMITS LIMlllitCt.IOND1IEM_f_0LOPERATION (Continurd)
- 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50 percent of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95%
or greater RATED THERMAL POWER. j l
- c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
- 1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to i 55% of RATED THERMAL POWER within the next 4 hours.
- 2. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50 percent of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified at 95% or greater RATED THERMAL POWER.
518VflitatiCE_REQUIJgMENIX 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is ,
OPERABLE. ]
- b. Calculating the ratio at least once per 12 hours during steady state operation when the alarm is inoparable.
- c. Using the movable detectors to determine the QUADRANT POWER TILT l
RATIO at least once per 12 hours when one Power Range Channel is inoperable and THERMAL POWER is > 75 percent of RATED THERMAL POWER. l l l 9 BEAVER VALLEY - UNIT 2 3/4 2-12
k
/"~' POWER DISTRIBUTION LIMITS k
' DNB PARAMETERS' LIMITING CONDU lQN FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on T6hle 3.2-1:
- a. Reactor Coolant System T avg
- b. Pressurizer Pressure
- c. Reactor Coolant System Total Flow Rate APPLICABILITY: MODE 1*
ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5 percent of RATED THERMAL POWER within the next 4 hours. SJAVflLLANCE REQUIREMENTS 4.2.5.1.1 Each of the parameters of Table 3.2-1 shall be verified to be indi-cating within their limits at least once per 12 hours. 4.2.5.1.2 The provisions of Specification 4.0.3 and 4.0.4 are not applicable for the reactor startups following the initial fueling for Reactor Coolant System total flow rate to allow a calorimetric flow measurement and the cali-bration of the Reactor Coolant System total flow cate indicators. 4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.
)
1 i
*The provisions of Specification 3.0.2 are not applicable for the reactor
( startup following the initial fueling for Reactor Coolant System total flow rate to allow a calorimetric flow measurement and the calibration of the I i Reactor Coolant System t7tal flow rate indicators. { BEAVER VALLEY - UNIT 2 3/4 2-13 l
TABLE 3.2-1 DNB PARAMETERS 3 Loops in PARAMETER Operation Reactor Coolant System T, < 580.2 F Pressurizer Pressure > 2220 psia
- Reactor Coolant System ~> 274,800 gpm**
Total flow Rate O I l
- Limit not applicable during either a THERMAL POWER ramp increase in excess of I 5 percent RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER. ,
** Includes a 3,5% flow measurement uncertainty. J BEAVER VALLEY - UNIT 2 3/4 2-14 ,
1
i i h
~
3/4.3 INSTRUMENTATION ! D 1 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMIII!1G CONDII10N FOR OPERATI0tL I 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1 EURVEILL/gCE RE0VIREMENTS 4.3.1.1.1 Each reactor trip si d em instrumentation channel and interlock and automatic trip logic shall be demonstrated OPERABLE by the performance of the R* actor Trip System Instrumentation Surveillance Requirements
- during the MODES and at the frequencies shown in Table 4.3-1.
4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE during l t the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock opera-
\ tian. The total interlock function shall be demonstrated OPERABLE at least once per 18. months during CHANNEL. CALIBRATION testing of each channel affected by interlock operation.
4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function j shall be demonstrated to be within its limit at least once per 18 months. Each ! test Uall include at least one logic t. rain such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of r ed" w. ' channels in a specific reactor trip function as shown in the " Total t of s .v els" column of Table 3.3-1. l l
)
"For the automatic trip logic, the surveillance requirements shall be the
\ application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the requira logic output including, as a minimum, a continuity check of output devices.
BEAVER kALLEY - UNIT 2 3/4 3-1
N O I 9 T C 2 5 7 7 7 7 A 1 22 2 2 3 4 4
- 4 E 3* 5 L
B *
,5
- d AS 3 2 2d 3 n CE
- n
- a ID ,5 , ,a ,5 2 2 ) ) 4 2 2 2 2 LO 2 3 y 1 , 2 d PM ,d ( * ( n , , , , ,
P n g , 1 14 2a3 1 1 1 1 A 1a I1 1 SE MLL UEB MNA INR NAE N IHP 2 2 2 2 2 2 2 O MCO 2 33 3 3 I T A T N E M U S R LP 1 3 3 T S N I M E EI NR NT A HO CT 1 22 2 2 1 1 1 2 2 2 2 9 E T L S B Y A S S T .L P OE I NN R N T LA AH R TC O O T TF 2 2 3 3 3 3 C O 2 4? 4 4 2 A E R n
, o x h x x x r u g u u u t l w l l u F o i l
F F F e n n L H N n o o - e n n n o i i e p o o o , r t t r r r re re e t Ta a u u i r tt tt tt g u Ar r s s ua ua n e e e s s T unt ep p e e T ein eR eR a N r r r Noi N N R ) rO O I po e e ,0 u P P N o ,t p ,v ,v e e1 n t p T p U t ei t g- p w ao Ao r) r _ c eet ei gt gt a nP u o ro o e7 e _ L a gS e na a t d eL rL z- 7 A e n S ni i t p e iP i N R ah as ag d R wr we r r O R gw R o R e e oa u me oe ue u I l i o P N m el t h ee pr sv s T a rHL r r r ceS S t r rh so s u e eh eh ex rB rh 8_ C wg u( eT eb e N n w wg t u eT v rA r U a o . . oi oi nl o . . v PH S a b O O P( P F M P ab PH IF
. . . . . . . 0 5 6 7 8 9 1 1 2 3 4
= sP? h5, " g Y"
N _ O I T C . A 7 7 7 7 7 7 7 E L B A C IS LE 2 2 2 PD PO , , , AM 1 1 1 1 1 1 1 g g n n i i l l t a t a e v wre oov w o nr nr e l e l SE ie i e l fhl fh MLL p p - - c- c UEB po po p p pt p pt MNA o o o o oao oa INR NAE oh p l co oh p l co l o o l dl sl dl s omo om N I HP /ao /ao / / n/i/n/i O MCO 2 2 el 2 el 2 1 a2 m2 a1 m 2 2 I T A g g T n n l
)
d N E i t i t e vt wn oi e M na na en l p u U S i r i r l e fh o n R LP e e d - co i T EI pp pp p pi ptl t n S NR oo p oos o oc oa N NT o o p o onh ome o I A l yo l oo l l itl sm C HO / no / wo / / oi/ia ( M CT 2 2 al 2tl 2 1 cw1 ms 2 2 E T 1 S
- Y l 3 S S e w
. t v o 3 P .E e l I ON l fh E R NN - - c s s L T A p p p p pt u u B LH o o o o oa b b A R AC o c o o om / /
T O T l l l l dl s 1 1 T OF / / / / n/i - - C TO 3 3 3 3 2 a2 m 3 3 A E R h t g n i p a H o ) l
- o 8 o )
l L p- o 9 e oP m C r v e o r al o P e l g L w e wee r t L o t ot v o c e n ol a l S e t a v r e i S we W F L c) e o Tb w a7 R b T t r ror e- - A I a - - d o eL e RP y ( N W n t w t t - cs U w wa ao ad a ee np p L r) e7 o) l 8 o l 7 rL e-wnW d a gv ao em) uu7 i r A N z - F - F - nw eo e r tb qP - T i P P P eh o l A e P O r f f GL F ct o( rt e I ue oe oe - /t a v f ne n T sv v v ml mar rs rav i ame ( C N U so eb rA so sb oA so sb oA ae ev t e esn ti e ep d m nu el o d ob noA b r u F P( L( L( SL SMG UP UC( T 1 2 3 4 5 6 7 8 1 1 1 1 1 1 1 1 M3E E sEC
- 2 :
h5w"3 d 4* Y'"
N O I 04 4 0 9 T 4 C 1 ,9 9 4 4 4 4 A 1 1 13 13 4 4 4 4 4 5 5 d d E n n L a a B A , , C *
- IS 4 4 LE 2 2 2 2 PD ,
PO , ,* , AM 1 1 13 13 2 1 1 1 1 g n i t ra SE er MLL k e ap UEB MNA eo INR r p NAE bro N IHP /eo 3 2 O MCO 2 1 pl 22 22 2 3 3 I T A - T
) N d E e M u U S n R LP i
t C ( n o T S N I M EI NR NT A HO CT 1 2 11 11 1 2 2 2 1 9 E T 1 S
- Y 3 S S
. L 3 P .E r I ON e E R NN k L T A a B LH e A R AC r T O T b T OF /
4 2 C TO 2 1 22 22 2 4 4 A E R r e 3 k a 1 e P r t B e 0 u s g6 8 9 1 , p p r c n- - - - e n m e i m aP P P P er I u k g e R su P a o t , , , , l s us n e L s ex x x x T o t r y t u eu eu eu pe I i n B p S al gl gl nF gl nF mr IP ap) iF nF N t i U c li 7 p r p d a a a e or- i T i en R n R n R n er L j oTP r rsmo o o o ne A n C T c Tk rr rr rr rr ib N IF ne i cet et et et bm O S rov r t rot u wu wu wu ra I T C U F N t yE em f o Sf ar oio ttb cia as( eo RP t R o c a e t A a m o u ol ne trIN ce at en . RI a oe PN b oe PN c d oe PN uh TC e 9 9 0 1 2 3 1 2 2 2 2 mb9 5PO . gg "
- T'o
TABLE 3.3-1 (Continued) TABLE NOTATION I 1
*With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal, i
(1) Trip function may be manually bypassed in this MODE above P-10.
]
(2) Trip function may be manually bypassed in this MODE above P-6. ACTION STATEMENTS l ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within ! 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 2 - With'the number of OPERABLE channels one less than the Total i Number of Channels and with the THERMAL POWEP level: ,
- a. Less than or equal to 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour and restore the inoperable channel to OPERABLE status within D 24 hours after increasing THERMAL POWER above 5% of RATED
( THERMAL POWER; otherwise reduce thermal power to less than 5% RATED THERMAL POWER within the following 6 hours,
- b. Above 5% of RATED THERMAL POWER, operation may continue provided all of the following conditions are satisfied: <
- 1. The inoperable channel is placed in the tripped condi-tion within 1 hour, j l
- 2. The Minimum Channels OPERABLE requirement is met; how-ever, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1.
- 3. Either, THERMAL POWER is restricted to <75% of RATED I THERMAL and the Power Range, Neutron Flux trip setpoint is reduced to <85% of RATED THERMAL POWER within 4 hours; or, tiie QUADRANT POWER TILT RATIO is monitored i at least once per 12 hours per Specification 4.2.4.C. !
ACTION 3 - I With the number of channels OPERABLE one less than required by l the Minimum Channels OPERABLE requirement and with t?.e THERMAL l POWER level: i
- a. Below P-6, restore the inoperable channel to OPERABLE status 1 I
prior to increasing THERMAL POWER above the P-6 setpoint. BEAVER VALLEY - UNIT 2 3/4 3-5 i
l TABLE 3.3-1 (Continued)
- b. Above P-6 but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.
- c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.
ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL F0WER level:
- a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above P-6 setpoint and suspend positive reactivity operations.
- b. Above P-6, operation may continue.
ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers, suspend all operations involving positive reactivity changes and verify 'lalve 2CHS-91 are closed and secured in position within the next hour. ACTION 6 - This Action is not used. ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:
- a. Less than or equal to 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within I hour; restore the inoperable channel to operable status within 24 hours after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the following 6 hours.
- b. Above 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour; operation may continue until performance of the next requirad CHANNEL FUNCTIONAL TEST.
1 f ' ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-9, place the inoperable channel in the tripped condition within 1 hour; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST. ACTION 9 - This ACTION is not used O BEAVER VALLEY - UNIT 2 3/4 3-6
l /x ( \ TABLE 3.3-1 (Continued) l'O l ACTION 10 - This Action is not used.' i' ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may. continue provided the inoperable channel is placed in the tripped condition within 1 hour. j ACTION 12 - With the number of channels OPERABLE one less than required by ' the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the reactor trip breakers. ACTION 39 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the reactor trip breakers within the next hour. ACTION 40 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status j within 48 hours or declare the breaker inoperable and apply ACTION 1 or ACTION 39 as applicable. Neither breaker shall be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status. ! O ACTION 44 - With less than the Minimum Number of channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required ; state for the existing plant condition, or apply Specification i 3.0.3. I I BEAVER VALLEY - UNIT 2 3/4 3-7 i
TABLE 3.3-2 ; REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME
- 1. Manual Reactor Trip NOT APPLICABLE
- 2. Power Range, Neutron Flux 5 0.5 seconds *
- 3. Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE <
- 4. Power Range, Neutron Flux, High Negative Rate 5 0.5 seconds *
- 5. Intermediate Range, Neutron Flux 40T APPLICABLE
- 6. Source Range, Neutron Flux NOT APPLICABLE (Below P-10)
- 7. Overtemperature AT $ 4.0 seconds *
- 8. Overpower AT s 4.0 seconds *
- 9. Pressurizer Pressure--Low -< 2.0 seconds (Above P-7)
- 10. Pressurizer Pressure--High 5 2.0 seconds
- 11. Pressurizer Water Level--High NOT APPLICABLE (Above P-7)
- 12. Loss of Flow - Single Loop (Above P-8) 5 1.0 seconds
- 13. Loss of Flow - Two Loop 5 1.0 seconds (Above P-7 and below P-8)
- 14. Steam Generator Water Level--Low-Low ~< 2.0 seconds (Loop Stop Valves Open)
- 15. Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE
- 16. Undervoltage-Reactor Coolant Pumps 5 1.5 seconds (Above P-7)
- 17. Underfrequency-Reactor Coolant Pumps 5 0.9 seconds (Above P-7)
- Neutron detectors are exempt from response time testing. Response time shall be measured from detector output or input of first electronic component in channel.
BEAVER VALLEY - UNIT 2 3/4 3-8
f~N TABLE 3.3-2 (Continued) ( REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL' UNIT RESPONSE TIME
- 18. Turbine Trip (Above P-9)
A. Emergency Trip Header NOT APPLICABLE Low Pressure : B. Turbine Stop Valve Closure NOT APPLICABLE !
- 19. Safety Injection Input from ESF NOT APPLICABLE l
- 20. Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE (Above P-7)
- 21. Reactor Trip Breakers NOT APPLICABLE 2.2. Automatic Trip Logic NOT APPLICABLE
- 23. Reactor Trip System Interlocks NOT APPLICABLE b
G i I O BEAVER VALLEY - UNIT 2 3/4 3-9 L______________-________
5 H C , IE
- HC 4 WN
- A ,
4 NLD
- ILE 3 2 3 IR SEI , , ,5
- 5 EVU 2 2 ) 2 2 2 3 2 2 2 2 DRQ 7 ,
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. . . . . . . . . 0 1 2 3 4 5 6 7 8 9 1 n 9< x <E [m e Zr o w
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- N 3 E M
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S S S S S N N N N. N N Y S d P n t I R al n F e t a e S T h v n l r E R h g
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l tL l o s m e i p e a o C o o k T H o ) v mr o r l r a) C - o 8 e se C r ) e C f e7 A l L p- L it o 9 d r-E e oP Ma r t - a e t BP R v e o r W o c P e v u e l L w e w t a H l p pe L g o t or c e e a n mv n ol a l o a) R) v p V I uo r e i we W Ft ?7 7 o i Pb S Tb a L - - - b rep n A T t r rr P P A T ro o t( I a - - d o ee - y ( ut i n N W n t t n e ce ysS t ap U r) w o) wa o a r ae ev go nv p cs c l i wG eo i nee e or L e7 l 8 l 7 e d ab ub r ern j oT A N z - F - F - n e em tA qA T gPi n C i F r P P ea l ( e( r b I n O f f G w F e o r e ewr ro I ue oe oe /o /t v f n mou y oi T sv v v mL mS rs rs i ELT t tt C so so so a- a ep ep b e ci N U eb rA sb oA sb oA ew ew d m nu d m nu r u f as F P( t o t o . a eo L( L( SL SL UP UP T A. B S RP 1 2 3 4 5 6 7 8 9 0 1 1 1 1 1 1 1 1 1 2
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) I N d E O e V I u R T n U LA i S ER t NB n N NI .
o O AL A C I HA . ( T CC R R R R R N A 1 T
- N 3 E
. M 4 U R
E T L S B N L A I E T NK M NC . . A A E AE A. A A A T HH . . . . . . S CC N N N N N N Y S P I R s T k c s R o r O l 3 e T r 1 k C e - a A t P e E ne 0 r R I g6 8 9 1 , B n- - - - e map P P P er s eR su s t , , , , l s a sex x x x us p T yt u eu eu eu pe y I S al gl gl gl mr B N iF nF nF nF I P U pd a a a p i en R n R n R n er i L rmo o o o ne r A Trr rr rr rr ib T N et et et et b m O rt u wu wu wu ra r o I one oe oe oe uh T TIN PN PN PN TC t C c c N a a U e . . . . . e F RA B C D E R 3 4 2 2
<Nh [z m ,; Em
(
I j \, I- i G TABLE 4.3-1 (Continued)
, TABLE NOTATION !
With the reactor trip system breakers closed and the. control rod-l l drive system capable of. rod withdrawal. (1) - If not performed in previous 7 days. (2) -- Heat balance only, above 15% of RATED THERMAL F0WER. (3) - Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrates if absolute difference > 3 percent. (4) - (Not used) (5) - Each train tested every other month on a STAGGERED TEST BASIS. (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION. (7) - Below P-10. 1 (8) - Below P-6. (9) - Required only when below Interlock Trip ! atpoint. 1 (10) - The CHANNEL FUNCTMNAL TEST shall independently verify the ! t.
OPERABILITY ne undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).
(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers. (12) - Local manual shunt trip prior to placing breaker in service. (13) - Automatic undervoltage trip. l 1 l v BEAVER VALLEY - UNIT 2 3/4 3-13
INSTRUMENTATION i 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDIT10N FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip , Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value stown in the Allowable Value column of Table 3.3-4, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours that Equation 2.2-1 was :
satisfied for the affected channel, or
- 2. Declare the channel inoperable and spply the applicable ACTION state-ment requirements of Table 3.3.3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
1 EQUATION 2.2-1 Z + R + S < TA j where: Z = The value for Column Z of Table 3.3-4 for the affected channel, l R = The "as measured" value (in percent span) of rack error l for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Drift) of Table 3.3-4 for the affected channel, and 1 I TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.
- c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
I BEAVER VALLEY - UNIT 2 3/4 3-14
i 1
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3/4.3.2 ENGINEERED SAFETY' FEATURE ACTUATION SYSTEM INSTRUMENTATION h1VflLLMCfLRE@JREMENTS i 1 4.3.2.1.1 Each engineered safety feature actuation system instrumentation ' channel and interlock and the autcmatic actuation logic with master and slave relays shall be demonstrated OPERABLE by the performance of the ESFAS Instru- i mentation Surveillance Requirements
- during the MODES and at the frequencies shown in Table 4.3-2.
4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock. opera- i tion. The total interlock function shall be demonstrated OPERABL'E at least o once per 18 months during CHANNEL CALIBRATION testing of each channel affected i by interlock operation. , 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated to be within the limit at least once per 18 months. Each , test shall include at least one logic train such that both logic trains are l tested at least once per 36 months and one channel per function such that all ) channels are tested at least once per N times 18 menths where N is the, total ' l number of redundant channels in a specific ESF function as shown in the " Total l p No. of Channels" Column of Table 3.3-3. l l u ( l
*For the automatic actuation legic,,the survelmnce requirements shall be the application of various simulated input conditions in conjunction with each l i
possible interlock logic state and verification of the required logic output including, as a minimum, a continuity check of output devices. For the actuation relays, the surveillance requirements shall be the clergi2ation of each master and slave relay and verification of OPERABILITY of each relay. g The test of master relays shall include a continuity check of each associated
; slave relay. The test of slave relays (to be performed at le:tst once per 92 V) days in lieu of at least once per 31 days) shall include, as'a nmimum, a continuity check of associated actuation devices that are r.ot Mstalle.
BEAVER VALLEY - UNIT 2 3/4 3 .t5
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l Ih ld TABLE 3.3-3 (Continued) TABLE NOTATION
# Trip function may be bypassed in this MODE below P-11.
## Trip Function automatically bypassed above P-11, and is bypassed below P-11 when Safety Injection on low steam pressure is not manually bypassed.
ACTION STATEMENTS ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of Channels, be in HOT STANDBY within six hours and in COLD SHUTDOWN within tho following 30 hours; however, one chanrel may be bypassed for up to two hours for surveillance testing per Specification 4.3.2.1.1. , provided the other channel is operable. ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels:
- a. Below P-Il or P-12, picce the inoperable channel in the ,
tripped condition within 1 hour; restore the inoperable ! channel to OPERABLE status within 24 hours after exceeding P-11 or P-12; otherwise be in at least HOT STANDBY within the following six hours.
- h. Above P-11 and P-12, place the inoperable channel in the tripped condition within 1 hour; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.
ACTION 15 - (This ACTION is not used) ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels:
- a. Below P-11 cr P-12, place the inoperable channel in the bypass condition; restore the inoperable channel to OPERABLE status within 24 hours after exceeding P-11 or P-12; other-wise be in at least HOT SHUTDOWN within the following 12 hours.
- b. Above P-11 or P-12, demonstrate that the Minimum Channels OPERABLE requirement is met within 1 hour; operation may continue with the inoperable channel bypassed and one channel may be bypassed for up to 2 hours for testing per Specification 4.3.2.1.
BEAVER VALLEY - UNIT 2 3/4 3-21
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l TABLE 3.3-3 (Continued) ACTION 17 - With less than the Minimum Channels OPERABLE, operation may j J continue provided the containment purge and exhaust valves are maintained closed. lCTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. ACTION 33 - With the number of OPERABLE Channels one less than the Total Number of Channels, the Emergency Diesel Generator associated with the 4kV Bus shall be declared inoperable and the ACTION Statements for Specifications 3.8.1.1 or 3.8.1.2, as appropriate shall apply. ACTION 34 - With the number of OPERABLE Channels one less than the Total Nuniber of Channels, STARTUP and/or POWER OPERATION may proceed until the performance of the next required Channel Functional Test provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 36 - The block of the automatic actuation logic introduced by a reset of safety injection shall be removed by resetting (closure) of the reactor trip breakers within one hour of an inadvertent initiation of safety injection providing that all trip input signals have reset due to stable plant conditions. Otherwise, the requirements of ACTION Statement 13 shall have been met. ACTION 37 - (This ACTION is not used) ACTION 38 - With less than the Minimum Number of Channels OPERABLE, within one hour determine by observation of the associated permissive annunciator window (s) (bistable status lights or computer checks) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ACTION 41 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5. ACTION 42 - With the number of OPERABLE Channels one less than the minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; howeve*, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.1 provided the other Channel is OPERABLE. ACTION 45 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within tne following 6 hours. BEAVER VALLEY - UNIT 2 3/4 3-22
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IT 7 st cl 7 s c 9 st sl 9 st sl A P RE uiee u0 ui e ui e I TS >Hwsd >H22 >Hw5d >Bw5d N. R T N O I T ) A S T ( N R E OT M SF 0 A U NI 0 0 0
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- t e
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- t r t r vn yt naSt o )
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- N 3 O . I 3 T A )
E U A L T T B C ( A A T E S C E N R A U LW T AO . . A TL A A 0 E OL . . . F TA N N 4 Y T E F A S D E , E R e R U r E T u 2 E A 4 s 1 N E - s - I F P e P G r N Y , P , E T p g E i r v F r e a A T z T S i r r w T DS o u o I EK t s L N RC c s1 - U EO a e1 w EL e r - o L NR R PP L A IE N GT O NN . . I EI a. b c T C N U . F 8 s EE9<f ' zZ m Ma E
,r TABLE 3.3-5 i
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIAIING SIGNAL AND FUNCTION RCSPONSE TIME IN SECONDS
- 1. Manual
- a. Sarety Injection (ECCS) Not Applicable feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable 1
- b. Containment Quench Spray Pumps Not Applicable Containment Quench Spray Valves Not Applicable Containment Isolation-Phase "B" Not Applicable
\ c. Containment Isolation-Phase "A" Not Applicable
- d. Control Room Ventilation Isolation Not Applicable
- 2. Containment Pressure-High
- a. Safety Injection (ECCS) 1 27.0*
- b. Reactor Trip (from SI: 12.0
- c. Feedwater Isolation 5 7.0(1)
- d. Containment Isolation-Phase "A" 5 61.5(4)/115.5(5)
- e. Auxiliary Feedwater Pumps 5 60.0 )
- f. Service Water System 5 72.5(2)/181.5(3)
- 3. Pressurizer Pressure-Low
- a. Safety Injection (ECCS) 1 27.0*/27.0#
- b. Reactor Trip (from SI) _<
2.0
- c. Feedwater Isolation 1 7.0(1)
- d. Containment Isolation-Phase "A" $ 61.0(4)/115.0 (5)
BEAVER VALLEY - UNIT 2 3/4 3-29 I L___ _ _ -- --
t TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 3. Pressurizer Pressure-Low (Continued)
- e. Auxiliary Feedwater Pumps 5 60.0
- f. Service Weter System 5 72.0(2)/181.0(3)
- 4. Steam Line Pressure-Low
- a. Safety Injection (ECCS) $ 37.0##/27.0#
- b. Reactor Trip (from SI) 5 2.0
- c. Feedwater Isolation 5 7.0(1)
- d. Containment Isolation-Phase "A" 5 61.0(4)/115.0(5)
- e. Auxiliary Feedwater Pumps 5 60.0
- f. Service Water System 5 72.0(2)/181.0(3)
- g. Steam Line Isoiation 5 7.0
- 5. Containment Pressure--High-High
- a. Containment Quench Spray 5 85.5(5)
- b. Containment Isolation-Phase "B" Not Applicable
- c. Control Room Ventilation Isolation 5 22.0(4)/77.0(5)
- 6. Steam Generator Water Level--High-High
- a. Turbine Trip 5 2.5 ;
- b. Feedwater Isolation 5 7.0(1)
- 7. Containment Pressure--Intermediate High-High 1
l a. Steam Line Isolation 5 7.0
- 8. Steamline Pressure Rate--High Negative )
- a. Steamline Isolation 57.3 l
BEAVER VALLEY - UNIT 2 3/4 3-30
TABLE 3.3-5 (Continued) U ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS I
- 9. Loss of Power
- a. 4.16kv Emergency Bus Undervoltage 1 1 0.1 sec.
(Loss of Voltage) (Trip Feeder)
- b. 4.16kv and 480v Emergency Bus Under- 90 1 5 sec.
voltage (Degraded Voltage)
- 10. (Intentionally blank)
- 11. Steam Generator Water Level-Low-Low
- a. Motor-driven Auxiliary ~< 60.0 Feedwater Pump **
- b. Turbine-driven Auxiliary -< 60.0 Feedwater Pump ***
- 12. Undervoltage RCP
- a. Turbine-driven Auxiliary 5 60.0 Feedwater Pump d 13. Trip of Main Feedwater Pumps
- a. Motor-driven Auxiliary 5 60.0 Feedwater Pumps
- 14. Turbine Driven Auxiliary Feedwater Pump Discharge Pressure Low
- a. Motor-driven Auxiliary 5 60.0 Feedwater Pumps
- 15. Control Room High Radiation
- a. Control Room Ventilation Isolation 5 180(6)
**on 2/3 in 2/3 Steam Generators
( ***on 2/3 any Steam Generator BEAVER VALLEY - UNIT 2 3/4 3-31
TABLE 3.3-5 (Continued) , TABLE NOTATION
- Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps and Low Head Safety Injectior. pumps. Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST)
(RWST valves open, then VCT valves close) is not included.
# Diesel' generator starting and sequence loading delays not included.
Offsite power available. Response time lirait includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suctian from the volume control tank (VCT) to the refueling water storage tank (RWST) (RWST valves open, then VCT valves close) is included.
- Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment i of discharge pressure for centrifugal charging pumps. Sequential transfer 1 of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST) (RWST valves open, then VCT valves close) is included.
(1) Feedwater system overall response time shall include verification of valve stroke times applicable to the feedwater ulves shown for penetrations 76, 77 and 78 shown in Table 3.6-1. ; (2) Diesel generator starting end sequence loading delays included. Response I time limit iricludes attainment of discharge pressure for service water pumps. 1 (3) Diesel generator starting and sequence loading delays not included. Response time limit only includes opening of valves to establish the flowpath to the diesel coolers. (4) Diesel generator starting and sequence loading delays not included. Offsite power available. Response time limit includes operation of valves / dampers. (5) Diesel generator starting, and sequence loading delays included. Response time limit includes operation of valves / dampers. (6) Diesel generator starting and sequence loading delays not included. Response time limit includes operation of dampers. O BEAVER VALLEY - UNIT 2 3/4 3-32
, WN 4 4 4 4 A
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- OU 3 IQ
. TE 4 AR U
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H e C c IE HC n WN 4 4 4 4 a 4 4 4 A , , l NLD , , l 3 3 3 - ILE 3 3 3 3 3 i IR e SEI , , , , , v , , , EVU 2 2 2 2 2 r 2 2 2 DRQ u , OUE , , , , , S , , 1 1 1 1 1 MSR 1 1 1 n o i t c e j n L I A y N LO t EI e NT f NCT ) ) ) ) a ) ) ANS 1 2 1 2 S 1 2 ( ( ( ( ( N HUE ( M O CFT M M M M M M M I T 'a A r T N o E f M e U v R N T O o S I b N T a I LA ER . MS NB 1
) ET NI .
A. t A A d TN AL A A. A. e SE HA . . . i n N N R u YM CC N N R N N n SE U i R l t NI n OU a . o I Q ns C TE ot ( AR i n U t e TE cm 2 CC L ne
- AN E ur 3 A NK Fi
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- g l D h o e -
E g L R - R i s e E d H ny nn r E ns - oa oo u N ay - n il ii s I a e o t e n n tt s G nl r n i aR o n o aa e N oe u o t u i o i uu r E n iR s N i a t n t i t tt P o t s O t i co c t a cc i an e I a t Ai e a t AA t t uo r T l i t j l i n Y a ti P A o n ca n o n cd eh mg A i ct L s I i u I s I i n R t Aa t O I tt I t a ni P i u n S l ac y l a iH S n ct e I a mA t a mc a-I i c m "A u o e "B u oi th T tA n T " n td f " n t g ng T N l a i N a un a a uo oi I E a mc a E e M A a S e s M AL CH N M u oi th M s U N n t g ng N a a I a uo oi I h . . . h P 1 2 3 L A M AL CH A P 1 2 3 A T T N N N O O . . . O . . I C a b c C a b T C N U . . F 2 3 EE9 <MG Eam w1 aL^ >
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'y t
NT e NCT ) ) ) ) f ANS 1 1 2 2 a N HUE ( ( ( ( S O CFT M M M M M M M M I l T l A T a N r E o M f U R N e T O v S I o N T b I LA ER a MS NB 1
) ET NI . . .
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HC WN 4 4 4 4 A NLD , , , , I LE 3 3 3 3 3 3 3 I R , , , , SEI , EVU 2 2 2 2 2 2 2 2 DRQ , , OUE , , , , , , MSR 1 1 1 1 1 1 1 1 L A N LO EI NT NCT ) ANS 2 N HUE ( O CFT M M M M M M M M I T A T N E M U R N T O S I N T I LA ER MS NB
) ET NI .
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)
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- t s n RP c g g cV ea R ay r i r ow
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)
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) ET NI .
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4. RL AE e A A UI HH e . A. A. . A. E TE CC_ S N N N N L AV N B ER A FU T S Y T S E K F - C A r r O _ S o o L t ht R D ) o t o ) E 1 E d M iM s T 1 R e W p N - E u t t m I P E N n r a wr oa u i P E , I t t Lt R e G n S p S n U r N o ( me( e T u 2 E c ur v A s ( n 4 1 _ P en i E - s - _ o se sr F P e P _ R i nsp pD r F t) cs eeO) vr m- Y , P , T A ep s nur T p g iD ep iP o E i r v _ W D jm nu r vm a t F r e a _ d el u Mro A T z T E IP - gap eM S i E erV ft r r w T F yn na n oat D o u o I N Y t e ev ih me wr E t s L U R fi b cav pd a R c s - rsei i et E a e w A ar uit r reS E e r o _ L I SD TDSD TF( N R P L A L I N I G O X . . N . . . I U d e f. E a b c T A C N U 7. F 8 E E39 hEf EZ m w1 wO"
TABLE 4.3-2 (Continued) TABLE NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNCTIONAL TEST at least once per-31 days. (2) Each train or logic channel shall be tested at least every other 31 days. O i i O\ BEAVER VALLEY - UNIT 2 3/4 3-38
i i
/ \ INSTRUMENTATION V 3/4.3.3 MONITORING INSTRUMENTATION j
RADIATION MONITORING I LIMITING CI)MITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 .- j shall be OPERABLE with their alarm / trip setpoints within the specified limits. q y APPLICABILITY: As shown in Table 3.3-6. l j ACTION: l i
- a. With a radiation monitoring channel alarm / trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours or declare the channel inoperable. ,
i
- b. With one or more rediation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.
1
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENT3 (v-4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3. l BEAVER VALLEY - UNIT 2 3/4 3-39
N 7 O 4 I T 0 0 1 C 9 6 6 A 1 3 4 2 2 2 c c c c c / c r r / i / h h i C i
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) M e eR lR E4 gA 1 kl l - b - 0 r1 d 3 t ae bS oS e1 a0 e 1 n eR oV NV gQ h1 u - e L NH H rR cQ n I u d 2 e2 u- sR i () l yn e( g( PR i-t n
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TABLE 3.3-6 (Continued) ACTION STATEMENTS ACTION 19 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area sur-veys of the monitored area with portable monitoring instrumentation at least once per 24 hours. ACTION 20 - With the number of channcis OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. ACTION 21 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the appli-cable ACTION requirements of Specifications 3.9.12 and 3.9.13. ACTION 22 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9. ACTION 36 - With the number of OPERABLE channels less than required by . the Minimum Channels OPERABLE requirement, either restore ! the inoperable channel (s) to OPERABLE status within 72 hours, oe:
- 1) Initiate the preplanned alternate method of monitoring the appropriate parameter (s), and
- 2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. j ACTION 43 - With the number of OPERABLE channels less than required by the Minimum channels OPERABLE requirement, either restore the inoperable channel (s) to OPERABLE status within 72 hours, or:
1 1) Initiate the prepianned alternate method of monitoring the appropriate parameter (s), and
- 2) Return the channel to OPERABLE status within 30 days or l
explain in the next Semi-Annual Effluent Release Report l why the inoperability was not covered in a timely manner. ACTION 46 - With the number of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, either resture the l inoperable channel to OPERABLE status within 7 days or close I the control room series normal air intake and exhaust isola-tion dampers. ACTION 47 - With no OPERABLE channels either restore one inoperable channel to OPERABLE status within 1 hour or close the control room series normal air intake and exhaust isolation dampers. BEAVER VALLEY - UNIT 2 3/4 3-42
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l ,- ~ INSTRUMENTATION ! '( MOVABLE INCORE DETECTORS LIMITING CONDIIl01LfDR.JEBATIOJ. 3.3.3.2 The movable incore detection system shall be OPERABLE with:
- a. At least 75% of the detector thimbles,
- b. A minimum of 2 detector thimbles per core quadrant, and
- c. Sufficient movable detectors, drive, and readout equipraent to map these thimbles.
APPLICABILITY: When the movable incore detection system is assed for: A. Recalibration of the axial f k x offset detection system, B. Monitoring the QUADRANT POWER TILT RATIO, or C. Measurement of F and Fq (Z). ACTION: fG -With the movable incore detection system inoperable, do not use the system for Q the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. I SURVEILLANCE RE0UIR$ENTS 4.3.3.2 The incore movable detection system shall be demonstrated OPERABLE by normalizing each detector output to be used within 24 hours prior to its use when required for: l
- a. Recalibration of the excore axial flux offset detection system, or i l
- b. Monitoring the QUADRANT POWER TILT RATIO, or i
- c. Measurement of F and Fq (Z).
)
BEAVER VALLEY - UNIT 2 3/4 3-45
', (.} ,
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?
i x , INSTRUMENTATION j b Q o
, , a
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t s! ,c,< ," ,,.
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d lll EtJ,g(jgjpN FOR OPERAT10N'
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i _ _ _f, . m. a 3.3.3.3 The seismic monitoring instromerdation shown in T0.ble 3.3-7 shall be fu s, , 4 OPERABLE. x's _
< t- .
t h APPLIypILITY: )Atalltims.
, 3
ACTION: . y, l , _p a. With the nilnberfof OPERAEiLE seismic monitoring instruments less than reqaireo%Tucle 3.3-7, restore the inoperable instrument (s) to OPERABL.E st'atus within 30 days, g c
- b. With one'or more seismic moaitoring instruments inoperable for more n ,'" than'30 days, prepare and submit a Special Report to the Commission
,L pursuant to Specification 6.0 2'withhi the next 10 days outlining the i,"'7 cause of the malfunction and the plans for restoritig the instrument (s)
~
to OPERABLE status. , i 1
, e: The pi'ovisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- < 1 S M L%,W E EEC E U6 w n L
- e _
3 4.3.3.3.C ' ' Each of the ahove seismic monitoring instruments sha!F be 'denon- ^ stratec;0PERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and Cf?.?NEL FUNCTIONAL)TESY' operations at the frequencies shown in Table 4:3-4. 4.3.3ds2 A seismic event greater than or equal to 0.01g shall be repcrted to the l'oamission withir. I hour. Each of the abovesseismic monitoring instruments actuattd i ting a seismic event greater than or equal to 0.01g shall Le restored ito OPERABLE status within 24 hours and a CHANNEL CALIBRATION pe" formed withir,
'Je da,Ms following the seismic evn.0.. Data shall be retrieved fnom actuated init'wants and analyzed to detetine the ' magnitude of the vibratory ground mo' nn,- A Srs.eial Report shall be' prepared and submitted to the Con' mission pursuant to @ edification 6.9.2 within 30 o'ays describing the magni.tude, frequency spectrum ar.d resultant effect . yon facility features impprtant ta safety. .
O BEAVFR VALLEY UNIT 2 3/4 3-46
? !? ? 1' ?' } 'A l '
\L y '
1 i
) ,q TABLE 3.3-7 j
_( ) SEISMIC MONITORING INSTRUMENTATION
%J .
MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. TRIAXIAL TIME-HISTORY ACCELER0 GRAPHS (1)(2)(4) j
- a. Containment Mat el. 692'-11" i1g 1
- b. Containment O 1g 1 el. 767'-10" perating Floor c, Switchyard i1g 1 (
i
- d. Containment Building - Steam i1g 1 Generator Support Cubicle No. 1 el. 710'6"
- e. Aux. Building - at center of i1g 1 Mat, el 710'6"
- f. Aux. Building - at base of i1g 1 480 volt MCC (MCC-2-E03),
el. 755'6"
- 2. TRIAXIAL PEAK ACCELER0 GRAPHS
- a. Containment Bulding - 12g 1 RHS heat exchanger (2RHS-E21A) el. 715'-6"
- b. Containment Building - Six Inch i2g 1 SI Pipe (2 SIS-006-269-1(A))
el 741'-5"
- c. Aux. Building - MCC-2-E03 15g 1 ;
el. 755'-6" l 3. TRIAXIAL SEISMIC SWITCH (3)
- a. Containment mat N/A 1
- 4. RESPONSE SPECTRUM ANALYZER l
- a. Control Room N/A 1 NOTES (1) Units a, b, c are wired to accelerograph recorders in the Control Room. Units d, e, and F are self contained units '
(2) Each accelerograph trigger setpoint is set at 0.01g. p (3) Triaxial seismic switch setpoints:
- horizontal sensor; 0.0499 vertical sensor: 0.0379 (4) Triaxial time-history accelerograph - Units a and c are input directly to the response spectrum analyzer in the Control Room.
