Text

Proposed Tech Specs,Incorporating Improvements in Scope & Content Endorsed by NRC in Final Policy Statement on TS Improvements for Nuclear Power Reactors,58FR39132,dtd 930722
	ML20069B568
Person / Time
Site:	Wolf Creek
Issue date:	05/24/1994
From:	; WOLF CREEK NUCLEAR OPERATING CORP.
To:	;
Shared Package
ML20069B566	List: ML20069B563; ML20069B568;
References
	FRN-58FR39132 NUDOCS 9405270291
	Download: ML20069B568 (275)
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Attachment IV to NA 94-0089 Page-1 of X l

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I ATTACHMENT IV -l

PROPOSED TECHNICAL SPECIFICATION CHANGES

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9405270291 940524 PDR-P ADOCK 05000482' .1 PDR l t

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LIMITING CONDITIONS FOR OPERATION-AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY............................................... 3/4 0 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown' Margin ......................................... '3/4 1-1 Moderator Temperature Coefficient........................- 3/4 1-3 FIGURE 3.1-1 BOL MODERATOR TEMPERATURE COEFFICIENT VS. POWER LEVEL ........................... 3/4 1-5 Minimum Temperature for Criticality...................... 3/4 1-6 3/4.1.2 BORATION SYSTEtiS ,

Fl ow P ath - Sh u tdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3f4-14bcld:;

Flow Paths 0perating................................... -314-+ pld' Charging Pump - Shutdown....'............................. -3f4- De l' t '

Charging Pumps - Operating............................... -314- Nic 4 (d>[

Borated Water Source - Shutdown..........................- -3f4-141- D4' I' Borated Water Sources - Operating........................ 3/41-12 Del' 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................- 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00......................................... 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown.................... 3/4_1-;c Deld'I Rod Drop Time............................................ 3/4,1- 10 De ldd Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits............................. 3/4 1-21 ,

WOLF CREEK - UNIT 1 IV Amendment No. 61 .

LIMITING CONDITION 3 FOR OPERATDN AND SURVEILLANCE REOUIREMENTS SECTION E8fd INSTRUMENTATION (Continued) 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................ 3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS................................ 3/4 3-40 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS........................................ 3/4 3-42 Movabl e Incore Detectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -S/4--S De l d Seismic Instrumentation.................................. -9/4-9-44 h W TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3 -45 De h TABL2 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... -3f4 3-46 De Wj Meteorological Instrumentation. . . . . . . . . . . . . . . . . . . . . . . . . . . -3/4 3-47 DcId' TAB! E 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. - 3/4 3 Di I TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION

, SURVEILLANCE REQUIREMENTS........................... -314-349- Di b Uj Remote Shutdown Instrumentation.......................... 3/4 3-50 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-51 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-52 Accident Monitoring Instrumentation...................... 3/4 3-53 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-54 TABLE 4.3-7 ACCIDENT HONITORING INSTRUMENTATION i SURVEILLANCE REQUIREMENTS........................... 3/4 3-55 l Chlorine Detection Systems............................... DELETED I Loose-Part Detection System.............................. -3/4 3 ';7 DdeTed 1

Radioactive Liquid Effluent Monitoring Instrumentation... DELETED i TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING !

INSTRUMENTATION.................................... DELETED WOLF CREEK - UNIT I VI Amendment No. M,42,66

L1MITIN3 CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

TABLE 4.3-8 RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS DELETED Radioactive Gaseous Effluent Monitoring

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Instrumentation DELETED , ,h E xplos i ve Gas Monitori ng Instrumentation . . . . . . . . . . . . . . . . . 3/4 3 LO TABLE 3.3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION............. 3/4 3-50 O lle Tf TABLE 4.3-9 EXPLOSIVE GA5 MONITORING INSTRUMENTATION M Nj SUR V E I L L AN C E RE QUI R EME NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . -314 3- 01 3/4.3.4 TUc.EINE OVERSPEED PROTECTION............................... 3/4 3-C3 bl 6d 3 /4. 4 REACTOR COOLANT SYSTEM 3/4.4.1 RE;; TOR COOLANT LOOPS AND COOLANT CIRCULATION Sta-tur anc Power Operiition...... ........................ 3/4 4-1 r.: S t a ::. .

... . .... . ... .... ...... ............. 3/4 4-2 H: 5 .tc;.- , .......... ................... 3/4 4-3 CLi c S h *. :: .. - - i. o c ; s F i 1 1 e d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4 - 5 Ccic Shutdeer - Loops Not Filled. . ....................... 3/4 4-6 3.> 4. 4. 2 54rETY VALVES D

P. s t e : .e . . . . . .... ............. ................. 4/4-4-+-

Ope-atin;. .. ................................... 3/4 4-8 3'4.4.3 FRE55UR:2Er.. . ....................... .................. 3/4 4-9 3 /4. 4. 4 RELIEC VALVES. . . . . . . . ' . ........................... 3/4 4-10 3/4.4.5 ST E AM G E N E R AT O R 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3f4--de D cl e 7AE E 4.4-1 MINIMJM NUMBER OF STEAM GENERATORS TO BE INSPECTED O'"Id'j D U R I N G I N S E R V I C E I N S P E C T I ON . . . . . . . . . . . . . . . . . . . . . . . . 4 /4-4-1fr :

b TABLE 4.4-2 STE AM GENERATOR TUBE INSPECTION. . . . . . . . . . . . . . . . . . . . . . . Sh+1t 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection 5ystems................................. 3/4 4-18 Operational Leakage....................................... 3/4 4-19 i f

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1 VII Amendment No.15, 42 l WOLF CREEK - UNIT I -

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w LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION EAiE TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES............................................... 3/4 4-21 3/4.4.7 CHEMISTRY.............................................. 3/4 4-22 Od TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS................ 3/4 4-23 DC U TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS......................................... 3/4 4--24 Dep0 5 3/4.4.8 SPECIFIC ACTIVITY...................................... 3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

> 1 #C1/ GRAM DOSE EQUIVALENT I-131.................. 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Re a c t o r Cool ant Syst em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 13.6 EFPY.......................... 3/4 4-30 l FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 13.6 EFPY.......................... 3/4 4-31 l TABLE 4.4-5 DELETED Pressurizer.......................................... 3/4 4--33 b' U "

Overpressure Protection Systems......,............... 3/4 4-34 FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM...................... 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY...................................... 3/4 4-37 UC 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................. 3/4 4-30 DC b 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.S.1 ACCUMULATORS.............................................. 3/4 5-1 WOLF CREEK - UNIT 1 VIII Amendment No. 40,67,71

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

SECTION PAGE I i

3/4.5.2 ECCS SUBSYSTEMS - T avg

> 350*F............................. 3/4 5-3 l 3/4.5.3 ECCS SUBSYSTEMS - T avg

< 350*F............................. 3/4 5-7 :

3/4.5.4 200'F.............................

i ECCS SUBSYSTEMS - T <

3/4 5-9 avg l 3/4.5.5 REFUELING WATER STORAGE TANK............................... 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT 1 Containment Integrity...................................... 3/4 6-1 Containment Leakage........................................ 3/4 0 2 D6klED

Containment Air Locks...................................... 3/4 6-4 Internal Pressure.......................................... 3/4 6-6 l Air Temperature............................................ 3/4 6-7 Containment Vessel Structural Integrity.................... 3/4 0--O DCSU b !

Containment Ventilation System............................ 3/4 6-11 l i

1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS I Containment Spray System................................... 3/4 6-13 i Spray Additive System...................................... 3/4 6-14 )

Containment Cooling System................................. 3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES........................ ...... 3/4 6-16 1

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES............................ 3/4 6-18 l l

3/4.6.4 COMBUSTIBLE GAS CONTROL !

Hydrogen Analyzers......................................... 3/4 G-Ji- DELGEDl Hydrogen Control Systems................................... 3/4 6-32 ,

1 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves.............................................. 3/4 7-1 i 1

WOLF CREEK - UNIT 1 IX ,

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-1 MAXIMUM ALLOWABLE-POWER RANGE'NEU1RON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION............................................. 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P................... .... 3/4 7-3 Auxilia$yFeedwaterSystem.................................. 3/4 7-4 <

Condensate Storage Tank.....................................

3/4 7-6 Specific Activity........................................... 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY I SAMPLE AND ANALYSIS PR0 GRAM............................ 3/4 7-8 Main Steam Li ne Isol ation Va1ves. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9 STw 4 (m (" NAM- Al"cs/4Wh hlief' (/ Sun .

3M 7 'ir M 4in thelws(rv c Q] < m . . . . . . . .

3/4 Mb 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION....... ..... 3/4 710 DCLEM 3/4.7.3 COMPONENT COOLING WATER SYSTEM.............................. 3/4 7-11 1

3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM.............................. 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK.......................................... 3/4 7-13 1

i 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM. . . . . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.7 EMERGENCY EXHAUST SYSTEM.................................... 3/4 7-17 j 3/4.7.8 SNUBBERS.................................................... -3/4 + 19-TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL....................... 3/4 7 24 @ N FIGURE 4.7-1 SAMPLING PLAN 2) FOR SNUBBER FUNCTIONAL TEST. . . . . . . . . . . . 3/4 7-26 DN 3/4.7.9 SEALED SOURCE CONTAMINATION................................ 4/4 7-27 N !

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j WOLF CREEK - UNIT 1 X Amendment No. 44 j s

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LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.10 DELETED TABLE 3.7-3 DELETED ..

3/4.7.11 DELETED 3/4.7.12 AREA TEMPERATURE MONITORING.............................. 3/ 7 0 06 d .1 TABLE 3.7-4 AREA TEMPERATURE MONITORING........................... -3/4 7 30 M W 3/4.8 ELECTRICAL POWER SYSTEMS i 3/4.8.1 A.C. SOURCES 1

Operating................................................ 3/4 8-1 l TABLE 4.8-1 DIESEL GENERATOR TEST SCHE 0VLE........................ 3/4 8 I i

Shutdown................................................. 3/4 8-8 3/4.8.2 0.C. SOURCES Operating................... ............................ 3/4 8-9 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-11 l i

4 Shutdown................................................. 3/4 8-12 3/4.8.3 ONSITE POWER DISTRIBUTION Operating................................................ 3/4 8-13 Shutdown................................................. 3/4 8-15 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES {

l Containment Penetration Conductor Overcurrent Protective 0evices..................................... -3/4 0-15-blN l 1

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WOLF CREEK -' UNIT 1 XI Amendment No. 15. 39, 44

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i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION.................... ..................... 3/4 9-2 3/4.9.3 DECAY TIME............................................... 3/4 9-3 3/4.9.4 CONTAINMENT CUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 C OM4U N I CAT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3/4 0-5 00 M O 3/4.9.6 REFUELING MACHINE........................................ -3/4 3-0 W 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY............... 3/' 3-3 @N 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-9 Len Water Level.......................................... 3/4 9-10 3/4.9.9 CONTAINMENT VENTILATION SYSTEM........................... 3/4 9-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL Fuel Assemblies.......................................... 3/4 9-12 Control Rods............................................ -3/" 3 13 DkM@

3/4.9.11 WATER LEVE L - STO RAGE POOL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-14 3/4.9.12 SP ENT FUE L ASS EMB LY ST0 RAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-15 F;3URE 3.9-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2........................................ 3/4 9-16 3/4.9.13 EMERGENCY EXHAUST SYSTEM................................. 3/4 9-17 WOLF CREEK - UNIT 1 XII