BEAVER VALLEY - UNIT 2 3/4 3-47 L_-_------------.--- - - - - - -
l
-TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. TRIAXIAL TIME-HISTORY ACCELERCGRAPHS
- a. Containment Mat. el. M* R SA 692'-11" l
- b. Containment Operating M* R SA >
floor el. 767'-10"
- c. Switchyard M* R SA
- d. Containment Building - Steam N/A R N/A Generator Support Cut >icle No. 1 el. 718'6"
- e. Aux. Building - at center of N/A R N/A Mat, el 710'6"
- f. Aux. Building - at base of N/A R N/A 480 volt MCC (MCC-2-E03),
el. 755'6"
- 2. TRIAXIAL PEAK ACCELER0 GRAPHS
- a. Containment Building - N/A R N/A RHS heat exchanger (2RHS-E21A) el. 715'-6"
- b. Containment Building -Six inch N/A R N/A SI pipe (2 SIS-006-269-1(A))
el. 741'-S"
- c. Aux. Building - MCC-2E03 N/A R N/A el 755'-6"
- 3. TRIAXIAL SEISMIC SWITCHES
- a. Containment mat N/A N/A R
- 4. RESPONSE SPECTRUM ANALYZER
- a. Control Room N/A R N/A i
- Except seismic trigger BEAVER VALLEY - UNIT 2 3/4 3-48
i 1
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[1
\g INSTRUMENTATION l
METEOR 0 LOGICAL' INSTRUMENTATION " LIMITING CONDITION FOR OPERATIOR 3.3.3.4 The meteorological monitoring instrumentation. channels shown in Table 3.3-8 shall & OPERABLE. APPLICABILITY: At all times. ACTION: f
- a. With the number of OPERABLE meteorological monitoring channels less than required by Table 3.3-8, suspend all release of gaseous radio-active material from the radwaste gas decay tanks antil the inoperable channel (s) is t. Stored to OPERABLE status,
- b. With one cr more required meteorological monitoring channels inoper-able for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next. 10' days outlining the cause of the malfunction and the plans for restoring-the channel (s) to OPERABLE status, l
s c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. S E EILLANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the f requencies shown in Table 4.3-5. 1' f
\
BEAVER VALLEY - UNIT 2 3/4 3-49
'I l
l i TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT ; MINIMUM MINIMUM INSTRUMENT LOCAfION ACCURACY OPERABLE ,, i
- 1. WIND SPEED
]
- a. Nominal Elev. 500' i 0.5 mph
- 1
- b. Nominal Elev. 150' O.5 mph
- 1 J i
- c. Nominal Elev. 35' i 0.5 mph
- 1
]
1
- 2. WIND DIRECTION 1
- a. Nominal Elev. 500' i5 1
- b. Nominal Elev. 150' i5 1
- c. Nominal Elev. 35' i5 1
- 3. AIR TEMPERATURE AT
- a. AT Elev. 500'-35' i 0.1 C 1
- b. AT Elev. 150'-35' i 0.1 C 1 O
- Starting speed of anemometer shall be < 1 mph.
BEAVER VALLEY - UNIT 2 3/4 3-50
1 i I f TABLE 4.3-5
)
l s METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. WIND SPEED
- a. Nominal Elev. 500' D SA
- b. Nominal Elev. 150' D SA
- c. Nominal Elev. 35' D SA
- 2. WIND DIRECTION
- a. Nominal Elev. 500' D SA
- b. Nominal Elev. 150' D SA
- c. Nominal Elev. 35' D SA j 7 3. AIR TEMPERATURE AT l
( a. AT Elev. 500'-35' D SA
- b. AT Elev. 150'-35' D SA l l
I BEAVER VALLEY - UNIT 2 3/4 3-51
INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CQNMIl0E_ EOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY: MODES 1, 2 and 3. ACTION: With the number of OPESABLE remote shutdown monitoring channels less than required by Table 3.? 9, either:
- a. Restore the inoperable channel to OPERABLE status within 7 days, or
- b. Be in H0T SHUTDOWN within the next 12 hours.
SURVEILLAtiCE RE0VIERiEN_TS ,_ , 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demon-strated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6. i O BEAVER VALLEY - UNIT 2 3/4 3-52
(1. t r o t r o t r o a a a r r r e e e n n n e e e g g g SE - MLL m m m UEB a a a MNA e e e INR t t t NAE s s s IHP / / / MCO 1 1 1 1 l 1 1 1 1 1 1 1 N O I T A s T p g N m M i E a P s M D S M p U 3 P P g R
- 0 C D 0 i -
T T 0 s S N 1 5 6 5 0 5 p M N F P E + 0 + F F 2 % G I M E o 1 % 0 0 0 G t o o 0 0 o 0 0 0 0 0 R t o t 0 0 t 0 2 1 N UE 1 t 4 0 7 7
, 9 3
I R O SG AN EA 1 0 5 0 0 5 0 0 0 7 1 1
- t o
0 4 T MR 1 - 1 - 0 0 1 0 0 0 5 0 3 I N E O L M B A L T E N A P N W O D T U H S E T O x e u t M l a - - e E F R R t r p e e t a r r e R a u x e u u l e t u t t t e t w l r l a a a r u o c a F R r r u O u t e e s l l F l N S r p p p e s e X e a u m m r e v H r n e g e e t e e u r e e a g l r T T s P L n n c a l - t P a a u s e a R R N t t n t e v r r e w n S n r e o o r d w a a P l t t u e o
- e e e e l l a a t e d S t t g g o o r r r r T a a n n o o a F t N a a e e e e r u i i C C z z n n e y h E d d R R g e M e e rg i i e p r S U m m e e re r r G G m a r r oe ol u u e i y R c c tl t s s m m T l c T e e r r c cd s s a a i n S t t u u at al e e e e R x e N n n o o eo eo r r t t H u g I I I S S RH RC P P S S R A r e
m 0 1 2 E 1 2 3 4 5 6 7. 8 9 1 1 1
- s sPQ h5] R^
I TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENTS
- CHECK CALIBRATION
- 1. Intermediate Range Nuclear Flux M N.A.
- 2. Intermediate Range Startup Rate M N. A.
- 3. Source Range Nuclear Flux (2) M(4) N.A.
- 4. Source Range Startup Rate (2) M(4) N.A.
- 5. Reactor Coolant Temperature - Hot Le9 M R
- 6. Reactor Coolant Temperature - Cold Leg M R
- 7. Pressurizer Pressure M R
- 8. Pressurizer Level M R
- 9. Steam Generator Pressure M R
- 10. Steam Generator Level M R
- 11. RHR Temperature - HX Outlet (3) M R
- 12. Auxiliary Feedwater Flow Rate S/U(1) R
- Emergency shutdown Panel (1) Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.b following an extended plant outage.
(2) Operability required in accordance with Specification 3.3.1.1. (3) Operability required in accordance with Specification 3.4.1.3. (4) Below P-6. I O BEAVER VALLEY - UNIT 2 3/4 3-54
INSTRUMENTATION 3/4.3.3.6 (This Specification number is not used.) l BEAVER VALLEY - UNIT 2 3/4 3-55 _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . i
INSTRUMENTATION ' CHLORINE DETECTION SYSTEMS LittUltiLCDHDlIIDN FOR OfffUiT10tL_ 3.3.3.7 Three independent chlorine detection systems, with their alarm / trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, 4, 5 and 6. ACTION:
- a. With one chlorine detection system inoperable, operation may continue provided the inoperable detector is placed in the tripped condition within 1 hour.
- b. With two chlorir..' detection systems inoperable, restore one of the inoperable detection systems to OPERA 3LE status within 7 days, or within the next 6 hours, isolate outside air inlet to the control room.
- c. With no chlorine detection system OPERABLE, within 1 hour isolate outside air inlet to the control room.
- d. The provisions of Specification 3.0.4 are not applicable.
ELR'LElllMCE BEWIREMENT$ _ 4.3.3.7 Each chlorine detection system shall be demonstrated OPERABLE by per-formance of a CHANNEL FUNCTION TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months. 1 i i O; i BEAVER VALLEY - UNIT 2 3/4 3-56
h p INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FAR OPE 3AT10N 3.3.3.8 The accident monitoring instrumentation chanhels shown in Table 3.3.11 shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3.11, either restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUIDOWN within the next 12 hours except for the PORV(s) which may be isolated in accordance with Specification 3.4.11.a.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.11, either restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
- c. With the nunber of OPERABLE Reactor Vessel Level Indication System channels less than the required number of channels or the Minimum Channels OPERABLE Q[N requirement, restore the inoperable channel (s) to OPERABLE status as per ACTION a or b above as applicable if repair is not feasible, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days that provides action taken, cause of the inoperability, and the plans anti schedule for restoring the channels to GPERABLE status.
This ACTION statement applies to the first fuel cycle only.
- d. With the number of OPERABLE Reactor Coolant System Subcooling Margin l Monitor instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.11, either restore the inoperable channel (s) to OPERABLE status within 7 days or be in a least HOT SHUTDOWN within the next i 12 hours,
- e. The provisions of Specification 3.0.4 are not applicable.
EURVElLLMCE RE0VIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies snown in Table 4.3-7. ( BEAVER VALLEY - UNIT 2 3/4 3-57
TABLE 3.3-11 ACCIDENT MONITORING INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS INSTRUMENT OF CHANNELS OPERABLE ACTION
- 1. Pressurizer Water Level 3 2 a, b
- 2. Auxiliary Feedwater Flow Rate 2 per steam 1 per steam a, b I generator generator l
- 3. Reacter Coolant System Subcooling Margin Monitor 2 1 d
- 4. PORV Limit Switch Position Indicator 1/ valve 0/ valve a, b
- 5. PORV Block Valve Limit Switch Position Indicator 1/ valve 0/ valve a, b
- 6. Safety Valve Position Indicator 1/ valve 0/ valve a, b
- 7. Safety Valve Temperature Detector 1/ valve 0/ valve a, b !
- 8. Containment Sump Wide Range Water Level 2 1 a, b
- 9. Containment Wide-Range Pressure 2 1 a, b
- 10. Reactor Vessel Level Indication System 2 1 a, b , c*
- 11. Core Exit Thermocouple 4/ core 2/ core a, b quadrant qua# ant
- 0nly ACTION statement c is applicable during the first fuel cycle. ACTION statements a and b are applicable thereafter.
BEAVER VALLEY - UNIT 2 3/4 3-58
j 1 /N TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBPsATION
- 1. Pressurizer Water Level M R
- 2. Auxiliary.Feedwater Flow Rate S/U* R
- 3. Reactor Coolant System Subcooling M R !
Margin Monitor
- 4. PORV Limit Switch Position Indicator M R
- 5. PORV Block Valve Limit Switch Position Indicator M R
- 6. Safety Valve Position Indicator M R
- 7. Safety Valve Temperature Detector M R
- 8. Containment Sump Wide-Range Water Level M R y 9. Containment Wide-Range Pressure N/A R
- 10. Reactor Vessel Level Indication System M R
- 11. Core Exit Thermocouple M R i
i
- Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.b following an extended plant outage.
BEAVER VALLEY - UNIT 2 3/4 3-59
t INSTRUMENTATION , RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LililllNG_CONDIT10!LEDR OPERA 110N 3.3.3.9 The radioactive liquid effluent monitoring instrurrentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm / trip retpoints of the radiation monitoring channels shall be determined in accordance with the Offsite Dose Calculation Manual (00CM). APPLICABILITY: During releases through the flow path. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or correct the alarm / trip setpoint.
- b. With one or more radioactive liquid effluent monitoring instruments-tion channels inoperable, take the ACTION shown in Table 3.3-12 or conservatively reduce the alarm setpoint. Exert a best effort to return the channel to operable status within 30 days, and if unsuc-cessful, explain in the next Semi-Annual Effluent Release Report why i the inoperability was not corrected in a timely manner.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
MRVEILLANCERE0MIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12. O BEAVER VALLEY - UNIT 2 3/4 3-60
i 77 ( ) TABLE 3.3-12 v' RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION
- 1. Gross Radioactivity Monitor Providing Alarm '
and Automatic Termination of Release a) Liquid Waste Process Effluent Monitor (2SGC-RQ100) 1 23 '] l
- 2. Gross Radioactivity Monitors Providing Alarm But i
-Not Providing Termfr,ation of Release None
- 3. Flow Rate Measurement Devices a) Liquid Radwaste Effluent (2SGC-FS100) 1 25 b) Cooling Tower Blowdown Line (2CWS-FT101) 1 25
- 4. Tank Level Indicating Devices (For tanks outside plant buildings)
None 1 a BEAVER VALLEY - UNIT 2 3/4 3-61
TABLE 3.3-12 (Continued) ACTION STATEMENTS ACTION 23 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed provided that prior to initiating a release;
- 1. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and;
- 2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valving; Otherwise, suspend release of radioactive effluents via this pathway. ]
ACTION 24 - (This ACTION is not used) ACTION 25 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases j via this pathway may continue provided the flow rate is esti- j mated at least once per 4 hours during actual releases. Pump 4 curves may be used to estimate flow. f O l l I 1 l l O I l BEAVER VALLEY - UNIT 2 3/4 3-62
(N TABLE 4.3-12 I k' u
)
RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL )
~
CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST. 1
- 1. Gross Radioactivity Monitor Providing Alarm and Automatic Termination of Release a) Liquid Waste Process D P(5) R(2)(3) Q(1)
Effluent (2SGC-RQ100) 1
'2. Gross Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release l None
- 3. Flow Rate Measurement Devices a) Liquid Radwaste 0(4) N/A R Q
, Effluent (2SGC-FS100) b) Cooling Tower Blowdown D(4) N/A R Q Line (2CWS-FT101)
- 4. Tank Level Indicating Devices (For tanks outside plant buildings)
None 1 1
/O V
BEAVER VALLEY - UNIT 2 3/4 3-63
TABLE 4.3-12 (Continued) ' TABLE NOTATION (Continued) (1) - The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room Alarm Annunciation occurs if the instrument indicates measured levels above the alarm / trip setpoint. (2) - The CHANNEL CALIBRATION shall also demonstrate that Control Room alarm annunciation occurs if either of the following conditions exist:
- 1. Downscale failure.
- 2. Instrument controls are not sot in operate mode.
(3) - The initial CHANNEL CALIBRATION for radioactivity measurement instru-mentation shall be performed using one or more of the reference stan-dards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measure-ment assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen months. (4) - CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made. (5) - A source check may be performed utilizing the installed means or flashing the detector with a portable source to obtain an upscale increase in the existing count rate to verify channel response. l 9 BEAVER VALLEY - UNIT 2 3/4 3-64
_ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ - _ _ _ _ _ ___ _ _ _ - _ i v] [ INSTRUMENTATION ' ] RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION _ LIMITING COBDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be.0PERABLE with their alarm / trip setpoints set to ensure that the limits of 3.11.2.1 are not exceeded. The alarm / trip setpoints of the radiation monitoring channels shall be determined in accordance with the Offsite Dose Calculation Manual (00Cl<). APPLICABILITY: During releases through the flow path. ACTION:
- a. With a radioactive gaseous process or effluent monitoring instruments-tion channel alarm / trip setpoint less con:,ervative than a value which will ensure that the limits of 3.11.2.1 are met, immediately suspend the release of rariloactive gaseous effluents monitored by thc affected channel or correct the alarm / trip setpoint.
- b. With one or more radioactive gaseous effluent monitoring instruments-tion channels inoperable, take the ACTION ~shown in Table 3.3-13 or conservatively reduce the alarm setpoint. Exert a best effort to
/' return the channel to operable status within 30 days, and if unsuc-( cessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SbREEILLAliCE_ REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13. 1 Ov BEAVER VAELEY - UNIT 2 3/4 3-65
N 0 0 O 3 3 I T , C 7 2 8 8 9 2 8 8 A 2 3 2 2 2 3 2 2 e e t e t e e a e t a t t R t a R a a a R R R y R y t w t w w it wt ot it ot ot vn on l n vn l n l n R i e l e F e i e F e F e E t m F m m t m m m T ce e re ce se re N E ar mr er ar sr er O M ou eu l u ou eu l ps u I A i s t s ps i s cs T R d a sa ma d a oa ma A A ae ye ae ae re ae T P RM SM SM RM PM SM N E M U R T S N N T I I L G I N B I A R C O I T L 3 I P 1 N P
- O A * * * * * * *
- 3 M 3 T N
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/ t f l y t l l e& s a f g F SA s a F F t a l E n a l sA G u i r n1 G u s r T a8 c mr ee c
o0 e c s e e N W0 e i eu l i1 i l p E 1 l t t s pi tQ l t c M s - b r sa mv aR b r o m U R uW oG N o P a ye SM ae SD l iS N o P a P r S a T e- tV S sM nH N aR . . . . e2 . . . . I G( a b c d V( a b c d 1 2 5 3" 9 e ?"}'sZ,n i ; sm w3 wac
. f N 0 O 3 I T , C 9 2 8 8 9 2 8 8 A 2 3 2 2 2 3 2 2 e e t e e t e e a t t a t t R a a R a a R R R R y y t w w t w w it ot ot it ot ot vn l n l n vn l n l n R ie F e F e .i e F e F e E t m m m t m m m T ce ar se re ce se re E sr er ar sr er M ou eu l u ou eu l u A is cs ps i s cs ps R d a oa ma d a oa ma A ae re ae ae re ae P RM rM SM RM PM SM Y T I L
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- t M M n t M M o S y i i y i st V t v e e d t v e e i c H i i t t l i i t t h a 2 v t a a i v t a a t e
( i c R R u i c R R r t A B) t A a e c w w B c w w if s A e o o n A e o o vo a t l l o& t l l e s a F F i s a F F eg l a l tA a l sn e G u s r a1 G u s r ai T R c s ee n0 c s e eg N e i e l c i3 e i e l l r E d l t c pi mQ l t c p eu M e b r o mv aR b r o m rp U t o a r ae t - o a r a R a N P P SD nQ4 N P P S gg a T v ot nn S e cR ii N l . . . . e2 . . . rr I E a b c d D( a b c d. uu D0 3 4
- ER 9 :.N }{ 'E:) ~
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N O I T 2 8 8 1 C 9 2 8 8 9 A 2 3 2 2 2 3 2 2 3 e e t e e t e e a t t a t t R a a R a a R R R R y y t w w t w w it ut ot it ot ot vn l n l n vn l n l n F e R i e F e F e i e F e E t m m m t m m m T ce se re ce se re E ar sr er ar sr er n M ou eu l u ou eu l u e ps cs ps g A i s cs i s R d a oa ma d a oa ma y x A ae re ae ae re ae P RM PM SM RM PM SM 0 Y T I L ) I d B e A u C n I i L t P
- n P o A
- C
( 3 1 3
. SE 3 MLL UEB E MNA L I NR B NAE A I HP 1 1 1 1 1 2 T MCO 1 1 1 t
n r e r V o o r t r t g o i r r o i r r n t n o o t n o o i i o t t i o t t d n M i i n M i i l o n n o n n i M y o o M y o o u t M M t t M M B y i l y i u v e e ) t v e e t me B a~ i i t t a i i t t a eg n v t a a V v t a i c R R i c R R t r r& i h) t A e) t A sa o c w w gB c w w yh tA sB e o o S c i0 i A e o o a A s n0 l & t l l r& t l l o s a F F o s a F F ei o1 PA a l tA a l tD MA 2 G u s r S3 G u s r s 0 c s e 0 c s e ak n-T N e1 t1 e i e l s3 e i t e c l p Wna eS gW E aQ l t c p aQ l yG M sR b r o m GR b r o m sT U n- o a r a - o a r a u x2 R eL N P P S eQ N P P S ne eg O( T dV tM sr S nH sR N o2 . . a2 . . . . au . I C( a b c d V( a b c d GS a 5 6 7 "rE 9 <N E 7 E Z m y
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1 l l l tQ TABLE 3.3-13_(Continued) U ACTION STATEMENTS ACTION 27 - 'With the. number of channels OPERABLE less than required by the j Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release:
- 1. At least two independent samples of the tank's content are analyzed, and at least two technically qualified members of j the Fecility Staff independently verify the release rate 3 calculations and discharge valve lineup, or 1
- 2. Initiate continuous monitoring with a comparable alternate '
monitoring channel. Surveillance requirements applicable to the inoperable channel _shall apply to the comparable ; alternate monitoring channel when used to satisfy this 1 technical specification requirement, i Otherwise, suspend releases of radioactive effluents via this I pathway. 1 ACTION 28 - With the number of channels OPERABLE less than required by the ,i Minimum Channels OPERABLE requirement, effluent releases via i this pathway may continue provided the flow rate is estimated at least once per 4 hours. ( ACTION 29 - With the number of channels OPERABLE less than required by the V Minimum Channels OPERABLE requirement, effluent releases via 3 this pathway may continue provided: j
- 1. Grab samples are taken at least once per 8 hours and these i samples are analyzed for gross activity within 24 hours, or l
- 2. Initiate continuous monitoring with a comparable alternate l monitoring channel. Surveillance requirements applicable !
to the inoperable channel shall apply to the comparable l l alternate monitoring channel when used to satisfy this ' technical specification requirement. l ACTION 30 - With the number of channels OPERABLE less than required by Mini-I mum Channels OPERABLE requirement, immediately suspend PURGING l i i of Reactor Containment via this pathway. ! ACTION 31 - With the number of channels OPERABLE one less than required by the MINIMUM Channels OPERABLE requirement, operation of this system may continue provided grab samples are obtained every 4 hours and analyzed within the following 4 hours during additions to a tank. ACTION 32 - With the number of channels OPERABLE less than required by the , Minimum Channels OPERABLE requirement, effluent releases via ' this pathway may continue provided samples are continuously f ( collected with auxiliary sampling equipment as required in Table 4.11-2 or sampled and analyzed once every 8 hours.
]
ACTION 35 - (This ACTION is not used) BEAVER VALLEY - UNIT 2 3/4 3-69 l
L A N LO EI NT ) NCT ) ANS 1 A 2 A HUE ( / ( / CFT Q N Q Q Q N Q Q S T N E N M O E I R T ) LA I U ER 6 Q NB ( E NI ) ) R AL 3 A 3 A HA ( / ( / E CC R N R R R N R R C N A L L I E V
- R E
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. N NC 4 I AE HH
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I N O M T e N c E i
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F e r ) E t o 8 g o s] r t 0 n r t S y1 o i e1 i o i r U S t n L - r t n o O t i n o aW u s i n M o t E ti M RG i S nn o - a o n A eU M y wR e M y o G V t oF M t M S y i l( y i E sP t v F e t v e e V sV i i e t i i t t I eB v t t c a v t a a T c[ i c ni R m i c R R C o t A ev e) t A A r) c ue w tB c w w 0 PB A e l D o s A e o o I / t f l y& t l l D e& s a f g F S s a F F A t a l E n A a l R sA G u i r n1 G u s r T a8 c mr e o0 c s e N W0 e i eu l i1 e i e l E 1 l t t s p tQ l t c p M s - b r sa m aR b r o m U R uW oG N o P a ye SM S a l - iS N o P a P r S a T e- tV S sM nH N aR . . . e2 . . . . I G( a b c d. V( a b c d 1 2
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& o e r r o V o t r t e r t A o i r r t o i t r g o i r 9 t n o o n t n a o n t n o 0 i o t t e i o R t i i o t 1 n M i i V n M n Q o n n o i d M i R M y o o g w n l o n M y o o i M y o y
t M M n t l M u t M 5 i i y F y V v i B i H t e e d t v e t v e i i t t l i i t t g t 2 v t a a i v t n a n i v i t a ( i c R R u i c e R i i c R t A B) t A u e c w w B c h) t A s A e o o n A e l w o sB c u
- a t l o& t f
f i A e o l e s a l l t l F F i s a E F o s a F l a l tA a l PA a e G u s r a1 G u sr r 2 G l u s T R c s e n0 c so e N e i e 3 e e1 c s E d l t c l p i mQ i t et l p t1 e i. e M e b r o m l r ci aQ l t c U t o a r a aR b on m sR b r o R a N P P t - o a ro a n- o a r m T v S nQ N P PM S eL N P P S S e oM dV
/ cR nH N l . . . e2 E a c o2 .
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)
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- i 3 i l 1 CC D W D D D 3
4 E L _. B A e g T r a h c s i D r o k r t n o i r r a t n o o T i o t t n M i i e o n n g M y o o r t t M M u _ l y i S _ u t v e e ) _ a i i t t m B _ V v t a a e _ i c R R t r& _ e) t A s o gB c w w y tA _ a A e o o S i0 _ r& t l l n0 o s a F F e o1 a t MA _ tA l S3 G u s r s 0 T 0 c s e a n-N s3 e i e c l p W eS gW E aQ l t yG M GR b r o m s U - o a r a u x2 R eQ N P P S o O( T tM e S sR s N a2 . . . . a . I W( a b c d G a 6 7 wM _ g< 9 g. E[ E o r f'
1 i : ( Table 4.3-13 (Continued) I A TABLE NOTATION :.
- During releases via this pathway l
~)uring purging of Reactor Containment via this pathway.
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room Alarm Annunciation occurs if any of.the following conditions exist:
- a. Instrument indicates measured levels above the alarm / trip setpoint.
- b. Downscale failure.
- c. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room Alarm Annunciation occurs if the instrument indicates measured levels above the alarm / trip setpoint. (3) The initial CHANNEL CALIBRATION for radioactivity measurement instru-mentation shall be performed using one or more of the reieeence standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in
.{
measurement assurance activities with NBS. permit calibrating the system over its intended range of Energy and These standards should rate capabilities. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen months. This can normally be accomplished during refueling outages. (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- 1. One volume percent oxygen, balance nitrogen, and
- 2. Four volume percent oxygen, balance nitrogen (5) A source check may be performed utilizing the installed means or flashing the detector with a portable source to obtain an upscale increase in the existing count rate to verify channel response.
(6) The CHANNEL CALIBRA110N shall also demonstrate that Control Room Alarm Annunciation occurs if either of the fcilowing conditions exist:
- 1. Downscale failure.
- 2. Instrument controls are not set in operate mode.
A BEAVER VALLEY - UNIT 2 3/4 3-73
'p3
)
il . l , { INSTRUMENTATION TURBINL-DVERSPEED Pf0TECTION _ ,1,glIINLCQMQlTION FOR OPERATION ,,.3, __
- a. 3.3.4, At least one Turbine Overspeed Protection System shall bs OPERABLE.
, '(
;, , f, APPLICABILITY MODES 1, 2*, and 3*.
l i.
'[','!
N ArTiON:
, n. With one stop n.lve or one governor vah e per high pressure turbine ste.wline ingetable gnd/or with one reheat stop valve or one i rehtat intercept ~ velve per lovi pressure turbine steam line inoperabie, restore the inoperable valve (i) to OPERABLE status within 72 hours, or close at leas', one valve ih the affected steam line(s) cr isolate the turbine from the steam supply wil':in the next 6 hours.
l .g
- b. With de above reouired Turbire Overspeed Protection System otherwise ,
l inoperable, within 5 hours isolate the turbine from the steam supply.
- c. The provisions of Specification 3.0.4 are not applicable.
~
}URVEILLANCILXt0VIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.
d.3.4.2 The above required Turbine Overspeed Protection Syste'm shall be demonstrated OPERABLE: 2 t , s e i L a),~ At least or.ce per 31 days by cycling each of the following valves 1 through at least one complete cycle from the running position: 7 P. 1) Four high pressure turbine stop valves, ( 2) Four high pressure turbina governor valves,
> 3) Four high pressure turbioc reheat stop valves, and
- 4) Four low pressure turbiae reheat intercept valves.
- b. t.t least once per 31 days by direct observation of the movement of each of the above valves through ong complete cycle from the running position,
- c. At least once per 18 months by performing a CHANNEL CALIBRATION on the Turbine Overspeed Protection Systemi, and
- d. At least once per 40 months by disassere.bling at least one of each of the above valves and performing a visail and surface insrection of valve seats, disks, and stems and verifying no unacceptrole flaws or excessive' corrosion. If unacceptable flaws or excessive corrosion are found, all ctrer .vakes of that type shall be inspecLed.
*SpW: ication not aoolicable with all main steam isolatior, valves rnd asso- o clated bypass valves in the closed re,sition and all other stestn fli.wr paths to the turbine isolated.
W - BEAVER VALLEY - UNIT 2 3/4 3-74
rm ) 3/4.4 REACTOR COOLANT SYSTEM
\w/ 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION NORMA 1 OPERATION LIMITING CONDlH ON FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.
APPLICABILITY: MODES 1 and 2. ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours. SEyf]LLANCE RE0VIREMENTS 4.4.1.1.1 All reactor coolant loops shall be verified in operation and circulat-ing reactor coolant at least once per 12 hours. 4.4.1.1.2 The power to each of the Reactor Coolant System loop stop valves shall be verified to be removed at least once per 31 days during operation in , MODES 1 and 2. t O BEAVER VALLEY - UNIT 2 3/4 4-1
l REACTOR COOLANT SYSTEM H0T STANDBY LIMIJJE1_f0@ITION FOR OPERATION 3.4.1.2 a. At least two reactor coolant loops and associated steam generators and reactor coolant pumps shall be in operation
- when the rod control system is capable of control bank rod withdrawal.
- b. At least two reactor coolant loops and associated steam generators and reactor coolant pumps shall be OPERABLE and one reactor coolant loop shall be in operation
- when the rod control system is incapable of control bank rod withdrawal.
APPLICABILITY: MODE 3** ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to vPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. With less than two reactor coolant loops in operation, immediately deenergize all control rod drive mechanisms, or align the rod control system so that it is incapable of control bank rod withdrawal.
- c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.
SufW111 LANCE RE0VIREMENTS 4.4.1.2.1 With the rod control system capable of rod withdrawal, at least two cooling loops shall be verified to be in operation and circulat'ng reactor ' coolant at least once per 12 hours. 4.4.1.2.2 With the rod control system incapable of rod withdrawal, at least two cooling loops, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.3 With the rod control system incapable of rod withdrawal, at least one cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
*All reactor coolant pumps may be deenergized for up to 1 hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration and (2) core outlet temperature is maintained at least 10 F below saturation temperature. This does not preclude natural circulation cooldown under abnormal cooldown conditions.
**See Special Test Exception 3.10.4.
BEAVER VALLEY - UNIT 2 3/4 4-2
L l-g- 'RLACTOR' COOLANT SYSTEM i
- SHUTDOWN LIMITING CONDIl10N F0P OPERATION
- 3.4.1.3 a. At.least two of the coolant loops listed below shall be OPERABLE.
- 1. Reactor Coolant Loop (A) and its associated steam generator and reactor coolant pump,
- 2. Reactor Coolant Loop (B) and its assc:iated steam generator and reactor coolant pump, l 3. Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump, '
- 4. Residual Heat Removal Pump (A) and the (A) RHR heat exchanger,**
- 5. Residual Heat Removal Pump (B) and the (B) RHR heat exchanger.**
- b. At least one of the above coolant loops shall be in operation.***
( APPLICABILITY: MODES 4 and 5. ACTION:
- a. With less than the above required loops OPERABLE, immediately i
initiate corrective action to return the required loops to OPERABLE J status as soon as possible; t>e in COLD SHUTDOWN within 20 hours. !
- b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolar.t System and immediately initiate corrective action to return the required coolant loop to operation. Refer to Specification 3.4.1.6 for additional limitations.
l 1
**The normal or emergency power source may be inoperable in MODE 5.
f i ***All reactor coolant pumps and Residual Heat Removal pumps may be deenergized V for up to 1 hour provided: 1) no operations are permitted that would cause 1 dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10 F below saturation temperature. BEAVER VALLEY - UNIT 2 3/4 4-3
l l l l REACTOR COOLANT SYSTEM SURVEILLANCE RE041REMENTS 4.4.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5, and by verifying that each residual heat removal pump develops a differential pressure of > 126 psid on recirculation flow. 4.4.1.3.2 The required reactor coolant pump (s), if not in operation, shall be detemined to bt OPERABLE once per 7 days by verifying correct breaker align-ments and indicaud power availability. 4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level greater than or equal to 15.5 percent narrow 4 range at least once per 12 hours. 4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. O O BEAVER VALLEY - UNIT 2 3/4 4-4
REACTOR COOLANT SYSTEM 3.4.1.4 (This specification number is not used.) i O 4 l l t O i BEAVER VALLEY - UNIT 2 3/4 4-5
REACTOR COOLANT SYSTEM 3.4.1.5 (This specification number is not used.) ) I l O O BEAVER VALLEY - UNIT 2 3/4 4-6
;f y
Y t' n (
.x)
, ;' ( l REACTOR COOLANT SYSTEM REACTOR COOLANT PUMP-STARTUP Q1111t#LCfdiDLT10fLE0lLQP E R AT1011 3.4.1.61 An idle reactor coolant' pump in a non-isolated loop shall not be.
. started, unless:. the secondary water temperature
- of each steam generator. is
.less than 50 F:above each of the inservice RCS cold leg temperatures.
APPLICAP,ILITYi When' the temperature of one or it. Ore of the non-isolated loop . cold legs is < 350 F. ACTION: With the temperature of the steam generator in the loop associated with the reactor coolant pump being started greater than 50 F above the cold leg tempera-ture of the other non-isolated loops, suspend the startup of the reactor coolant pump. 1UlVfil.1MC.E . REQUIREMENTS
-4 4.1.6.1 The secondary water temperature of the non-isolated steam generators shall be determined within 10 minutes prior to starting a reactor p coolant pump.
I 1 l i
/~n '
(s
*The secondary water temperature is to be verified by direct measurement of the fluid temperature, or contact temperature readings on the steam generator secondary. or blowdevn piping after purging of stagnant water within the piping.
BEAVER VALLEY - UNIT 2 3/4 4-7
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES - SHUTDOWN i 11BlIJBGl0NDITION FOR OPERAIJDIL f 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting
- of 2485 psig i 1 percent.
APPLICABILITY: MODES 4 and 5 ACTION: With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdr>wn cooling mode. SURVEILLANCE RE0VlREMENTS 4.4.2 No additional requirements other than those required by Specification 4.0.5. O { l l
*The lift setting pressure shall correspond to ambient conditions of the valve j
at nominal operating temperature and pressure, l
\
l BEAVER VALLEY - UNIT 2 3/4 4-8 l 1
1
/ REACTOR COOLANT SYSTEM a
3/4.4.3 SAFETY VALVES - OPERATING. 1 l UBillR(LCDHDIll01Lf.0R OPERAI10f1 ,_, , { 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift - setting of 2485 psig i 1%.* j q APPLICABILITY: MODES 1, 2, and 3 ACTION:
- a. With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in :j liOT SHUTDOWN within 12 hours.
1 l 1UMUllaMCLREQUlREtiENTS ) 4.4.3 No additional requirements other than those required by Specification 4.0.5. 1 I O i 1 i l ['j\
*The lift setting shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
l j BEAVER VALLEY - UNIT 2 3/4 4-9 l
l 1 REACTOR COOLANT SYSTEM 3/4.4.4 PRESSURIZER LIMURG.10M1110tLfDILDEERATION 3.4.4 The pressurizer shall be OPERABLE with at it:ast 150 kW of pressurizer heaters and with a steam bubble. APPLICABILITY: MODES 1, 2, and 3 l l ACTION: With the pressurizer inoperable due to less than 150 kW of heaters supplied by an emergency bus, be in at least HOT STANDBY within the next 6 hours and in HOT SHUT 00WN within the'following 12 hours. With the pressurizer otherwise inoperable, be in at least H0T STANDBY with the reactor trip breakers open within 6 hours and in the HOT SHUTDOWN within the following 6 hours. SUMEILLMCLEEg{1JEMff4TS 4.4.4.1 The power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by energizing the heaters supplied by the emergency bus, i 1 ! ) i i O BEAVER VALLEY - UNIT 2 3/4 4-10
('] - REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIH1IINGl0ffTION FOR DPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: With one or more steam generators inoperable, restore the inoperable generator (s)- to OPERABLE status prior to increasing T avg above 200 F. SURVEILLANCE REQUIRE! R TS 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sainple Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The in-service inspection of steam generator tubes shall be performed at the frequen-Q cies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Steam generator tubos shall be examined in accordance with the method prescribed in Artcle 8 -
" Eddy Current. Examination of Tubular Products," as contained in ASME Boiler and Pressure Vessel Code, Section V "Non-destructive Examination," and referenced in ASME Boiler and Pressure Vessel Code - Appendix IV of the 1980 Edition through Winter 1980 Addenda of Section XI " Inservice Inspection of Nuclear Power Plant 3 Components." The tubes selected for each inservice inspection shall include at l least 3% of the total number of tubes in all steam generators; the tubes selected f for these inspections shall be selected on a random basis except:
l a. Where experience in similar plants with similar water chemistry indicates l critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas; l b. The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
- 1) All nonplugged tubes that previously had detectable wall penetrations (> than 20%),
- 2) Tubes in those areas where experience has indicated potential problems.
I BEAVER VALLEY - UNIT 2 3/4 4-11 I i w__- - - 1
REACTOR COOLANT SYSTEM SilRVEILLANCE RE0VlREMENTS (Continued)
- c. The second and third inservice inspections may be less than a full tube inspection by concentrating (selecting at least 50% of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.
The results of each sample inspection shall be classified into one of the j following three categories: Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubec defective. C-2 One or more tubes, but not more than 1 percent of the total tubes inspected are defective, or between 5 percent and 10 percent of the total tubes inspected are degraded tubes. C-3 More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (>10 percent) further wall penetrations to be included in the above percentage calculations. 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies.
- a. The firct inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under All Volatile Treatment (AVT) conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
- b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose i results fall in Category C-3, the inspection frequency shall be in- l creased at least once per 20 months. The increase in the inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.
BEAVER VALLEY - UNIT 2 3/4 4-12
- __________a
REACTOR COOLANT SYSTEM V SURVEILLANCE RE0VIREMENTS (Continued) i c. Additional, unscheduled inservice inspections shall be performed on l each steam generator in accordance with the first sample inspection i specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions.
- 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,
- 2. A seismic occurrence greater than the Operating Basis Earthquake,
- 3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or
- 4. A main steam line or feedwater line break.
l 4.4.5.4 Acceptance Criteria
- a. As used in this Specification:
- 1. Imperfection means an exception to the dimensions, finish or
/
( contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.
- 2. Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube.
t
- 3. Degraded Tube means a tube containing imperfections > 20 percent i of the nominal wall thickness caused by degradation.
- 4. % Degradation means the percentage of the tube wall thickness affected or removed by degradation.
- 5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.
Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube. l
- 6. Plugging Limit means the imperfection depth at or beyond which '
the tube shall be removed from service because it may become un-serviceable prior to the next inspection and is equal to 40 per-cent of the nominal tube wall thickness. v l l BEAVER VALLEY - UNIT 2 3/4 4-13
l REACTOR COOLANT SYSTEM SURVEILLANCE RE001REMENTS (Cantinued)
- 7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above. 1 i
- 8. ' Tube Inspection means an inspection of the steam generator tube !
from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reports
- a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
- b. The complete results of the steam generator tube inservice inspection shall be included in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This report shall include:
- 1. Number and extent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection.
- 3. Identification of tubes plagged.
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. The written report shall provide a description of investigations con-ducted to cetermine cause of the tube degradation and corrective measures taken to prevent recurrence.
O BEAVER VALLEY - :lT 2 3/4 4-14
l 1 J [^N TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Three Three First Inservice Inspection All Two Second & Subsequent Inservice Inspections One1 One2 Table Notation
- 1. The inservice inspection may be limited to one steam generator on a $
rotating schedule encompassing 9 % of the tubes if the results of the ! first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating condi more severe tha,tions in in n those one or more other steamsteam generators may be found to be generators. n Under such circum-( stances the sample sequence shall be modified to inspect the most severe conditions. '
- 2. The other steam generator not inspected during the first inservice inspec-tion shall be inspected. The third and subsequent inspections should follow the instruction described in 1 above.
I i f'\ U BEAVER VALLEY - UNIT 2 3/4 4-15
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!/ REACTOR COOLANT SYSTEM V 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- a. The containment atmosphere particulate radioactivity monitoring system, l
- b. The containment sump discharge flow measurement system or narrow range level instrument, and
- c. Containment atmosphere gaseous radioactivity monitoring system.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the above required radioactivity monitoring leakage
. g* detection systems inoperable, operations may continue for up to 30 days provided:
qd
- 1. The other two ;ve required leakage detection systems are OPERABLE, and
- 2. Appropriate grab samples are obtained and analyzed at least once per 24 hours:
1 otherwise, be in at least HOT STANDBY within the next 6 hours and in { COLD SHUTDOWN within the following 30 hours. l
- b. With the containment sump discharge flow measurement system and narrow range level instrument inoperable, restore at least one inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. The provisiens of Specification 3.0.4 are not applicable in MODES 1, 2, and 3.
SURVEILLANCE RE0VIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: p} U
- a. Containment atmosphere particulate and gaseous monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, BEAVER VALLEY - UNIT 2 3/4 4-17
i REACTOR COOLANT SYSTEM i SURYflLLMCflE@lREMENTS (Continued) ,
- b. Containment sump discharge flow measurement system performance of CHANNEL CALIBRATION at least once per 18 months.
- c. Logging the narrow range level indication every 12 hours i
O l l 1 \ l 1 1 O f i f BEAVER VALLEY - UNIT 2 3/4 4-18
i l
.g
/ R_EACTOR COOLANT SYSTEM \ OPERATIONAL LEAKAGE LIMITINGCONDITI0gl0ROPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE B0UNDARY LEAKAGE,
- b. 1 GPM UNIDENTIFIED LEAKAGE,
- c. 1 GPM total-reactor-to secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam' generator not isolated from the Reactor Coolant' System,
- d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
- e. 28 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the next 30 hours.
b, With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
- a. Monitoring the containment atmosphere particulate and gaseous radioactivity monitor at least once per 12 hours.
I
- b. Monitoring the containment sump discharge at least once per 12 hours.
- c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve full open.
s
- d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours during steady state operation, and BEAVER VALLEY - UNIT 2 3/4 4-19
REACTOR COOLANT SYSTEM jjjjVEILLANCE RE0VIREMENTS (Continued)
- e. Monitoring the reactor head flange leakoff temperature at least once per 24 hours. 1 l
O l l O BEAVER VALLEY - UNIT 2 3/4 4-20
l I i / ] REACTOR COOLANT SYST3 ' l PRESSURE ISDTION VALVES ) l i LIMPING CON _DITIO:.! FOR OPERATION 3.4.6.3 Reactor Coolant System pressure isolation valves as shown in .I
. Table 4.4-3 shall be OPERABLE. i APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
- 1. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit stated in Table 4.4-3, isolate the high i pressure portion of the affected system from the low pressure portion within 4 hours by use of a closed manual or deactivated automatic valve, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.*
- 2. The provision of Specification 4.0.4 is not applicable for entry into MODE 3 or 4.
[ SURVEILLANCE RE0VIREMENTS
%/
4.4.6.3.1 Leakage testing ** of each valve listed in Table 4.4-3 shall be accomplished prior to entering MODE 2 after every time the plant is placed in the COLD SHUTDOWN condition for refueling and prior to returning the valve to service after each maintenance, repair or replacement work is performed; and 4.4.6.3.2 Additional leakage testing of each valve identified by note (d) listed in Table 4.4-3 shall be accomplished prior to entering MODE 2 after each time the plant is placed in COLD SHUTDOWN for 72 hours if testing has not been accomplished in the preceding 9 months. j
- Motor operated valves shall be placed in the closed position and power supplies de-energized.
** To satisfy ALARA requirements, leakage may be measured indirectly (as from the perfoniance of pressure indicators) if accomplished in accordance with p approved procedures and supported by computations showing that the method is capable of demonstrating compliance within the valve leakage criteria.
BEAVER VALLEY - UNIT 2 3/4 4-21
TABLE 4.4-3 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Leakage Rates (a) System Valve No. A"owable/ Maximum Loop 21, Cold leg, LHSI 2 SIS-107 5 3.0/5 5.0 gpm(b)(d) Loop 22, Cold leg, LHSI 25I5-108 5 3.0/5 5.0 gpm(b)(d) Loop 23, Cold 1.eg, LHSI 2S15-109 < 3.0/< 5.0 gpm(b)(d) Common, Cold leg, LHSI 2515-132 < 5.0/< 5.0 gpm(d) 2 SIS-133 < 5.0/< 5.0 gpm(d) Loop 22, Hot leg, LHSI 2 SIS-128 < 3.0/< 5.0 gpm(b) Loop 23, Hot leg, LHSI 2515-129 5 3.0/5 5.0 gpm(b) Common, Hot leg, LHSI 2 SIS-130 < 5.0/< 5.0 gpm Loop 21, Cold leg, SIACC 2 SIS-151 5 5.0/5 5.0 gpm(b) 25I5-148 5 5.0/5 5.0 gpm Loop 22, Cold leg, SIACC 2 SIS-145 5 5.0/5 5.0 gpm(b)(d) 2S15-147 < 5.0/< 5.0 gpm Loop 23, Cold leg, SIACC 2 SIS-141 5 5.0/5 5.0 gpm(b)(d) 2515-142 5 5.0/5 5.0 gpm Loop 21, Hot leg, RHS-A 2RHS-MOV702A < 5.0/< 5.0 gpm(b) 2RHS-MOV701A < 5.0/< 5.0 gpm(b) Loop 22, Cold leg 2RHS-MOV720A 5 5.0/5 5.0 gpm(b)(c) Loop 21, Hot leg, RHS-B 2RHS-MOV702B < 5.0/< 5.0 gpm(b) 2RHS-MOV701B < 5.0/< 5.0 gpm(b) Loop 23, Cold leg 2RHS-MOV720B 5 5.0/5 5.0 gpm(b)(c) (a) At function pressure:
- 1. Leakage rates less than or equal to 0.5 gpm/ inch diameter are acceptable.
- 2. Leakage rates greater than 0.5 gpm/ inch diameter but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50 percent or greater.
BEAVER VALLEY - UNIT 2 3/4 4-22
s 'l
/ 3. Leakage rates greater than 0.5 gpm/ inch diameter but less than or equal
.('- to 5.0 gpm are considered unacceptable if the latest measured rate ex-ceeded the rate determined by the previous test by an amount that re- 1 duces the margin between measurea leakage rate and the maximum permis-sible rate of 5.0 gpm by 50 percent or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable. !
- 5. Observed leakage rates shall be adjusted to the function maximum -!
pressure in accordance with ASME XI IWV 3423. (b) Minw.:a test differential pressures shall not be less than 150 psid. (c) Leakage rate continuously monitored during plant operatior,, no other leakage
-rate testing required. Leakage rate acceptance criteria shall be as st6ted in (a) and (b) above and shall be recorded at intervals as noted in para-graph 4.4.6.3.1 as a minimum.
(d) Both surveillance 4.4.6.3.1 and 4.4.6.3.2 are required. t r ( 1 1 1 l BEAVER VALLEY - UNIT 2 3/4 4-23 l 1
REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITl0N FOR OPERATl0N 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1. APPLICABILITY., At all times. ACTION.. MODES 1, 2, 3, and 4
- a. With any one or more chemistry parameters in excess of its Steady State Limit but within its Transient Limit, restore the Parameter to within its Steady State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
At all other times With the concentration of either chloride or fluoride in the Reactcr ' Coolant System in excess of its Steady State Limit for more than 24 hours ' or in excess of its Transient Limit, reduce the pressurizer pressure to 5 500 psig, if applicable, and perform an analysis to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4. SURVE11 LANCE RE0VlREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within l the limits by analysis of those parameters at the frequencies specified in Table 4.4-10. l O' , BEAVER VALLEY - UNIT 2 3/4 4-24 i
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\ REACTOR COOLANT SYSTEM 3 s s.: CHEMISTRY LIMITS
?< 1 x i STEADY-STATE- TRANSIENT )
PARAMETER LIMIT LIMIT DISSOLVEDOXYGhllA < 0.10 ppm * $ 1.00 ppm ^ 1 g' CHLORIDE < 0.15 ppm - 5 1.50 ppm l FLUORIDE 5 0.15 ppm 5 1.50 ppm a (p
~'
it i t 4 e ' h g
- Limit not applicable wito T avg 5 250 F.
f I BEAVER VALLEY - UNIT 2 3/4 4-25 . - _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ z
P 1
,l[ TABLE 4.4-10 m > A PE_AC10R COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS MINIMUM MAXIMUM TIME CONTAMINANT SAMPLING FREQUENCIES BETWEEN SAMPLES DISSOLVED DXYGEN 3 times per 7 days
- 72 hours CHLORIDE 3 times per 7 days 72 hours FLUORIDE 3 times per 7 days 72 hours l' :i.
i - ,
*Not required with T avg 5 250 F.
O s O BEAVER VALLEY - UNIT 2' 3/4 4-26
(~'N REACTOR COOLANT SYSTEM
!l 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDIT101LfDR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
- a. 5 1.0 pCi/ gram DOSE EQUIVALENT I-131, and
- b. 5 100 /E pCi/ gram APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION: H0 DES 1, 2 and 3*:
- a. With the specific activity of the primary coolant > 1,0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T < 500 F within 6 hours, avg
- b. With the specific activity of the primary coolant > 100 /E pCi/ gram, be in HOT STANDBY with T avg < 500 F within 6 hours.
C' MODES 1, 2, 3, 4, and 5
- a. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100 /E pCi/ gram, perform the sampling analysis requirement of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored to within its limits.
SURVEILLANCE RE0HIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the performance limits of the sampling and analysis program of Table 4.4-12.
/
(\
*With T avg 1 500 F.
BEAVER VALLEY - UNIT 2 3/4 4-27
[ , 3 ) i
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TABLE 4.4-12 I : PRIMARY C0OLANT SPECIFIC ACTIVITY SAMPLE I ~ AND ANALYSIS PROGRAM ~~ [ 4 TVN OF MEMUi<b4ENT MINIMUM , i MODES IN WHICH l' _AND AMAli;/S- FREQUENCY ' SURVEILLANCE REQUlr<ED
- 1. Gross Activity 3 times per(7 daysi 1,2,3,4 Determination with a maxi.num time of 72 houritbetwesn samples. *
]
- 2. Isotopic Ana?ysis for 1 per 14 days 1, DOSE EQUIVE,ti.NT I-131, 9* Concentrate 6n 1,
~
- 3. Radiochemical for E 1 per 6 months CeO!rmination-
- 4. fsotopic AC.alysis for a)' Once per 4 hours, 1#,2#,3#,4#, 5#
, Iodine inc M ing I-13] whenever the
't 1-133, and I-135 spedlic activity exceeds 1.0 pCi/ gram DOSE EQUIVALENT I-131 or 100</E pCi/gn,m, and b) One sample between 1, 2, 3 2 and 6 hours follow-ing a THERMAL POWER change exceeding 15 percent of the Rt.TED THERMAL P0'4ER within a 1-hv i period.
4 k y r
~' #Until;the specific act.ivity of the primary coolant systen h restored to O
within its lirrfts.
, BEAVER VALLEY - (4 NIT 2 3/4 4-28
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- i 4 i i . 1 I i 6 i . 6 . i i i e a i 6 20 30 40 50 00 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 N
DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL MWER with the Primary Coolant Specific Activity > 1.0 pCi/ gram Dose Equivalent I-131 BEAVER VALLEY - UNIT 2 3/4 4-29
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM _LIMLUNG_CONDlIl0N FOR_0PERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and j pressure shall be limited in accordance with the limit lines shown on i Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with
- a. A maximum heatup of 60 F in any 1-hour period,
- b. A maximum cooldown of 100 F in any 1-hour period, and i i
- c. A maximum temperature change of < 5 F in any 1-bour period during hydrostatic testing operations above system design pressure.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an analysis to determine the ( effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T avg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours. MRVflLLANCE RE0VlREMENTS 4.4.9.1
- a. The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
l
- b. The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.
- c. The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5. The results of these examina-tions shall be used to update Figures 3.4-2 and 3.4-3.
O 4 BEAVER VALLEY - UNIT 2 3/4 4-30 i
M A TERI AL - P RO P ERTY-- B A SIS [\ k Controlling Material Copper Content
- Plate Metal
- Conservatively Assumed to be 0.10 wt%
Phosphorus' Content. : 0.010 wt% RTNDT Initial : 600F RTNDT Af ter 10 EFPY : 1/4T,1390F 3 /4 T, 114*F 3000 , , CURVE APPLICABLE FOR HEATUP I RATES UP TO 60*F/HR FOR THE
. SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.
LEAK TEST LIMIT b b l 6.2000 UNACCEPTABLE OPERATION ACCEPTABLE OPERATION W E D (f) Cf) O t '$ 1500 a NE ATUP CURVli
/
O W l-
< J 1000 e
/ y CRITICALITY LIMIT BASED ON INSERVICE HYDROSTATIC TEST TEMPER ATURE (279eF) 500 FOR THE SERVICE -
PERIOD UP TO 10 EPPY. O I
- 0. 100 200 300 400 500 1 I
INDIC ATED TEMPER ATURE (DEG.F) FIGURE 3.4-2 BEAVER VALLEY UNIT 2 REACTOR C30LANT SYSTEM HEATUP O- LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY BEAVER VALLEY - UNIT 2 3/4 4-31
l M ATERIAL PROPERTY B ASIS Controlling Material : Plate Metal Copper Content Conservative 41y Assumed to be 0.10 wt% Phosphorus Conter.t 0.010 wt% RT NDT InHial
. 600F
- 1/ 4 T,13 90F RTNDT Af ter 10 EFPY 3/ 4 T. 1 14 *F
'3000 , .
I CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100eF/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 2500 m 2000 UNACCEPTABLE ACCEPTABLE v OPER ATION O PER ATION W C D U) C 1500 Q. O W H 9 1000 Q 3 - [ r COOLDOWN RATES (oF/HR) 500 0 -- 100 # l 0 0 100 200 300 400 500 INDIC ATED TEMPER ATURE (DEG.F) FIGURE 3.4-3 BEAVER VALLEY UNIT N0. 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY BEAVER VALLEY - UNIT 2 3/4 4-32
,S ,- TABLE 4.4-5 \_ REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE Estimated Vessel Lead Withdrawal Capsule Fluence Capsule Location Factor Time (EFPY) (n/cm2 )
U 343 3. 5 1st Refueling 0.8 x 1019 V 107 3.5 3 2.13 x 1019 X 287 3.5 6 4.26 x 1019(a) W 110 2.9 11 6.48 x 1019(b) Y 290 2.9 20 11.77 x 1019 Z 340 2.9 Standby -- (a) Approximate fluence at 1/4 T vessel wall thickness at end-of-life. (b) Approximate fluence at vessel inner wall at end-of-life, b v O b BEAVER VALLEY - UNIT 2 3/4 4-33
REACTOR COOLANT SYSTEM PRESSURIZER LIMll1NG_t0EDlIl0E_10R OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup of 100 F in any 1-hour period,
- b. A maximum cooldown of 200 F in any 1-hour period, and
- c. A maximum auxiliary spray water temperature differential of 625 F.
APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an analysis to determine the of fects of the out-of-limit condltion on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. SUMf1LLANCE REQUIREMENTS ___ O 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. i i 1 l l l O BEAVER VALLEY - UNIT 2 3/4 4-34
.i l 1 l s
[N) REACTOR COOLANT SYSTEM U OVERPRESSURE PROTECTION SYSTEM.S. UMIIING CONDIT10llf.OR OPERATION , 3.4.9.3 At-least one of the following Overpressure Protection Systems (OPPS) shall be OPERABLE:
- a. Two power-operated relief valves (PORVs) with nominal maximum allow- i able lift settings which vary with the RCS temperature and which do i not exceed the 1 mits established in FIGURE 3.4-4, or
- b. A Reactor Coalant System vent of > 3.14 square inches.
APPLICABILITY: When the temperature of one or more of the non-isolated RCS coldlegs.isf350F. ACTION:
- a. With one PORV inoperable, either restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent (s) within the next 12 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status. Refer to Technical Specification 3.4.1.6 for further limitations.
- b. With both PORVs inoperable, depressurize and vent the RCS through !
a 3.14 square inch vent (s) within 12 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status,
- c. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report snall describe the circumstances initiating the transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable. ;
EURYGLLANCE REQUIREMENTS , 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE BY:
- a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, I but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel Os. at least once per 18 months.
BEAVER VALLEY - UNIT 2 3/4 4-35
REACTOR COOLANT SYSTEM SEVEILLANCE RE0lllRE.EENTS(continued)
- c. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure pr>tection. ,
1
- d. Stroking the operable PORV(s) each time the plant enters MODE 5, unless tested within the preceding 3 months. ,
4.4.9.3.2 The > 3.14 square inch RCS vent (s) shall be verified to be coen at least once per 12 hours
- when the vent (s) is being used for overpressure.
protection. 9 l 1
- Except when the vent pathway is provided with a valve which is locked, or provided with remote position indication, sealed, or otherwise secured in the open position, then verify these valves open at least once per 7 days.
BEAVER VALLEY - UNIl 2 3/4 4-36 l 1
- f~~)
( )' 750 0 3 f O ] g 700 E o N u 650 l E i o UNACCEPTABLE ' ACCEPTABLE Q- Ol'E R ATION OP ERATI DN 3 i m 600 it; o d
< J 550 i
3 /
/
500-j E 3 Q z
' 450 X
X cn a. 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400 TRTC-AUCTIONEERED LOW-MEASURED RCS TEMPERATURE ('F) O g Figure 3.4-4 Maximum Allowable Nominal PORV Setpoint For The Overpressure Protection System BEAVER VALLEY - UNIT 2 3/4 4-37
REACTOR COOLANT SYSTEM 3/4.4.10- STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2, AND 3 COMPONENTS LIMITING COROJTION F0P,0PERATIOR , 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components j shall be maintained in accordance with Specification 4.4.10. I APPLICABILITY: All MODES. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum tempera- i ture required by NDT considerations,
- b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F.
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected cortponent(s) from service,
- d. The provisions of Specification 3.0.4 are not applicable.
SUMEILLANCLEEQUlREMENTS 4.4.10 Each ASME Code Class 1, 2, and 3 component shall be demonstrated OPERABLE in accordance with Specification 4.0.5. l l l i BEAVER VALLEY - UNIT 2 3/4 4-38
l i l
- l. )
. (] REACTOR COOLANT' SYSTEM 3/4.4.11 RELIEF VALVES
{; LIMITING _ CONDITION FOR OPERAIION i 3.4.11 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. 1 APPLICABILITY: MODES 1, 2, and 3 ACTION:
- a. With one or more PORV(s) inoperable, because of excessive seat !
leakage, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within'the next 6 hours and in COLD SHUTDOWN within the following 30 hours. i
- b. With one or two PORV(s) inoperable as a result of causes other than excessive seat leakage, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s). .l
- c. With all three PORVs inoperable due to causes other than excessive seat leakage, within 72 hours either restore one PORV to OPERABLE V[_] status or be in HOT STANDBY within the next 6 hours rad COLD SHUTDOWN within the following 30 hours.
- d. With one or more block valve (s) inoperable, within 1 hour: (1) re-store the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV; and (2) apply the ACTION b. or c. above, as appropriate, for the isolated PORV(s),
- e. The provisions of Specification 3.0.4 are not applicable.
SJJRV.IILLAN.CE_RE@lREENIS 4.4.11.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:
- a. Performance of a CHANNEL CALIBRATION, and '
- b. Operating the valve through one complete cycle of full travel.
4.4.11.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. , c. or d. in Specification 3.4.11. p b BEAVER VALLEY - UNIT 2 3/4 4-39
R_EACTOR COOLANT SYSTEM REACTOR C00LANi SYSTEM HEAD VENTS MMITING CONDITIM_E0B_0fEILAll01l _m 3.4.12 All Reactor Coolant System head vent valves, powered from emergency buses shall be OPERABLE
- and closed ** for each of the reactor vessel head vent paths.
APPLICABILITY: MODES 1, 2, 3, and 4 , ACTION:
- a. With at least one vent path from the above location OPERABLE and one or more power operated vent valves inoperable, STARTUP and/or POWER 4 OPERATION may continue provided the inoperable valve (s) is maintained closed with power removed. Power operation may continue until the next scheduled outage, at which time all Reactor Coolant System head vent valves shall be OPERABLE prior to e ry into MODE 1. The provi- l sions of Specification 3.0.4 are not applicable,
- b. With all vent paths from the above location inoperable maintain the inoperable valves closed with power removed or close the manual isola-tion valves, and restore at least one vent path from the above locations to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLMCLEEQUIREMENTJ 4.4.12 Each Reactor Coolant System head vent path shall be demonstrated OPER-ABLE at least once per 18 months by:
- 1. Verifying the manual isolation valve in the vent path is locked or sealed in the open position.
l
- 2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room.
- 3. Verifying flow through the Reactor Coolant System Head vent path to the Pressurizer Relief Tank.
l l *For purposes of this Specification, an inoperable vent valve is defined as: a valve which exhibits leakage in excess of Specification 3 4.6.2 limits, or cannot be opened and closed on demand, or does not have its normal emergency power supply OPERABLE.
**These valves may be operated for required venting operations and leak testing in MODES 3 and 4.
BEAVER VALLEY - UNIT 2 3/4 4-40
i L
/
O 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:
- a. The isolation valve open,
- b. Between 7532 and 7802 gallons of borated water, l
- c. Between 1900 and 2100 ppm of boron, and
- d. A nitrogen cover pressure of between 585 and 665 psig.
APPLICABILITY: MODES 1, 2 and 3.* ACTION:
- a. With one accumulator inoperable, except as a result of a closed isola-tion valve, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours g and in HOT SHUTDOWN within the following 6 hours.
- b. With one accumulator inoperable due to the isolation valve being 4 closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours.
)i i
SURVflLLANCEEQUIREMENTS ! 4.5.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1. Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and
- 2. Verifying that each accumulator isolation valve is open.
- b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution.
- Pressurizer Pressure above 1000 psig.
BEAVER VALLEY - UNIT 2 3/4 5-1
I EMERGENCY CORE COOLING SYSTEMS , i SURVEILLANCE REQUIREMENTS (Continued)
- c. At least once per 31 days when the RCS pressure is above 1000
} psig by verifying that power to the isolation valve operator l control circuit is disconnected by removal of the plug in the lock out jack from the circuit.
- d. At least once per 18 months by verifying that each accumulator isolation valve opens automatically under each of the following J conditions:
- 1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection)
Setpoint, and
- 2) Upon receipt of a Safety Injection test signal.
4.5.1.2 Each accumulator water level and pressure alarm channel shall be demonstrated OPERABLE:
- a. At least once per 31 days by the performance of a CHANNEL FUNCTIONAL TEST.
- b. At least once per 18 months by the performance of a CHANNEL ,
CALIBRATION. 4.5.1.3 During normal plant cooldown and depressurization, each accumulator discharge isolation valve 2 SIS-MOV 865 A, B and C shall be verified to be closed and de-energized when RCS pressure is reduced to 1,000 1 100 psig. l l l l l l l BEAVER VALLEY - UNIT 2 3/4 5-2
.1 i
- n. fMERGENCY' CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tava > 350 F LIMITING CONDITION FOR OPERATION l
3.5.2 Two separate and independent ECCS subsystems shall be OPERABLE with each ; subsystem comprised of:
- a. One OPERABLE centrifugal charging pump, i
- b. One OPERABLE low head safety injection pump, and "
I
- c. One OPERABLE recirculation spray pump
- capable of supplying the safety injection flow path during recirculation phase, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of .;
operation. ' APPLICABILITY: MODES 1, 2 and 3.** 1 l n ACTION: !
- a. With one ECCS subsysten inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in HOT SHUTOOWN within the next-12 hours.
I
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days !
describing the circumstances of the actuation and the total accumu-lated actuation cycles to date.
$#fWElLLANCE RE0UIREMENTS
.)
4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: ! l
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operator l control circuits disconnected by removal of the plug in the lock out circuit from each circuit:
]
^ Recirculation spray pump 2RSS-P21C or 2RSS-P210.
**The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pumps declared inoperable pursuant to Specification 4.5.3.2 provided the centrifugal charging pumps are restored 1 y/ to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first. j BEAVER VALLEY - UNIT 2 3/4 5-3
1 EMERGENCY CORE COOLING SYSTEMS Valve Number Valve Function Valve Position O ! I
- a. 2 SIS-MOV 8889 LHSI to hot legs Closed
- b. 2 SIS-MOV 869A HHSI to hot leg Closed
- c. 2 SIS-MOV 869B HHSI to hot leg Closed
- d. 2 SIS-MOV 841 HHSI to cold leg Open
- e. 2CHS-MOV 8132A HHSI pump disch x-conn Open
- f. 2CHS-MOV 81328 HHSI pump disch x-conn Open
- g. 2CHS-MOV 8133A HHSI pump disch x-conn Open
- h. 2CHS-MOV 8133B HHSI pump disch x-conn Open
- b. By verifying that each of the following pumps develop the required differential pressure on recirculation flow when tested pursuant to Specification 4.0.5.
1.) Centrifugal charging pump > 2437 psid 2.) Low head safety injection pump, [103psid
- c. At least once per 31 days by:
- 1) Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- 2) Verifying that each ECCS subsystem is aligned to receive elec-trical power from separate OPERABLE emergency huses.
I O 1 BEAVER VALLEY - UNIT 2 3/4 5-4 l l
1 I EMERGENCY CORE COOLING SYSTEMS e 1
$UAVEILLANCE RE0VIREMENTS (Continued)
- d. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performec:
- 1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and
- 2. Of the areas affected within containment at the completion of ;
each containment entry when CONTAINMENT INTEGRITY is established.
- e. At least once per 18 months by:
- 1. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted.by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.
- f. At least once per 18 months, during shutdown, by:
'd 1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel. !
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection signal.
- 3. Verifying that the centrifugal charging pump and low head safety injection pumps start automatically upon receipt of a safety injection signal.
- g. The containment recirculation spray subsystem shall be demonstrated OPERABLE per the applicable portions of Specification 4.6.2.2.
I 1 A BEAVER VALLEY - UNIT 2 3/4 5-5 l
f EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T < 350 F av0 LIMITlHG CONDITION FOR OfERaI10N 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be ! OPERABLE:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE Low Head Safety Injection Pump, and
- c. One OPERABLE recirculation spray pump
- capable of supplying the safety injection flow path during recirculation phase, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pomp or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE ,
status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days describing the circumstances of the actuation and the total ,
accumulated actuation cycle to date, jjjRVEILLANCE RE0Q1REMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2. 4.5.3.2 All charging pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable ** by verifying that the control switches are plaCEd in the PULL-TO-LOCK position and tagged within 4 hours after entering j t MODE 4 from MODE 3 prior to the temperature of one or more of the RCS cold legs decreasing below 325 F, whichever comes first, and at least once per 12 hours thereafter.
- Recirculation spray pump 2RSS-P21C or 2RSS-P21D. l
**An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with '
l power removed from the valve operator, or by a manual isolation valve secured in the closed position. BEAVER VALLEY - UNIT 2 3/4 5-6
i [%j' , 3/4.6 CONTAINMENT SYSTEMS-3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY I l LIMITING CONDITION FOR OPERAIl0N i 3.6.1.1 Primary CONTAINMENT. INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN , within the following 36 hours. I 4 SURVEILLANCE RE0VIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that: !
i
- 1. All penetrations
- not capable of being closed by OPERABLE contain-ment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or i deactivated automatic valves secured in their positions, except l as provided in Table 3.6-1 of Specification 3.6.3.1.
- 2. All equipment hatches are closed and sealed,
- b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.
*Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
BEAVER VALLEY - UNIT 2 3/4 6-1
I CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to: ,
- a. An overall integrated leakage rate of < L , 0.10 percent by weight of the containment air per 24 hours at P a (44.7 psig).
- b. A combined leakage rate of < 0.60aL for all penetrations and ,
valves subject to Type B and C tests as identified in Table 3.6-1, when pressurized to Pa (44.7 psig). APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L8 , or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L . restore the leakage rate (s) to within the 1.imit(s) prior to increasing he Reactor Coolant System temperature above 200 F. S)REliLANCE_RE0VlREMENTS - _ 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria ' specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972:
- a. A Type-A test (0verall Integrated Containment Leakage Rate) shall be conducted at 40 1 10 month intervals during shutdown at P a (44.7 psig).
- b. If any Periodic Type A test fails to meet 0.75 L a, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commissien. If two consecutive Type A tests fail to meet 0.75 L a, l a Type A test shall be performed at least every 18 months until two I consecutive Type A tests meet 0.75 aL at which time the above test l schedule may be resumed.
I l e I j
I, ,\ CONTAINMENT SYSTEMS 1 J SURVEILLANCE REQUIREMENTS (Continued)
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the Type A test by verifying that the difference between supplemental and Type A test data is within 0.25 La '
- 2. Has a duration sufficient to accurately establish the change in leakage rate between the Type A test and the supplemental test.
- 3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage rate at Pa (44'7 IUi9)'
- d. Type B and C tests shall be conducted with gas at Pa (44.7 Psig) at intervals no greater than 24 months except for tests involving:
7s 1. Air locks,
- 2. Penetrations using continuous leakage monitoring systems, and
- 3. Valves pressurized with fluid from a seal system.
- e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
- f. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the Provisions of Appendix J, Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressur. zed to at least 1.10 P a
(49.2 psig) and the seal system capacity is adequate to maintain system pressure for at least 30 days.
- g. All test leakage rates shall be c 'culxted using observed data con-verted to absolute values. Error analyses shall be performed to J determine the inaccuracy of the measured leakage rates due to maximum l measurement accuracy and instrument repeatability; the measured j leakage rates shall be adjusted to include the measurement error.
- Applicable valves may be tested using water as the pressure fluid in accordance with the Inservice Testing Program l BEAVER VALLEY - UNIT 2 3/4 6-3 l 4
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMlll11LC08DlDDlif") OPERATJON 3.6.1.3 Each containment air lock shall be OPERABLE with: 3 I
- a. Both doors closed except when the air lock is being used for normal l transit entry and exit through the containment, then at least one ;
air lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 La at Pa (44.7 psig).
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain the associated OPERABLE air-lock door closed and either restore the associated inoperable air lock door to OPERABLE status within 24 hours or lock the associated OPERABLE air lock door closed.
- 2. Operation may then continue until performance of the next requirt.d overall air lock leakage test provided that the associated OPERABLE air lock door is verified to be locked closed at least once per 31 days.
- 3. 'itherwise, Le in at least HOT STANDBY within the next 6 hours and in COLD 5HUTDOWN within the following 30 hours.
- 4. The provisions of Specification 3.0.4 are not applicable.
1
- b. With a containment air lock inoperable, except as a result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
j O 1 BEAVER VALLEY - UNIT 2 3/4 6-4 _ _ _ _ _ __ __ _A
[ l i CONTAINMENT SYSTEMS I jpRVEILLANCERE0VIREMENTS l 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE: i
- a. Within 72 hours following each containment entry, except when-the air lock is being used for multiple entries, then at least once per 72 hours, by verifying no detectable seal leakage when the gap between the door seals is pressurized for at least 2 minutes to:
- 1. Personnel airlock 1 44.7 psig
- 2. Emergency air lock 2 10.0 psig
)
or, by quantifying the total-air lock leakage to insure the requirements of 3.6.1.3.b are met.
- b. By conducting overall air lock leakage tests, at not less than P (44.7 psig), and verifying the overall air lock leakage rate is a within its limit:
- 1. At least once per 6 months, # and
; 2. Upon completion of maintenance which has been performed on_the
( air lock that could affect the air lock sealing capability.*
- c. At least once per 18 months during shutdown verifying: ;
- 1. Only one door in each air. lock can be opened at a time, and
- 2. No detectable seal leakage when the volume between the emergency air lock shaf t seals is pressurized to greater than or equal to 44.7 psig for at least 2 minutes.
I l 1 I l
# The provisions of Specification 4.0.2 are not applicable.
- Exemption of Appendix J of 10 CFR 50 BEAVER VALLEY - UNIT 2 3/4 6-5
r. i CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING _ CONDITION FOR_0PERATION 3.6.1.4 Primary Containment internal air partial pressure shall be maintained 2 9.0 psia and within the acceptable operation range (below and to the left of the RWST water temperature limit line) shown on Figure 3.6-1 as a function of service water temperature. APPLICABILITY: MODES 1, 2, 3 aad 4. ACTION: With the containment internal air partial pressure < 9.0 psia or above the RWST water temperature limit line shown on Figure 3.6-1, restore the internal pressure to within the limits within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. 4 i SURVEII, LANCE RE0111REMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours. O BEAVER VALLEY - UNIT 2 3/4 6-6 i _.__._____._____j
4 /
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i 10.5 m-- -- i-. ' I
^ 10,4,4 THE OPER ATING CURVE REQUIRES TH AT THE
)
h AVER AGE CONT AINMEN T TEMPER ATURE DOES NOT LIE BELOW 85'F OR ABOVE 105=F. M ' (1. L w 10.3 , SET POINT VALUE IN CONT AINMENT V ACUUM - W AL ARM INSTRUMENTATION 3HOULO BE SET AT Z 0.25 PSI BELOW THE RWST TEMPER ATURE L LouiT CURVE ANO 29.0 PSI A, D I D..? i _ D f g , w '
,(
c 10.1 -- n.
$'10.0 3
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$. RW ST TEMPE RATURE550*F<
k f '9.9 e UNACCEPTAB LE
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ACCEPTABLE Il 2 OPER ATIO N 3 3 9. 2 - - E
< 9.1 y -: . - -
9.0 32 35 40 45 50 55 60 65 70 75 80 85 90 SERVICE WATER TEMPER ATURE (*F) t
/
O FIGURE 3.6-1
\) MAXIMUM ALLOWABLE PRIMARY CONTAINMENT AIR PRESSURE VERSUS SERVICE WATER TEMPERATURE AND RWST WATER TEMPERATURE l BEAVER VALLEY - UNIT 2 3/4 6-7 l I
G i-cm L CONTAINMENT SYSTEMS AIR TEMPERATURE , LIMIIIMCL.CDEDIT10dQR OPERATI.0N 3.6.1.'5" Primary containment average air temperature shall be maintained < 105 F and > 85*F. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment average air temperature > 105 F or < 85 F restore the average a b temperature to within the limit within 8 hours or be in at least HOT STANDBY A ithin the next 6 hours and in COLD SHUTDOWN within the following 30 hours. j I SUREEJLl& CE REQUIREMENTS ! o '
- t. ,
4.6.1.5 The primary containment average maximum and minimum air temperatures shall be the arithmetical average of the temperatures at the following locations and shall be determined at least once per 24 hours. The nearest alternate detector may be used for temperature determination up to a maximum of one per location. Location
- a. Reactor Head Storage Area - Elev, 802'-0"
- b. Pressurizer Cubicle - Elev 802'0"
- c. RC Annulus - Elev. 777'-4"
- d. RHR Heat Exchanger - Elev. 801'-6"
- e. RC Annulus - Elev. 701'-6" O
BEAVER VALLEY - UNIT 2 3/4 6-8
~ ____ - _ _ _ _ _ _ _ _ - _ _ ___-
l t h : p CONTAINMENT SYSlEMS
? i \ CONTAINMENT STRUCTURAL INTEGRITY '
l LIMITING C0HDlIIDH.J_0R OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1. APPLICABILITY: MODES 1, 2, 3 and 4. ACT'DN:
- With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.
SUB11LLANCf RE0VIREMENIS 4.6.1.6.1 Liner Plate and Concrete The structural integrity of the contain-ment liner plate and concrete shall be determined during the shutdown for each Type A containment leakage rate test'(reference Specification 4.6.1.2) by: r (N a. a visual inspection of the accessible surfaces and verifying no apparent changes in appearance or other abnormal degradation.
- b. a visual inspection of accessible containment liner test channels prior to each Type A containment leakage rate test. Any containment liner test channel which is found to be damaged to the extent that channel integrity is impaired or which is discovered with a vent plug removed, shall be removed and a protective coating shall be applied to the liner in that area.
- c. a visual inspection of the dome area prior to each Type A containment leakage rate test to insure the integrity of the protective coating.
4.6.1.6.2 Reports An initial report of any abnormal degradation of the contain-ment structure detected during the above required tests and inspections shall be made within 10 days after completion of the surveillance requirements of this specification, and the detailed report shall be submitted pursuant to Specification 6.9.2 within 90 days after completion. This report shall include a description of the condition of the liner plate and concrete, the inspection procedure, the tolersinces on cracking and the corrective actions taken. BEAVER VALLEY - UNIT 2 3/4 6-9
s. e.g
> q i< CONTAINMElTSYSTEMS .
3/4.6.2 01PR$55URIZATION AND C00l.itG SYSTEMS ,3 K CONTAINMitiT'QUENCF SPRAY SYSTEH
< l I .
li ,
! LIMITING CORDJTION FCB OPERATSt{ g.
p 3.6.2.1 Two separate and independent containment' quench spray subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3: and 4. ACTION: I With one containment quench spray subsystem inoperable, restore the inoperable subsystem to OPERABLE status sithin 72 hours;or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the followiag 30 hours. j jDE$ElliflLfL REQUIREMENTS , _ , _ _ i
- 4.6.2.1 Sach containment quench spray subsystem shall be demonstrated OPERABLE:
- a. At least once per 31 days bj: O1
- 1. Verifying that each valve'(manual, power operated or automatic)
} in the' flow path that is not locked, sealed, or otherwise secured in pos'. tion, is in its correct acsition.
- 2. Verifying the temperature of the barated water in the refueling water storage tank is within thu likits of: Specification 3.1.2.8.b.3.
- b. By verifying, that on a recirculitiori flow, each pump develops a differential pressure of > 138 psid at a flow of > 3000 gpm when tested i ~
l pursuant to Specification 4.0.5.