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION- PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... 3/4 10-1 DilE N 3/4.10.2 GROUP HEIGNT, INSERTION, AND POWER DISTRIBUTION LIMITS.... 3/4 10-2 3/4.10.3 PHYSICS TESTS............................................. 3/4 10-3 3/4.10.4 REACTOR CDDLANT L00PS..................................... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN .................... 3/? 10-5 I)lt N 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration DELETED TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM DELETED Dose DELETED Liquid Radwaste Treatment System DELETED Li qu i d Hol dup Tan ks. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 3/4 11- 1 N 3/4.11.2 GASEOUS EFFLUENTS Dose Rate DELETED TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM DELETED Dose-Noble Gases DELETED Dose-Iodine-131 and 133, Tritium and Radioactive Material in Particulate Form DELETED Gaseous Radweste Treatment Systee DELETED y

Expl o s i v e Gas Mi xture. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3/4 11-2 M Ga s S to ra ge Tan ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -9/4 11- 3 DE b WOLF CREEK - UNIT 1 XIII Amendment No. 42 -

9 BASES SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.......................................... C 2/51-2 DElGtc0 3/4.1.3 MOVA8LE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR. . . . . . . B 3/4 2-2 FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX OIFFERENCE VERSUS THERMAL P0WER.................................. B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 0 0/4 : C Ck E 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 WOLF CREEK - UNIT 1 XV

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BASES ,

SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 STEAM GENERATORS.......................................... 0 J/' a CL LEllD 3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. 0 3/4 4 5 - DL R ll [>

3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS......................... B 3/4 4-10 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF SERVICE LIFE (EFFECTIVE FULL POWER YEARS)...... B 3/4 4-11 3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/4 41G Dttell0 3/4.4.11 REACTOR COO LANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . 0 J/44 10 (ycteu 0 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 A C C UMU LATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-1 3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS..................... B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK....... ...................... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 i

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l WOLF CREEK - UNIT 1 XVI Amendment No. 40 l

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BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 0 3/4 7 13-DSflCO 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... B 3/4 7-3 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM........................... B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK....................................... B 3/4 7-3 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................ B 3/4 7-4 3/4.7.7 EMERGENCY EXHAUST SYSTEM................................. B 3/4 7-4 3/4.7.8 S NU B B E RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -B-3/4-t-5 DC l6 K D 3/4.7.9 SEALED SOURCE CONTAMINATION.............................. 0 3/4 7 0 Delfl0 3/4.7.10 DELETED 3/4.7.11 DELETED 3/4.7.12 AREA TEMPERATURE MON!T061NG.............................. 4HS/4 7 7 CLtcT(C) 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION.............................. B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES.................. 0 3/4 S-3 DEk'0D' 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 '- l ' -l 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ B-3/4 9-I 3/4.9.5 COMMUNICATIONS........................................... B 3/4 3-1 Dkl YI'O l

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i WOLF CREEK - UNIT 1 XVII Amendment No. 15

BASES SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING MACHINE........................................ 0 0/4 0 2 D6N 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY. . . . . . . . . . . . . . . -0 /4 0 2 NO 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............ B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION SYSTEM........................... B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L........................................... B 3/4 9-3 3/4.9.12 SPENT FUEL ASSEMBLY ST0 RAGE.............................. B 3/4 9-3 3/4.9.13 EMERGENCY EXHAUST SYSTEM................................. B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................... 0 3/410-1 [fMT@

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 R EACTO R COO LANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... -03/4101(E@l@

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......................................... O ;/4 11-1 % T @

3/4.11.2 GASEOUS EFFLUENTS........................................ 0 3/4 11 1 I)lEIbD 3/4.11.3 DELETED 3/4.11.4 DELETED 3/4.12 RADIOACTIVE ENVIRONMENTAL MONITORING B 3/4 12-1 3/4.12.1 DELETED 3/4.12.2 DELETED l l

3/4.12.3 DELETED l 1

WOLF CREEK - UNIT 1 XVIII Amendment No. 42

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident cenditions are either:

1) Capable of being closed by at OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of

. Specification 3.6.1.3, t k does e A. Thecontainmentleakageratesarewithinthelimitskf Specification 3.6.1 7, and i

J g. The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

[0RE OPERATING LIMITS REPORT 1.10 .The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. !

These cycle-specific core operating limits shall be determined for each reload i cycle in accordance with Specificatic., 6.9.1.9. Plant operation within these l operating limits is addressed in individual Specifications. l l

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WOLF CREEK - UNIT 1 1-2 Amendment No. 61 P

1/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTOOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1g;1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/kfee-APPLICABILITY: MODES 1, 2*, 3, d 5. 1 WM/o 15 O rc5 With the SHUTDOWN MARGIN less than 1.3% Ak/k, i =:di:t: b initiate and continue boration at greater than or equal to 30 gpm of abolution containing greater than or equal to 7000 ppm boron or equiv .* ant until the required SHUTDOWN MARGIN is restored. ,

SURVEILLANCE RE0VIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal ~

to 1.3% Ak/k: ,

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); [

b. When in MODE 1 or MODE 2 with K least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by veriYy# greater ing_that controlthan bankor equal to 1 at withdraw is within the limits of Specification 3.1.3.6;

c. less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or to achen in MODE 2 with K 'cality by verifying that vingreactorcrifi e predicted crit control rod position is within t mits of Specifica 3.1.3.6;

d. Prior to initial oper ve 5% RATED THERMAL POWER after each fuel loading, by cons a of the factors of Specification 4.1.1.1.le belo th the con anks at the' maximum insertion-limit of S cation 3.1.3.6; and

[ Special Test Exception Specification 3.10.1.

WOLF CREEK - UNIT 1 3/4 1-1 Amendment No. 61 a

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' REACTIVITY CONTROL SYSTEMS ,

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-l SURVEllLANCE RE0VIREMENTS (Continued) ] ,

h ,a'. Sh: in "000 0, 4, , , / least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

P

4. . . . overall core reactivity balance shall be compared to values ' to demons ment within i.1% ok/k at least +

Effective Full Power Days (

. com a consider at least those factors stated-in Specifica . ... ve. The predictr.d reactivity values sha s ed (normalized) to co o the actual core condi or to exceeding a fuel burnup of 60 EFPD after .