- c. At least owe per 18 months during shutdown, by:
l l 1. Cycli g each power operated (eveluding automatic) valve in the flow patT that is notitestable during plaat operation, through l at least one complete cycle of full travsl-1' ' I 2. Verifying that each authmatic valve in the flow path actuates to its correct position on a test signal. 1 3. Verifying that each spray pump starts automatically on a test ' ' signal. 1 BEAVER VALLEY - UNIT 2 3/4 6-10 i , _ _ _ - _ _ _ _ _ _ _ _ _ _ - .
\
.i
_ CONTAINMENT SYSTEMS U SURVEILLANCE RE0VI_REMERIS (Continued) i
- d. At 1 sc once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unob-structed.
O i i C/ BEAVER VALLEY - UNIT 2 3/4 6-11
CONTAINMENT SYSTEMS CONTAINMENT RECIRCULATION SPRAY SYSTEM LIBlDRG_CMDlIl0Bl0LDEEILAIION _ l
- 3. 6. 2. 2 Four separt.- and independent containment recirculation spray subsystems, each composed of a spray pump, associated heat exchanger and flow path shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. For subsystems containing recirculation spray pumps 2RSS-P21A or 2RSS-P21B: With one containment recirculation spray subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours; restore the inoperable spray system to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 30 hours.
- b. For subsystems containing recirculation spray pumps 2RSS-P21C or 2RSS-P210: See action statements in Specification 3.5.2 or 3.5.3.
SURVEILLANCE RE0VIREMEtlTS 4.6.2.2 Each containment recirculation spray subsystem shall be demonstrated , OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position; 1
l b. When tested pursuant to Specification 4.0.5, manually start each ! recirculation spray pump and verify the pump shaft rotates; I
- c. At least once per 18 months by verifying that or, a Containment Pressure-High-High signal, each recirculation spray pump starts automatically after a 628 1 5 second delay.
- d. At least once per 18 months, during shutdown, by verifying, that on recirculation flow, each recirculation spray pump develops a differential pressure of 1 112 psid at a flow of 1 3500 gpm.
- e. At least once per 18 months during shutdown, by:
- 1. Cycling each power operated (excluding automatic) valve in the ficw path not testable during plant operation, through at least one complete cycle of full travel.
O BEAVER VALLEY - UNIT 2 3/4 6-12 _ ._ __ _-__Y
[j^)
\
CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
- 3. Initiating flow through each Service Water subsystem and its two associated recirculation spray heat exchangers, and verifying a flow rate of at least 12,000 gpm.
- f. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
~%
(Q l i i BEAVER VALLEY - UNIT 2 3/4 6-13
CONTAINMENT SYSTEMS CHEMICAL ADDITION SYSTEM LIMITING CONDITION FOR OPERATION l 3.6.2.3 The chemical addition system shall be OPERABLE with;
- a. A chemical addition tank containing at least 8500 gallons of between 23 and 25 percent by weight Na0H solution, and
- b. Two chemical injection pumps each capable of adding NaOH solution from the chemical addition tank to a containment quench spray system pump flow.
APPLICABILITY: MODES 1, 2, 3 and 4. > ACTION: With the chemical addition system inoperable, restore the system to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 hours; restore the chemical addition system to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE RE0VIREMENTS 4.6.2.3 The chemical addition system shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. By verifying that on recirculation flow, each injection pump develops a flow between 40 and 60 gpm when tested pursuant to Spec.ficatior 4.0.5.
- c. At least once per 6 months by:
- 1. Verifying the contained solution volume in the tank, and
- 2. Verifying the concentration of the Na0H solution by chemical analysis.
- d. At least once per 18 months, during shutdown, by:
- 1. Cycling each valve in the chemical addition system flow path that is not testable during plant operation, through at least one complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.
BEAVER VALLEY - UNIT 2 3/4 6-14
_. - . _ -_______ ______ _ _- -_____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ L I L
' CN -
CONTAINMENT SYSTEMS
'(
3/4.6.3 CONTAINMENT ISOLATION VALVES LIM:TI_NG CONDU10N FOR OPERATION _ 3.6.3.1' The containment isolation valves'specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more of the isolation valve (s).specified in Table 3.6.1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate the affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
- c. Isolate the affected penetration within 6 hours by use of at least
, one closed manual valve or blind flange; or
- d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, i
' SURVEILLANCE RE0VIREMENTS 4.6.3.1.1 The isolation valves specified in Table 3.6-1 shall be demonstrated )
OPERABLE: j
- a. At least once per 92 days by: I
- 1. Cycling each OPERABLE power operated or automatic valve testable during plant operation through at least one complete cycle of full travel.
- 2. Cycling each weight or spring loaded check valve testable during plant operation, through one complete cycle of full travel and verifying that each check valve remains closed when the differ-ential pressure in the direction of flow is < 1.2 psid and opens when the differential pressure in the direction of flow is
> 1.2 psid but less than 6.0 psid.
- b. Immediately prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated i actuator, control or power circuit by performance of the applicable 4
\ cycling test, above, and verification of isolation time. ]
l l BEAVER VALLEY - UNIT 2 3/4 6-15 i
l ( CONTAINMENT SYSTEMS SUEVELLIAELREMEME_MI_1 ( Co n ti n u e d ) 4,6.3.1.2 Each isalation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
- a. Verifying that on a Phase A containment isolation test signal each ,
Phase A isolation valve actuates to its isolation position. I
- b. Verifyint that on a Phase B containment isolation test signal, each Phase B i;olation valve actuates to its isolation position.
- c. Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation position.
- d. Cycling each power operated or automatic valve through at least one complete cycle of full travel and measuring the isolation time pursuanc to Specification 4.0.5.
- e. Cycling each weight or spring loaded check valve not testable during plant operation, through one complete cycle of full travel and verifying that each check valve remains closed when the differential pressure in the direction of flow is < 1.2 psid and opens when the differential pressure in the direction of flow is > 1.2 psid but less than 6.0 psid.
- f. Cycling each manual valve not locked, sealed or otherwise secured in the closed position through at least one complete cycle of full travel.
1 O BEAVER VALLEY - UNIT 2 3/4 6-16
E M S UE( MK IOE 0 0 0 0 0 0 0 XRM 6 6 6 6 A 6 6A 1 ATI .
/ /
MST < < < < N < <N < 1 1 1 1 1
- - - - A -
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C E M S UE( S MK N IOE 0 0 0 0 0 0 O XRM 6A 6A 6A 6A A 6 6 A I ATI / / / / / / T MST <N <N <N <N N < < N A 1 R
- T 2 2 2 2 2 O3.
6 E L E N E P T 7 55 10 V1 OV 0 52 10 V1 OV 1 53 10 V1 OV 6 54 10 V1 OV 3 3 3 1 V O 3 5 1 V O 1 B N MR MR MR MR 8 M M 3 A E - - - - - - - - - - - - T M PP PP PP PP S C S S N CC CC CC CC I A W H I CC CC CC CC S I S C A 22 22 22 22 2 2 2 2 T N E ) ) ) ) ) ) O DE B B B B 2 2 C IV ( ( ( ( ( ( SL ) ) ) ) ) ) ) ) NA 1 1 1 1 3 A B 3 IV ( ( ( ( ( ( ( ( N O I T P I R C s rc S s t r E e e Wi D R s s R n c
/ e e o ee N m R R m i r cR O o o t i i s I r o o r c A vrl T f t t f e rii A j t eA o C l h l h l h l h d n n S c I oc oc oc oc aI e . G F ox ox ox ox e m &t g N I CE CE CE CE H y u nn I T E t E r E l oi G E N pt pt pt pt R h e R t R l Cl R R E ma ma ma ma A gf A s A i o A A D oe oe oe oe P ia P n P h oo H P I CH CH CH CH S HS S I S CtC C S A
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C E M S UE( S MK 0 N IOE 0 0 0 O XRM A 6A A 6 A 6 6 A
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t T MST N <N N < N < < 111NN n A o R 2 2 C T 2 ( E - - - ABC N 8 5 4 2 0002 1 E 7 5 5 5 00043
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E M S SS S S S S S SSSSS L N I HH I W H W W HHHHH B I S CC S 5 R S 5 CCCCC A A 2 22 2 2 2 2 2 22222 T T N E ) O DE 2 C IV ( )))) SL ) ) ) ) ) NA 3 A B B B AAA1 IV ( ( ( ( ( (((( N O I T k n P n w I l a c o R p s aT r d C m r l v ri rc t S u p t i or t c t r e E P u W o me We R Wi c L D n ne C et e ee t
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N . 7 8 EO 7 9 0 1 2 3 4 5 PN 1 1 2 2 2 2 2 2 2 2 O9 sPQ i eg " g TM
C E M S UE( MK IOE 0 0 0 0 XRM 6A AA A A A 6A 6 6 ATI / // / / / / MST <N NN N N N <N < < 1 1 B A B C B A B 8 60 8 8 8 0 0 0 05 34 0 0 0 00 0 0 11 88 3 3 3 11 1 1 V1 VV V V V V1 V V 0V O0 0 0 O 0V 0 0 AR MM M M M AR A A SS SS S S S SS G G GG II H H H AA D D DD S0 C C C DD B B 22 22 2 2 2 22 2 2 E D )) ) ) ) IE 22 3 3 3 SV (( ( ( ( TL ) )) ) ) ) ) ) ) UA A 33 2 2 2 A 2 2 OV ( (( ( ( ( ( ( (
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C E M S UE( S MK N IOE 0 0 O XRM 6 A A A A 6 A A ) I ATI / / / / / / / t T MST < N N N N < N N n A o R C T m m ( E A A e e N 8 0 t t 1 E 0 0 s s
- P 1 1 y y 6 V 4 6 5 V S S . T 0 4 7 7 7 0 3 N A 9 4 4 4 A d d E - - - - - - e e E M S S S S S S s s L N G I H H H A o o B I D S C C C D l l A A 2 2 2 2 2 2 C C T T N E ) ) ) )
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2 3 9 6 9 1 1 0 5 10 1 0 0 0 1 1 50 1 1 1 V V V1 V V V V 4 O 0 0V C 0 0 0 1 S 5 AR F A A A S S S SS S S S S A V V CC H R C N S C C RR C V R G 2 2 2 22 2 2 2 2
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/ / / / /
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2 2 2 2 2 2 A A B A A A A B A B 3 0 9 6 2 8 0 7 0 7 2 5 0 5 3 07 3 08 20 09 1 5 1 11 3 3 9 1 11 1 11 12 11 1 9 1 12 1 1 V V V1 V V1 V1 V1 V V V V1 V V 0 0 0V 0 0V 0V 0V 0 0 0 0V 0 O 5 5 AR S AR 5R AR A 5 A AR 5 M S R RR S RR RR RR R S R RR S C M S SS C SS SS SS S M S SS C A L S SS H SS SS SS S L S SS H I 2 2 22 2 22 22 22 2 2 2 22 2 2 E D ) ) IE A A SV ( ( TL ) ) ) ) ) ) ) ) ) ) ) ) ) UA 2 1 A 1 A 1 A 2 2 2 A 1 A OV ( ( ( ( ( ( ( ( ( ( ( ( (
)
C E M S UE( S MK N IOE 0 0 0 0 0 0 O XRM 6 6 A 6 6 6 A A 6 A A
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t T MST < < N < < < N N < N N n A . o R 1 1 1 1 1 1 C T - - - - - m m -
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/ i a t z e q e e p z N t T a y l e i l n l a y r O c e f W l a
p m l p L e p o p V l i I T t e r n a m i m e a A A e m rl a t a rl n i o A S a ep S c S ep A t C D l e t a n g S zm e zm n I e i a n t n i a n e F R l e e g rS w e w rS e m I g u g L e u o D o u g u T a s m o l se d d se o r N k s u r d sc w k w sc r t E a e c d l t ea o a o ea d s D e r c y o o rp i e l rp y n I L P A H C H PS B L B PS H I A E R
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C ) ) ) ) ) ) ) E ) M S 4 4 4 4 4 4 4 4 UE( ( ( ( ( ( ( ( ( MK 0 0 0 0 0 0 IOE 0 0 XRM A A A 6A 6A 6 6 6 6 6A 6A ATI / / / / / / MST N N N <N <N < < < < <N < A 8 A C 8 9 8 A 8 A C D B 8 3 8 1A 1B 5 5 5 5 6A 6C 8 8 8 01 01 5 5 5 5 56 56 8 8 8 10 10 1 1 1 1 15 15 V V V V1 V1 V V V V V1 V1 0 O 0 OV OV O O O O OV OV M M M MR MR M M M M MR MR S S S SS SS S S S S SS SS I I I SS SS S S S S SS SS RR S S S QQ QQ R R R R RR 2 2 2 22 22 2 2 2 2 22 22
)
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/ / / / / / / / /
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. IV C C C C C C C C C C C C N O I T P I R "B C " m S e E m t m _ D e s e
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) ) ) ) )
SL NA 2 2 2 5 5 ( ( ( ( ( IV _ N O I T P I R r C r r S e e t e E i n n s y n D i i u l i b b a p b
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C E M S UE( S MK N I 0E 0 O XRM A A 6 A A A A
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/ b V n i e - z l D e N m o t l y p r y O o . i c p e l p .A t I c t t e m l a u t i T e n c t a p n S n . v A R o u e S mS A ot a C n C n S D aH k C n C I no o n SR n c e F ei ri r e w e a rP r I gt ot o g o d& g R o o T oa t a t E a d i o E E t e t N rl cl c R k w up r R e R cl p c E d o ao e A a o qm d A s A ab m a D ys es j P e l iu y P o P eau e I HI RI E S L B LS H S H S RCP R O
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)
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)
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CONTAINMENT SYSTEMS N 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR' OPERATION 3.6.4.1 :Two separate and independent' wide-range containment hydrogen analyzers shall be OPERABLE. APPLICABII.ITY: MODES 1.and 2. ACTION:
- a. With one wide-range hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours.
- b. With both wide-range hydrogen analyzers inoperable, restore at least.
, one wide-range hydrogen analyzer to OPERABLE status within 72 hours or be in.H0T STANDBY within the next 12 hours.
jgglll.ANCE RE0VIREMENTS O 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once i per 92 days on a STAGGERED' TEST BASIS by:
- a. Performing a CHANNEL CALIBRATION using sample gases containing:
. 1. One volume percent hydrogen, balance nitrogen, and
- 2. . Four volume percent hydrogen, balance nitrogen., i l
l l O BEAVER VALLEY - UNIT 2 3/4 6-31
i CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS LIMIENB_COND1 HON _EDLDEEMUDN 3 6.4.2 Two separate and independent containment hydrogen recombiner systems , shall be OPERABLE. j l APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours. SJBYELLL8NCE_REQUIREBEtill _ 4.6.4.2 Each hydrogen recombiner system shall be demonstrated OPERABLE:
- a. At least once per 6 months by verifying during a recombiner system functional test using outside atmospheric flow rate of > 42 scfm that the heater outlet temperature increases to '> 700 F within 90 minutes and is maintained for at least 2 hours.
- b. At least once per 28 months by:
- 1. Performing a CHANNEL CALIBRATION of all recombiner instruments-tion and control circuits.
- 2. Verifying through a tisual examination that there is no evidence of abnormal conditions within the recombiners (i.e., loose wiring cr structural connections, deposits of foreign materials, etc.).
- 3. Vetifying during a recombiner system functional test using con-tainr+nt atmospheric air at a pressure of < 13 psia and a flow i at e 01 4 2 sc f ra , that the heater temperature increased to lltM within 5 hours and is maintained for at least 4 hours.
- 4. Verifyinu +he integrity of all heater electrical circuits by pei f orm s w a : ontinui ty and resistance to ground test immediately f following the nbo e required functional test. The resistance to ground f ar any bodan.r phase shall be > 10,000 ohms,
- c. Verifying that the hydregen recor.biner isolation valves (2HCS-MOV110A&B and 2HC5 'm J W.C) wadlosed and de-energized af ter every surveil-lance test fner 4. f4 4.2.4.) is completed or after their use, post-accidentt ,tr mcm.ine hydroaen in t he containment is c^mpleted.
~
O BEAVER VALLEY UW ' , ' 4 fr 3 2
o .. ,p [' g -, 4 .- CONTAINMENT SYSTEMS 3/4.6.4.3 .(This specification number is' not used)
'f I
O O BEAVER VALLEY - UNIT 2 3/4 6-33
CONTAINMENT SYSTEMS 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM , I STEAM JET AIR EJECTOR j LJMITING CONDITION FOR OPERATIO11 3 6.5.1 The inside and outside manual isolation valves in the steam jet air i ejector suction line shall be closed. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the inside or outside manual isolation valve in the steam jet air ejector suction line not closed, restore the valve to the closed position within 1 hour or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. SURVEILLANCE _RE0VIREMENTS 4.6.5.1 1 The steam jet air ejector suction line outside manual isolation valve shall be determined to be in the closed position by a visual inspection prior to increasing the Reactor Coolant System temperature above 350 F and at least once per 31 days thereafter. 4.6.5.1.2 The steam jet air ejector suction line inside manual isolation valve ' shall be determined to be sealed or locked in the closed position by a visual insgectionpriortoincreasingtheReactorCoolantSystemtemperatureabove 350 F. t O BEAVER VALLEY - UNIT 2 3/4 6-34
A e- ,
.z i ;-
]
i []- 3/4.7 PLANT SYSTEMS ,
\_/
3/4.7.1 TURBINE CYCLE a SAFETY VALVES i LIMITIM_CQHDJTION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
,) 4, APPLICABILITY: MODES 1, 2 and 3. [ o .
ACTION: f s j
- a. With 3 reactor coolatri loops and associated steam generators in operation and with one'or more main steam line code safety;va7ves inoperable, operation in MODES 1, 2 and 3 may proceed prpvided, that within 4 hours,.either the inoperable salve istresto' red to OPERABLE status or the Power Range Neutron Flux High setpoint trip
- is reduced per Table 3.7-1; otherwise, be in at least F0T STANDBY within the next G hours.and in COLD SHUTDOWN within the following 30 hours. '
, b. The provisions of Spech'Jication 3.0.4 are not applicable. " \
J E RVEILLANCL RE0VIREMENTS , 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 3.7-2 ! by performance of the surveillance required by Specification 4.0.5. I ! l
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f PLANT SYSTEMS AUXILIARY FEE 0 WATER SYSTEM l LQjlUliG_;61$ TION FOR OPERATIOL_. .,, l .s L 3.7.1.2 At least three steam generator auxiliary feedwater pumps and l- associated flow paths shall be OPERABLE with: l
, a. Two motor-driver auxiliary feedwater pumps, each capable of being powerei from siparate emergency husses, and i;'
- b. yne steam turbineu driven auxiliary feedw u r pump capable'of bding powred from an rJERABtf steam supply synen, APPLICABILITY MODES 1, 2 and.3.
1 ALTIO_th
- a. With one auxiliary feedwater pump inoperable, restore at leass three
" auxiliary feedwater pumps (two capable of being powered from separata emergency bussettand one capable of being powered by an OPETABLE steam supply system) to OPERABLE status within 72 nours or be.in H0T STAND 53Y whhir :.he next S hours and in HOT SHUTDOWN within the 1cllow-
, ing 6 hourt,
- b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN withir. the following ti hours.
- c. With three auxiliary feedwater pumps inoperable. immediately initiate correctivt action to restore at least one auxiliary feedwater pump to OPERAELE status as soon as possible.
g Q LLANCE.lE0VIREMENTS ,_ ,_ _ i 4.7 3,2 Each auxiliary feedwater pump sha'll bd demonstrated OPERABLE:
- a. At least once pr.v 31 days on a STAGGERED TEST BASIS by:
- 1. Verifying that:
l
- a. Each motor driven pump develops a differential. pressure of l 1 1290 psid on recirculation flow of 1 100 gpm, and i l
- b. The steam turbine driven pump develops a differential pree- 1 sure of 21310 psid on recirculation flow of 2 220 gpm when the secondary steam pressure is greater than 600 psig.
The provisions of Specification 4.0.4 are not applicable j foN entry into MODE 3. ' BEAVER VALLEY - i(NIT 2 3/t T-4
b PLANT SYSTEMS U BBEI1LANCERE001BfMENTS(Continued)
- 2. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- 3. Reverifying the requirements of Tech Spec. surveillance 4.7.1.2.a.2 by a second and independent operator.
- 4. Establish and maintain constant communications between the control room and the auxiliary feed pump room while any normal discharge valve is closed during surveillance testing.
- 5. Verifying operability of each Service Water auxiliary supply valve by cycling each manual Service Water to Auxiliary Feed-water System valve through one complete cycle.
- b. Following a plant outage of 30 days or greater, verify Auxiliary Feedwater Flow from TK-210 to the Steam Generators with the Auxiliary l Feedwater Valves in their normal alignment.
- c. At least once per 18 months during shutdown by:
- 1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.
- 2. Verifying that each automatic valve in the flow path actuates to its correct position on an auxiliary feedwater actuation test signal.
- 3. Verifying that each auxiliary feedwater pump starts automatically upon receipt of an auxiliary feedwater actuation test signal.
) /%
s BEAVE-l VALLEY - UNIT 2 3/4 7-5
P1 ANT SYSTEMS PRIMARY PLANT DEMINERALIZED WATER (PPDW) kIBIIIN(LCOND1Il0N F0R 0PIRallQN 3.7.1.3 The primary plaat demineralized water storage tank shall be OPERABLE with a minimum contained volume of 127,000 gallons. APPLICABILITY: MODES 1, 2 and 3. ACTION: With less than 127,000 gallons of water in the PPOW storage tank, within 4 hours either:
- a. Restore the water volume to within the limit or be in HOT SHUTDOWN within the next 12 hours, or
- b. Demonstrate the OPERABILITY of the service water system as a backup supply to the auxiliary feedwater pumps and restore the PPDW storage tank water volume to within its limit within 7 days or be in HOT SHUTDOWN within the next 12 hours. I SUREElLLANCE_Rf._QUIRDiENTS O1 4.7.1.3 The PPDW storage tank shall be demonstrated OPERABLE at least once per 12 hours by verifying the water level.
O BEAVER VALLEY - UNIT 2 3/4 7-6 ) I
l l I i n PLANT SYSTEMS
" ACTIVITY LIMIIltfG_.COND.ITION FOR OPERATIQX _
.)
3.7.1.4 . The specific activity of the secondary coolant system shall be-
-1.0.10 pCi/ gram DOSE EQUIVALENT I-131. f 1
APPLICABILITY: H0 DES 1, 2,-3, and 4. ACTION: With the specific-activity of thu secondary coolant system > 0.10 pCi/ gram l l DOSE EQUIVALENT I 131, be in at least HOT STANDBY within 6 hours and in COLD l SHUTDOWN within the next 30 hours. EURV_EILLARCE_ REQUIREMENTS
- 4. 7.1. 4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-2.
O O BEAVER VALLEY - UNIT 2 3/4 7-7
)
TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM AND ANALYSIS FREQUENCY
- 1. Gross Activity Determination 3 times per 7 days with ,
a maximum time of 72 hours 1 between samples
- 2. Isotopic Analysis for DOSE a) 1 per 31 days, when-EQUIVALENT I-131 Concentration ever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit.
b) 1 per 6 months, when-ever the gross activity determination indicates iodine concentrations below 10% of the allow-able limit. O l l O f BEAVER VALLEY - UNIT 2 3/4 7-R
r- PLANT SYSTEMS
- p k MAIN STEAM LINE ISOLATION VALVES LIMITINGCONDITIONFOROPERATIQN 3.7.1.5 'Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3. ACTION: MODES 1 - With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; Otherwise, be in HOT SHUTDOWN within the next 12 hours. MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed after:
- a. The inoperable isolation valve is-restored to OPERABLE status, or
- b. The isolation valve is maintained closed; (m)
%/
Otherwise, be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE RE0VIREMENTS 4.7.1.5 Each main steam line isolation valve that is open shall be demonstrated l OPERABLE by: l i
- a. Part-stroke exercising the valve at least once per 92 days, and
- b. Verifying full closure within 5 seconds on any automatic closure I actuation signal while in HOT STANDBY with T,yg > 515 F during each reactor shutdown except that verification of full closure within 5 seconds need not be determined more often than once per 92 days.
l l BEAVER VALLEY - UNIT ? 3/4 7-9 l
PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION Ll!illlRGl0MDil101 LEO.fL0ffEA11011_ _ .m 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 70 F when the pressure of either coolant in the steam generator is > 200 psig. APPLICABILITY: At all times. ACTION: With the requirements of the above specification not satisfied:
- a. Reduce the steam generator pressure of the applicable side < 200 psig within 30 minutes, and
- b. Perform an analysis to determine the effect of the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable f or continued operation prior to increasing its temperatures above 200 F.
SMIWflLLANCE REQUIR[MERIS 4.7.2.1 The pressure in each side of the steam generator shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant in the steam generator is < 70 F.
}
O BEAVER VALLEi - UNIT 2 3/4 7-10
l J I O ' PLANT SYSTEMS l \ L
'") 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM LIMITIN_G CONDITION FOR OPERATION I
3.7.3.1 At least two primary component cooling water subsystems _shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. i ACTION: With one less than two primary component cooling water subsystems OPERABLE, restore at least two subsystems to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0UIREMENTS 4.7.3.1 At least two primary component cooling water subsystems shall be m demonstrated OPERABLE. [ h V a. Verify that each pump develops the required differential pressure and flow rate when tested in accordance with the requirements of Section 4.0.5.
- b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not j locked, sealed, or otherwise secured in position, is in its correct j position. 1
- c. At least once per 18 months during shutdown, by cycling each power operated valve servicing safety related equipment that is not test-able during plant operation, through at least one complete cycle of full travel.
j i v BEAVER VALLEY - UNIT 2 3/4 7-11
PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM (SWS) i LIMITlNG10NDJ110Bl0LOPERATION 3.7.4.1 At least two service water subsystems supplying safety related equip- ! ment shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one less than two SWS subsystems OPERABLE, restore at least two subsystems to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. f ( SilREILLAtiCE RE0VIREMENIS 4.7.4.1 At least two SWS subsystems shall be demonstrated OPERABLE: I
- a. Verify that each pump develops the required differential pressure and flow rate when tested in accordance with the requirements of Section 4.0.5.
- b. At least once per 31 days by verifying that each valve (manual, power operated or automatic servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct f position.
i
- c. At least once per 18 months during shutdown, by cycling each power operated valve servicing safety related equipment that is not testable during plant operation, through at least one complete cycle of full travel.
O BEAVER VALLEY - UNIT 2 3/4 7-12
,[' Pl. ANT SYSTEMS l\ 1 3/4.7.5 ULTIMATE HEAT SINK - OHIO RIVER l LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPERABLE with:
- a. A miniman water level at or above elevation 654 Mean Sea Level, at the intake structure, and
- b. . An average water temperature of 1 86 F.
A_ APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
.With the requirements of the above specification not sctisfied, be in at least i HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.7.5.1 The ul+imate heat sink shall be determined OPERABLE at least once per
/( 24' hours by verifying the average water temperature and water level to be within their. limits.
l 1 1 1 l
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fJ' l 1 BEAVER VALLEY - UNIT 2 3/4 7-13
PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION 1.ltRIING CONDITION FORE.ER&Il0N 3.7.6.1 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Ohio River exceeds 1 695 Mean Sea Level at the intake structure. APPLICABILITY: At all times. ACTION: With the water level at the intake structure above elevation 695 Hean Sea Level:
- a. Be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours, and
- b. Initiate and complete within 8 hours, the following flood protection measures:
- 1. Install and seal the flood doors in the intake structure.
SURVEILLANCE RE0VIRBENTS 4.7.6.1 The water level at the intake structure shall be determined to be j within the limits by;
- a. Measurement at least once per 24 hours when the water level is below elevation 690 Mean Sea Level, and
- b. Measurement at least once per 2 hours when the water level is equal to or above elevation 690 Mean Sea Level.
O l BEAVER val. LEY - UNIT 2 3/4 7-14
f3 _ PLANT SYSTEMS- I (b) 3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP AND PRESSURIZATION SYSTEM LIMITING _ CONDITION FOR OPERATION 3.7.7 The Control Room Emergency Air Cleants and Pressurization System com- l prised of the following shall be OPERABLE: 1
- a. A pressurization filtration unit comprised of two trains of fans and filters, and flow path control dampers.**
- b. A bottled air pressurization system comprised of 5 subsystems with two bottles in each subsystem.*
- c. Two isolation dampers in series in each of four normal air flow paths (two intake and two exhaust) with each damper OPERABLE by automatic actuation or OPERABLE by being secured in a closed position with power removed.
APPLICABILITY: All MODES. ACTION: MODES 1, 2, 3 and 4: ~ With one train of the pressurization filtration unit, or one subsystem of the bottled air pressurization system, or one of two isolation dampers in series inoperable, restore the system to OPERABLE status within 7 days or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ( l MODES S and 6: I 1
- a. With one train of the pressurization filtration unit or one seUsystem of the bottled air pressurization system, or one of two isolation dampers in series inoperable, restore the inoperable system to OPERABLE status within 7 days or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel and movement of loads over irradiated fuel.
- b. With both trains of the pressurization filtration unit, or more than one subsystem of the bottled air pressurization system, or two of two !
isolation dampers in series inoperable suspend all operations involv-ing CORE ALTERATIONS, positive reactivity changes, movement of irradi-ated fuel and movement of loads over irradiated fuel.
*The air bottles may be isolated for up to 8 hours for performance of instru-mentation and control systems testing. \ ** Emergency backup power for one train of dampers and fans of the pressuriza-tion filtration unit not required in MODES S and 6.
1 BEAVER VALLEY - UNIT 2 3/4 7-15 j l
PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP AND PRESSURIZATION SYSTEM j l XVMEILL6HCE REQUIBEMfiMIS _ 4.7.7.1 The Control Room Emergency Air Cleanup ar.d Pressurization System j shall be demonstrated OPERABLE: )
- a. At least once per 12 hours by verifying that the control room air temperature is 1 88 F.
- b. At least once per 31 days by:
- 1. Initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for 15 minutes with the heaters in operation.
- 2. Verifying that the outtled air pressurization system contains a minimum of 10 bottles of air each pressurized to at least 1825 psig and that each soleoid operated valve is powered from an operable emergency bus.
- c. At least once per 18 months or (1) after each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after i any structural maintenance on the HEPA filter or charcoal adsorber l housings by:
l 2 Verifying that the charcoal adsorbers remove > 99.95% of a halo-q genated hydrocarbon refrigerant test gas when they are tested { in place in accordance with ANSI N510-1980 while operating the pressurization filtration system at a flow rate of 800 to 1000 cfm. i 2. Verifying that the HEPA filter banks remove > 99.95% of the D0P when they are tested in place in accordance with ANSI N510-1980 while operating the pressurization filtration system at a flow rate of 800 to 1000 cfm.
- 3. Verifying a system flow rate of 800 to 1000 cfm during system operation.
l
- d. At least once per 18 months or (1) after 720 hours of system opera-tion, or (2) following painting, fire or chemical release in the vicinity of control room outside air intakes while the systems is operating subjecting the carbon contained in at least one test canis-ter or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a re-moval efficiency of > 99% for radioactive methyl iodide at an air flow velocity of 0.7 ft/sec + 20% with an inlet methyl iodide con-I 3 > 70% relative humidity, and 30 C i l centration 1/2 C; other test conditions sha be in accordance with ANSI of 1.5 to 2.0 mg/m , T1 N510-1980. The carbon samples not obtained from test canisters shall be prepared by either:
- BEAVER VALLEY - UNIT 2 3/4 7-16 l
l 1 PLANT. SYSTEMS \j _ SURVEILLANCE REQUIREMENTS (Continued) l a) Emptying one entire bed from a removed adsorber tray, mixing ) the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly,'and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed,
- e. At least once per 18 months by:
- 1. Verifying that the pressure drop for the combined HEPA filters and charcoal adsorber banks is less than 5.6 inches Water Gauge ;
while operating the pressurization filtration system at a flow rate of 800 to 1000 cfm.
- 2. Verifying that on a Containment Isolation Phase B/ Control Room High Radiation test signal, the system automatically closes all the series isolation ventilation system dampers which isolate the control room from the outside atmosphere ar.d the system automatically starts 60 minutes later and supplies air to the control room through the HEPA filters and charcoal adsorber banks.
U 3. Verifying that on a chlorine test signal from the system auto-matic311y closes all the series isolation ventilation system dampers which isolate the combined control room from the out-side atmosphere.
- 4. Verifying that the pressurization filtration system maintains the control room at a positive pressure of > 1/8 inch Water Gauge relative to the outside atmosphere during system operation. l
- 5. Verifying that the heaters dissipate 5 0.5 kw when tested in accordance with ANSI N510-1980.
- 6. Verifying that a chlorine / control room high radiation / containment phase B isolation signal will initiate operation of the bottled air pressurization system.
- 7. Verifying by a partial discharge ted from four out of five sub-systems of the bottled air pressurization system at a discharge flow of less than 1000 cfm that the bottled air pressurization j system will pressurize the control room to > 1/8 inch Water Gauge relative to the outside atmosphere during system operation.
BEAVER VALLEY - UNIT 2 3/4 7-17
PLANT SYSTEMS 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS) LIMIIJHG CONDITION FOR OPERATION 3.7.8.1 Two SLCRS exhaust air filter trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one SLCRS exhaust air filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. B![NEILLANCf_RE@.IREMENTS 4.7.8.1 Each SLCRS exhaust air filter train shall be demonstrated OPERABLE:
- a. At least once per 31 days by initiating, from the control room, flow through the " standby" HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes with the heater controls operational,
- b. At least once per 18 months and (1) after each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (3) following painting, fire or chemical release in any ventilation zone communicating with the system by:
- 1. Verifying that the charcoal adsorbers remove > 99.95% of a halo-
~
genated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 59,000 cfm + 10%.
- 2. Verifying that the HEPA filter banks remove > 99.95% of the DOP when they are tested in place in accordance Qith ANSI N510-1980 while operating the ventilation system at a flow rate of 59,000 cfm i 10%.
, 3. Subjecting the carbon contained in at least ore test canister or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory caroon sample analysis and verifying a removal efficiency of > 99% for radioactive methyl iodide at an air flow velocity of 077 ft/sec i 20% with an inlet methyl iodide concentration of 1.5 to 2.0. mg/m3 , > 70% relative humidity, and 30 C i C; other test conditions shall be in accordance with ANSI N510-1980. The carbon samples not obtained from test canisters shall be taken with a slotted tube sampler in accord-ance with ANSI N509-1980. BEAVER VALLEY - UNIT 2 3/4 7-18
I PLANT SYSTEMS v' SURVEILLANCE RE0VIREMENT1
- 4. Verifying a system flow rate of 59,000 cfm 10% during system operation.
- c. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.8 inches Water Gauge while operating the ventilation system at a flow rate of 59,000 cfm i 10%.
- 2. Verifying that the exhaust from the contiguous area is diverted through the SLCRS filter train on a Containment Isolation -
Phase "A" signal in less than 5 minutes.
- d. Verifying that the air flow distribution to each HEPA filter and charcoal adsorber is within + 20% of the averaged flow per unit after initial installation and after any maintenance affecting the flow distribution.
- e. At least once per 4 months of system operation, perform the surveil-lance requirement of 4.7.8.1 b.3.
O v s 1 1 O : BEAVER VALLEY - UNIT 2 3/4 7-19 I i j
PLANT SYSTEMS 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CON _D11J0N FOR OPERAll0N 3.7.9.1 Each sealed source containing radioactive material either in excess .I of those quantities of byproduct material listed in 10 CFR 30.71 or > 0.1 micro-curies of any other material, including alpha emitters, shall be free of l
> 0.005 microcuries of removable contamination.
APPLICABILITY: AT ALL TIMES. ACTION:
- a. Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:
- 1. Either decontaminated and repaired, or
- 2. Disposed of in accordance with Commission Regulations.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not
- applicable.
MRVEILLANCLREQUIREMENTS 4.7.9.1.1 Test Requirements - Each sealed source shall be tested for leakage i I and/or contamination by:
- a. The licensee, or
- b. Other persons specifically authorized by the Commission or e Agreement St. ate.
The test method shall have a detection sensitivity of at least 0.005 microcuries 1 per test sample. l l 4.7.9.1.2 Test Frequencies - Each category of sealed sources shall be tested , l ! at the frequency described below,
- a. Sources in use (excluding startup sources previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive materials.
- 1. With a half-life greater than 30 days (excluding Hydrogen 3) and
- 2. In any form other than gas.
O BEAVER VALLEY - UNIT 2 3/4 7-20
rw -\ / PLANT SYSTEMS--- U SURVEILLANCE RE001REMENTS (CONTINVED)
- b. Stored sources not in use - Each sealed source shall be tested prior to use or transfer to another licensee unless tested within the pre-vious six months. Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use.
- c. Startup sources - Each sealed startup source shall be tested prior to being subjected to core flux and following repair or maintenance to the source.
4.7.9.1.3 Reports - A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days if source leakage tests reveal the presence of > 0.005 microcuries of removable contamination. A i l l t^
\
s BEAVER VALLEY - UNIT 2 3/4 7-21
PLANT SYSTEMS 3/4.7.10 (This specification number is not used) l 4 l O O BEAVER VALLEY - UNIT 2 3/4 7-22 l i
PLANT SYSTEMS 3/4.7.11 (This specification number is not used) 4 0 . O
' BEAVER VALLEY - UNIT 2 3/4 7-23
PLANT SYSTEMS 3/4.7.12 SNUBBERS , l RiiIIING_COSSil101LEDR OPERATION ., l l' 3.7.12 All snubbers shall te OPERABLE. The only snubbers excluded from this requirement are those installed on non-safety-related systems and then only if their failure or failure of the system on whicn they are installed, would have no adverse effect on any safety-related system. APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems
- required OPERABLE in those MODES).
ACTION: , With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.12.c on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system. . SURVEILLANCE RE0MIREMENIS 4.7.12 Each snubber shall be demonstrated OPEi!ABLE by performance of the fol-lowing augmented inservice inspectior. program and the requirements of Specifi-cation 4.0.5.
- a. Visual Inspections The first inservice visual inspection of snubbers shall be performed after four months but within 10 months of commencing POWER OPERATION and shall include all snubbers. If less than two (2) snubbers are found inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months i 25%
from the date of the first inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule: No. Inoperable Snubbers Subsequent Visual per Inspection Period Inspection Period. ** # 0 18 months i 25% 1 12 months i 25% 2 6 months 25% 3, 4 124 days i 25% 5,6,7 62 days i 25% 8 or more 31 days i 25%
*These systems are defined as those portions or subsystems required to prevent releases in excess of 10 CFR 100 limits.
- The inspection interval shall not be lengthened more than one step at a time.
#The provisions of Specification 4.0.2 are not applicable.
BEAVER VALLEY - UNIT 2 3/4 7-24
m . l (/ PLANT SYSTEMS l g '_ SURVEILLANCE REQUIREMENTS N ON INUED) ___ The snubbers may be' categorized:into two groups: those accessible and those ! -inaccessible during reactor operation. Each group may be inspected independently in accordance'with the above schedule. b .- Visual Inspection Criteria Visual inspections shall verify (1) that there are no visible indica-l tions of damage or impaired OPERABILITY, (2) attachments to the
' foundation or supporting structure are secure, and (3) in those loca-tions where snubber movement can be manually induced without discon- l necting the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular.
snubber and for other snubbers'that may be generically susceptible;
- i. and (2) the affected snubber is functionally tested in the as-found l condition and determined OPERABLE per Specification 4.7.12.d. How-ever, when a fluid port of a hydraulic snubber is found to be un-covered, the snubber'shall be determined inoperable and cannot be
~
determined OPERABLE via functional testing for the purpose of.estab-- ,[ .lishing the next visual inspection interval.
- c. Functional Tests At least once per.18 months during shutdown, a representative sample (of at least 10%) of the total of each type of. snubber in use in the plant shall be functionally. tested.either in place or in a bench test.
For Functional Testing type of snubber shall mean a group or combina- , tion of groups by load size and kind (i.e. , hydraulic or mechanical). ' or any other combination of load size and kind. For each snubber that does not meet the functional test acceptance criteria of Speci-fication 4.7.12.d, an additional 10% shall be functionally tested. l I i C BEAVER VALLEY - UNIT 2 3/4 7-25
PLANT SYSTEMS SURVEILLANCE RE0VIBEMENLS (Continued) The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At least 25% of the snubbers in the representative sample shall include snubbers from the following three categories: 1, The first snubber away from each reactor vessel nozzle.
- 2. Snubbers within 5 feet of heavy equipment (valve, pump, turbine, motor,etc.).
- 3. Snubbers within 10 feet of the discharge from a safety relief valve.
Snubbers that are especially difficult to remove or in high radiation zones during shutdown shall also be included in the representative sample.* If a spare snubber has been installed in place of a failed snubber, the spare snubber shall be retested. Test results of this snubber ' may not be included for the re-sampling. If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement shall be independent of the require-ments stated above for snubbers not meeting the functional test acceptance criteria. For the snubber (s) found inoperable, an engineering evaluation shall be perfortred on the components which are supported by the snubber (s). The purpose of this engineering evaluation shall be to determine if the components supported by the snubber (s) were adversely affected by the inoperability of the snubber (s) in order to ensure that the supported component remains capable of meeting the designed service.
- Permanent or other exemptions from functional testing for individual snubbers in these categories may be granted by the Commission only if a justifiabh basis for exemption is presented and/or snubber life destructive testing was performed to qualify snubber operability for all design conditions at either the completion of their fabrication or at a subsequent date.
BEAVER VALLEY - UNIT 2 3/4 7-26
i I' 1 L PLANT SYSTEMS O -gjRVEIL'LANCEREOUIREMENTS(Continued)
.i
- d. Snubber Functional Test Acceptance Criteria I The' snubber functional test shall verify that: i
- 1. Activation (restraining action) is-achieved within the specified range of velocity or acceleration in both tension and compression.
- 2. Snubber ~ bleed, or release rate, where required, is within the specified range in compression or tension.
- 3. The force that' initiates free movement of the snubber-rod in either tension'or compression is less than the specified maximum drag force, j
- e. _ Service Life Monitoring l
The service life of hydraulic and mechanical snubbers'shall be monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for
~
various seals, springs, and other critical. parts shall be determined and established based'on engineering information and may be extended or shortened based on monitored test results and failure history, i Critical parts shall be replaced so that the maximum service life j will not be exceeded during a period when the snubber is required to be OPERABLE. .The parts replacements shall.be documented and the documentation shall be retained in accordance with Specification j 6.10.2. Service life will be oefined to commence at plant startup subsequent to-initial fuel load. 1 O i o BEAVER VALLEY - UNIT 2 3/4 7-27
PLANT SYSTEMS 3/4.7.13 STANDBY SERVICE WATER SYSTEM (SWE). J.lMITING CONDITION FOLOPERATION 3.7.13.1 At least one standby service water subsystem shall be OPERABLE. 1 APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With less than one SWE subsystem OPERABLE, restore at least one subsystem to OPERABLE status within 7 days cr be in at least HT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following thirty hours.
}DRVEILLAlg __ REQUIREMENTS 4.7.13.1 At least one SWE subsystem shall ue demonstrated OPERABLE:
- a. By verifying that each pump develops at least 109 psid differential pressure while pumping through its test flow line when tested pur-suant to Specification 4.0.5. ;
i
- b. At least once per 18 months during shutdown by starting a Standby Service Water System Pump, shutting down one Service Water System i Pump, and verifying that the Standby Service Water Subsystem provides I at least 8584 gpm cooling water to that portion 01 the Service Water System under test for at least 2 hours.
1 I l l l l I i BEAVER VALLEY - UNIT 2 3/4 7-28
x U d 3 2 Y 3/4.8 ELECTRICAL g ER SYSTEMS l )jqs
*a
.() 3/4.8.1 A.C. SOURCES ., e OPERATING l i LIMITIRG CONDITION FOR OPERATION _ L 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be - OPERABLE: -
/
- a. Two physically independent circuits between the offsite transmission i network and the onsite Class IE' distribution system, and
- b. Two separate and independent diesel generators each with:
N,i
- 1. Separate day tank containing a minimum of 35'O gallons of fuel,
- 2. A separate fuel storage system containing a minimum of 53,225 A ,
gallons of fuel, j f 3 .. A separate. fuel transfer pump, y 1
- 4. Lubricating oil storage containing a minimum total volume of j f "504 gallons of lubricating oil, and ;
I ( 5. Capability to transfer lubricating oil from-storage to tne diesel generator unit. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With either an offsite circuit or diesel generator of the above re- I quired A.C. electrical power sources inoperable, demonstrate the !
OPEMBILITY of the remaining A.C. sources by performing Surveillance Regdqrements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at least once per 8 hours thereafter; restore at least two offsite cir-cuits and two diesel generators to OPERABLE status within 72 hours or be in COLD SHUTDOWN within the next 36 hours. 1
- b. With one offsite circuit and one diesel generator of the above re-quired A.C. electricai power sources inoperable, demonstrate the. ;
OPERABILITY of the remaining A.C. sources by performing Surveillance d Regoirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at I leart. once per 8 hours thereafter; restore at least one of the in-operable sources to OPERABLE status within 12 hours or be in COLD ] 1 SHUTDOWN within the next 36 hours. Restore at least two offsite l circuits and two diesel generators to OPERABLE status within 72 hours I from the time of initial loss or be in COLD SHUTDOWN within the next {
^ 36 hours. , i BEAVER VALLEY - UNIT 2 3/4 8-1 1
ELECTRICAL POWER SYSTEMS LltilllffCLCQHD1 TION FOR OPERAJ1QR (Continue 1D O
- c. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.5 within one hour and at least orce per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to
..,, OPERABLE status within 24 hours or be in at least HOI STANDBY witnin l the next 4 hours. With only one offsite source restored, restore at !
) least two offsite circuits to OPERABLE status within 72 hours from j j time of initial loss or be in COLD SHUTDOWN within the next 36 hours. L l d.1 With two of the above required diesel generators inoperable, demon- ]
, strate the OPERABILITY of two offsite A.C. circuits by performing j Surveillance Requirement 4.8.1.1.1.a within one hour and at least l once per 8 hours thereafter; restore at least one of the inoperable f diesel generators to OPERABLE status within 2 hours or be in COLD l SHUTDOWN within the next 36 hours. Restore at least two diesel gen- I f eratcrs to OPERABLE status within 72 hours from time of initial loss {'
or be in COLD SHUTDOWN within the next 36 hours. I SURVELLLehCE_REDIRBEt4Ts j 4.8.1.1.1. Twa physically independent circuits between the offsite trans-mission network and the onsite Class 1E distribution system shall be: t a. Determine OPERABLE at least once per 7 days by verifying correct breaker alignment, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months by transferring (manually and automatically) unit power supply from the unit circuit to the system circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1. Verifying the fuel level in the day tank, l
- 2. Verifying the fuel level in the fuel storage tank, f
O l BEAVER VALLEY - UNIT 2 3/4 8-2
O -ELECTRICAL POWER SYSTEMS SURYI1LLatiCEEQUIREMENTS (Continued) l
- 3. Verifying that a sample of diesel fuel from the fuel storage tank is within the acceptab'le limits specified in Table 1 of ASTM 0975 when checked for viscosity, water and sediment,
- 4. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank,
- 5. Verifying the diesel starts from ambient condition,
- 6. Verifying the generator is synchronized, loaded to 1 4,238 kw, and operates for at least 60 minutes, and
- 7. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
- 8. Verifying the lubricating oil inventory in storage.
- b. At least once per 18 months during shutdown by:
- 1. Subjecting the diesel to an inspection in accordance with pro-
, cedures prepared in conjunction with its manufacturer's recom- \ mendations for this class of standby service,
- 2. Verifying the generator capability to reject a' load of > 825 kw without tripping,
- 3. Simulating a loss of offsite power in conjunction with a safety injection signal, and:
a) Verifying de-energization of the emergency busses and load shedding from the emergency busses. b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with per-manently connected loads, energizes the auto-connected emergency loads through the load sequencer and operates for 1 5 minutes while its generator is loaoed with the emergency loads. ; l
- 4. Verifying that on a loss of power to the emergency busses, all ]
diesel generator trips, except engine overspeed, generator j differential current, and generator overexcitation are 4 automatically disabled.
- 5. Verifying the diesel generator operates for at least 60 minutes while loaded to 2 4,238 kw.
{ V BEAVER VALLEY - UNIT 2 3/4 8-3
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. t ELErlRICAi1POWELSYSTEMS -
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EURVEffJAk'EJ0VIREMENIs (continued) .
' j
? ?, / 6. Verifying that the auto-contected loads to each diesel generatcr do not exceed the 2000 hour rating cf 4,535 kw.
- 7. Verifying that the automatic load seagence timer is OPERABLE with-each load sequence time within *10% yf its required value.
,. c. Check for and remove accumulated water:
/,
/ l ,
- 1. FW.m the day tark,,< at le'ast once per $1 days and af ter each opkation'of the diesel where the period of operation was
' greater than 1 hour, and d.t fre.n'the fuel oil storage tank, at least once per 92 days.
, - 3 dm ,dt least. once per,92 hys and from new fuel oil prior to its addi-5 , lt'on t'o the storage tanks by verifying that a sample obtained in i .,
accordance with ASTM J270-1975 meets the following minimum require-ments and is Lested within the specified time limits: ): i < .
- 1. As s/on as a sample iP taken /or prior to adding new fuel to a' the' storage tank) verify in sciordance with the tests specified e in ASTM D975-1977 that t,he sarr)le has:
,' a) A water and sediment content of less than or equal to 0.05 t volume percent; b) A kinematic v,iscos'ity at 40 C of greater than or equal to M centistokes, but less than or equal to 4.1 centistokes;
/g) An API Gravity of within 0.3 degrees of 60 F, or a specific l/ , gravity of within 0.0016 at 60/66 F, when compared to the
) ,, sup41ier's certificate or an absciute specific gravity at j
' 60(6R F of greater than or equal to 0.83 but less than or j
,r aerdal to 0.89, or an API Gravity of greater than or equal I j' to 27 degrees but less than or equal to 39 degrees; and I
- 2. Within one week af ter oWining the smple, verify an impurity l l 4 level of less than 2 rill'i; rams of insolubles per 100 milliliter l is met when tested in eccordance with ASTM D2274-1970. The j l
' analysis on the sample may be performed after the addition of i new fuel oil.
- 3. Within two weeks of obtaining the sample, verify that the othcr l lf oroperties specified in Table 1 of ASTM D975-1977 and Regulatory Guide 1.137 Position 2.a are met (when tested in accordance with ASTM 0975-1977). An analysis for sulfur shall be performed in accordance with ASTM D129, ASTM D1552-1979 or ASTM D2622-1982. ;
BFAVER VALLEY.- UNIT 2 3/4 8-4
A i =v ( [ ELECTRICAL POWER SYSTEMS-SURVEILLANCE RE0VIREMENTS (Continued)
- e. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting ** both diesel' generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 10 seconds.
- f. At least once per 10 years by:
- 1) Draining each main fuel oil storage tank, removing the accumu-lated sediment, and cleaning the tank using a sodium hypcchlorii,?
solution or other appropriate cleaning solution, and
- 2) Performing a pressure test, of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code, at a test pressure equal to 110% of the system design pressure.
V r **This test shall be conducted in accordance with the manufacturer's recommen-l ( dations regarding engine prelube and warmup procedures, and as applicable regarding loading recommendations. BEAVER VALLEY - UNIT 2 3/4 8-5
ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITIQt4 FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit between the offsite transmission netwurk and the onsite Class 1E distribution system, and
- b. One diesel generator with:
- 1. Day tank containing a minimum of 350 gallons of fuel,
- 2. A fuel storage system containing a minimum of 53,225 gallons of fuel,
- 3. A fuel transfer p"mp.
APPLICABILITY: MODES 5 and 6. ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status. SLIRy111 LANCE REQUIREMENTS q (
)
4.8.1.2 The above required A.C. electrical power sources shall be f demonstrated OPERABLE by the performance of each of the Surveillance ! Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2.a.6. j l 4 l I 1 Ol : BEAVER VALLEY - UNIT 2 3/4 8-6
l ['v ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEM A.C. DISTRIBUTION - OPERATING LIMITING CONDLIl0N FOR OPERAIION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other.than the diesel generators with tie breakers open between redundant busses: 4160 volt Emergency Bts #2AE and 480V Emergency Bus #2N 4160 volt Emergency Bus #20F and 480V Emergency Bus #2P 120 volt A.C. Vital Bus #I 120 volt A.C. Vital Bus #II 120 volt A.C. Vital Bus #III 120 volt A.C. Vital Bus #IV
\ APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus to OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVELLLARCE RE0VIREMENTS ___ 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated power availability. i i
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j V BEAVER VALLEY - UNIT 2 3/4 8-7 J l
ELECTRICAL POWER SYS1 EMS A.C. DISTRIBUTION - SHUTDOWN LIlilTING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel generator ! but aligned to an OPERABLE diesel generator. 1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Bus 2 - 120 volt A.C. Vital Busses APPLICABILITY: MODES 5 and 6. i ACTION: With less than the above complement of A.C. busses OPERABLE and energized, establish CONTAINMENT INTEGRITY within 8 hours. O 51!RV11LLANCE REOU.LREMENTS ,, 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker alignment and indicated power availability, f ! e BEAVER VALLEY - UNIT 2 3/4 8-8
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A i I ELECTRICAL POWER SYSTEMS V-D.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.3 The following D.C. bus trains shall be energized and OPERABLE: TRAIN "A" (orange) consisting of 125-volt D.C. busses No. 2-1 & 2-3, 125-volt D.C. battery banks 2-1 & 2-3 & charger 2-1 and rectifier 2-3. TRAIN "B" -(purple) consisting of 125-volt D.C. busses No. 2-2 & 2-4, 125-volt D.C. battery banks 2-2 & 2-4 and charger 2-2 and rectifier 2-4. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
- a. With one of the required battery banks inoperable, restore the inoperable battery bank to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one of the required full capacity chargers or rectifiers inoper-ab'le, demonstrate the OPERABILITY of its associated battery bank by performing Surveillance Requirement 4.8.2.3.2.a.1 within one hour.
Within 4 hours place in . service spare charger 2-7 or continue the Surveillance Requirement of 4.8.2.3.2.a.1 at least once per 8 hours thereafter. If any Category A limit in Table 3.8-1 is not met, declare the battery inoperable. SURVEILLANCE RE0VIREMEHIS 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indi-cated power availability. 4.8.2.3.2 Each 125-volt battery bank and charger / rectifier shall be demon-strated CPERABLE-
- a. At least once per 7 days by verifying that: )
)
- 1. The parameters in Table 3.8-1 meet the Category A limits, and
- 2. The total battery terminal voltage is greater than or equal to 127.8 - volts on float charge.
- b. At least once per 92 days and within 7 days af ter a battery discharge with battery terminal voltage below 110 -volts, or battery overcharge with battery terminal voltage above 150 -volts, by verifying that:
BEAVER VALLEY - UNIT 2 3/4 8-9
i ELECTRIC POWER SYSTEMS SURVEILLANCE _REQMREMENTS 1 i
- 1. The parameters in Table 3.8-1 meet the Category B limits.
- 2. There is no visible corrosion at either terminals or connectors,
]
or the connection resistance of these items is less than 150 x i 10 6 ohms, and
- 3. The average electrolyte temperature of every tenth cell of connected cells is above 60 F.
- c. At least once per 18 months by verifying that:
- 1. The cells, cell plates, and battery racks show no visual indica-tion of physical damage or abnormal deterioration,
- 2. The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material,
- 3. The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 6 ohms; and
- 4. The battery charger will supply at least 100 amperes at 140-volts for at least 4 hours.
- d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the 2-hour design duty cycle when the battery is subjected to a battery service test.
- e. At least once per 60 months, during shutdown, by verifying that the l
battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval, . this performance discharge test may be performed in lieu of the battery service test.
- f. At least once per 18 months, during shutdown, performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 35% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on I previous performance tests, or is below 90% of the manuf acturer's l rating.
1 O BEAVER VALLEY - UNIT 2 3/4 8-10
-m TABLE 3.8-1 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2)
Parameter Limits for each Limits for each Allowable (3) designated pilot connected cell value for each cell connected cell Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and < " above and < \" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (c) > 2.07 volts Specifi > 1.200(b) > 1.195 Not more than Gravity a) .020 below the average of all connected cells Average of all Average of all ' A connected cells connected cells f > 1.205
> 1.195(b)
(a) Corrected for electrolyte temperature and level. (b) Or battery charging current is less than (2) amps when on charge. (c) Corrected for average electrolyte temperature. (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category 8 para-meters are within their allowable values and provided the Category 8 parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an ! inoperable battery. ' l I I
!n h V
i i l BEAVER VALLEY - UNIT 2 3/4 8-11 i
ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUT 00WN LIHlIING CONDIU ON OF OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall ; be energized and OPERABLE: l 2 - 125-volt D.C. bus systems, and 2 - 3.25-volt battery bank and chargers / rectifiers associated with the abe/e D.C. bus systems. APPLICABILITY: MODES 5 and 6. ACTION: With less than the above complement of D.C. equipment and bus system OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours. S1!RVEILLANCE RE0VIREMENTS 4.8.2.4.1 The above required 125-volt D.C. bus system shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability. 4.8.2.4.2 The above required 125-volt battery bank and chargers / rectifiers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2. O l BEAVER VALLEY - UNIT 2 3/4 8-12
4 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIBJTING CONDlIJ0N FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, t..e boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:
- a. Either a K eff f 0.95 or less, which includes a 1% ok/k conservative allowance for uncertainties, or
- b. A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservative allowance for uncertainties.
Additionally, valve 2CHS-91 shall be closed and secured in position. APPLICABILITY: MODE 6* ACTION:
- a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or
/' positive reactivity changes and initiate and continue boration at
> 30 gpm of greater than or equal to 7000 ppm boric acid solution or its equivalent until K e is reduced to < 0.95 or the boron con-centration is restored g > 2000 ppm, whichever is the more restric-tive. TheprovisionsofSEecification3.0.3 are not applicable.
- b. With valve 2CHS-91 not closed and secure in position, immediately close and secure in position.
SURVEILLANCE REQUIREMENTS , 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position.
- 4. 9.1. P. The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours.
4.9.1.3 Valve 2CHS-91 shall be verified closed and locked secure in position at least once per 31 days. I J *The reactor shall be maintained in MODE 6 when the reactor vessel head is , U unbolted or removed. I 1
)
l BEAVER VALLEY - UNIT 2 3/4 9-1 I
REFUELING OPERATIONS 1 INSTRUMENTATION LitdJIUiq_ CONDITION 0[. OPERATION _ 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY: MCDE 6. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable. SUBVEILLAliCE_REQU.1REMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance cf:
- a. A CHANNEL FUNCTIONAL TEST at least once per 7 days, and
- h. A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. A CHANNEL CHECK at least once per 12 hours during CORE ALTERATIONS.
i 1 I 1 l O l BEAVER VALLEY - UNIT 2 3/4 9-2
/^% . REFUELING OPERATIONS \~ DECAY 7TME ~ '~
l 1 \ 1-l LIMITING CDIGITION OF OPERATION 3.9.3 .The reactor shall be subcritical for at least 150 hours. APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel ACTION: 1 With the reactor subcritical for less than 150 houis, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. 1 i EURVELLLatlCE RE0VIREMENTS l 4.9.3 The reactor shall be determined to have been subcritical for at least 150 hours by verification of the date and time of subcriticality prior to move-ment of irradiated fuel in the reactor pressure vessel. (/ 1 l 1 A BEAVER VALLEY - UNIT 2 3/4 9-3
REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDlILQN FOR OP_ERAIIM - 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door closed and held in place by a minimum of four bolts,
- b. A minimum of one door in each airlock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual vt ve, or i
- 2. Exhausting at less than or equal to 7500 cfm through OPERABLE Containment Purge and Exhaust Isolation Valves with isolation times as specified in Table 3.6-1 to OPERABLE HEPA filters and charcoal adsorbers of the Supplemental Leak Collection and Release System (SLCRS).
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment. The provisions of Specification 3.0.3 are not applicable. 511RVE111MCLRE0VIREMENTS 4.9.4.1 Each of the above required containment penetrations shall be determined to be in its above required condition within 150 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containrant. 4.9.4.2 The containment purge and exhaust system shall be demonstrated OPERABLE by;
- a. Verifying the flow rate to the SLCRS at least once per 24 hours when the system is in operation.
- b. Testing the Containment Purge and Exhaust Isolation Valves per the applicable portions of Specification 4.6.3.1.2, and
- c. Testing the SLCRS per Specification 4.7.8.1 with the exception of item 4.7.8.1.c.2.
BEAVER VALLEY - UNIT 2 3/4 9-4
/
c' REFUELING OPERATI0iis
\
l COMMUNICATIONS LIMITING C0:4DITION FOR OPERATION i
' 3. 9. 5 Direct communications shall be maintained between the control room and f personnel at the refueling station. j APPLICABILITY: During CORE ALTERATIONS.
1 ACTION:
.I When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable, j SURVEILLANCE RE0VIREMENTS _
4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of g and at least once per 12 hours during CORE ALTERATIONS. m) i 1 1 BEAVER VALLEY - UNIT 2 3/4 9-5 l
REFUELING OPERATIONS MANIPULATOR CRANE OPERABILITY LIMITING CONDITIO.N FOR OPERAU Q _ 3.9.C The manipulator crane and auxiliary hoist shall be used for movement of , control rcds or fuel assemblies and shall be OPERABLE with:
- a. The manipulator crane used for movement of fuel assemblies having:
- 1. A minimum capacity of 3250 pounds, and
- 2. An overload cut off limit 5 2700 pounds.
- b. The auxiliary hoist used for movement of control rods having:
- 1. A minimum capacity of 700 pounds, and
- 2. A load indicator which shall be used to prevent lifting loads in excess of 600 pounds.
APPLICABILITY: During movement of control rods or fuel assemblies within the reactor pressure vessel. ACTION: With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use ,f any inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of control rods and fuel assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. LVRyflLLANCE RE001REMENTS 4.9.6.1 Each manipulator crane used for movement of fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 150 hours prior to the start of such operations by performing a load test of at least 3250 pounds and demonstrating an automatic load cut off when the crane load exceeds 2700 pounds. 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of control rods within the reactor pressure vessel shall be demonstrated OPERABLE within 150 hours prior to the start of such operations by performing a load test of at least 700 pounds. O BEAVER VALLEY - UNIT 2 3/4 9-6
. _ _ . - - _ _ _ _ . . _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ ._ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . _ _ _ _ = _ - _ . ___
p () REFUELING OPERATIONS i i CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING 1 LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 3000 pounds shall be prohibited from travel over fuel assemblies in the storage pool. ] j i
' APPLICABILITY: With fuel assemblies in the storage pool. l
{ ACTION: With the requirements of the above specification not satisfied, place the crane l load in a safe condition. The provisions of Specification 3.0.3 are not j applicable. ! SURVEILLANCE RE0VIREMENTS 4.9.7 Crane interlocks and phys" 1 stops which prevent crane travel with loads in excess of 3000 pounds over fu , assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. 6
\
l I l l I i l BEAVER VALLEY - UNIT 2 3/4 9-7 l I i _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - - - - _ - - . __ ---- --- - --- - -- - - J
REFUELING OPERATIONS , t 3/4 9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION LIMLIING CONDIIl0N FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* APPLICABILITY: MODE 6. ACTION:
- a. With less than one residual heat removal loop in operation, except as provided in b below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations pro-viding direct access from the containment atmosphere to the outside atmosphere within 4 hours.
- b. The residual heat ramoval loop may be removed from operation for up to 1 hour per 8 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.
- c. The residual heat removal loop may be removed from operation for up to 4 hours per 8 hour period during the performance of Ultrasonic In-service Inspection inside the reactor vessel nozzles provided there is at least 23 feet of water above the top of the reactor vessel flange.
- d. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQQ1REMENTS 4.9.8.1 At least one residual heat removal loop shall be verified to be in j operation and circulating reactor coolant at a flow rate of > 3000 gpm at least ] once per 4 hours when making boron dilution changes and > 1060 gpm for decay . heat removal when the Reactor Coolant System is in the drained down condition within the loops. 3
- Prior to initial criticality, the RHR loop may be removed from operation for up to 1 hour per 2-hcur period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
1 1 BEAVER VALLEY - UNIT 2 3/4 9-8
/g REFUELING OPERATIONS \
/ LOW WATER LEVEL LIMITING CONDITION FOR OPERAUON 3.9.8.2 _Two Residual Heat Removal (RHR) loops'shall be OPERABLE.*
APPLICABILITY: MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet. ACTION: ;
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE ,O per specification 4.0.5. (. I I O
*The normal or emergency power source may be inoperable for each RHR loop.
BEAVER VALLEY - UNIT 2 3/4 9-9
REFUELING OPERATIONS CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING _COSDlIl0N FOR OPERAT10L. 3.9.9 The Containment Purge e d Exhaust isolation system shall be OPERABLE. APPLICABILITY: During CORE .;LTERATIONS or movement of irradiated fuel within the containment. ACTION: With the Containment Purge and Exhaust isolation system inoperable, close each of the purge and exhaust penetrations providing direct access from the contain-ment atmosphere to the outside atmosphere. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE0VIREMENTS 4.9.9 ne Containment Purge and Exhaust isolation system shall be demonstrated v OPERAbLc within 150 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment Purge and Exhaust isolation occurs on manual initiation and on a high radiation signal from each of the containment radiation monitoring instrumentation channels. O BEAVER VALLEY - UNIT 2 3/4 9-10
r I l i l
/~'N REFUELING OPERATIONS
'\' 3/4 9.10 WATER LEVEL - REACTOR VESSEL k.IBITINGCONDITIONFOROPERATIDN 3.9.10 At least 23 feet of water shall be uaintained over the top of the reactor pressure vessel flange. APPLICABILITY: During-movement of fuel assemblies or. control rods within the Entainment when either the fuel assemblies being moved or the fuel assemblies i seated within the reactor vessel are irradiated while in MODE 6.
. ACTION:
With the requirements of the above specification not satisfied, suspend all
' operations involving movement of fuel assemblies or control rods within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.
S BYE]1 LANCE RE0UIREMENTS 4.9.10 The water. level shall be determined to be at least its minimum required ('l depth within 2 hours prior to the start of and at least once per 24 hours () thereafter during movement of fuel assemblies or control rods. l I BEAVER VALLEY - UNIT 2 3/4 9-11
REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION F0lL0EERATION 3.9.11 As a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. ACTION: With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE0VIREMENTS _ 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. l O BEAVER VALLEY - UNIT 2 3/4 9-12
l 1 i i i REFUELING OPERATIONS l (N) i
'v ;
FUEL BUILDING VENTILATION SYSTEM - FUEL MOVEMENT ' l LIMITING CONDITION 10R OPERATION j 3.9.12 The fuel building portion of the Supplemental Leak Collection and Release System (SLCRS) shall be operating and discharging through at least one train of the SLCRS HEPA filters and charcoal adsorbers during either:
- a. Fuel movement within the spent fuel pool, or
- b. Crane operation with loads over the spent fuel storage pool.
APPLICABILITY: When irradiated fuel which was decayed less than 60 days is in the fuel storage pool. ACTION: With the requirement of the above specification not satisfied, suspend all operations involving movement of fuel'within the storage pool or crane operation with loads over the storage pool The provisions of Specification 3.0.3 are not applicable. 1 A SURVEILLANCE RE0VIREMENTS 4.9.12 The fuel building portion of the SLCRS shall be verified to be operat- ' ing with all building doors closed within 2 hours prior to the initiation of and at least once per 12 hours during either furl movement within the fuel storage pool or crane operation with loads over the fuel storage pool. l V l BEAVER VALLEY - UNIf 2 3/4 9-13 I
REFUELING OPERATIONS FUEL BUILDING VENTILATION SYSTEM - FUEL STORAGE O4 LIMITING CORDll10N FOR OPERAIl0B ._ 3.9.13 The fuel building portion of the Supplemental Leak Collection and l Release System (SLCRS) shall be OPERABLE l APPLICABILITY: Whenever irradiated fuel is in the storage pool. l ACTION: Without the fuel building portion of the SLCRS OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until the SLCRS portion cf the fuel building ventilation system is restored to OPERABLE status. The provisions of Specification 3.0.3 are not applicable. SMRVEILLMCE RE0_VIREMENIS 4.9.13 The fuel building portion of the SCLRS shall be demonstrated OPERABLE by testing the SLCRS per Specification 4.7.8. 1 l 1 1 1 l l i O BEAVER VALLEY - UNIT 2 3/4 9-14
[] 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITINR l0llDITION FOR OPERATION _ 3.10.1 The SHUTDOWN MARGIN requirements of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided the reactivity equivalent to at least the highest estimated control rod worth is available far trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2 ACTION:
- a. With the reactor critical (Keff 1 1.0) and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at > 30 gpm of > 7000 ppm boric acid solution or its equivalent until the SHUTOOWN RARGIN required by l Specification 3.1.1.1 is restored.
- b. With the reactor subcritical (K,ff < 1.0) by less than the above s reactivity equivalent, immediately initiate and continue boration at 2 30 gpm of 2 7000 ppm boric acid solution or its equivalent until O\ -
the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. SURVEILLANCE RE0_UIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. O BEAVER VALLEY - UNIT 2 3/4 10-1
i SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS j LJMITING CDNDITION FOR OPERATI0h_ _ 3.10.2 The group height, insertion and power distribution limits of ! Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided . i
- a. The THERMAL POWER is maintained _< 85% of RATED THERMAL POWER, and
- b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and deter-mined at the frequencies specified in Specification 4.10.2.2 below. i APPLICABILITY: MODE 1 ACTION:
With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either;
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
- b. Be in HOT STANDBY within 6 hours.
SufLVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMAL
~
l POWER at least once per hour during PHYSICS TESTS. l l 4.10.2.'2 The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS:
- a. Specification 4.2.2 - At least once per 12 hours.
- b. Specification 4.2.3 - At least once per 12 hours.
O BEAVER VALLEY - UNIT 2 3/4 10-2 (.
g!,. [ b _]
~
,/~ ' SPECIALTFJTEXCEPTIONS \ k - PHYSICS TESTS sv Ii ls ., - ' LIMITING CCNDITION FOR OPERATION i ., y - e 1 J, , 3.10.3' The limitations of Specifications 3.1.1.4,.3.1.1.5, 3.1.3.1, 3.1.3.5, o and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERHAL POWER does not exceed 5% of RATED THERMAL POWER, ed
- b. The reactor trip set. points on the OPERABLE Intermediate and Power 41 Range Nuclear Channels are set at 5 25% of RATED THERMAL POWER, and c.
The is > Reactor 531 F. Coolant System. lowest. operating loop temperature (T"V9) APPLICABIll]f; MODE 2. ACTION:
- a. With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers. ,
,Q b. With a Reactor Coolant System operating loop temperture a (Tavg) 'd 531 F, restore Tavg[ *ithin its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutbs. ' q SURVEILLANT E QUIREMENTS i i 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL - { POWER at least once per hour during PHYSICS TESTS. 1 1 4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a 1 CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating FHYSICS TESTS. ,
'9 4.10.3.3 The Reactor Coolant System temperature (Tavg) shall be determined St -
to be > 531 F at least once per 30 minutes during PHYSICS TESTS. h
/~'N 'l l
l BEAVER VALLEY - UNIT 2 3/4 10-3
]
s ________--._____________-_--___-__---------_----A
qit - 1; . SPECIAL TEST EXCEPTIONS i 3/4.10.4 REACTOR < COOLANT LOOPS LIMITING _f0NDITION FOR OPERATIDF f 3.10.4 The limitations of Specification 3.4 2 2 may be suspended during the 1 performance of hot rod drop time measurements in MODE 3 previded at least two l reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE. APPLICABILITY: During performance of hot rod drop time measurements. ACTION: Witt less than the above required reactor coolant loops OPERABLE during performance of hot rod drop time measurements, immediately open the ! I' reactor trip breakers and comply with the provisions of the ACTION state-ments of Specification 3.4.1.2. SlRVELLLANCE RE0VIREME1El 4.10.4 At least-the above required reactor coolant loops shall be determined OPERABLE within 4 hours prior to initiation of the hot rod drop time measure-ments and at least once per 4 hours during the hot rod drop time measurements by verifying correct br,eaker alignments and indicated power availability and by verifying secondary side narrow range water level to be greater than or equal to 15.5 %. L 1 i O BEAVER VALLEY - UNIT 2 3/4 10-4
E p J
~N SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERhTION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full-length shutdown and control rod drop time measurements provided; 1
- a. Only one shutdown or control bank is withdrawn from the fully in-l serted position at a time, and L
- b. The rod position indicator is OPERA 3LE during the withdrawal of the
, rods.* APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements. ACTION _: I With the Position Indication Systems inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.
. t*
511RVEILLANCE REQUIREMDUS 4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:
- a. Within 12 steps when the rods are stationary, and
- b. Within 24 steps during rod motion.
I 1 l
*This requirement is not applicable during the initial calibration of the Digital Rod Position Indication System provided: (1) K is maintained eff less than or equal to 0.95, and (2) only one shutdown or control rod bank s is withdrawn from the fully inserted position at one time.
BEAVER VALLEY - UNIT 2 3/4 10-5 i _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ - . _ _ . - _ - - _ -- - - - - - - ~ ~ - ' - - ~ - ' ' - ~ ~ - - ~ ~ ~ ~
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/ 3/4.11 RADI0 ACTIVE EFFLUENTS I Q'/
i 3/4.11.1 LIQUID EFFLUENTS , e CONCENTRA,TJON
,1 ' -
;, MS $ $ .f0BDITION FOR OPERATION 1
- 3. 11, 1. 1 The concentration of radioactive n.6tenf al released at anytime from
( j 'th'e site (See Figure 5.1-1) shall be limited to the concentrations specified s '"in 10 CFR Part 20, Appendix B, Table 11, Column 2 /or radionuclides other than dissolved or entrained noble gases. For dists1ved or entrained noble gases, the concentration shall be limited to 2 x 16f pCi/ml total activit', y APPLICA8ILITY: At all time 4 , ACTION: "/
- a. With the concentration of radioactive material released from the site to unrestricted areas exceeding the sbove limits; immediately restore concen-
, tration within the above limits,'ind t >
- b. Submit a Special Report to the Commission within 30 days in accordance with Specification 6.9.2.
(t) (/ c. The provitilons of Specific.ations 3.0.3 and 3.0.4 are not applicable.
. ?
ggElj. LANCE RE0018EMENIS_..E. _ ___ _ u 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*. 4 . u .1.1. 2 The results of radioactive analysis shall be use Min accordance with the methods of the ODCM to assure that the concentrations at the point of release are maintained within. the limits of Specification 3.11.1.'1.
?
?
/- i
/
- Radioactive liquid aischarges are normally via batch moco. Tur.>ine Building l
(' Drains shall be aonitored as specified in Section 4.11.1.1.3. Recirculation ; drain pump discharge Aall be monitored as specified in Section 4.11.1.1.4. ' i BEAV;.R VALLEY - UNIT 2 3/4 11-1 i i r ,
$11RVEILLANfE REQUIREMENTS (Continued) 4.11.1.1.3 When the activity of the secondary coolant is greater than 10 6 pCi/mi gross and the Turbine Building transfer pumps (2DB5+P42, 20BS-P43, , 2DBS-P44) are not pumping their sumps to t.he steam generator blowdown tank (2SGC-TK21B), grab samples shall be taken for each sump discharge from the tur-bine building. The sample shall be analyzed for gross activity at a sensitivity of at least 10 7 pCi/ml and recorded in plant records. Water volume discharged shall be estimated from the number of pump operations unless alternate flow or volume iq.ctrumentation is provided. 4.11.1.1.4 Prior to the Recirculation Drain Pump (s) (2DAS-P215A/B) discharging to catch basin 16, a grab sample will be taken. The samples will be analyzed , for gross activity at a sensitivity of at least 10 7 pCi/ml. Water volume dis-charged shall be est.imated from the number of pump operations unless alternate flow or volume instrumentation is provided. O O BEAVER VALLEY - UN11 2 3/4 11-2 _ _ _ _ _____D
- r, _ n t o ii mt a _ i c ) 7 6 5 5 8 6 7 6 5 5 7 8 _ L e l - - - - - - - - - - - - t m 0 0 0 0 0 0 0 0 0 0 0 0 re)/ 1 1 1 1 1 1 1 1 1 1 1 1 eDDi w ofL p LC x x x x x x x x x x x x L o(( 5 1 1 1 5 1 5 1 1 1 1 5
# f s s r d r d M e e ) e e)
A t n s t ns R t i r t i r G i a e i ae O m r t m rt R E t t E tt P n i ni a E m a E m S m E m E 1 I m d m d
- S a n a 0 a na 0 1 Y y G am 9 G am a 9 1 L A
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L D cy c 1 o s 9 5 c 1 os s 9 B N eAl n 3 s e 8 5 n 3 se s 8 A A p a i 1 s s 3 - - i 1 ss 3 o - T G yf n r -1 i a - r e r -l ia - r r ToA P D G H S F P DG H G S N I L h b P h b c c C M y c e e e e A sc t t t t t t S min a i i i i i E use myu PB M Ms o 0s o Wso M Ms o Qs T il q h p p p p S nae c m m m m A i nr a o o o o W MAF E C C C C C D P I J Q' h 9 9 9 9 I h M h h e e e e L h / h h l l l l y c h c c p p p p E gc t c t t m m m m V nn a t a a a a a a I i e PB a PB PB S S S S T l u PB C pq h h h b b b b A me c e c c a a a a 0 ar a n a a r r r r I SF E O E E G G G G D A R d s 9 k en g' t a sT u e a ns p y Wes ue ns T h a i a ce t e
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TABLE 4.11-1 (Continued) TABLE NOTATION
- a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. l For a particular measurement system (which may include radiochemical separation):
LLD = 4.66 s b l (E) (V) (2.22) (Y) exp(-AAT) where: LLD is the lower limit of detection as defined above (as pCi per unit mass or volume); s is the standard deviation of the background counting rate b or of the counting rate of a blank sample as appropriate (as counts per minute); E is the counting efficiency (as counts per transformation); V is the sample size (in units of mass or volume); 2.22 is the number of transformations per minute per picocurie; Y is the fractional radiochemical yield (when applicable); A is the radioactive decay constant for the particular radionuclides; AT is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environment' samples, not plant effluent samples). The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclides determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium in milk samples). Typical values of E, V, Y and AT should be used in the calculations. The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.