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WOLF CREEK - UNIT 1 3/4 1-2 Amendment No. 61

REACTIVITY CONTROL SYSTEMS

' 3/4.1.2 BORATION SYSTEMS -

b FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION

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3.1.2.1 As a minimum, one of the following boron injecti flow paths shall be OPERABLE and capable of being powered from an OPERABL emergency power source:

a. A flow path from the Boric Acid Storage 5 tem via a boric acid transfer pump and a centrifugal chargin pump to the Reactor Coolant System if the Boric Acid Storage Syste is OPERABLE as given in Specification 3.1.2.5a. for MODES 5 d 6 or as given in Specification 3.1.2.6a. for MODE 4; or

b. The flow path from the refuelin water storage tank via a centrifugal charging pump to the Reactor C lant System if the refueling water storage tank is OPERABLE as ven in Specification 3.1.2.5b. for MODES 5 and 6 or as given i Specification 3.1.2.6b. for MODE 4.

APPLICABILITY: MODES 4, 5, and 6.

ACTION:

With none of the above flow aths CPERABLE or capable of being powered from an OPERABLE emergency power urce, suspend all operations involving COPE ALTERATIONS or positive activity changes.

SURVEILLANCE REQUIR ENTS 4.1.2.1 At I ast one of the above required flow paths shall be demonstrated OPERABLE at east once per 31 days by verifying that each valve (manual, I power-oper ted, or automatic) in the flow path that is not locked, sealed, r otherwise secured in position, is in its correct position.

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WOLF CREEK - UNIT 1 3/4 1-7

Alternativa REACTIVITY CONTROL SYSTEMS CORE REACTIVITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The measured core reactivity shall be within +1%

Ak/k of predicted values.

APPLICABILITY: Modes 1 and 2 ACTION:

With the measured core reactivity not within limits, within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s:

a. reevaluate core design and safety analysis, and determine that the reactor core is acceptable for continued operation, and

b. establish appropriate administrative operating-restrictions and surveillance requirements, or

c. be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.1.5.1 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i1% ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in' Specification 4.1.1.1.1b. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

4.1.1.5.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k_ prior-to initial-operation above 5% RATED THERMAL POWER after each fuel loading,-by consideration of the factors.of Specification 4.1.1.1.lb, with the control banks at the maximum insertion. limit of Specification 3.1.3.6.

WOLF CREEK - UNIT 1 3/4 1-7

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0 REACTIVITY CONTROL SYSTEMS @

FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow aths shall be OPERABLE:

a. The flow path from the Boric Acid Storage System vi a boric acid transfer pump and a cent-ifugal charging pump to t Reactor Coolant System, and

b. Two flow paths from the refueling water stora tank via centrifugal charging pumps to the Reactor Coolant System APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

With only one of the above required boron inj tion flow paths to the Reactor Coolant System OPERABLE, restore at least tw boron injection flow paths to the Reactor Coolant System to OPERABLE sta s within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTD0 MARGIN equivalent to at least 1.3% Ak/k at 200*F within the next 6 ho s; restore at laast two flow paths to j OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next i 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0VIREMENTS _

4.1.2.2 At least two of the ove required flow pnas shall be ~ demonstrated OPERABLE:

1

a. At least once p r 31 days by verifying that each valve (manual, '

power-operate , or automatic) in the flow path that is not locked, ;

sealed, or o erwise secured in position, is in its correct I position;

b. At leas once per 18 months during shutdown by verifying that each i autom ic valve in the flow path actuates to its correct position on i a Sa ty Injection test signal; and

c. A least once per 18 months by verifying that the flow path required y Specification 3.1.2.2a. delivers at least 30 gpm to the Reactor !

Coolant System. ;

The rovisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry in o MODE 3 for the centrifugal charging pump declared inoperable pursuant to ecification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status withir. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

WOLF CREEK - UNIT 1 3/4 1-8 Amendment No. 61

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REAC'!v:Tv OCNTROL SYSTEMS CHARG!N3 PUMP - SHUT 00WN -

s L:M:':NG CONDITION FOR OPERATION 3.1.2.3 One centrifugal charging pump in the boron injection f w path reovired by Specification 3.1.2.1 shall be OPERABLE and capable of bei powered from an OPERABLE emergency power source.

A00LICASILITY: MODES 4, 5, and 6.

ACTION:

Witn no centrifugal charging pump OPERABLE or cap e of being powered from an OPERABLE emergency power source, suspend all erations involving CORE ALTERATIONS or positive reactivity changes.

SURVE!LLANCE REOUIREMENTS

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4.1.2.3.1 The above require centrifugal charging pump shall be demonstrated OPERABLE by verifying, on circulation flow, that the pump develops a dif-ferential pressure of gr ter than or equal to 2400 psid when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All cent fugal charging pumps, excluding the above required OPERABLE p urc.0 , shall be de nstrated inoperable

at least once per 31 days, except when the reactor vess head is removed, by verifying that the motor circuit breakers are secured in he open position.

An i operable pump may be energized for testing or for filling accumulators pr ided the discharge of the pump has been isolated from the RCS by a closed 4

olation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

WOLF CREEK - UNIT 1 3/4 1-9

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REACTIVITY CONTROL SYSTEMS f(5' 0

CH,4RGING PUMPS - OPERATING b

<c}-

LIMITING CONDITION FOR OPERATION

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3.1.2.4 -At least two centrifugal charging purps shall be OPER LE.