- b. A composite sample is one in which the quantity of liquid sampled is pro-portional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.
BEAVER VALLEY - UNIT 2 3/4 11-4
l-I f] TABLE 4.11-1 (Continued) TABLE NOTATION
- c. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
l d. A batch relene exists when the discharge of liquid wastes is from a dis-l crete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.
- e. A continuous release exists when the discharge of liquid wastes is from a nondiscrete volume; e.g., from a volume of a system having an input flow during the continuous release. This is applicable to the Turbine Building drains when the secondary coolant gross radioactivity (beta and gamma) is greater than 10 5 pCi/ml.
- f. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the Lt.D O
Q for the analyses should be reported as "less than" the nuclide's LLD, and should not be reported as being present at the LLD level far that nuclide. The "less than" values should not be used in the required dose calculations. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semi-annual Radioactive Effluent Release Report.
- g. Whenever there is primary to secondary leakage, sampling is done for tur-bine building drain effluents by means of grab samples taken every 4 hours during the period of discharge and analyzed for gross radioactivity (beta and gamma) at a sensitivity of at least 10 7 pCi/ml and recorded in the plant records, along with the flow rate. Primary to secondary leakage is considered to be occurring whenever measurements indicate that secondary coolant gross radioactivity (beta and gamma) is greater than 10 5 pCi/ml.
In addition, two (2) plant personnel shall check release calculations to verify that the limits of 3.11.1.1 and 3.11.1.2 are not exceeaed.
- h. Whenever the Recirculation Drain Pump (s) are discharging to catch basin 16, sampling will be performed by means of a grab sample taken every 4 hours during pump operation.
O h BEAVER VALLEY - UNIT 2 3/4 11-5
RADI0 ACTIVE EFFLUENTS DOSE j.1MIIING_CORDJTION FOR OPERATION 3.11.1.2 The dose or dose commitm-ant to MEMBER (S) 0F THE PUBLIC from radio- f active materials in liquid effluents released from the reactor unit (see Figure 5.1-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, p.:rsuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases, and the proposed '
corrective actions to be taken to assure the subsequent releases will be within the above limits. (This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act).*
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable SMRVEILL ANCERERl)lEEMENIS 4.11.1.2.1 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.
O
- Applicable only if drinking water supply is taken from the receiving water body.
BEAVER VALLEY - UNIT 2 3/4 11-6
i f~} - RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT-LLMITING CONDITION FO.fLOPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be used to reduce the radioactive materials in each liquid waste batch _ prior to its discharge when the projected doses due to liquid effluent releases from the reactor unit (See Figure 5.1-2) when averaged over 31 days would exceed 0.06 mrem.to the total body or 0.2 mrem to any organ.
. APPLICABILITY: At all times.
ACTION:
- a. With liquid waste.being discharged without treatment and exceeding the limits specified, prepare and submit to the Commission within 30 days !
pursuant to Specification 6.9.2 a Special Report which includes the follow-ing information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
,O 2. Action (s) taken to restore the inoperable equipment to opera-Q tional status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTL
)
4.11.1.3.1 Doses due to liquid releases shall be projected at least once per 1 31 days, in accordance with the ODCM. i i a BEAVER VALLEY - UNIT 2 3/4 11-7
RADI0 ACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION f 3.11.1.4 The quantity of radioactive material contained in each miscellaneous temporary outside radioactive liquid storage tank shall be limited to < 10 curies, excluding tritium and dissolved or entrained noble gases. APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any of the above tanks ex-ceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit, and
- b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2 and include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when l radioactive materials are being added to the tank. O BEAVER VALLEY - UNIT 2 3/4 11-8
l
-) ;
(~} RADI0 ACTIVE EFFLUENTS ) 3/4.11.2 GASE0US EFFLUENTS i DOSE RATE l LIMITING CONDITION FOR OPIRATIOR 3.11.2.1 The dose rate in the unrestricted areas (see Figure 5.1-1) due to I radioactive materials released in gaseous effluents from the site shall be I limited to the fol'iowing values: f
- a. The dose rate limit for noble gases shall be < 500 mrem /yr to the total body and 1 3000 mrem /yr to the skin *, and
- b. The dose rate limit, inhalation pathway only, for I-131, tritium and all radionuclides in particulate form (excluding C-14) with half-lives greater than 8 days shall be i 1500 mrem /yr to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately decrease the release rate to comply with the above limit (s), and
/ '~}
- b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
EURVf1LLMCE RE0VIRE!dENIS I 4.11.2.1.1 The dose rate due to noble gaseous effluents shall be determined to l be within the abcVe limits in accordance with the methods and procedures of the ODCM. l l 4.11.2.1.2 The dose rate, inhalation pathway only, for I-131, tritium and all radionuclides in particulate form (excluding C-14) with half-lives greater than 8 days in gaseous effluents, shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by obtaining represen-tative samples and performing analyses in accordance with the sampling and anal-ysis program specified in Table 4.11-2.
*During containment purge the dose rate may be averaged over 960 minutes.
BEAVER VALLEY - UNIT 2 3/4 11-9 l
)
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TABLE 4.11-2 (Continued) l I TABLE NOTATION
- a. The Lower Limit of Detection (LLD) is defined in Table Notation (a) of Table 4.11-1 of Specification 4.11.1.1.
- b. Sampling and analysis shall also be performed following shutdown, startup, ,
or a THERMAL PJWER change exceeding 15% of RATED THERMAL POWER within a l 1-hour period. This requirement does not apply if (1) analysis shows that the Dose Equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
- c. Trit! m grab samples shall be taken at least once per 24 hours (from the Tppr priate ventilation release path) when the refueling canal is flooded.
- d. sait sles 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 fellowing each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This rec,uirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor 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.
- e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel l is in the spent fuel pool.
- f. The average 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 cal-culation made in accordance with Specification 3.11.2.1, 3.11.2.2 and 3.11.2.3.
- g. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semi-annual effluent report. I h. Only when this release path is in use. l l l l BEAVER VALLEY - UNIT 2 3/4 11-12 l
L i RADI0 ACTIVE EFFLUENTS
}'] '
DOSE-NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose from the reactor unit in unrestricted areas (See Figure 5.1-1) due to noble gases released in gaseous effluents shall be limited to the following:
- a. During any calendar quarter, to < 5 mrad for gamma radiation and < 10 mrad
~
for beta radiation.
~
L
- b. During any calendar year, to < 10 mrad for gamma radiation and < 20 mrad
~
for beta radiation.
~
' APPLICABILITY: At all times.
ACTION:
- a. With the calculated air dose from radioactive noble gases in gaseous effluent.c exceeding any of the above limits, prepare and submit to the Commissia,within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and de-fines the corrective actions taken to reduce the releases and the proposed O( corrective actions to be taken to assure the subsequent releases will be within the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i
}1!RVEILLARCE_ REQUIREMENTS 4.11.2.2 Dose Calculations. Cumulative dose contributions shall be determined in accordance with the ODCR at least once every 31 days.
l v l
]
3 BEAVER VALLEY - UNIT 2 3/4 11-13
RADI0 ACTIVE EFFLUENTS DOSE-RADIOI0 DINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR QPIRATION _ , _ . 3.11.2.3 The dose to MEMBER (S) 0F THE PUBLIC from radiciodines and radioactive materials in particulate form (excluding C-14), and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous ef fluents released from the reactor unit (see Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter to 1 7.5 mrem to any organ, and
- b. During any calendar year to i 15 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, (excluding C-14), and radionuclides (other than noble gases) with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report, which identifies the cause(s) for exceetiing the limit and defines the corrective actions taken to reduce the releases and the preposed corrective actions to be taken to assure the subsequent releases will be within the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not appliceble, i
l SUFXE1;LLANCE RE0lflREE.VNTS 4.11.2.3 Dose Calculations. Cumulative dos contributions shall be determined in accordance with the ODCM at least once every 31 days. l [ 1 O BEAVER VALLEY - UNIT 2 /4 11-14
- /' RADI0 ACTIVE EFFLUENTS I ?.
k GASE0US RADWASTE TREATMENT l 1 l LIMITINLCOND.ITION FOR OPERATION i i 3.11.2.4 -The Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste l prior to their discharge when the projected gaseous. effluent air doses due to ; gaseous effluent releases from the reactor unit (see Figure 5.1-1), when aver- j aged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The' appropriate portions of the Ventilation Exhaust Treatment i System shall be used to reduce radioactive materials in gaseous waste prior to j their discharge when the projected doses due to gaseous effluent releases from 3 the reactor unit (see Figure 5.1-1) when averaged over 31 days would exceed i 0.3 mrem to any orgam APPLICABILITY: At all times. ACTION:
- a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which in:1udes the rollowing information.
- 1. Identification of the inoperable equipment nr subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the' inoperable equipment to operational status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCEREQUIREMENTS 4.11.2.4 Doses due to gaseous releases from the site shall be projected at least once per 31 days, in accordance with the 00CM. l l i l BEAVER VALLEY - UNIT 2 3/4 11-15
RADI0 ACTIVE EFFLUENTS GASE0US WASTE STORAGE TANKS pf[UfjCt_ CONDITION FOR OPER&UON ,_ i 3.11.2.5 The quantity of radioactivity contained in any connected group of { gaseous waste storage tanks shall be limited to < 19,000 curies noble gases (considered as Xe-133). APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any connected group of gaseous waste storage tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tanks and within 48 hours re-duce the tanks contents to within the limit, and
- b. Submit a Special Report to the Commission within 30 days pursuant to Specification 6.9.2 and include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCLREQUIREMElfl$ O 4.11.2.5.1 The quantity of radioactive material contained in any connected group of gaseous waste storage tanks shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tanks. i i l i 1 l O l BEAVER VALLEY - UNII 2 3/4 11-16
q
/ RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE ' LIMITING _f0HDlJ10N FOR OPERATIOX 3.11.2.6 The concentration of oxygen in the waste gas holdup system shall be limited to 1 2% by volume whenever the hydrogen concentration exceeds 4% by volume.
APPLICABILITY: At all times. ACTION:
- a. With'the concentration of oxygen in the waste gas holdup system > 2% by volume, immediately suspend all additions of waste gases to the gaseous waste decay tank and reduce the concentration of oxygen to 1 4% within 48 hours.
- b. With the concentration of oxygen in the waste gas holdup system greater than 4% by volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the affected tank and reduce the concentration of oxygen to less than or equal to 2% by volume n within twelve hours.
, c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVELLLMCE RE0VIREMENTS 4.11.2.6 The concentrations of oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10 or monitoring in conjunction with its associated action statement. l I ( BEAVER VALLEY - UNIT 2 3/4 11-17 , I t - - - - - - - - - - - - - - - _ - - _ -
RADI0 ACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE LIMITING CQEDITION FOR OPERATION 3.11.3.1 The solid radwaste system shall be used, as applicable, to solidify j and package radioactive wastes, and to ensure meeting the requirements of i 10 CFR Part 20, 10 CFR Part 61 and of 10 CFR Part 71. Methods utilized to meet these requirements shall be described in facility procedures and in the Process Control Prograrr. (PCP). APPLICABILITY: At all times. , ACTION:
- a. With the applicable requirements of 10 CFR Part 20, 10 CFR Part 61 and 10 CFR Part /1 not satisfied, suspend affected shipments of solid radio-active wastes from the site.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are net aoplicable.
EURVflLiANCE RE0V1flEMEHIS O' 4.11.3.1.1 Prior to shipment, solidification shall be verified in accordance with Station Operating Procedures. 4.11.3.1.2 Reports. The semi-annual Radioactive Effluent Release Report in Specification 6.9.1.12 shall include the following information for each type of solid waste shipped offsite during the report period: a, container volume;
- b. total curie quantity (determined by measurement or estimate);
1 l
- c. principal radionuclides (determined by measurement or estimate);
- d. type of waste (e.g., spent resin, compacted dry waste evaporator bottoms);
l e. type of container (e.g., LSA, Type A, Type B, large Quantity);
- f. solidification agent (e.g., cement); and l g. classification and other requirements specified by 10 CFR Part 61.
Oi BEAVER VALLEY - UNIT 2 3/4 11-18 l l l _______________ -
;/'] RADI0 ACTIVE EFFLUENTS
~
;V4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4.1 The' dose or dose commitment to MEMBER (S) 0F THE PUBLIC from all facility releases is limited to < 25 mrem to the total body or any organ (except the thyroid, which is liiiited to < 75 mrem) for a calendar year.
APPLICABILITY: At all times. ACTION: .
- a. With the calculated dose from the release of radioactive materials in
-liquid or gaseous effluents exceeding twice the limits of Specifica-tions 3.11.1.2.a. 3.11.1.2.b, 3.11.2.2.a. 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, prepare and submit a Special Report to the Commission within
.30 days pursuant to Specification 6.9.2 defining the corrective action and limit the subsequent releases such that the dose or dose commitment to MEMBER (S) 0F THE PUBLIC is limited to < 25 mrem to the total body or any organ (except thyroid, which is limite3 to < 75 mrem) for a calendar year.
This special report.shall describe the steps to be taken or modifications necessary to prevent a recurrence. Otherwise, obtain a variance from the Commission to permit releases which exceed the-40-CFR 190 Standard. [ b. The provisions of Specification 3.0.3 and 3.0.4 are net applicable. SURVEILLANCE _RE0_UIREMENTS 4.11.4.1 Dose Calculations. Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifica-tions 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a. 3.11.2.2.b, 3.11.2.3.a, and 3.11.2.3.b and in accordance with the ODCM. i f
-f BEAVER VALLEY - UNIT 2 3/4 11-19
A 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING' PROGRAM LIMITING CONDITION FOR OPERATION i 3.12.1 -The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1. APPLICABILITY: At all times. l
' ACTION:
- a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological' Environmental Report,'a description of the reasons for not conducting the program as required and the plans for preventing a re-currence. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal un-availability, or to malfunction of automatic sampling equipment. If.the latter, every effort shall be made to complete corrective action prior to the end of the next. sampling period.
~
- b. With.the level of radioactivity in an environmental sampling medium at one e or more of the locations specified in Table 3.12-1 exceeding the limits of
--Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from..the end of affected calendar quarter .1 a Special Report pursuant to Specification 6.9.2 which includes an evalua-tion of any release conditions, environmental factors or other aspects which caused the limits of Table 3.12-2 to be exceeded This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report.
When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if: Concentration (1) Concentration (2)
+ + .... > 1.0 Limit Level (1) Limit Level (2)
- c. With milk or fresh leafy vegetable suples unavailable from the required number of locations selected in accordance with Specification 3.12.2 and listed in the ODCM, obtain replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table
< 3.12-1 and the ODCM provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations, if available.
- d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
BEAVER VALLEY - UNIT 2 3/4 12-1
1 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM (Continued)
. SURVEILLANCE REQUIREliEllIs 4.12.1.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1.
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- a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represerds a "real" signal.
For a particular measurement system (which may include radiochemical separation):
= 4.66 S b LLD (E) (V) (2.22) (Y) exp(-AAT) >
l where: LLD is the lower limit of detection as defined above (as pCi per unit mass ' or volume); S is the standard deviation of the background counting rate or of the b counting rate of a blank sample as appropriate (as counts per minute); E is the counting efficiency (as counts per transformation); V is the sample size (in units of mass or volume); 2.22 is the number of transformations per minute per picocurie; Y is the fractional radiochemical yield (when applicable); A is the radioactive decay constant for the particular radionuclides; AT is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environ-mental samples, not plant effluent samples). The value of S used in the calculation of the LLD for a detection system shall be basedbon the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclides determined by gamma-ray spectrometry, the back-ground shall include the typical contributions of other radionuclides normally present in the samples (e.g. , potassium-40 in milk samples). Typical values of E, V, Y and AT should be used in the calculations. The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measunment.
- b. LLD for drinking water.
- c. If parent and daughter are totaled, the most restrictive LLD should be applied.
BEAVER VALLEY - Ut41T 2 3/4 12-8
h- . RADIOLOGICAL ENVIRONMENTAL MONITORING \v '/ 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall indentify the location of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles. For elevated releases as defined in Regulatory Guide 1.111, (Rev.1) July 1977, the land use census shall also identify the locations of all milk animals and all gardens of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteo-rological sectors within a distance of three miles.
APPLICABILITY: At all times. ACTION:
- a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently beinc calculated in Specification 4.11.2.3, prepare and submit to the Commission .ithin 30 .
days, pursuant to Specification 6.9.2, a Special Report, which identifies the new location (s). (' b. With a land use census identifying a milk animal location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1 prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report, which identifies the new location. The new location shall be added to the radio-logical environmental monitoring program within 30 days, if possible. The milk sampling program shall include samples from the three active milk animal locations, having the highest calculated dose or dose commitment. Any replaced location may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.2.1 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 using that information which will provide the best results, such as by a door-to-door survey **, aerial survey, or by consulting local agriculture authorities. Q
- Broad leaf vegetation sampling may be performed at the site boundary in Q the direction secter with the highest D/Q in lieu of the garden census.
** Confirmation by telephone is equivalent to door-to-door.
BEAVER VALLEY - UNIT 2 3/4 12-9
e
) f
) . g
, RADIOLOOJCAL ENVIRONMENTAL MONITORING.
3/4.12.3 IATERLABORATORY COMPARISON PROGRAM LI MI TI flG_CONDI TIQtLE0JLQP E R ATION l 3.12.3 Analyses shall be performed on radioactive materials supp7aied as part l of an Interlaboratory Comparison Program. APPLICABILITY: At all times. ACTION:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Report.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SUBYEILLAtfCE RE0VIREMENTS 4.12.3.1 The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Report. l l l 1 O BEAVER VALLEY - UNIT 2 3/4 12-10 f
)
( BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
)
l l j] 3/4.0 APPLICABILITY LJ flASES The' specifications of this section provide the general requirements appli-cable to each of the Limiting Condition for Operation and Surveillance Requirements within Section 3/4. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is l provided to delineate specifically when each specification is applicable. 3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement. 3.0.3 This specification delineates the ACTION to be taken for circum-stances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 3.5.1 calls for each Reactor Coolant System accumulator to be OPERABLE and provides explicit ACTION requirements if one accumulator is inoperable. Under the terms of Specification 3.0.3., if more than one accumulator is inoperable, the unit is required to be in at least H0T STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. As a further ex role, Specifict-
/ tion 3.6.2.1 requires two Containment Quench Spray Subsy aems, to be OPERABLE
( and provides explicit ACTION requirements if one spray system is inoperable: Under the terms of Specification 3.0.3., if both of the required Containment Quench Spray Subsystems are inoperable, the unit is required to be in at least H0T STANDBY within 6 hours, in a least HOT SHUTDOWN within the following 6 hours and in at least COLD SHUTDOWN in the next 24 hours. It is assumed that the unit is brought to the required MODE within the required times by promptly initiating and carrying out the appropriate ACTION statement. 3.0.4 This specification provides that entry into an OPERABLE M09E, or other specified applicability condition :nust be made with (a) the full comple-ment of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements. The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded. Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications. ( 3.0.5 This specification delineates what additional conditions must be ( satisfied to permit operation to continue, consistent with the ACTION state-ments for power sources, when a normal or emergency power source is not OPERABLE. It specifically prohibits operation when one division is inoperable because its l BEAVER VALLEY - UNIT 2 B 3/4 0-1
APPLICABILITY BASES normal or emergency power source is inoperable and a system, subsystem, train, component or device in another division is inoperable for another reason. The provisions of this specification permit the ACTION statements associated with individual systems, subsystems, trains, components, or devices to be con-sistent with the ACTION statements of the associated electrical power source. It allows operation to be governed by the time limits of the ACTION statement associated with the Limiting Condition for Jperation for the normal or emergency power source, not the individual ACTION statements for each system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its normal or emergency power source. For example, Specification 3.8.1.1 requires in part that two emergency diesel generators be OPERABLE. The ACTION statement provides for a 72 hour out-of-service time when one emergency diesel generator is not OPERABLE. If the definition of OPERABLE were applied without consideration of Specifica-tion 3.0.5, all system subsystems, trains, components and devices supplied by the inoperable emergency power .,ource would also be inoperable. This would dictate invoking the applicable ACTION statements for each of the applicable Limiting Conditions for Operation. However, the provisions of Specifica-tion 3.0.5 permit the time limits for continued operation to be consistent with the ACTION statement for the inoperable emergency diesel generator instead, provided the other specified conditions are satisfied. In this case, this would mean that the corresponding normal power source must be OPERABLE, and all re-dundant systems, subsystems, trains, components, and devices must be OPERABLE, or otherwise satisfy Specification 3.0.5 (i.e, be capable of performing their design function and have at least one normal or one emergency power source OPERABLE). If they are not satisfied, action is required in accordance with this specification. As a further example, Specification 3.8.1.1 requires in part that two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system be OPERABLE. The ACTION statement pro-l vides a 24-hour out of-service time when both required offsite circuits are not OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable normal power sources, both of the offsite circuits, t would also be inoperable. This would dictate invoking the applicable ACTION l statement for the inoperable normal power sources instead, provided the other specified conditions are satisfied. In this case, this would mean that for one division the em &gency power source must be OPERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices in the other division must be OPERABLE, or like-wise satisfy Specification 3.0.5 (i.e., be capable of performing their design functions and have an emergency power source OPERABLE). In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems, trains, components and devices in both divisions must also be OPERABLE. If these conditions are not satisfied, action is required in accordance with this specification. BEAVER VALLEY - UNIT 2 B 3/4 0-2
. (N - APPLICABILITY -(
V)~ EASES , In MODES 5 or 6 Specification 3.0.5 is not applicable, and thus the indi-vidual ACTION statements for each applicable Limiting Condition for Operation j in these MODES must be adhered to. 4.0.1 This specification provides that surveillance activities necessary to ensure the Limiting Conditions for Operation are met and will bc performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification. 4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The tolerance values, taken either individually or consecutively over 3 {s test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval. 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with OPERABILITY requirements of the Limiting Ccndi-tions for Operation. Under this crit ria, equipment, systems or components are i assumed to be OPERABLE if the ar;ociated surveillance activities have been sat-isfactorily performed within the specified time interval. Nothing in this pro-vision is to be construed as drfining equipment, systems or components OPERABLE, { when such items are found or known to be inoperable although still meeting the Surveillance Requirements. 4.0.4 This specification ensures that the surveillance activities associ-ated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MODE or other appli-cable condition. The intent of this provision is to ensure that surveillance i activities have been satisfactorily demonstrated on a current basis as required i to meet the OPERABILITY requirements of the Limiting Condition for Operation. Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activi- ; ties must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status. 4.0.5 This specification ensures that inservice inspection of ASME Code ! (s Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated ! version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda i BEAVER VALLEY - UNIT 2 B 3/4 0-3 [ _ _ _ _ _ _ _ - - _ _ _ - -
APPLICABILITY , MSES as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications. This specification includes a clarification of the frequencies for perform-ing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarifi-cation is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specification.s take precedence over the ASME Boiler and Pres-sure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes pricedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declar'd inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which Ellows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. l l l I O BEAVER VALLEY - UNIT 2 B 3/4 0-4
O BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION-AND SURVEILLANCE REQUIREMENTS 6
A 3/4.1 REACTIVITYCONTJ0LSYSTEMS LV l MSES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T The most restrictive avg condition occurs at EOL, with T at no load operating temperature, and is avg associatod with a postulated steam line break accident and resulting uncontrolled i RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of l 1.77% Ak/k is initially required to control the reactivity transient. Accord-
- ingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR accident analysis assumptions. With T <200 F, avg the reactivity transients resulting from a postulated steam line break cooldown
- (E are minimal and a 1% ok/k SHUTDOWN MARGIN provides adequate protection.
The purpose of borating to the COLD SHUTDOWN boron concentration prior to blocking safety injection is to preclude a return to criticality should a steam line break cccur during plant heatup or cooldown. 3/4.1.1.3 BORON DILUTION A minimum flew rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron ' concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9370 cubic feet in approximately 30 minutes. The reactivity change rate associate with boron reductions will therefore be within the capability for operator recognition and control. 3/4.1.1.4 MODERATOR TEMPERATURE C0 EFFICIENT (MTC) The limitations on MTC are provided to ensure that the assumptions used in_the accident and transient analyses remain valid through each fuel cycle. The t, surveillance requirement for measurement of the MTC at the beginning and near the end of each fuel cycle is adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
\
BEAVER VALLEY - UNIT 2 B 3/4 1-1
i l REACTIVITY CONTROL SYSTEMS MSES j l 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALIT)Y This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the pressurizer is capable of being in an OPERABLE status with a steam bubble, 3) the reactor pressure vessel is above its minimum RI NDT temperature and 4) the protective instrumentation is within its normal operating range, l 3/4.1.2 B0 RATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each MODE of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 350 F, a minimum of two boron in-jection flow paths are provided to ensure single functional capability in the r event an assumed failure renders one of the flow paths inoperable. A.llowable ' out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. With the RCS average temperature less than 200 F, low Head Safety Injection pump may be used in lieu of the operable charging pump with a minimum cpen RCS vent of 3.14 square inches. This will provide latitude for maintenance and ISI examinations on the chstging system for repair or corrective action and will ensure that boration and makeup are available when the charging pumps are out-of-service. An open vent insures that RCS pressure will not exceed the shutoff head of the Low Head Safety Injection pumps. 2 SIS-MOV8888A and B are the Low Head Safety Injection Pump discharge isolation valves to the RCS cold legs, the valves must be closed prior to reduc-ing RCS pressure below the RWST head pressure to prevent draining into the RCS. Emergency backup power is not required since these valves are outside contain-ment and can be manually operated if required, this will allow the associated diesel generator to be taken out of service for maintenance and testing. The technical specification limit on the refueling water storage tank has been established at 859,248 gallons to account for reactivity considerations and the NPSH requirements of the ECCS system and the water required for contain-ment spray operation. O BEAVER VALLEY - UNIT 2 B 3/4 1-2
i l 4 l REACTIVITY CONTROL SYSTEMS d i BASES l 3/4.1.2 B0 RATION SYSTEMS (Continued) The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for in-jection by the ECCS in the event of either a LOCA or a steamline break. The l limits on RWST minimum volume and boron concentration ensure that: 1) suf-ficient water is available within containment to permit recirculation cooling flow to the core, 2) the reactor will remain subcritical in the cold condition (68 to 212 degrees-F) following a small break LOCA assuming complete mixing of the RWST, RCS and ECCS water volumes with all control rods inserted except the most reactive control rod assembly (ARI-1), 3) the reactor will remain subcriti-cal in the cold condition following a large break LOCA (break flow area > 3.0 ft2) assuming complete mixing of the RWST, RCS, ECCS, chemical addition tank, con-tainment spray system piping, and other water-volumes that may eventually reside in the sump Post-LOCA with all control rods assumed to be out (AR0),
- 4) long term subcriticality following a steamline break assuming ARI-1 and to preclude fuel failure.
The maximum allowable value for the RWST boron concentration forms the basis for determining the time (post-LOCA) at which operator action is required P to switch over the ECCS to hot leg recirculation in order to avoid precipi-tation of the soluble boron. The limitations for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 350 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. Subs" tuting a Low Head Safety Injection pump for a charging pump in MODES 5 and 6 will not increase the probability of an overpressure event since the shutoff head of the Low Head Safety Injection pumps is below the setpoint of the overpressure protection system. The boration capability of either system is sufficient to provide'a SHUT- ) DOWN MARGIN from all operating conditions of 1.77% ak/k after xenon decay and cooldown to 200 F. The maximum boration capability requirements occur at E0L from full power equilibrium xenon conditions and requires 13,390 gallons of 7000 ppm borated water from the boric acid storage tanks or 58,965 gallons of ' 2000 ppm borated water from the refueling water storage tank. With the RCS temperature below 350 F, one boron injection flow path is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting . CORE ALTERATIONS and positive reactivity change in the event the single injec- l tion system becomes inoperable, j l O O BEAVER VALLEi - UNIT 2 B 3/4 1-3
REACTIVITY CONTROL SYSTEMS DASES 3/4.1.2 B0 RATION SYSTEMS (Continued) The boration capability required below 200 F is sufficient to provide a SHUIDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200 F to 140 F. This condition requires either 2315 gallons of 7000 ppm borated water from the boric acid storage tanks or 10,196 gallons of 2000 ppm berated water from the refueling water storage tank. 3/4.1.3 M0VABLE CONTROL ASSEMBLIES The specifications of this section ensure that 1) acceptaSle power distri-bution limits are maintained, 2) the minimum SHUTDOWN MARGIN it maintained, and
- 3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned <ithin i 12 steps (indicated position), of the respective group demand counter pasition. The OPERABILITY of the control rod position indication system is required to deter-mine control rod positions and thereby ensure compliance witn the control rod alignment and insertion limits.
The ACTION statements which permit limited variations from the basic re-quirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement j of peaking factors and a restriction in THERMAL POWER. These restrictions pro-vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm l that the results remain valid during future operation. 1 Continuous monitoring of rod position with respect to insertion limits and rod deviation is provided by the rod insertion limit monitor and rod deviation monitor, respectively. If the rod deviation monitor o* the rod insertion limit monitor is inoperable, the frequency of manual comparison of indicated rod posi-tion is increased to an interval of at least once per 4 hours. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T avg greater than or 89" I to 541 F and with all reactor coolant pumps operating ensures that the l measured drop times will be representative of insertion times experienced dur-ing a reactor trip at operating conditions. l l BEAVER VALLEY - UNIT 2 B 3/4 1-4
l I ip 3/4.2 POWER DISTRIBUTION LIMITS U E31fs j The specifications of this section provide assura'.:e of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core > 1.30 during normal , operation and in short term transients, and (b) limiting the fission gas release, fuel pellet- temperature and cladding mechanical properties to within assumed de-sign criteria. In addition, limiting the peak linear power density during Con-dition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded. The definitions of hot channel factors as used in these specifications are as follows: F9 (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods. F H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated
/" power to the average rod power.
( 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound 9 envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associcted core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the 7% target band about the target flux differeno during rapid plant THERMAL POWER reductions, control rod motion will cause e AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RAlED THERMAL POWER (with the AFD within the target band) provided the time O BEAVER VALLEY - UN!T 2 8 3/4 2-1
POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (AFD) (Continued) duration limit of the deviation is limited. Accordingly, a 1 hour penalty de-viation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THER-MAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL r0WER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% of RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical manthly target band near the beginning of , core life. 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS F9(Z)and FN AH The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200 F. Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion differing by more than 12 steps from the group demand position.
- b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.
e BEAVER VALLEY - UNIT 2 B 3/4 2-2
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Figure B 3/4 2-1 [mi TYPICAL INDICATED AXIAL FLUX DIFFERENCE (AFD) VERSUS THERMAL POWER AT BOL t v/ BEAVER VALLEY - UNIT 2 B 3/4 2-3
l J l, POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4 2.3 HEAT FLUX AND NUCLEAR ENIMALDY HOT CHANNEL FACTORS F q (Z) AND F H (Continued)
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.
The relaxation in F H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. F H will be maintained within its limits provided conditions a thru d above, are maintained. When an F measurement is taken, both experimental error and manufacturing 9 tolerance must be allowed for. 5% is the appropriate experimental m or allow-ance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance. The specified limit of F"H contains an 8% allowance for uncerta nties which ! means that normal, full power, three loop operation will result in FAH $ 1.55/1.08. Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this reduction in the generic margin. The generic design margins, totaling 9.1% DNBR, and completely offsets any rod bow penalties (< 3% for the worst case which occurs at a burnup of 33,000 MWD /MTU). This margin includes the following:
- 1. Design Limit DNBR of 1.30 vs. 1.28
- 2. Grid Spacing (Ks ) f 0.046 vs. 0.059
- 3. Thermal Diffusion Coefficient of 0.038 vs. 0.059
- 4. DNBR Multiplier of 0.865 vs. 0.88
- 5. Pitch reduction The radial peaking factor Fxy (Z) is measurqd periodically to provide assurance that the hot channel factor, Fq (Z), rarr ains within its limit. The
) as provided in the Radial Peaking xy limit f r Rated Thermal Power (F F
Factor Limit Report per Specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core. 3/4.2.4 QUADRANT POWER TILT RATIO The Quadrant Power Tilt Ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. BEAVER VALLEY - UNIT 2 B 3/4 2-4
I-t
,q . POWER DISTRIBUTION LIMITS U BASES
'3/4.2.4 QUADRANT POWER TILT RATIO (Continued)
Radial power distribution measurements are made during startup testing and periodically during power operation. The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x y plane power tilts. The two-hour time allowance for operation with a tilt condition greater than 1.02 but~less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainly on F isq reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0. 3/4.2.5 DNB PARAMETERS' The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in g the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour basis. O BEAVER VALLEf - UNIT 2 B 3/4 2-5 '
i
'l 3/,.3 INSTRUMENTATION ~
v , BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Protection System and Engineered Safety l Featura Actuation System Instrumentation and interlocks ensure that 1) the associated action and/or reactor trip will be initiated when the parameter moni- ] ; tored by each channel or combination thereof reaches its setpoint, 2) the- j specified coincidence logic is maintained, 3) sufficient redundancy is main- ; tained to permit a channel to be out of service for testing or maintenance, 'and- d
- 4) sufficient system functional capability is available from diverse parameters. .
j The OPERABILITY of these systems is required to provide the overall relia-bility, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The inte-l grated operation of each of these systems is consistent with the assumptions used 4 in the accident analyses. The surveillance requirements specified for these j systems ensure that the overall system functional capability is maintained com-parable to the original design standards. The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability. O The Engineered Safety Feature Actuation System Instrumentation Trip Set-V points specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A setpoint is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the
. band allowed for calibration accuracy.
1 To accommodate the instrument drift assumed to occur between operational ' tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the setpoints have been specified in Table 3.3-4. Opera-tion with setpoints less conservative than the Trip Setpoint but within the l Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its trip setpoint is found to ) exceed the Allowable Value. The methodology of this option utilizes the "as ! measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncer-tainties of the instrumentation to measure the process variable and the uncer-tainties in calibrating the instrumentation. In Equation 2.2-1, Z + R + S 5 TA, the interactive effects of the errors in the rack and the sensor, and the "as I measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of ! their measurement. TA or Total Allowance is the difference, in percent span, ) between the trip setpoint and the value used in the analysis for the actuation. R or Rack Error is the "as measured" deviation, in percent span, for the affected
/
5 channel from the specified trip setpoint. S or Sensor Drift is either the "as measured' deviation of the sensor from its cal bration point or the value speci-fied in Table 3.3-4, in percent span, from the analysis assumptions %e of ' Equation 2.2-1 allows for a sensor drif t f actor, an increased rack orift factor, and provides a threshold value for REPORTABLE EVENTS. BEAVER VALLEY - UNIT 2 , B 3/4 3-1 ,
i 3/4.3 INSTRUMENTATION BASES _
'3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift ' in excess of the Allowable Value exhibits the behavior that the rack has not met its.allowante. Being that there is a small statistical chance that this Rack or sensor drift, will happen, an infrequent excessive drift is expected. in excess of the allowance that is more then occasional, may be indicative of more serious problems and should warrant further investigation. The surveillance requirements for the Manual Trip Function, Reactor Trip Breakers, and Reactor Trip Bypass Breakers are provided to reduce the poshi-bility of an Anticipated Transient Without Scram (ATWS) event by ensurinc, OPERABILITY of the diverse trip features (
Reference:
Generic Letter 85-09). The measurement of response time at the specified frequencies prevides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taFen in the analyses for those channels with response times indicated as not applicable. ESF response times specified in Table 3.3-5 which include sequential operation of the RWST and VCT valves (Notes # and ##) are based on values assumed on the non-LOCA safety analyses. These analyses take credit for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves. When sequential operation of the RWST and VCT valves is not included in the response times The LOCA (Note *), the values specified are based on the LOCA analyses. analyses take credit for injection flow regardless of the source. Verification of the response times specified in Table 3.3-5 will assure that the assumptions used for the LOCA and Non-LOCA analyses with respect to operation of the VCT and RWST valves are valid. l The maximum response time for control room isolation on high radiation l is based on ensuring that the control room remains habitable following a small line break outside the containment. From a control room habitability aspect, l l ' the worst case accident that does not initiate a Containment Isolation - Phase B signal is the small line break outside the containment. This response time J includes radiation monitor processing delays associated with the monitor averaging techniques. Diesel Generator starting and sequence loading delays are not included since these delays occur prior to the control room environ-ment exceeding the high radiation setpoint. Response time may be demonstrated by any series of sequential, overlapping or tetal channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or
- 2) utilizing replacement sensors with certified response times.
1 BEAVER VALLEY - UNIT 2 B 3/4 3-2
i 3/4.3 INSTRUMENTATION
' k.j '
' MSES l
3/4.'3.1 and'3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES j ACTUATION SYSTEM INSTRUMENTATION (Continued) The Engineered Safety Feature Actuation System interlocks perform the following functions: P Reactor tripped - Actuates turbine trip, closes main'feedwater valves on T avg below setpoint, prevents the opening of the main feedwater i valves which were closed by a safety injection or high steam generator water level signal, allows safety injection block so that components can be reset or tripped. Reactor not tripped. prevents manual block of safety injection. P -11 Above the setpoint, P-11 automatically reinstates safety injection 3 actuation on . low pressurizer pressure, automatically blocks steamline isolation on high steam pressure rate, and enables safety injection and steamline isolation (with Loop Stop Valve Open) on low steamline pressure. Below the setpoint, P-11 allows the manual block of safety injection actuation on l_ow pressurizer pressure, allows manual block
/ of safety injection and steamline isolation (with Loop Stop Valve
( Open) on Low steamline pressure and enables steamline isolation on high steam pressure rate. P-12 Above the setpoint, P-12 automatically reinstates an arming signal to the steam dump system. Below the setpoint P-12 blocks steam dump and allows manual bypass of the steam dump block to cooldown condenser dump valves. l l 1 l O BEAVER VALLEY - UNIT 2 B 3/4 3-3
s v INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that: 1) the radiation levels are continually measured in the areas served by the individual channels; 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and 3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of NUREG-0737,
" Clarification of TMI Action Plan Requirements," October, 1980.
3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of (quipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each j detector used and determining the acceptability of its voltage curve. ForthepurposeofmeasuringF(Z)orFfH,afullincorefluxmapisused. 9 Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in re-calibration of the excore neutron flux detectioi system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the Quadrant Power Tilt Ratio when one Power Range Channel is inoperable. I BEAVE VALLC.Y - UNIT 2 8 3/4 3-4
I i p INSTRUMENTATION \) mu l 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of'the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. -This capabil-ity is required to permit comparison of the measured response to that used in the design basis for the facility and is consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes." . 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that suffi-cient meteorological data is available for estimating potential radiation doses ' to the public as a result of routine or accidental release of radioactive mate-rials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent witl the recommendations of Regulatory Guide 1.23, "0nsite Meteorological Progran,s." ( i 3/4.3.3.5 REMOTE SHUT]0WN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that suffi-cient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. 3/4.3.3.6 (This Specification number is not used). 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection systems ensures that sufficient l capability is available to promptly detect and initiate protective action in 1 the event of an accidental chlorine release. This capability is required to ; protect control room personnel and is consistent with the recommendations of ' Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators
-Against an Accidental Chlorine Release," January 1977.