APPLICABILITY: MODES 1,12,- and. 3. "

ACTION:

With only one centrifugal charging pump OPERABLE, rest re at least two cen-trifugal charging pumps to OPERABLE status within 72 ours or be in at least-HOT STANOBY and borated to a SHUTOOWN MARGIN equiva ent to at least Li ak/k' at 200*F within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: restore at; leas two charging pumosLto '

OPERABLE status within the next 7 days or be in T SHUT 00WN within-the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS --

4.1.2.4 At least two centrifu 1 charging pumps shall be demonstrated OPERABLE' by verifying, on recirculatio flow, that the pump develops a differential pressure of greater than or qual to 2400 psid when tested pursuant .to Specification 4.0.5.

d I

The pro isions of Soecifications 3.0.4. and 4.0.4 are not apolicable for; entry ' '

into'. OE 3-for the centrifugal charging pump declared inoperable cursuant to Spec'f f cation 4.1.2.3.2 provided that the centrifugal charging pumo is resterec to PERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temoerature of one or more of-

-t e RCS cold legs exceeding 375'F, whichever comes'first. '

WCLF CREEK - UNIT 1 3/4 1-10

-_ _ ~ __ _ , _ , , _ _ _ _ _ _ .- , ,

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REACTIVITY CONTROL SYSTEMS g ;

BORATED WATER SOURCE - SHUTOOWN -

l LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sour s shall be OPERABLE:

a. A Boric Acid Storage System with: .

1) A minimum contained berated water volume of 2968 gallons,

2) 8etween 7000 and 7700 ppa of boron, a

3) A minimum solution temperature of 'F. -

b. The refueling water storage tank (RW ) with:

1) A minimum contained borated ter volume of 55,416 gallons,

2) A minimum boron concentra on of 2400 ppa, and l ;

. 3) A minimum solution tem rature of 37'F. l APPLICABILITY: MODES 5 and 6.

ACTION:

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With no borated water source PERA8LE, suspend all operations involving C0RE '

ALTERATIONS or positive rea ivity changes.

SURVEILLANCE REQUIREME .

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4.1.2.5 The above equired borated water source.shall be demonstrated'0PERA8LE:.

a. At le t once per 7 days by:'

1) Verifying the boron concentration of thel water.

.{

Verifying the contained borated water volume,' and '

3) Verifying the Boric Acid Storage Systes' solution temperature when it is the source of borated water.

. At least once.per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by-verifying .the RWST temperature when it-is the source of borated water and the outside air temperature is-less than 37'F.

WOLF CREEK - UNIT 1 3/4 1-11 A**" * "t No. 23.

REACTIVITY CONTROL SYSTEMS SARATED WATER SOURCES - OPERATING g(

LIMITING CONDITION FOR OPERATION 9

3.1.2.6 As a minimum, the following borated water sources shall be OP BLE as required by Specification 3.1.2.2 for MODES 1, 2, and 3 and one of he following borated water sources shall be OPERABLE es required by Sp ifica-tion 3.1.2.1 for MODE 4:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of ,658 gallons,

2) Between 7000 and 7700 ppm of boron, and

3) A minimum solution temperature of 65* .

b. The refueling water storage tank (RWST ith:

1) A minimum contained borated wa r volume of 394,000 gallons,

2) Between 2400 and 2500 ppm o boron,

3) A minimum solution temp ature of 37'F, and

4) A maximum solution t perature of 100*F.

APPLICABILITY: MODES 1, 2, 3, d 4.

ACTION: l l

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a. With the Boric id Storage System inoperable and being used as one of the above quired borated water sources in MODE 1, 2 or 3, restore the torage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at leas HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN RGIN equivalent to at least 1.3% ok/k at 200'F; restore the Bo c Acid Storage System to OPERABLE status within the next 7 da or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . I

b. W h the RWST inoperable in MODE 1, 2, or 3, restore the tank to PERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

. With no borated water source OPERABLE in MODE 4, restore one borated water source to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

WOLF CREEK - UNIT 1 3/4 1-12 Amendment No. 23, 61

. . .-. - - - . = - -. - .- -- ..

N REACTIVITY CONTROL SYSTEMS-qt SURVEILLANCE-REQUIREMENTS 4.1.'2.6 -Esch required borated water-source shall e demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentr tion in the water,

2) Verifying the contained rated water volume of the water source, and

3) Verifying the Boric cid Storage System solution temperature when it is the so ce of borated water,

b. At least once per hours by verifying the RWST temperature. when :

the outside air t erature is'either less than 37'F or greater than 100*F.

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WOLF CREEK - UNIT 1 3/4-1 1

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All' full-length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indica'ed position) of their group step counter I demand position.

APPLICABILITY: MODES 1* and 2*. I ACTION: The ACTION to be taken is based on the cause of inoperability of control rods as follows:

l ACTION l l

More Than i CAUSE OF IN0PERABILITY One Rod One Rod j a) Immovable as a result of excessive (1) (1) friction or mechanical interference or known to be untrippable.

b) Misaligned from its group step (3) (2) counter demand height or from any other rod in its group by more than 12 steps (indicated position).

c) Inoperable due to a rod control urgent (4) (4) failure alarm or other electrical problem in the rod co'ntrol system, but trippable, vM I ACTION 1 - Dutuig.ine thet the 0;;UTDOWN M RGIN cecuiccuicnt .f t ccificeticn

, . it - y 1.1.1 is setisfied ithin I huur enc be in NOT STANOGY ithin 0 ACTION 2 - Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 3 - POWER OPERATION may continue provided that within I hour:

1. The rod is restored to OPERABLE status within the above slignment 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 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

WOLF CREEK - UNIT 1 3/4 1-14 Amendment No. 27, 4, 61

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INSFRT 1-14 i I

ACTION 1 - 1. Determine that the SitUTDOWN MARGIN is greater than or equal to 1.3% Ak/k, with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s), is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

2. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

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. REACTIVITY CONTROL SYSTEMS LIMITING CON 011 TON FOR OPERATION ACTION (Continued) g jg % A k"/g

3. The rod -_is dec,larede___p,in, operable and the SHUTDOWN MARGIN

...:_- ;,_.^.^

_ ' M _;. POWER OPERATION may then continue provided that; a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions;

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g2i-_: _2 g,___........,,m_.._ r bg) A power distribution map is obtained from the movable incoredetectorsandF(Z)andFhareverifiedtobe g

within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and C g) The THERMAL POWER level is re'duced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.