I l 1 O ; BEAVER VALLEY - UNIT 2 8 3/4 3-5
INSTRUMENTATION BASES 3/4.3.3.8 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations." 3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Th9 r dioactive liquid effluent instrumentation is provided to monitor and control, at applicable, the releases of radioactive naterials in liquid effluents during actul or potential releases of liquid effluents. The alarm / trip set-points.for these instruments shall be calculated in accordance with the proce-dures of ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. 3/4.3.3.10 RADI0 ACTIVE GASE0US 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 / I trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceed-ing the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup 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. i 3/4.3.4 TUkBINE OVERSPEED PROTECTION l This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine j excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures. O BEAVER VALLEY - UNIT 2 B 3/4 3-6
3/4.4 REACTOR COOLANT SYSTEM l() BASES , 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in opera-tion and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2, with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within l 6 hours. In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a subcritical condition, two operating coolant loops are required to meet the DNB design basis for this Condition 11 event when the rod control system is capable of control bank rod withdrawal. In MODES 4 and 5, a single reactor coolant loop or RHR subsystem provides sufficient heat removal capability for removing decay heat; but aingle failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. The operation of one Reactor Coolant-Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. j The reactivity change rate associated with boron reduction will, therefore, be l within the capability of operator recognition and control. The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs ltss than or equal to 350 F are provided to prevent RCS pressure tran-sients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against . overpressure transients and will not exceed the limits of Appendix G by restrict- i ing starting of the RCPs to when the secondary water temperature of each steam ) generator is less than 50 F above each of the RCS cold leg temperatures. ! l BEAVER VALLEY - UNIT 2 B 3/4 4-1
REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being j pressurized above its Safety Limit of 2735 psig. Each safety valve is designed ' to relieve 345,000 lbs. per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, pro-vides overpressure relief capability and will prevent RCS overpressurization. During operation, all pressurizer code safet" valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.4 PRESSURIZER The requirement that 150 kw of pressurizer heaters and their associated controls and emergency bus provides assurance that these heaters can be ener-gized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY. 3/4.4.5 STEAM GENERATORS { One OPERABLE steam generator in a non-isolated reactor coolant loop pro-vides sufficient heat removal capability to remove decay heat after a reactor shutdown. The requirement for two OPERABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate l l
)
O BEAVER VALLEY - UNIT 2 B 3/4 4-2 1
- REACTOR COOLANT SYSTEM
( 1
\_ -
BASES 3/4.4.5 STEAM GENERATORS (Continued) decay heat removal capabilities for RCS temperatures greater than 350 F if.one steam generator becomes inoperable due to single failure considerations. .Below 350 F, decay heat is removed by the RHR system. The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation.due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubiag also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. l The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant p chemistry is not maintained within these parameter limits, localized corrosion i may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube l leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during opera-tion will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and . plugged. ' Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube exami-nations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40% of the tube nomir.al wall thickness. Steam generator tube inspections of operating plants have demonstrated the capebility to reliably detect degradation that has penetrated 20% of the original tube wall thickness. j l Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.6 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in y a requirement for analysis, laboratory examinations, test, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. BEAVER VALLEY - UNIT 2 8 3/4 4-3
' REACTOR COOLANT SYSTEM BASEL _. _ __
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are pro-vided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.- These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems." 3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage. The 10 GPM IDENTIFIED LEAKAGE limitation provides allowanc'e for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems. The CONTROLLED LEAKAGE limitation restricts operation when .the total f'ow supplied to the reactor coolant pump seals exceeds 28 GPM with the modulaf.ng valve in the supply line fully open at RCS pressures in excess of 2235 p<ig. This limitation ensures that in the event of a LOCA, the safety injecti n flow will not be less than assumed in the accident analyses. The total steam generator tube leakage limit of 1 GoM for all steam genera-tors not isolated from the RCS ensures that the dosage contiit.ution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 limits in the event of either a steam generator tube rupture or steam line break. The i GPM limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure. 3/4.4.6.3 PRESSURE ISOLATION VALVE LEAKAGE The leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are BEAVER VALLEY - UNIT 2 B 3/4 4-4
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BASES 3/4.4.6.3 PRESSURE ISOLATION. VALVE LEAKAGE (Continued) important in preventing overpressurization and rnpture of the ECCS low pressure piping which could result in a LOCA, thesi valves should be tested )
periodically to ensure low probability of gross failure. . l The Surveillance Requirements for RCS pressure isolation valves provide l added assurance of valve integrity thereby reducing'the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure , isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. 3/4.4.7 CHEMISTRY l The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life. ! (m) of the plant. The associated effects of exceeding the oxygen, chloride and V fluoride limits are time and temperature dependent. Corrosion studies show that' operation may be continued with contaminant concentration levels in excess , of the Steady. State Limits, up to the Transient Limits, for the specified limited ! time intervals without having a significant effect on the structural integrity l of the Reactor Coolant System. The time interval permitting continued cperation j within the restrictions of the Transient Limits provides time for taking correc- I tive actions to restore the contaminant concentrations to within the Steady State Limits. The surveillance requirements provide adequate assurance that concentra-tions in excess of the limits will be detected in sufficient time to take correctiva action. j l 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appro-priately small fraction of 10 CFR Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to- I secondary steam generator leakage rate of 1.0 GPM. l l The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accom-modates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval BEAVER VALLEY - UNIT 2 B 3/4 4-5
ret.CTOR COOLANT SYSTEM j gg1ES __ 3/4.4.8 SPECIFIC ACTIVITY (Continued) or exceeding the limits shown on Figure 3.4-1 must be restricted since the ac-tivity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. Reducing T avg to < 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveil-lance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take cor-rective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location. The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the con-trolling location. The thermal gradients established during heatup produce ten-sile stresses at the outer wall of the vesscl. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling i nation, each heatup rate of interest must be analyzed on an individual basis. BEAVER VALLEY - UNIT 2 B 3/4 4-6
.[ REACTOR COOLANT SYSTEM
( BASES 3/4.4.9 PRESSURE / TEMPERATURE' LIMITS (Continued) The heatup limit curve, Figure 3.4-2, is a composite curve which was pre-pared by determining the most conservative case, with either the inside or out-side wall controlling, for any heatup rate up to 60 F per hour. The cocidown limit curves Figure 3 4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the
.inside wall where the cooldown thermal. gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cool-down curves were prepared based upon the most limiting value of 'Se predicted adjusted refe nce temperature at the end of 10 EFPY, The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor opera-tion and resultant fast neutron (E >1 Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the fluence, NDT.
copper conter and phosphorus content of the material in question, can be pre-dicted using s'igures B 3/4.4-1 and Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damar #as to Reactor Vessel Materials." O V The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT NDT as well as adjustments for possible errors in the pressura and temperature sensing instruments. Additionally, these curves are not impacted by the special 10 CFR Part 50 rules for closure flange regions due to the lo'.; initial RT NDT f the flange material.
]
Heatup and cooldown limit curves are calculated using the most limiting value of RTND7 (reference ni' ductility temperature). The most limiting RT NDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART RT is designated as the higher of either the drop weight NDT. NDT nil-ductility transition temperature (TNDT) r the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expan-sion (normal to the major working direction) minus 60 F. RT NDT increases as the material is exposed to fast-neutron radiation. Thus, to find the most limiting RT NDT at any time period in the reactor's life, ART NDT due to the radiation exposure associated with that time paried must be added to the original unirradiated RT The extent of the shift in RTNDT is enhanced NDT. by certain chemical elements (sur.h as copper and phosphorus) present in reactor i vessel steels. The Regulatory tiuide 1.99 Revisica 1 curves which show the effect l of fluence, copper content and phosphorus content on ART f r reador vessel NDT steels are shown in Figure B 3/4.4-2. j l BEAVER VALLEY - UNIT 2 B 3/4 4-7
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REACTOR COOLANT SYSTEM l BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) Given the copper and phosphorous contents of the most limiting material, the radiation-induced ARTNDT can be estimated from Figure B 3/4.4-2. Fast-neutron fluence (E > 1 Mev) at the 1/4 T (wall thickness) and 3/4 T (wall thick-ness) vessel locations are given as a function of full power service life in Figure B 3/4.4-1. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RT NDT' The preirradiation fracture-toughness properties of the Beaver Valley Unit 2 reactor vessel materials are presented in Table B 3/4.4-1. The fracture tough-ness properties of the ferritic mater!al in the reactor coolant pressure boundary are determined in accordance with the 1972 Summer Addenda to Section III of the ASME Boiler and Vessel Code. The ASME approach for calculating the allowable limit curves for various heatup ard cooldown rates specifies that the total stress intensity factor, Ky , for the combined thermal and pressure stresses at any time during hertup and cooldown cannot be greater than the reference stress intensity factor, KIR, f r b the metal temperature at that time. K is obtained from the reference fracture IR toughness curve, defined in Appendix G to the ASME Code.2 The K curve is given by the equation: IR KIR = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)] (4-1) whert K yg is the reference stress intensity factor as a function of the metal temperature T and the metal reference nilductility temperature, RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G to the ASME Code 2 as follows: CKIH + kit < KIR (4-2) where K IM is the stress intensity factor caused by membrane (pressure) stress K It is the stress intensity factor caused by the thermal gradients i K IR is a function of temperature to the RT f the material NDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical 2 ASME Boiler ar.d Pressure Vessel Code, Section III, Division 1 - Appendices,
" Rules for Construction of Nuclear Vessels," Appendix G. " Protection Against Nonductile Failure," pp. 559-569, 1980 Edition, American Society of Mechanical Engineers, New York, 1983.
BEAVER VALLEY - UNIT 2 B 3/4 4-11
_R.F. COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value fur RTNDT, and the. reference fracture toughness curve. The thermal stresses result ing from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factors, kit, f r the reference flew are computed. From equation 4-2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. For the calculation of the allowable pressure-versus-coolant temperature { during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is al- ; ways at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increases with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit ensures the constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material 9l!, temperature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel l location is at a higher tempt- .ture than the fluid adjacent to the vessel ID. j This condition, of course, is not true for the steady-state situation. It fol- { lows that, at any given reactor coolant temperature, the AT developed during ( cooldown results in a higher value of KIR at the 1/4 T location for finite cool j down rates than for steady-state operation. Furthermore, if conditions exist l ) such that the increase in K IR exceeds K It, the calculated allowable pressure ) during cooldown will be greater than the steady-state value. The above procedures are needed tm luse there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknow-ingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressurc-l temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gradients during heatup produce compres- , sive stresses at the inside of the wall that alleviate the tensile stresses l produced by internal pressure. The metal temperature at the crack tip lags the l coolant temperature; therefore, the KIR f r the 1/4 T crack during heatup is BEAVER VALLEY - UNIT 2 B 3/4 4-12
p REACTOR COOLANT SYSTEM EASES 3/4.4.9 -PRESSURE / TEMPERATURE LIMITS (Continued) lower than the K IR f r the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K 's do not IR offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of pressure-temeprature limitations for the case in which a 1/4 T deep outside sur-face flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of O heatup and the time (or coolant temperature) along the heatup ramp. Since the (' thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and iinite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three valuec taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. The actual shift in RT NDT f the vessel material will be established period-ically during operation by removing and evaluating, in accordance with 10 CFR 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vcssel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially iden-tical, the measured transition shift for a sample can be applied with confidence to the adjacent sectien of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART NOT cetermined from the surveillance capsule is different from the calculated ART NDT f r the equivalent capsule radiation s exposure. BEAVER VALLEY - UNIT 2 B 3/4 4-13
REACTOR COOLANT SYSTEM EASES , 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic i testing. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of spendix H to 10 CFR Part 50. The limitations imposed on the pressurizer heatup and cooldown rates 'nd 1 auxiliary spray water temperature W fferential are provided to assure that the pressurizer is operated within the decign criteria assumed for the fatigue analysis performed in accordance with t?e ASME Code requirements. The OPERABILITY of two PORVs or an RCi vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure tran-sients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are < 350 F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tempera-ture of the steam generator < 50 F above the RCS cold leg temperature or (2)thestartofachargingEumpanditsinjectionintoawatersolidRCS. OVERPRESSURE PROTECTION SYSTEMS The Maximum Allowed PORV Setpoint for the Overpressure Protection Systems (0PPS) is derived by analysis which models the performance of the OPPS assuming various mass input and heat input transients. Operation with a PORV setpoint less than or equal to t'ne maximum setpoint ensures that nominal 10 EFPY Appen-dix G limits will not be violated with consideration for: (1) a maximum pressure overshoot beyona the PORV setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50 F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for OPPS; (3) instrument uncertainties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed canact occur, Technical Specifications require lockout of all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and Jisallow start of an RCP if secondary coolant l temperature is more than 50 F above reactor coolant temperature. Exceptions to these requirements are acceptable as described below. Operation above 350 F but less than 375 F with only one centrifugal charg-ing pump dPERABLE is allowed for up to 4 hours. As shown by analysis LOCAs occurring at low temperature, low pressure conditions can be successfully miti- - gated by the operation of a single centrifugal charging pump and a sing!e LHSI pump with no credit for accumulator injection. Given the short time duration i BEAVER VALLEY - UNIT 2 B 3/4 4-14
a-1 1 n REACTOR COOLANT SYSTEM g b BASES 1 OVERPRESSURE' PROTECTION' SYSTEMS (Continued) j i that the' condition of having only one centrifugal charging pump OPERABLE is 1 allowed and the probability of,a LOCA occurrirs during this time, the failure of the single centrifugal charging pump is not assumed. {j Operation below 350 F but greater than 325 F with all centrifugal charging pumps OPERABLE is allowed for up to 4 hours. During low pressure, low tempera- 1 ture operation'all automatic Safety Injection actuation signals are blocked. ' In normal conditions a single failure of.the ESF actuation circuitry will: result-in the starting of at most one train of Safety Injection (one centrifugal _ charging pump, and one LHSI. pump). For temperatures above 325 F,'an overpress-ure event occurring as a result of starting these two pumps can be successfully mitigated'by' operation of both PORVs without exceeding Appendix G limit. Given i the short time duration that this condition is allowed and the low probability
.of a single failure causing an overpressure event during this time, the. single failure of a PORV is not assumed. Initiation of both trains.of Safety Injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered'to be credible accidents.
The maximum allowed PORV setpoint for the Overpressure Protection System-O will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part.50, Appendix H and in accordance with the schedule in Table 4.4-5. l 3/4.4.10 STRUCTURAL INTEGRITY l The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the J life of the plant. These programs are in accordance with Section XI of.the ASME Boiler and Pressure Vessel Code and applicable' Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g)(6)(i).
+
BEAVER VALLEY - UNIT 2 B 3/4 4-15
REACTOR COOLANT SYSTEM MSEs 3/4.4.11 REACTOR COOLANT SYSTEM RELIEF VALVES The relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is supplied from an emer-gency power source to ensure the ability to seal this possible RCS leakage path. The operability of at least one PORV will ensure the additional capability to vent the pressurizer steam space via the PORV's. 3/4.4.12 REACTOR COOLANT SYSTEM HEAD VENTS Reactor Coolant System Vents are provided to exhaust ncncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head or the pressurizer steam space via the PORV's ensures the capability exists to perform th., function. The valve reduHancy of the Reactor Coolant System Head vent paths serves to minimize the preoability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant ' System vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements", November 1980. O l BEAVER VALLEY - UNIT 2 B 3/4 4-16
Q 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)- V QASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each of the RCS accumulators ensures that a suffi-cient volume of bo,'ated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core i provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure i
.that the assumptions used for accumulator injection in the accident analysis j are met.
The limit of one hour for operation with an inoperable accumulator minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding ; temperatures. The RCS accumulators are isolated when RCS pressure is reduced to 1000 + ! 100 psig to prevent borated water frem being injected into the RCS during normal i plant con!down and depressurization conditions and also to prevent inadvertent overpressurization of the RCS at reduced RCS temperature. V 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two separate and independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consider- l ation. Either subsystem operating in conjunction with the accumulators is cap-eble of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe dovnward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period. The surveillance requirements provided to insure OPERABILITY of each component ensure that at a minimum, the assumptions used in the accident analyses are met and that subsystem 0PERABILITY is maintained. The limitation for a maximum of one charging pump to be OPERABLE and the surveillance requirement to verify all chargirg pumps except the required OPER-ABLE pump to be inoperable below 350 F providt s assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. BEAVER VALLEY - UNIT 2 B 3/4 5-1
- 3/4.6 ' CONTAINMENT. SYSTEMS L;
BASES _ 3/4.'6.1 PRIMARY. CONTAINMENT
'3/4.6.1.1 CONTAINMENT INTEGRITY
. Primary CONTAINMENT INTEGRITY ensures that the release of radioactive.
materials from the. containment atmosphere will be restricted.to those leakage p paths and associatea leak rates assumed in the accident analyses. This l restriction, in conjunction with the leakage rate limitation, will limit the j site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions'. 3/4.6.1.2- CONTAINMENT LEAKAGE The' ! imitations on containment leakage rates' ensure that the total I containment leakage' volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P a. As an added conservatism, the l
' measured overall integrated leakage rate is further-limited to < 0.75 La
- during performance of the periodic test to account for possible degradation of
'the containment leakage barriers between leakage tests.
1 ( The surveillance testing for measuring leakage rates are. consistent with the requirements of Appendix "J" of 10 CFR 50. 3/4.6.1.3 CONTAINMENT' AIR LOCKS The limitations on closure and leak rate for the containment air locks are required.to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. 3/4.6.1.4and3/4.6.1.5 INTERNAL PRESSURE AND AIR TEMPERATURE The limitations on containment internal pressure and average air temperature as a function of'RWST temperature ensure that 1) the containment structure is prevented from exceeding its design negative pressure of 8.0 psia, 2) the_ con-tainment peak pressure does not exceed the design pressure of 45 psig during LOCA conditions, and 3) the containment pressure is returned to subatmospheric conditions following a LOCA. 1 The containment internal pressure and temperature limits shown as a function of RWST and. service water temperature describe the operational envelope that will 1) limit the containment peak pressure to less than its design value BEAVER VALLEY - UNIT 2 B 3/4 6-1 u . .. .. _ - - -
l 1 i I CONTAINMENT SYSTEMS I BASES , i 3/4.6.1.4'AND 3/4.6.1.5 INTERNAL PRESSURE AND AIR TEMPERATURE (Continued) of 45 psig and 2) ensure the containment internal pressure returns I subatmospheric within 60 minutes following a LOCA. l The limits on the parameters of Figure 3.6-1 are consistent with the assumpticiis of the accident analyses. 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressu7 of 44.7 psig in the event of a LOCA. The visual and Type A leakage tests are sufficient to demonstrate this capability. 3/4.6.2 DEPRESSURIZATIUN AND COOLING SYSTEMS 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT QUENCH AND RECIRCULATION SPRAY SYSTEMS The OPERABILITY of the containment spray systems ensures that containment depressurization and subsequent return to subatmospheric pressure will occur in the event of a LOCA. The pressure reduction and resultant termination of containment leakage are consistent with the assumptions used in the accident analyses. 3/4.6.2.3 CHEMICAL ADDITION SYSTEM The OPERABILITY of the chemical addition system ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on Na0H minimum volume and concentration, ensure that 1) the iodine removal efficiency.of the spray water is maintained because of the increase in pH value, and 2) corrosion effects on components within containment are minimized. These assumptions are consistent with the iodine reme" 1 efficiency assumed in the accident analyses. 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the con-tainment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressuri-zation of the containment. Containment isolation within the time limits speci-fied ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for both a LOCA and major secondary system breaks. BEAVER VALLEY - UNIT 2 B 3/4 6-2
i i L i
/N CONTAINMENT SYSTEMS (v)- l 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection )
and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of con-trolling the expected hydrogen generation associated with 1) zirconium-water j reactions, 2) radiolytic decomposition of water, and 3) corrosion of metals within containment. These hydrogen control systems are consistent with the recommendations of Regulatory Guice 1 7, " Control of Combustible Gas Concentra-tions in Containment Following a LOCA." 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3/4.6.5.1 STEAM JET AIR EJECTOR 1he closure of the manual isolation valves in the suction of the steam jet air ejector ensures that 1) the containment internal pressure may be maintained within its operation limits by the mechanical vacuum pumps and ('~'% 2) the containment atmosphere is isolated from the outside environnient in the ( ,) event of a LOCA. These valves are required to be closed for containment 1 solation. l l a l
,e \s. J l
BEAVER VALLEY - UNIT 2 B 3/4 6-3 1
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Q 3/4.7 PLANT SYSTEMS BASES ; 3/4.7.1 TURBINE CYCLE j l 3/4.7.1.1 SAFETY VALVES ' The OPERABILITY of the main steam line code safety valves ensures that the secondary system pre.ssure will be limited to within itt design pressure of l 1085 psig during the most severe anticipated system operational transient. The mar,imum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink i (i.e., no steam bypass to the condenser). 1 The speelfied valve lift settings and relieving capacities are in accord- 1 ance'with tLe requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition and Winter 1972 Addenda. The total relieving capacity for all valves on a?1 of the steam lines is 12.7 x 106 lbs/hr which is 110 percent of , the total secondary steam flow of 11.6 x 106 lbs/hr at 100% RATED THERMAL POWER. l A minimum of 2 OPERABLE safety valves per operable steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1. ) b) V STARTUP ana/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction I in secondary system r. team flow and THERMAL POWER required by the reduced reactor I trip settings of the Power Range Neutron Flux channels. The reactor trip ' reductions are derived on the following bases: For N loop operation q cn (X) - (Y)(V) x (109) y I Where: SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V = maximum number of inoperable safety valves per steam linn I (109) = Power Range Neutron Flux-High Trip Setpcint for (N) loop operation i O BEAVER VALLEY - UNIT 2 B 3/4 7-1
p.7 PLANT SYSTEMS 4 MSIS ___ 3/4.7.1.1 SAFETY VALVES (Continued), X = Total relieving capacity of all safety valves per steam line in lbs/ hour (4,242,375) Y = Maximva relieving capacity of one safety valve in lbs/ hour (848,475) 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coulant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss of offsite power. Each electric driven auxiliary feedwater pump is capable of delivering a ; total feedwater flow of 350 gpm at a pressure of 1133 psig to the entrance of the team generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1133 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the Residual Heat Removal System may be placed into operation. 3/4.7.1.3 PRIMAP.Y PLANT DEMINERALIZED WATER (PPDW) The OPERABILITY of the PPDW storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours with steam discharge to atmosphere. 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 0.35 gpm primary-to secondary tube leak in the steam gene.ator of the affected steam line. These values are consistent with the assumptions used in the accident analyses. O i BEAVER VALLEY - UNIT 2 8 3/4 7-2 i (I - w
I' D '3/4.7 PLANT SYSTEMS l[j '\ l MSfl. m 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rup-ture. This restrict' ion is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown assc;iated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator average impact values taken at 10 F and are sufficient to prevent brittle fracture. ( ' ( 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM The OPERABILITY of the primary component cooling water system ensures that { sufficient tooling capacity is available for continued operation of safety ! related equipment during normal and accident conditions. The redundant cooling i capacity of this system, assuming a single failure, is consistent witn the assumptions used in the accident analyses. 3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during 3 normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the acci-dent conditions. 3/4.7.5 ULTIMATE HEAT SINK i The limitations on the ultimate heat sink level and temperature ensure that sufficient cooling capacity is available to either 1) provide normal cool-down of the facility, or 2) to mitigate the effects or ar.cident conditioils within i' acceptable limits. f% The limitations on minimum water level and maximum temperature are based l (A ] on providing a 30 day cooling water cupply to safety related equipment without l I BEAVER VALLEY - UNIT 2 B 3/4 7-3 l 1
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3/4.7 PLANT SYSTEMS l BASES 3/4.7.5 ULTIMATE HEAT SINK (Contfnued) exceeding their design basis temperature and is consistent with the recommenda-tions of Regulatory Guide 1.27. " Ultimate Heat Sink for Nuclear Plants." 3/4.7.6 FLOOD PROTECTION j i The limitation on flood level ensures that facility operation will be ter-minated in the event of flood conditions. The limit of elevation 695 Mean Sea Level was selected on an arbitrary basis as an appropriate flood level at which to terminate further operation and initiate flood protection measures for safety I related equipment. 3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP AND PRESSURIZATION SYSTEM The OPERABILITY of the control room emergency air cleanup and pressuriza-tion system ensures that the control room will remain habitable with respect to potential radiation and chlorine hazards for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radia- ! tion exposure to personnel occupying the control room to 5 rem or less whole i body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50. The control room air cleanup system includes two pressurization systems. The filtration pressurization system draws outside air through filters. The bottled air pressurization system pressurizes by discharge of air from bottles without filtration and with closure of intake and exhaust dampers. Although the bottles are shared with Unit 1, the discharge can be initiated by Unit 2 control systems in response to chlorine or radiation levels. Closure of the intake and exhaust dampers can be initiated by Unit 2 control systems. However, closure of dampers in one intake and in one exhaust is dependent upon w ailabil-ity of Unit 1 power sources. 3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS) The OPERABILIT' of the SLCRS provides for the filtering of postulated radio- 1 active effluents resulting from a Fuel Handling Accident (FHA) and from leakage of loss of coolant accident (LOCA) activity from systems outside of the Reactor ! Containment building, such as Engineered Safeguards Features (ESF) equipment, prior to their release to the environment. Tnis system also collects potential leakage of LOCA activity from the Reactor Containment building penetrations , into the contiguous areas ventilated by the S'.CRS except for the Emergency Air Lock. The operation of this system was assumed in calculating the postulated offsite doses in tt.e analysis for a FHA. System operation was also assumed in that portion of the Design Basis Accitient (DBA) LOCA analysis which addressed ESF leakage following the LOCA, however, no credit for CLCRS operation was taken in the DBA LOCA analysis for collection and filtration of Reactor Con-tainment bui! ding leakage even though an unquantifiable amount cf :ontiguous area penetration leakage would in fact be tollected and filtered. Based on the i results of the analyses, the SLCRS must be OPERABLE to ensure that ESF leakage ; following the postulated DBA LOCA and leakage resulting from a FHA will not exceed 10 CFR 100 limits. BEAVER VALLEY - UNIT 2 B 3/4 7-4 ,
1 I i b 3/4.7 PLANT' SYSTEMS j MSliS ; 3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or iriividual organ irradiation does not exceed allowable limits in the event of ings stion or inhalation of the source material.. The limitations on removable conta,ination for sources requiring leak testing, including alpha emitters, is based ,1 10 CFR 70.39(c) limits for plutonium. Leakage of sources excluded from the requirements of this specification represent less than one ! maximum permissible body burden for total body irradiation if the source material is inhaled or ingested. i 3/4.7.10 and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR) Deleted 1 3/4.7.12 SNUBBERS n All snubbers are required OPERABLE to ensure that the structural integrity V) i of the reactor coolant system and all other safety-related systems is main-tained during and following a seismic or other similar event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on i nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related system. l The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined q by the number of inoperable snubbers found during an inspection. Inspections J performed before that interval has elapsed may be used as a new reference point ' to determine the next inspection. When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically ( susceptible, and verified OPERABLE by inservice functional testing, that snubber ] may be exempted from being counted as inoperable. Generically, susceptible j snubbers are those which are of a specific make or model and have the same l design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation and vibration. When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to O determine if any safety-related component or system has been adversely affected 1 Q by the inoperability of the snubber. The engineering evaluation shall determine I whether or not the snubber mode of failure has imparted a significant effect or degradation en the supported component or system. j I i BEAVER VALLEY - UNIT 2 B 3/4 7-5 j
- - - - - - - - - - _ - - a
I I i PLANTSYSTEljS M1F3 ) I SNUBBERS (Continued) 1 To provide assurance of snubber functional reliability, a representative j sample of the installed snubbers will be functionally tested during plant shutdowns at refueling or 18 month intervals not to exceed two (2) years. Observed failures of these sample snubbers shall require functional testing of additional units. The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc...). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber life review are not intended to affect plant operation. 3/4.7.13 STANDBY SERVICE WATER SYSTEM (SWE) The OPERABILITY of the SWE ensures that sufficient cooling capacity is available to bring the reactor to a cold shutdown condition in the event that a barge explosion at the station's intake structure or any other extremely remote event would render all of the normal Service Water System supply pumps inoperable. l l l I t 1 1 O BEAVER VALLEY - UNIT 2 B 3/4 7-6
3/4.8 ELECTRICAL POWER SYSTEMS
,f n)~
V MSES 3/4.8.1, 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS
'The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related eqtipment required'for 1) the safe
-shutdown of the facility and 2) the mitigation and control of accident
! conditions within the facility. The minimum spec fied independent and i redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50. The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consis- l tent with the initial condition assumptions of the safety analyses and are based ; upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coin-cident with an assumed loss of offsite power and single failure of the other onsite A.C. source. l The OPERABILITY of the minimum specified A.C. and D.C. power sources and ; p associated distribution systems during shutdown and refueling ensures that ' ( 1) the facility can be maintained in the rhutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. The surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are based on the recommendations of Regulatory Guides 1.9, Revision 2, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," December 1979; 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977; and 1.137, " Fuel-011 Systems for Standby Diesel Generators," Revision 1, October 1979, Appendix A to Generic Letter 84-15 and Generic Letter 83-26, " Clarification of Surveillance Requirements for Diesel Fuel Impurity Level Tests." The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129, "tiaintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lcad Storage < Batteries for Generating Stations and Substations. l
)
Verifying average electrolyte temperature above the minimum for which the aattery was shed, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated i capacity. BEAVER VALLEY - UNIT 2 B 3/4 8-1
3/4.8 ELECTRICAL POWER SYSTEMS 1 flASEs 3/4.8.1, 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION (Continued) , 1 Table 3.8-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 3.8-1 is permitted for up to 7 days. During this 7 day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less j than the safety margin provided in sizing; 3) the allowable value for an j individual cell's specific gravity, ensures that an individual cell's specific j gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and 4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function. 1 I 1 j l i BEAVER VALLEY - UNIT 2 B 3/4 8-2
I l L 3/4.9 REFUELING OPERATIONS 1 / Q lWSES l 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensure that:
- 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. 'The limitation on K I eff ,
no greater than 0.95 which includes a conservative allowance for uncertainties, I is sufficient to prevent reactor criticality during refueling operations. Isolating all reactor water makeup paths from unborated water sources pre- i cludes the possibility of an uncontrolled boron dilution of the filled portions ! of the Reactor Coolant System. This limitation is consistent with the initial conditions assumed in the accident analyses for MODE 6. 3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the 1
--reactivity condition of the core.
r 1 3/4.9.3 DECAY TIME 1 The minimum requirement for reactor subcriticality prior to movement of i irradiated fuel assemblies in the reactor vessel ensures that sufficient time l has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment penetration closure limit leakage of radio-active material within containment to the environment to ensure compliance with 10 CFR 100 limits. The requirements on operation of the SLCRS ensure that trace amounts of radioactive material within containment will be filtered through HEPA l filters charcoal absorbers prior to discharge to the atmosphere. These require- 1 ments are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. 3/4.9.5 COMMUNICATIONS I The requirements for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. O Beaver Valley Unit 2 8 3/4 9-1
3/4.9 REFUELING OPERATIONS I flASES._ l 3/4.9.6 MANIPULATOR CRANE OPERABILITY l l The OPERABILITY requirements for the manipulator cranes ensure that:
- 1) manipulator cranes will be used for movement of control rods and fuel assem- 1 blies; 2) each crane has sufficient load capacity to lift a control rod or fuel assembly; and 3) the core internals and pressure vessel are protected from {
excessive lifting force in the event they are inadvertently engaged during lift-ing operations. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING i The restriction on movement of loads in excess of the normal weight of a fuel assembly over other fuel assemblies ensures that no more than the contents , of one fuel assembly plus an additional 50 rods in the struck fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumed in the accident analyses. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that 1) sufficient cooling capacity is available to remove 4 decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING MODE, and 2) sufficient coolant circulation is maintained througnout the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange encures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core. 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM j The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The integrity of the containment penetrations of this system is required to meet 10 CFR 100 requirements in the event of a fuel handling accident inside containment. Applicability in MODE 5, although not an NRC safety requirement, will provide additional protection against small releases of radioactive material from the containment during maintenance activities. . i l Beaver Valley - Unit 2 B 3/4 9-2 i
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t [ EASil_. , _ ,.,,._._, _ n t. a 7aw P , m q 3/4.9.10 and 3/4.9.11 WATER LEVFi - REACTOR VESSEL AND STORAGE POOL Ly @ . The'. restrictions on minimum water level erisure that sufficient water
' )6 f.; *U depth is available.to remcVe 99% of the assumed 10% iodine gap activity-V re. leased;from the rupture of an irradiated.fuell assembly. The minimum water
' depth is. consistent with the asssmptiions of.the accident analysis.
- 3/4.9.12 and 3/4.9.13 ?OEL BUILDI4G VENTILATION SYSTEM I
- The' limitations orcthe storage pool sentilattion system ensure that'all 1-
, radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adscroer prior to discharge to the
,4 atmosphere. The OPERABILITY of'this' system and'the resulting iodine removal ; b cepacity are consistent with'the assumptions o*f the acr.ident analyses. The ; m 4 pent. fuel pool area ventilation system is non-safety related and only recircu- ! [0 hates air through the fuel building. The fuel building portion of the SLCRS .
'is safety related and continuously filters the fuel building exhaust air. This maintains-a negative pressure in the fuel building.
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3]4.10 SPECIAL TEST EXCEPTIONS 3 u
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BASES _ , , __. t 3/4.10.1 SHUT 03WN MARG _IN_ i'
, This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This speciai test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition cceurring as a result,of fuel burnup or fuel cycling operations.
3/4.10.2,,,,_ GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION 8.IMITS ihis special test exception permits individual control rods to be positioned cutside of their normal group heights and insertico limits during the perfomance of such PHYSICS TESTS as those required to 1) meas Jre control rod worth and
- 2) detemine the reactor stability index and damping factor under xenon oscil-lation conditions.
3/4.10.3 PHYSICS TESTS f^ This special test exception permits PHYSICS TESTS to be performed at less ( j than or equal to 5% of RATED THERMAL POWER with the RCS T avg slig m y icwer than normally allowed so that the fundamental nuclear characteristics of the core and related instrumentation can be verified. In order for various charac-terictics to be accurately measured, it is at times necessary to operate out-side the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 and the RCS T avg m y fall slightly below the minimum temperature of Specification 3.1.1.5. 3/4.10.4 REACTOR COOLANT LOOPS This special test exception is required to perform certain startup tests. 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN This special test exception permits the Position Indication System to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normel voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE. p (Jf BEAVER VALLEY - UN]T 2 B 3/4 10-1
3/4.11 ROI0 ACTIVE EFFLUENTS
@MS. ___
33 11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided tc ensure that the con:e'ntration of radioactive materials released in Liquid waste effluents f rom the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Taale II, Columa 2. This limitation provides addi-tional assurance that the levels of radioactive materials ir bodies of water i outside the site will reFult in up;sure within (1) the Section 11. A design objectives of Aopendix 1, 10 CFR Part 50, to an individual and (2) the limits I of 10 CFR Part 20 'IO6(e) to the population. The concentration limit for dis-I solved or entraiv.d noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent cona ntration in water using the methods described in International Comnnssion on Radiological Pr,otection (ICRP) Publication 2. 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set fteth in Section II.A of Appendix !. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioac+ive material in liquid effluents will t'e kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology'provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October,1977, and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of F.ffluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix 1," April 1977. NUREG-0133 provides methods for dose calculations consisterit with Regulatory Guides 1.109 and 1.113. This specification, applies to the release of liquid effluents from Beaver
] Valley Power Station, Unit No. 2. For units with shared radwaste treatment j systems, the liquid effluents from the shared system are proportioned among the units sharing that system.
BEAVER VALLEY - UNIT 2 B 3/4 11-1
1 LIQUID EFFLUENTS 3/4.11.1.3 LIQUID WASTE TREATMENT
- The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive waterials in liquid effluents will be kept "as low as is reasonably , ;hievable." This speci-fication implernents the requirements of 10 CFR Part 50..Ja, General Design r Criterion 60 of Apperdix A to 10 CFR Part 50 and design objective given ir
Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing f the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectivos set forth in Section II./, of Appendix I, 10 CFR Part 50, for liquid effluents. This soeci-fication applies to Beaver VaYley Power Station, Unit No. 2. ! ) ', 3/4.11.1.4 LIRUIDHOLDUPTANfS 1 Restricting the quantit)/ of radioactive material contained in the specified l tanks provides assurarece that the event of an uncontrolled release of the tanks' I contents, the resulting concentrations would be less than the limits of 10 7R Part 20, Appendix B, Table II, Column 2, at the r.earest potable water supply and the nearest surface water supply in an unrestricted area. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at anytime at the site boundary from gaseous effluents from all units on the site will be within the anneal dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B. Table II of 10 CFP. Part 20 (10 CFR Part 20.106(b)). For individ-uals who may at times be within the site boundary, the occupancy of the ind1vid-ual will be sufficiently low to compensate for any increase in the atmospheric diffusion tactor above that for the site boundary. The speci'ied release rate limits restrict, at all times, the corresponding gamma and be:a dose rates above background to an individual at or beyond the site boundary to y 500 mrem / year to the total body or to < 3,000 mrem / year to the skin. These reh.Ne rate I limits also restrict, at all times, the corresponding thyroid dose iete above ' background to a child via the inhalation pathway to 5, 1,500 mrem / year. O GEAVER VALLEY - UNIT 2 0 3/4 11-2
QQUIDf.FFLUENTS MSf1,._. , _, =m -_.=m- _ 3/4.11.2.:._ DOSE RATE (Continued) This specification applies to the release of gaseous effluents from Beaver Valley Power Station, Unit lic. 2. For units with shared radwaste treatment system, the gaseous effluents from the shared systen are proportioned among the units sharir,g that system. 3/4.11.2.2 DOSEzjiOBLEgASES l This specification is provided to implement tne requirements of f Sections II.E, Ill.f. and IV.A of Appendix L, 10 CFR Part 50. The limiting g Conditicn for Operatica implements the guides. set forth in Section II.B of g l Appendix I. The ACTION statements provide the required operating flexibility 6 and at the same time implertent the guides set forth in Section IV.A of Appendix I to assure that the release of radioactive material in gaseous effluents vill be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conforr,ance with the guides of Appendix I be shown by calculational procedures bued on nodels and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the l actual release rates of radioactive noble gases in gaseous effluents are con-sistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October, 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef fluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July,1977. The ODCM equations provided for determining the air doses-at the exclusion area boundary.are based upon the historical average atmospheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111. This specifications applies to the release of gasec s effluents from Beaver Valley Power Station, Unit No. 2. l 3/4.11.2.3 DOSE, RADIOI0 DINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER T_HAN NOBLE GASES This specification is provided to implement the requirements of Sec-tions II.C., III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Con-ditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required cperating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kr:pt "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix 1 be PEAVER VAL.EY - UtilT 2 B 3/4 11-3
i 1
. LIQUID EFFLUENTS 1
MEL- -- __ _ m ) i l 3/4.31.2.3 00SE, RAD 101001NES RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND i M61720TI.T6pHER THAN @LE GASES (ContinuedT l shown by calculational procndures based on models and data such that the actual exposure of an individual through appropriate pathways is .unlikely to be sib-stantially underestiltated. The ODCM calculational methods for calculating the doses due to the actual celear e rates of the subject materials are consistent with the methodology provide 6 in Regulatory Guide 1.109, " Calculation of /cnnual Doses to Man from Routir,g Rekases of Reactor Effluents for the Purpose of Evaluating Compliance with 1] CFR Psrt 50, Appendix 1," Revir. ion 1, October, 1977 and Fegulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gasecus Ef fluenti in Rottine Roleeses frora Light-Water-Cooled l Reactors," Revision 1, July,1971, These equatiins also provide for determining the actual doses based upon tb2 historical averaae atmospheric conditions. The release rate specifications, for radioiodines, rt.dioactive material in particu-i late form, and radionuclides other than r.oble gnses s.re dependent on the e>.ist-ing radionuclides pathvays to man, in the urcest*icted area. The pathways which are examined in the development of these calculations are: 1) individua'l in- - halation of airborne radionuclides, 2) depositien of radionuclides onto vege- I tation with subsequent consumption by mar., 3) Jet >osition onto grassy areas where milk animals and meat producing animals graze with consumption of the / milk and meat by man, and 4) deposition on tho ground with subsequent exposure of man. This specification applies to radioactive material in particulate form and radionuclides other than noble gases released from Beaver Valley Power Statson, Unit No. 2. 3/4.11.2.4 GASE0US RADWASTE TREATMENT The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were Specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gueous ef fluents. This specification applies to gaseous radwaste from Beaver Valley Power Station. Unit No. 2 3/4.11.2.5 GASEOUS WASTE STORAGE TANKS l Restricting the quantity of radioactivity contained in any connected group of gaseous waste storage tanks provides assurance that in the event of an un-controlled release of the tanks' contents, the resulting total body expJsure to an individual located at the nearest exclusion area boundary for two hours im-mediately following the onset of the release will not exceed 0.5 rem. The l I BEAVER VALLEY - UNIT 2 B 3/4 11-4 l
9 LIQUID EFFLUENTS MKS _ 3/4.11.2. L GASEOUS WASTE STORAGE TANKS (Continued) specified limit restricting the quantity of radioactivity contained in any con-nected group of gaseous waste storage tanks was specified to ensure that the total body exposure resulting from the postulated release remained a suitable fraction of the reference value setforth in 10 CFR 100.11(a)(1). The curie content limit is applied indi/idually to each gaseous wasta storage tank and collectively to the number of unisolated .pweous waste storage tanh. 3/4.11.2.6 U P10SIVE GAS MIXTURE This specification is provided to er.sure that the concentration of poten-tially explosive gas nixtures containet.' in the waste gas toidup system is inain-tained below the flamtnability limits at tydrogen and exygen. Isolation of the affected tank f or purposes of purginc and/or discharge permits the flammable , gas concentrations of the tank to be reduced below the lower explosive limit in ; a hydrogen ricn system. Maintaining the concentration of hydrogen and oxygen be6 their flanrnability limits providas assurance that the releases of radio-active materials will be controlled in conformance with the requirements of O Ceneral Design Criterion 60 of Appendix A to 10 CFR Part 50. 3/4.11.3 SOLID RADI0 ACTIVE WASTE This specification knplements the requirements of 10 CFR Part 50,36a and General Design Criteria E0 of Aprendix A of 10 CFR /ut 50 and requires the system be used whene'ier solid r Wwastes require processing and packaging prior to being shipped offsi a. The process parameters used in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste cil content, waste principal chemical constituents mixing and curirg times. l 3/4.11.4 TOTAL 003 This specification is provided to meet the dose limitations of 40 CFR 190. l The Specification requires the preparation and submittal of a Special Report, in lieu of any other report, whenever the calculated dotes f rom plant radio-active effluents exceed twice the design cbjective doses of Appendix I. For sites containing up to 4 nuclear reactors, it is higbiy unlikely that the resul-tant dose to MEMBEF,(S) 0F THE PUBLIC will exceed the dose limits of 40 CFR 190 f the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the liraitation of dase to MEMBER (S) 0F lHE OUBLIC for the calendar yest to within the 40 CFR 190 limits. For the purpcses of the Spec;al Report, it may be p) V asc umed that the dose commitment to MEMBER (S) 0F THE PUBLIC fram other uranium feel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site oc within a radius of l 5 miles be considered. { FiAVER VALLEY - UNIT 2 B 3/4 11-5 \ _ _ - - _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ b
(N -3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING V l BASES __ , _ _ 3/4.12.1 MONITORING PROGRAM The radiological monitoring program required by this specification provides measurements of radiation and of r.adioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation expo-sures of MENBER(S) 0F THE PUBLIC resulting from the Station zoperation. This monitoring program thereby supplement.s the radiological effluent monitbring . program by verifying that the measurable concentrations-of radioactive materials ! and levels of radiation .are not higher than expected on the basis of the efflu- ) ent neastreiients and modding of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three ye ars of commercial operation. Following this period, program changes may be initiated based on operational experience. The datection capabil'ities required by Table 4.12-1 are state-of-the-art for routhe environmental measurements in industrial laboratories. The LLD's for drirking water meet the requirements of 40 CFR 141. i 1 3/4.12.2 LAND USE CENSUS 1 Thi's specification is provided to ensure that changes in the use of unre-stricted areas are identified and that modifications to the nnnttoring programs are made if required by the results of this census. The best survey information from the door-to-door survey, aerial survey or by consulting with local agri- j culture authorities sha'>l be used. This census satisfies the requirements of Section IV.B.3 of Appnidix I to 10 CFR Part 50. Restricting the census to- I gardens cf greater than 500 square feet provides assurance that significant j expnsure pathways via leafy vegetables will be identified and monitored sir.ct a j garden of this size is the minimum required to produce the quantity (26 kg/ year) j of leafy veget6bles assumed in Regulatory Guide 1.109 for consumption by a child. ! 10 determine this minimum garden size, the following asst:mptions were used: 1) j i that 20% of the garden was used for growing broad leaf vegetation (i e., similar l l Lo lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter. 3/4.12.3 INTERLABORATORY COMPARIS0N PROGRAM The requirement for participation in an Interlsboratory Comparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitorbg l in order te demonstrate that the results are reasonably valid. l O l BEAVER VALLEY - UNIT 2 B 3/4 12-1
l 0 1 l l l SECTION 5.0 dei 1GN FEATURES l 0 1
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Lb 5.1 SITE SITE BOUNDARY FOR GASEQUS EFFLUENTS. 5.1.1 The site boundary for gaseous effluents shall be as shown in Fig-ure 5.1-1. Release paths are shewn on Figure 5.1-2. SITE BOUNDARY FOR LIO_UID EFFLUENTS 5.1.2 The site boundary for liquid effluents shall be as shown in Fig-ure 5.1-L Release points are shown on Figure 5.1-2. EXCLUSION AREA 5.1.3 The exclusion area shall be as shown in Figure 5.1-3.
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LOW POPULATION ZONE j 5.1. 4 The low population zone shall be as thown in Figure 5.1-4. - FLOOD _. CONTROL 5.1. !. The flood cor, trol provisions (dikes, levees, etc.) shall be desit;ned ! Q and mainti.ined in a':cordance with the origiral design provisions contained in ; Q 5ec':. ion 3.4.1 of tne FSAR. ' 5 '.2 CONTAINMENT C0!! FIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete builcing of cylindrical shape, with a dome roof and having the following design features:
- a. flominal inside diarxeter = 126 feet
- b. Nominal inside hei[,ht = 185 feet.
- c. Minimum thickness of concrete walls = 4.5 feet,
- d. Minimum thickness of concrete roof = 2.5 feet,
- e. Minimum thickness of foundation mat = 10 feet,
- f. Nominal +~a ickness of vertical portion of steel liner = 3/8 inch.
- g. Ncminal thickness of steel liner, dame portion = 1/2 inch.
b h. Minimum free volume = 1.73 x 106 cubic feet. BEAVER VALLEY - UNIT 2 5-1
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MS_}RLEEAElRES _ , , DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for maximum internal pressure of 45 psig and a temperature of 280.0 F. PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained in Section G.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requireinents. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactcr core shall contain 157 fuel assemblies with each fuel assem-bly containing 264 fuel rods clad with zircaloy-4. Each fuel roa shall have e nominal active fuel length of 144 inches. Reload fuel shall be similar in physical design to the initial core ioading and shall have a maximum enrichment
- of 3.3 weight percent U-235.
CONTROL R0D ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of ebsorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. l 5.4 REACT 0.R COOLANT SYSTEM l ) DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Cociant System is designed and shall be faaintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650 F, except for the pressurizer which is 680 F.
VOLUME 5.4.2 The total water and steam volume ,of the Reactor Coolant System is 9370 cubic feet at a nominal 1 of '576 F. 1 BEAVER VALLEY - UNIT 2 5-6
e
, DESIGN FEATURES \
5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of , the FSAR with allowance for normal degradation pursuant tc the applicable ' Surveillance Requirements. 5.6 FUEL STORAGE ! CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum of 10.4375 inch center-to-center distance between fuel assemblies placed in the storage racks to ensure a k equivalent to ,<0.95 with the storage eff pool filled with unborated water. The k eff f $0.95 includes a conservative allowance of at least 1.4% Ak/k for uncertainties. DRAINAGE 5.6.2 The spent fuel storage pcal is. designed and shall be maintained to prevent inadvertent draining of the pool below elevation 751'-3". ( _ CAPACITY i 5.6.3 The fuel storage pool is designed and shall be maintained with- a storage capacity limited to no more than 1088 fuel assemblies. 5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I items in Section 3.7 of the FSAR shall be designed and maintained to the original de-sign provisions with allowance for normal degradation pursuant to the applicant Surveillance Requirements. 5.8 METEOROLOGICAL TOWER LOCATION
- 5. 8.1 The meteorological tower shall be located as shown on Figure 5.1-1.
BEAVER VALLEY - UNIT 2 5-7
Y I l r l l i SECTION 6,0 ADMINISTRATIVE CONTROLS I l
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1 L
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'6. 0 ADMINISTRATIVE CONTROLS ,
1 l 6.1' RESPONSIBILITY ! l 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence. l4 6.2 ORGANIZATION i ! 0FFSITE- 1 6.2.1 The corporate organization for facility management and technical support shall be as shown on Figure 6.2-1, 1 FACILITY STAFF 1 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:
- a. Each duty shift shall be composed of at least the mininum shift crew composition shown in Table 6.2-1.
)
- b. At least one licensed Operator shall be in the control room when fuel {
is in the reactor.
- c. At least two licensed Operators shall be in the control room during reactor start-up, scheduled reactor shutdown and during recovery from J reactor trips.
- d. An individual qualified in radiation protection procedures shall be onsite when fuel is in thb reactor,
- e. All CORE ALTERATIONS after the initial fuel loading shall be directly !
supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has ,no other concurrent responsibilities during this operation.
- f. Administrative procedures shall be ceveloped and implemented to limit the working hours of unit staf f wha perform safety-related functions; senior reactor operators, reactor operators, radiation control tech-nicians, auxiliary operators, meter and control repairman, and all personnel actually perforraing work on safety related eauipment.
The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is operating. However, in ; the event that anforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:
. fR BEAVER VALLEY - UNIT 2 6-1
i ADMINISTRAI1YEl0RIA011 _ - __ _ _
}
FACILITt STAFF (Continued)
- a. An individual should not be permitted to work more than 16 hours straight, o cluding shift turnover time.
- b. An individual should not be permitted to work more than 16. hours in any 24-hour period, nor more than 24 hours in eny 48-hour period, nor more than 72 hours in any seven day period, all excluding shift turnover time.
- c. A break of at least eight hours should be allowed between work periods, including shif t turncver time.
- d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines chall be authorized by the Plant Manager or predesignated citernate, or higher levels of manage-ment. Authorized deviations to the workirig hour guidelines shall be documented and available for NRC review. 6.2.3 INDEPENDENT SAFETY EVALUATION GROUP (ISEG) FUNCTION 6.2.3.1 The ISf3 shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources , ! of unit design and operating experience information, including units of simi-lar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, ihaintenance activities, operations activities, or other ineans of improving unit I safety to corporate management. If not otherwise implemented, all recommenda-tions shall then be made to the Vice President, Nuclear Group. COMPOSITION 6.2.3.2 The ISEG shall be cnmposed of at least five, dedicated, full-tise engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuciear fieid. RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surr N of unit activities to provide independent verification
- that these . ., X are performed correctly and that human errors are reduced as mue . rm ical.
l RECORDS 6.2.3.4 Records of activities performed by the ISEG shi .w p; rec ." tained, and and forwarded each calendar month to the Vi a m, . Group. , , , t 1 l
*Not responsible for sign-off function.
BEAVER VALLEY - UNIT 2 6-2
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TABLE 6.2.1 MINIMUM SHIFT CREW COMPOSITION # SINGLE UNIT FACILITY LICENSE CATEGORY APPLICABLE MODES QUALIFICATIONS 1, 2, 3 and 4 S and 6 SRO* 2 1** i R0 2 1 i l Non-Licensed Auxiliary Operator 2 1 , I Shift Technical Advisor 1(a) None Required { l
*Incluces the Licensed Senior Reactor Operator serving as the j Shift Supervisor.
**Does not include the Licensed Senior Reactor Operator or Senior I Reactor Operator Limited to Fuel Handling, supervising CORE OPERATIONS.
7,3 #Shfft crew composition may be one less than the minimum l
) requirements for a period of time not to exceed 2 hours in ss order to accommodate unexpected absenco of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. (#)The Shift Technical Advisor (STA) position may be filled by the same STA from the BVP5 Unit 1, if the individual is qualified for BVPS Unit 2. i i l l l i i 1 () l Q )' l l l
- BE HER kALLEY - UNIT 2 6-5
ADEM11NJIYLCORIRQLS . 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility and Radiation Protection staff shall meet or exceed the minimam qualifications of ANSI N18.1-1971 for comparable positions, i except for the Radiological Control Manager who shall meet or exceed the quali-fications of Regulatory Guide 1.8, September 1975, and the technical advisory engineering representative who shall have a bachelor's degree or equivalent in i a scientific or engineering discipline with specific training in plant design ! and response analysis of the plant for transients and accidents. 6.4 TRAINING
- 6. 4. '1 A retraining and replacement training program for the facility staff j shall be maintained under the direction of the Nuclear Training Manager and i shall neet or exceed the " requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and App adix "A" of 10 CFR Part 55, 6.5 REVIEW AND AUDIT t
6.5.1 ONSITE SAFETY COMMITTEE (OSC) 1 FUNCTION 6.5.1.] The OSC shall function to advise the Plant Manager on all matters re-lated to nuclear safety and shall provide review capability in the areas of:
- a. nuclear power plant cperations
- b. radiological safety
- c. maintenance
- d. nuclear engineering e, nuclear nower plant testing
- f. technical advisory engineering g, chemistry
- h. quality control
- i. instrumentation and control COMPOSITION 6.5.1.2 The plant Safety Review Director is the OSC Chairman and shall appoint '
all members of the OSC. The membership shall consist of a minimum of one individual from each of the ares designated in 6.5.1.1. OSC members and alternates shall meet or exceed the minic.dm qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions. The nuclear power plant , operations indiH dual shall meet the qualifications of Section 4.2.2 and the maintenance individual shall meet the qualifications of Section 4.2.3. i O' BLAVER VALLEY - UNIT 2 6-6
jMINISTRATIVECONTROLS , . . ,, , . , . , _
~)
s COMPOSITION (Continued) ALTERNATE _S 6.5.1.3 All alternate members shall be appointed in writing by the OSC Chairman to serve on a temporary basis; however, no'more than two alternates shall participate as voting members in OSC activities at any one tip MEETING FREQUENCY' 6.5.1.4 The OSC shall meet at least once per calendar month and as convenod by the OSC Chairman or his designated alternate. QUORUM 6.5.1.5 A quorum of the OSC shall consist of the Chairman or his design m d alternate and at least one half of the members including alternates.
' RESPONSIBILITIES 6.5.1.6 The OSC shall be responsible for:
- a. Review of 1) all procedures required by Specification 6.8 and changes of intent thereto, 2) any other proposed procedures or changes
[. thereto as determined by the Plant Manager to affect nuclear safety. C
- b. Review of all proposed tests and experiments that affect nuclear j safety.
- c. Review of all proposed changes to the Technical Specification:. !
l
- d. Review of all proposed changes or modifications to plant systems .r equipment that affect nuclear safety.
- e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence to the Senior Manager !
Nuclear.0perations and to the Chairman of the Offsite Review Committee.
- f. Review of all REPORTABLE EVENTS.
- g. Review of facility operations to detect potential safety hazards. j
- h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Offsite Review Committee. .
i 1 1 O : BEAVER VALLEY - UNIT 2 6-7 l i
~
Y , i i {, I [, -hidWDMEdi910M .-- ,, s. g AUTHOPIly W C. L "/ fin AC ( N 11:
- e. . Reco w nu )c the Plant Manager written approval or disapproval of i t em ice s A red und u 6.5.1.6(a) through (d) above.
I b R w e, ds ..nitrations in wri'.ing with regard to whether or not each ite.,em sideri>d under 6.5.1.4(a) through (e) above constitutes an orn "ewa sefety question. I e ProviG aitten notif; cation within 24 hours to the Senior Manager , [ Poelea: su rativ s and the Offsite Review Committee of disagreement i tu.t . Uw OSC ?.na the Plant Manager; however, the Plant Manager ) j 1 i,i mve espersibility for resolution of such disagreements t,u: msut in 12.1.1, above. l } .R_EC..O @._i i ! 6. E 1. 8 in+ CSC shall maintain written minutes of each meeting and copies shall ! be provided to the Senior Manager Nuclear Operations and Chairman of the Offsite
- i. Review Comii tee. ,
l > b . 5. 2 0FFMjE REVIEW COMMITTEE (0RC) j J TUNCTION ! 6.5.2.1 The ORC shall function to provide independent review and audit of designated activities in the areas of:
- a. nuclear power plant operations l
l b. nuclear engineering , c. chemistry and radiochemistry
- d. metallurgy l e. instrumentation and control
- f. radiological safety l
- g. mechanical and electrital engineering
- h. quality assurar,ce practices O
BEAVER VALLEY - UNIT 2 6-8
ADM11&&lIYLC0HIROLS , COMPOSITION 6.5.2.2 The chairman and all members of the ORC shall be appointed by the Senior Vice President, Nuclear Group. The membership shall consist of a minimum of five individuals who collectively possess a broad based level of experience and competence enabling the committee to review and audit those activities designated in 6.5.2.1 above and to recognize when it is necessary to obtain technical advice and counsel. An individual may possess expertise in more than one speciality area. The collective competence of the committee will be maintained as changes to the membership are made. ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall partic:pate as voting members in ORC activities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the ORC Chairman to provide expert advice to the ORC. MEETING FREQUENCY 6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter. QUORUM 6.5.2.6 A quorum of ORC shall consist of the Chairien or his designated alter-nate and at least four members including alternater. No more than a minority of the quorum shall have line responsibility for operatier. of the facility. REVIEW 6.5.2.7 The ORC shall review;
- a. The safety evaluations for 1) changes to procedures, equipment, or systems and 2) tests or experiments completed under the provision of i Section 50.59, 10 CFR, to verify that such actions did not consitute an unreviewed safety question.
l BEAVER VALLEY - UNIT 2 6-9
I ADMINISTgUyLf0HIROLS _ REVIEW (Continued)
- b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety quet, tion as defined in Section 50.59, 10 CFR.
- c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- d. Proposed changes in Technical Specifications or licenses.
- e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
- f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
- g. All Rt.PORIABLE EVENTS.
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components.
- i. Reports and meeting minutes of the OSC.
- j. The results of the Radiological Monituring Program prior to submittal l of the annual report provided in accordance with Specification 6.9.1.10.
AUDITS , l 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the ORC. These audits shall encompass:
- a. The conformance of facility operations to provisions contained within i the Technical Specifications and applicable license conditions at least once per 12 months.
- b. The performance, training, and qualifications of the entire facility staff at least once per 12 months,
- c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operation that affect nuclear safety at least once per 6 months.
- d. The performance activities required by the Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.
- e. The Facility Emergency Plan and irtplementing procedures at least once per 12 months.
BEAVER VALLEY - UNJi 2 6-10
ADMillIMILYLCONTROLS AUDITS (Continued)
- f. The Facility Security Plan and implementing procedures at least once per 12 months.
- g. Any other area of f acility operation considered appropriate by the ORC or the Vice President, Nuclear.
- h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
- i. An independeai fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified off-site licensee personnel or an outside fire protection firm.
- j. An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.
AUTHORITY 6.5.2.9 The ORC shall report to and advise the Senior Vice President, Nuclear Group on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8. RECORDS 6.5.2.10 Records of ORC activities shall be prepared, approved, and distri-buted as indicated by the following:
- a. Minutes of each ORC meeting shall be prepared for and approved by the ORC Chairman nr Vice Chairman within 14 days following each meeting.
- b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be documented in the ORC meeting minutes. I
- c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Senior Vice President, Nuclear Group and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
- a. The Commission shall be notified in accordance with 10 CFR 50.72 and/or a report be submitted pursuant to the requirements of Section 50.73 te 10 CFR Part 50, and
- b. Each REPORTABLE EVENT shall be reviewed by the OSC, and the results of this review shall be submitted to the ORC.
BEnVEP VALLEY - UNIT 2 6-11 l L_.____ _ _ ____ _ _. - - - - - - - - - - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - --
tD!iU!IET9JlVE C@IR0t5 _ i-6.7 SAFETY LIMIT VIOLATION 6.7.1 1he following actions shall be taken in tne event a Safety Limit is violated:
- a. The facility shall be placed in at least HOT STANDBY within one (1) hour,
- b. The Safety Limit violation shall be reported to the Commission within one hour. The Safety Limit violation shall be reported to the Senior Manager Nuclear Operttions and to the ORC within 24 hours.
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Dn-Site Safety Committee (OSC). This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, ano (3) correc-tive action taken to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the Commission, the ORC and the Senior Manager Nucisar Operations within 30 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.
l b. Refueling opera.tions.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
t l e. Emergency Plan implementation. ' \
- f. Fic' Protection Program implementation.
- g. PROCESS CONTROL PROGRAM implementation.
- h. OFFSIlE DOSE CALCULATION MANUAL implementation.
6.8.2 Each )rocedure and administrative policy of 6.8.1 above and changes thereto, shall Le reviewed by the OSC and approved by the Plant Manager, , i predesignated alternate or a predesignated Manager to whom the Plant Manager has assigned in writing the responsibility for review and approval of specific subjects considered by the committee, as applicable. Changes to procedures and administrative policies of 6.8.1 above that do not receive OSC review, such as correcting typographical errors, reformatting procedures and other changes not affecting the purpose for which the procedure is performed shall receive an , independent review by a qualified .ndividual and spproved by a cesignated ! manager or director. i BEAVER VALLEY - UNIT 2 6-12 i
) { .ADMlt@JBATIVECONIROLS _ PROCEDURE (Continued) 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided: I
- a. The intent of the original procedure is not alterede
- b. The change is approved by two (2) members of the plant management staff, at least one (1) of whom holds a Senior Reactor Operator's License on the unit affected.
c, lhe change is documented, reviewed by the OSC and approved by the Plant Manager within 14 days of implementation. 6.8.4 A Post-Accident monitoring program shall be established, implemented, and maintained. The program will provide the capability to obtain.and analyze reactor coolant, radioactive iodines and particulate in plant gaseous effluents, and-containment atmosphere samples following an accident. The pro-gram shall include the following: (i) Training of personnel, (ii) Procedures for sampling and analysis, and (iii) Provisions for maintenance of sampling and analysis equipment. m 6.8.5 A program for monitoring of secondary water chemistry to inhibit steam
/ generator tube degradation shall be implemented. This orogram shall be described in the station chemistry manual and shall include:
- a. Identification of a sampling schedule for the critical parameters and control points for these parameters;
- b. Identification of the procedures used to measure the values of the critical parameters;
- c. Identification for process sampling points;
- d. Procedures for the recording and management of data;
- e. Procedures defining corrective actions for off control point themistry conditions; and
- f. A procedure identifying:
- 1) the authority responsible for the interpretation of the data, and
- 2) the sequence and timing of administrative events "equired to l initiate corrective action.
6.9 REPORTING REQUIREMENTS l ROUTINE LPORTS ( 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Re0ulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted. BEAVER VALLEY - UNIT 2 6-13
ADMINISTRATIVE CQNTROLS STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power escalation testing will be submitted following (1) receipt of an operating license, (2) amendment to i the license involving a planned increase in power level, (3) installation of l fuel that has a different design or had been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details requested in license conditions based on other commitments shall be included in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following com-pletion of the startup test program, (2) 90 days following resumption or com-mencement of commercial power operations, er (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e. , initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed. ANNUAL REPORfS1 6.9.1.4 f.nnual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality. 6.9.1.5 Reports required on an annual basis shall include:
- a. A tabulation of the number of station, utility, and othe personnel (includ-ing contractors) receiving exposure greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions2 (e,g,,
reactor operations and surveillance, inservice inspection, routine mainten-ance, special maintenance (describe maintenance), waste processing, and refueling). The dose assignments to various duty functions may be esti-mated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions. 1A single submittal may be made for a multiple unit site. The s :bmittal should I combine those sections that are common to all units at the site. l 2This tabulation supplements the requirements of Section 20.407 of 10 CFR Part 20. BEAVER VALLEY - UNIT 2 6-14
- _ _ _ - _ _ - )
t [D i ADMINISTRATIVE CONTROLS
\s ANNUAL REPOR15 (Continued)
- b. Documentation of all challenges to the pressurizer power operated relief valves (PORVS) or pressurizer safety valves.
- c. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was .eeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radiciodine con-centrations; (3) Clean-up system flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentrat'on in microcuries per gram as a function of time for the duration of the specific activity above the steady state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
MONTHLY OPERATING REPORT
,3 6.9.1.6 Routine reports of operating statistics and shutdown experience shall /
V) be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC no later than the 15th of each month following the calender month covered by the teport. 6.9.1.7 This item intentionally blank 6.9.1.8 This item intentionally blank 6.9.1.9 This item intentionally blank
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ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT 3 1 6.9.1.10 Routine radiological environmental operating reports covering the j operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year and will include reporting any deviations not reported under 6.9.2 with respect to the Radiological Effluent Technical Specifications. 6.9.1.11 The annual radiological environmental reports shall include sumn. aries, 3 interpretations, and statistical evaluation of the recults of the radiological l environmental surveillance activities for the report period, including a com- j parison with preoperational studies, operational controls (as appropriate), and ' previous environmental surveillance reports, and an assessment of the observed { impacts of the plant operation on tne environment. The reports shall also l include the results of the land use censuses required by Specification 3.12.2. ' A If harmful effects or evidence of irreversible damage are detected by the 3A single submittal may be made for a multiple unit site. The submittal should ! combine those sections that are common to both units l r BEAVER VALLEY - UNIT 2 6-15 1 U_ _-__ _ _ _ _ _ _ _ -
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ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT (Continued) monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.
'The annual radiological environmental operating reports shall include l summarized and tabulated results in the format of Table 6.9-1 of all radio- l logical environmental samples taken during the report period. In the event i that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. <
The missing data shall be submitted as soon as possible in a supplementary j report. The reports shall also include the following: A summary description of the radiological environmental monitoring program; a map of all sampling loca-tions keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program required by Specification 3.12.3. i SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 4 6.9.1.12 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. 6.9.1.13 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluent- and solid waste released from the unit as outlined in Regulatory Guide 3.d, Revision 1, June 1974, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly f basis following the format of Appendix B thereof. In addition the radioactive effluent release report to be submitted 60 days after January 1 of each year shall include an annual summary of hourly meteoro-logical data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radion tive liquid and gaseous effluents released from the unit or station during the previous calendar year. Tnis report shall also include an assessment of the radiation doses from radioactive effluents to MEMBER (S) 0F THE PUBLIC due to their activities inside the site boundary I (Figure 5.1-1 and 5.1-2) during the report period. All assumptions used in l making these assessments (e.g., specific activity, exposure time and location) ' shall be included in these reports. The assessment of radiation doses shall be performed in accordance with 0FFSITE DOSE CALCULATION MANUAL (0DCM).
*A single submittal may be made for a multiple unit site. The submittal should
( combine those sections that are common to all units at the site; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. SEAVEP VALLEY - UNIT 2 6-17
ADMINISTRAllVLCOMIRQLs SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) The radioactive effluent release report to be submitted 60 days after January 1 J of each year shall also include an assessment of radiation doses to the likely 1 most exposed real individual from reactor releases for the previous calendar 1 year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1. The SKYSHINE Code (available from Radiation Shielding Information Center, (ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16. The radioactive effluent release reports shall include an assessment of radiation doset from the radioactive liquid and geseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21. In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shail be evaluated. The assessment of radiation doses shall be per-formed in accordance with ODCM. The radioactive effluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month periou. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The F xy limit for Rated Thermal Power (FRTP) x shall be provided to the Regional Administrator of the Regional Office of the NRC, with a copy to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, for all core planes containing bank "D" control rods and all unrodded core planes at least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission. P Any information needed to suport F will be by request from the NRC and l need not be included in this report. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report. These reports shall be submitted covering the activities identified below pur-suant to the requirements of the applicable reference specification:
- a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
- b. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
- c. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
BEAVER VALLEY - UNIT 2 6-18 j l t
ADMINISI RTIVE CONTROLS-(O) v
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- d. Seismic event analysis, Specification 4.3.3.3.2.
- e. Sealed source leakage in excess of limits, Specification 4.7.9.1.3.
- f. Miscellaneous reporting requirements specified in the Action State-ments for Radiological Effluent Technical Specifications.
- g. Containment Inspection Report, Specification 4.6.1.6.2.
- h. Steam generator tube inservice inspection, Specification 4.4.5.5.
- i. Inoperable accident monitoring, Specification 3.3.3.8.
- j. Operation of PORVs or RCS vents, Specification 3.4.9.3.
6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five (5) years;
- a. Records and logs of facility operation covering time interval at each power level.
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
- c. ALL REPORTABLE EVENTS.
- d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
- e. Records of reactor tests and experiments,
- f. Records of changes made to Operating Procedures,
- g. Rec +,rds of radioactive shipments,
- h. Records of sealed source leak tests and results.
- i. Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License: I
- a. Records and drawing changes reflecting facility design modifications
, made to systems and equipment described in the Final Safety Analysis
/ Report.
BEAVER VALLEY - UNIT 2 6-19
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' AD!ilHLSIMTIYLCGiTROL5 ,
RECORD RETENTION (Continued)
- b. Records of new irradiated fuel inventory, fuel transfers and assembly burnup histories.
l
- c. Records of facility radiation and contamination surveys. l
- d. Records of radiation exposure for all individuals entering radiation control areas.
- e. Records of gaseous and liquid radioactive material released to the environs,
- f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.
- g. Records of training and qualification for current members of the plant staff.
- h. Records of in-service inspections performed pursuant to these Technical Specifications.
- i. Records of Quality Assurance activities required by the QA Manual.
- j. Records of reviews performed for changes made to procedures or equip-ment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the OSC and the ORC.
- 1. Records of the service lives of all hydraulic and mechanical snubbers I including the date at which the service life commences and associated installation and maintenance records.
- m. Records of analyses required by the Radiological Environmental Monitoring Program.
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the j I intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological O BEAVER VALLEY - UNIT 2 6-20 j
m- m f 6 ADMINISTRATIVE CONTROLS ,, ,,_ f
'C J HIGH RADIATION AREA (Continued)
Work Permit
- or Radiological Access Control Permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by.one or more of the following: !
- a. A radiation monitaring device which continuously indicates the l radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radia-
- ion dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made af ter the dose rate level it, the area has been established and personnel have been made knowledgeable of them.
- c. An individual qualified in radiatk a protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible fer providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specifiec by a facility health physics supervisor in the Radiological Work Permit or Radiological Access Control Permit.
, 6.12.2 The requirements of 6.12.1, above, also apply to each high radiation i area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or a facility health physics supervisor.
6.13 This item intentionally blank i I i
- Health physics personnel, or personnel escorted by health physics personnel
( in'accordance with approved emergency procedures, shall be exempt from the RWP j ( issuance requirement during the performance of their radiation protection ; duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. BEAVER VALLEY - UNIT 2 6-21
eD111RISIRAI1YLCONTR0LS 6.14 PROCESS CONTROL PROGRAM (PCP) FUNCTION 6.14.1 The PCP shall be those manuals, procedures, or references to procedures containing the processing steps, a set of established process parameters and the steps detailing the program of sampling, analysis, and evaluation within which solidification of radioactive wastes is assured, consistent with Specifica-tion 3.11.3.1 and the surveillance requirements of these TecEnical Spe:ifications. 6.14.2 Licensee initiated changes
- 1. Shall become effective upon review and acceptance by the OSC.
,6.15 0FFSITE DOSE CALCULATION MANUAL (0DCM)
FUNCTION 6.15.1 The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm /+do setpoints consistent with the applicable LC0's contained in these Techni . Specifications. Methodologies and calculational procedures acceptable to the Commission are contained in NUREG-0133. 6.15.2 Licensee initiated changes:
- 1. Shall become effective upon review and acceptance by the OSC.
6.16 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid) FUNCTION 6.16.1 The radioactive waste treatment systems (liquid, gaseous and solid are those systems described in the facility Final Safety Analysis Report or Hazards Summary Report, and amendments thereto, which are used to maintain that control over radioactive materials in gaseous and liquid effluents and in solid waste packaged for offsite shipment required to meet the LC0's set forth in Specifica-tions 3.11.1.1, 3.11.1.2, 3.11.1.3, 3.11.1.4, 3.11.2.1,. 3.11.2.2, 3.11.2.3, 3.11.2.4, 3.11.2.5, 3.11.2.6, 3.11.3.1 and 3.11.4.1. 6.16.2 MAJOR CHANGES as defined in Section 1 to the radioactive waste systems (liquid, gaseous and solid) shall be made by the following method: A. Licensee initiated changes: I
- 1. If a permanent f acility change is made to a radioactive treatment system that could result in an increase in the volume or activity discharged, the Congnission shall be informed by the inclusion of a 6-22 I BEAVER VALLEY - UNIT 2
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I t ' [N gMINISTRATIVECONTROLS FUNCTION (Continued) suitable discussion of each change in the Annual 10 CFR 50.59 Report for the period in wnich the changer, were made. The discussion of each' change shall contain:
- a. A summary of.the evaluation that led to the determination that the change could be made (in accordance with 10 CFR 50.59);
- b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- c. A detailed description of the equipment, components and procestec 1 involved and the interfaces with other plant systems;
- d. An evaluation of the change will be subinitted which shows the predicted increase of releases of radioactive materials in liquid or gaseous effluents and/or quantity of solid waste from those previously predicted in the license application and amendments thereto;
- e. An evaluation of the change which shows the expected increase C1 in the maximum exposures to an individual in the unrestricted
() area from those previously predicted in the license application and amendments thereto;
- f. A comparison of the predicted increase of releases of radio-active materials in liquid and gaseous effluents and in solid waste to the actual releases for the period the changes were made;
- g. An estimate of the exposure to plant operating personnel as a result of the change; and
- h. Documentation of the fact that the change was reviewed and found acceptable by the OSC.
- 2. The change shall become effective upon review and acceptance by the OSC.
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1 6.16.3 Background of what constitutes MAJOR CHANGES to radioactive waste 1 systems (liquid, gaseous, and solid). ) A. Background
- 1. 10 CFR Dart 50, Section 50.34a(a) requires that each application to I l
construct a nuclear power reactor provide a description of the equip-ment installed to maintain control over radioactive material in gaseous i ( g and liquid effluents produced during normal reactor operations includ-ing operational occurrences. BEAVER VALLEY - UNIT 2 6-23 J l
l ADM1HISIFATJyE CONTROLS ,,, I FUNCTION (Continuea)
- 2. 10 CFR Part 50, Section 50.34a(b)(2) yequires that each application to construct a nuclear power reactor provide an estimate of the quan- !
t hy of radionuclides expected to be released annually to unrestricted areas in liquid and gaseous effluents produced during normal reactor 3
- operation.
- 3. 10 CFR Part 50, Section 50.34a(3) requires that each application to construct a nuclear power reactor provide a descri~ p tion of the provisions for packaging, storage and shipment offsite of solid waste containing radioactive materials resulting from treatment of gaseous and liquid effluents and from other -sources.
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- 4. 10 CFR Part 50, Section 50.34a(3)(c) requires that each application to operate a nuclear power reactor shall include (1) a description of the equipment and procedures for the control of gaseous and liquid effluents and for the maintenance and use of equipment installed $n radioactive waste systems and (2) a revised estimate of the information required in (b)(2) if the expected releases and exposures differ -
significantly from the estimate submitted in the application for a construction permit. ;
- 5. The Regulatory staff's Safety Evaluation Report and amendments thereto issued prior to the issuance of an operating license contains a description of the radioactive waste systems installed in the nuclear power reactor and a detailed evaluation (including est.imated releases of radioactive materials ir, liquid and gaseous waste and quantities of solid waste produced from normal operation, estimated annual maxi-mum exposures to an individual in the unrestricted area and estimated exposures to the general population) which shows the capability of these systems to meet the appropriate regulations. .
6.17 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Senior Manager Nuclear Operations delegates the responsibility for the Radiological Environmental Monitoring Program to the Radiological Control Manager (Figure 6.2-1) or his designated alternate. The Radiological Centrol Manager is responsible for administering the offsite Radiological Environmental Monitoring Program. He shall determine that the sampling program is being implemented as described to verify that the environ-ment is adequately protected under existing procedures. He shall also have the responsibility for establishing, implementing, maintaining and approving offsite environmental program sampling, analyses and calibration procedures. O l l 4 l BEAVER VALLEY - bNIT 2 6-24 1
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MONTM YEAR August 1987 7 PtHFORMtNG ONG&NilifION NAME AND MalltNG ADDAt $$ f/tw*,de /@ Opfm/ ~ 3 PR0gf CT/T ASK WONK UNIT NUMBER Division of Reactor Projects I/II Office of Nuclear Reactor Regulation . P, Om oaA,.r NuM en i U.S. Nuclear Regulatory Commission ' Washington, D.C. 20555 10 $ PUN 3uM.NG QHGANIZ ATIGN NAME AND Malt ING ADOME55 ttevar te Corser via Tvet OF REPOR T Same as 7 above p 88600 COVE RED fla$vssre ,erssi 12 $UPPLEMtNT AR r NOTE 5 Docket No. 50-412 SJ AO ST R Ac T (20d words or ress/ p) The Beaver Valley Power Station, Unit 2, Technical Specifications were prepared by t the U.S. Nuclear Regulato r y Correission to set forth the limits, operating conditions, V and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. I 1 I 1e potvMt NT AN A L v5'5 - e a f *wC**D5'DE stR:P 7 0 As
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