ACTION 4 - Restore the inoperable rods to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at

. least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position. j deviation monitor is inoperable, then verify the group positions at least once '

per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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4.1.3.1.2 Each full-length rod not fully inserted in the core shall be !

determined to be OPERABLE by movement of at least 10 steps in any one direction !

at least once per 31 days. j

-l, l, 3, f,3 INhAT l-85 WOLF CREEK - UNIT 1 3/4 1-15 Amendment No. 27

t INSERT 1-15 4.1.3.1 . 3 Prior to reactor criticality, verify that the rod drop time-of the full-length shutdown and control rods is in accordance with USAR Section 16.1.3.2, with Tgya 3 551*F, and all reactor- ;

coolant pumps operating

a. For all rods following each removal of the reactor vessel 't head, and

b. For specifically affected-individual rods following any maintenance on or modification to the control. rod drive <

system which could affect.the drop. time.of those specific rods.

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POSITION INDICATION SYSTEM-SHUTOOWN q- L

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LIMITING CONDITION FOR OPERATION

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3.1.3.3 One digital rod position indicator (excluding d nd position-indica-tion) shall be OPERA 8LE and capable of determining the c ntrol rod position within 2 12 steps for each shutdown or control rod not ully inserted.-

APPLICABILITY: MODES 3*#, 4*#, and 5*f.

ACTION: /L ^

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With less than the above required position ind cator(s) 0PERABLE, immediately open the Reactor Trip System breakers. .

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SURVEILLANCE REQUIREMENTS

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4.1.3.3 Each of the above r utred digital rod position indicator (s) shall be determined to be OPERABLE b verifying that the digital rod position indicator agrees with the demand po tion indicator within 12 steps when exercised over '

the full range of rod tr el at'least once per 18 months.

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With the actor Trip System breakers in the closed position.

See Spe al Test Exception Specification 3.10.5.

. WOLF CREEK - UNIT.1 3/4 1-18

REACTIVITY CONTROL SYSTEMS '

W ROD DROP TIME Q

LIMITINGCONDITIONFOROPERATION '

1 3.1.3.4 The individual full-len th shutdown and control rod rop time from the physical fully withdrawn pos tion shall be less than or qual to 2.7 seconds l from beginning of decay of stationary gripper coil voltag to dashpot entry with:

a.

T,yg greater than or equal to 551'F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With the rod drop time of any ful length rod determined to exceed the above limit, restore the ro drop time to within the above limit prior to proceeding to MODE 1 2.

b. With the rod drop times wit n limits but determined with three ,

reactor coolant pumps oper ting, operation may proceed provided THERMAL POWER is restric d to less than or equal to 66% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS

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4.1.3.4 The rod drop t'.e of full-length rods shall be demonstrated through measurement prior to r actor criticality:

a. For all r ds following each removal of the reactor vessel head,

b. For s cifically affected individual rods following any maintenance on o modification to the Control Rod Drive System which could aff et the drop time of those specific rods, and

c. t least once per 18 months.

WCLF CREEK - UNIT 1 3/4 1-19 Amendment No. 22, 47

REACTIVITY _ CONTROL SYSTEMS CONTROL R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1* and 2*f.

l ACTION: I With the control banks inserted beyond the insertion limits specified in the (d # COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:

I it A --P 6 y. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or c #. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL' POWER which is allowed by the bank position using the insertion limits specified in the COLR, or J g. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS V

4.1.3.6,lThe position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K g , greater than or equal to 1.

WOLF CREEK - UNIT 1 3/4 1-21 Amendment No. 61

1 INSERT 1-21A

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , verify that the SHUTDOWN MARGIN is greater than or equal to 1.3% Ak/k ' or initiate boration until the SHUTDOWN MARGIN is restored to >

greater than or equal to 1.3% Ak/k, and INSERT 1-21B 4.1.3.6.2 When in Mode 2 with Keff less than 1, verify that the predicted critical control rod position is within insertion limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality.

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INSTRUMENTATION MOVABLE INCORE DETECTORS 9@

l!MITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABL with:

a. At least 75% of the detector thimbles,

b. A minimum of two detector thimbles per core adrant, and

c. Sufficient movable detectors, drive, and r adout equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detecti n System is used for:

a. Recalibration of the Excore Neutr Flux Detection System,

b. Monitoring the QUADRANT POWER LT RATIO, or

c. Measurement of F,(X,Y,Z) an F3 ,(X,Y) .

ACTION:

a. With the Movable Inco e Detection System inoperable, do not use the system for the abov applicable monitoring or calibration functions,

b. The provisions o Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIR ENTS 4.3.3.2 The Mo able Incore Detection System shall be demonstrated OPERABLE at least once pe 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

a. ecalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Measurement of F,(X,Y,Z) and F,,(X,Y).

/

WOLF CREEK -_ UNIT 1 3/4 3-43 Amendment No.61

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INSTRUMENTATION jQC C5 SEISMIC INSTRUMENTATION ; -

LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3 shall be OPERABLE.

APPLICABILITY: At all. times.

ACTION:

a. With one or more of the above required seis c monitoring instruments inoperable for more than 30 days, prepare nd submit a Special Report to the Commission pursuant to Spe fication 6.9.2 within the next 10 days outlining the cause of th malfunction and the plans

, for restoring the instrument (s) to OP RABLE status,

b. The provisions of Specifications .0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

/

4.3.3.3.1 Each of the above equired seismic monitoring instruments shall be demonstrated OPERABLE by t performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG C NNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of e above required seismic monitoring instruments actuated during a seismic ent greater than or equal to 0.01 g shall be restored to OPERABLE status ithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 10 days follo ng the seismic event. Data shall be retrieved from actuated instruments nd analyzed to determine the magnitude of the vibratory ground motion. A pecial Report shall be prepared and submitted to the Commission pursuant o Specification 6.9.2 within 14 days describing the magnitude, fre-quency pectrum, and resultant effect upon facility features important to safe .

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WOLF CREEK - UNIT 1 3/4 3-44

l

>%e TA?tE 3.3-7 j SEISMIC 5'DNII ' RING INSTRUMENTATION P.NIMUM MEASUREMENT 'aSTRUMENTS INSTRUMENT 5 AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Peak Recording Accelerographs

a. Radwaste Base Slab i 1. 0g I

c. Control Room 1.0g 1

c. ESW Pump Facility 1.0g 1

c. Ctmt Structure ! 2.0g i

e. Auxiliary Bldg. SI Pump Suctions 1 1. 0g i f, SGB Piping

5.0 1

g. SGC Support t 1. g i

2. Triaxial Time. History and Response Spectrum Recording System, Monitoring tne Following Accelerometers (Active)

a. Ctmt. Base Slab 1. 0g i

b. Ctmt. Oper. Floor 2 1. 0g I

c. Reactor Support 2 1.0g i

d. Aux. Bldg. Base Slab i 1.0g i

e. Aux. Bldg. Control Room Air Filter 2 1. 0g i

f. Free Field 2 0.5g 1

3. Triaxial Response-Spectrum ecorder (Passive)

Ctmt. Base Slab i 1. 0g 1 4 Triaxial Seismic Swi ches ACCELERATION LEVEL North East Vertical

a. OBE Ctmt. ase Slab 0.06g 0.06g .' ;

0.069 1

b. SSE Ctm Base Slab 0.15g 0.15g 0.16g A 1

c. OBE C . Oper. F1 0.07g 0.07g 0.07g i

d. SSE C t. Oper. F1. 0.16g 0.16g 0.179 I

e. Sys m Trigger 0.01g 0.01g 0.01g . 1 WOLF CREEK - UNIT 1 3/4 3-45

. , ,. - - , . ~ . . - .. - ... . . . ~.

. ol lL 4

. , $r TABLE 4.3-4 8

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REOUIREM f75 ,

ANALOG CHANNEL- i CHANNEL .

CHA EL OPERATIONAL- ,

INSTRUMENTS'AND SENSOR LOCATIONS CHECK CAL RATION ' TEST' i

1. Triaxial' Peak Recording Accelerographs

a. Radwasue Base Slab N. A. R N.A.

b. Control Room N. A. R N.A. -i

c. ESW Pump Facility N.A. R N.A. ;

d. Ctat Structure N. . R N.A.

e. Auxiliary Bldg. SI Pump Suction N A. R N. A.

f. SGB Piping .A. R N.A.

g. SGC Support N.A. R N.A. ;

2. Triaxial Time History and Response Spectrum Recording System, Monitoring -

the Following Accelerometers (Active

a. Ctmt. Base Slab- M R SA ;

b. Ctmt. Oper. Floor M R SA. !

c. Reactor Support M R JSA"" i

d. Aux. Bldg. Base Slab M - R SA*"

e. Aux. Bldg. Control Room flters M .R SA"*- i

-f. Free Field M R SA*" j

3. Triaxial Response-Spectr Recorder (Passive) ,

i Ctmt. . Base Slab -N.A. R N.A."- :

4. Triaxial Seismic itches ,

a. OBE Ctmt. se Slab H R SA'

b. .SSE Ctmt. ase Slab -M R SA

c. OBE Ctmt Oper. Fl. M R SA

d. SSE C . Oper. Fl. M R SA

e. Syste Trigger M R SA

Checki at the Main Control Board Annunciation for contact closure out:ut.

.in th Control Room shall be performed at-least once per 184. cays.

"*The i-stable Trip Setpoint'need not be determined during the performance of n ANALOG CHANNEL:0PERATIONAL TEST.

^0LF a CREEK.- UNIT l'- 3/4 3-46

- - , , g ,

4 c

N57Ru'ENTATION b+$

>ETE:RC'_0GIC',L :'STRUuENTA':0N

_It': IN3 CONOI ION FOR OPERATION i

3,3.3.4 The meteorological monitoring instrumentation chann s in Table 3.3-5 snall be OPERABLE.

L!: ABILITY: At all times.

A: :0N:

a. With one or more required meteorological monitoring channels inoperaole for more than 7 days, prepare and subm' a Special Report to tne Commission pursuant to Specification .9.2 within the next 10 days outlining the cause of the malfunct n and the plans for restoring the channel (s) to OPERABLE status. ,

b. The provisions of Specificatior 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.3.3.4 Each of the above eteorological monitoring instrumentation channels shall be ceconstrated OPE BLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION at e frequencies shown in Table 4.3-5.

Y WOLF CREEK - UNIT 1 3/4 3-47

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0 4

Q TABLE 3.3-8 i\

METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUNENT LOCATION OPERABLE

1. Wind Speed Nominal Elev 10m 1 Nominal Elev. 6m 1

2. Wind Direction Nominal Ele . 10m 1 Nominal ev. 60m 1

3. Air Temperature - AT Nomi al Elev. 10m-60m 1 -

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'a0LF 'REE.( - UNIT 1 3/4 3-48 um_;.

/

.\

zk TABLE 4.3-5 ME~EOROLOGICAL MONITORING INSTRUMENTATI N SURVEILLANCE REQUIREMENTS.

CHANNEL CHANNEL

!NSTRUMENT CHECK ALIBRATION

1. Wind Speed

a. Nominal Elev. 10m 0 SA

b. Nominal Elev. 60m 0 SA

2. Wind Direction

a. Nominal Elev. 10m SA

b. Nominal Elev. 60m 6 SA

3. Air Temperature - ST

a. Nominal Elev.10-60m 0 SA WOLF CREEK - UNIT 1 3/4 3-49.

i ,

A IN5 RUMENTATION

EN' 'CN!T 0:NG :NSTRUMENTAT :N LD':':N3 CONDITION FOR OPERATION-

+

3.3.3.6 The accident monitoring instrumentation channels shown in. Table 3.3-10 shall be OPERABLE.

ADPLICABILI'Y: MODES 1, 2, and 3.

ACT:3N: _

With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table'3.' 0, tore the inoperable channel (s) to OPERABLE status within 7 s;

)[NSb{b 3-h othe ise, be in at least HOT STANDBY within the next 6 hou in HOT ' TOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

and M '

b. With the numbe f OPERABLE accident monitoring i rumentation channels, except containment radiation le monitor and the unit vent - high rang oble gas monitor, s than the Minimum-Channels OPERABLE requir nts of Tabl . 3-10, restore the . inoper-able channel (s) to OPERABLE

tus in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ;'otherwise, be 'in at-least HOT STANDBY within the t 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in. HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With the numbe'r of.0PE LE channels'for containment radiation level monitor or th nit vent ~- high range n le gas monitor less than the Minimu annels OPERABLE requirements Table 3.3-10, initiate the aplanned alternate method of monitori the appropriate parameter within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and either restore the ino able channel to OP" LE status within 7 days, or prepare and submit a cial Re t to the Commission pursuant to Specification 6.9.2 with days that provides actions taken, cause of the inoperability an plans and schedule for restoring the channels to OPERABLE status. o ejd The provisions of Specification 3.0.4 are not applicable. ^

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at thel frecuencies snown in Table 4.3-7.

WOLF CREEK - UNIT 1 3/4 3-53 1

INSERT'3-53

a. With the number of OPERABLE accident monitoring instrumentation channels less than the. Total Number of Channels shown in Table 3.3-10,~ restore the inoperable channel (s) to OPERABLE status within 30 days or prepare and submit a Special Report to the Commission pursuant to Technical Specification 6.9.2 within.the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status,

b. With the number of OPERABLE accident monitoring instrumentation channels, -i except for instrument functions 10, 16 and.18 (Containment Hydrogen Concentration Level, Containment Radiation Level, and the Reactor Vessel Level Indicating System), less than the Minimum. Channels OPERABLE-requirements of Table 3.3-10, restore one channel to OPERABLE status within 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With the number of OPERABLE channels for instrument functions 16 and 18 (Containment Radiation Level and the. Reactor Vessel Level Indicating System), less than the Minimum Channels OPERABLE requirements of Table 3.3-10, initiate the preplanned alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and either restore one inoperable channel to OPERABLE status within 7 days, or prepare and submit'a Special Report to the Commission pursuant to Technical Specification . 6.9.2 within the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

d. With the number of OPERABLE channels for the containment hydrogen.

concentration level monitor less than the Minimum Channels . OPERABLE requirement of Table 3.3-10, restore one channel to OPERABLE status within 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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TABLE 3.3-10 D~ ACCIDElli MONITORING INSTRllMENTAll0N o

In TOTAL HINIMUM n

NO 0F CilANNELS s., -lil51RilMENT CllANNELS OPERABLE

1. Containment Pressure - IVM *L MI C g g cg a) Sr.;a! Lag 2 1 Z id O tc..Jcd Eng: 1.

2. Reactor Coolant Outlet Temperature - Til0T (Wide Range) 2 1

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1

4. Reactor Coolant Pressure - Wide Range 2 1

5. Pressurizer Water Level 2 1

6. Steam Line Pressure 2/ steam generator 1/ steam generator Mu

7. Steam Generator Water Level - Harrow Range ,2.2/ steam generator 1/ steam generator ta ' 8. Steam Generator Wate.- Level - Wide Range 1/ steam generator

- 1/ steam generator

'f. 9. Refueling Water Storate Tank Water Level 2 1

10. Containment flydrogen Cancentration Level 2 1

11. Auxiliary feedwater Flaw Rate 1/ steam generator 1/ steam generator

12. PC; .' : , ! t ; .. i..d k at-. * . DELETED :/va:.c is.;..

13. ; u.-: . J i mm .o!.m P u a . m . v.. i..u.uoiu. " DC i/Voive iivaive

14. R ftty W hc " ;i'!:r '-dit t - NenTc" bh 1/Volvc 1 1/Ve :-T I .+

15. Containment Water level 2 1

16. Containment Radiation' Level (Iligh Range) -ft-*- 1 1 t/. Thermocouple / Core Cooling Detection System 4/ core quadrant 2/ core quadrant lit. ":' "- -t "ig " ..,c L L;m C a 3 ;;m... i o ,

e2 g R eccic.* t/rsje t (_ e ve t. .Tu ketaj Syurm

^

. .., ; . .. .. . : ; c .J, k ;; m ; ., .. ;aci LJ m:ack ..:.m a a . . . . G.m c : ,m a ,, . i . u...

" " ;. , i m .. . . . ..:. k .; U .m ..:uci ..:.m i. ; ; k- . .. m;.s u - , , a , ,

. . u . . ...u

.m. . ,,u.c. .2 .c-u,cJ

TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS' E

!;; CHANNEL CHANNEL n INSTRtMENT CHECK CALIBRATION

1. Containment' Pressure - NceaaI /* vjc- M R n

, 2. Reactor. Coolant Outlet Temperature - T I"Id' "*"8'} " "

HOT e

g- 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R

- 4. Reactor Coolant Pressure - Wide Range M . R

.I

5. Pressurizer Water Level M R

6. Steam Line Pressure M R

^ .l

7. Steam Generator Water Level - Narrow Range M R

8. Steam Generator Water Level - Wide Range M R.

9. Refueling Water Storage Tank Water Level M R w

i 10. Containment Hydrogen Concentration Level M R y 11. Auxiliary Feedwater Flow Rate M R E 12. i as 7-. ;i;-. ; J;s.t.. " Deleted -#- J.A.

13. Nr 7.. ;t;.;. hdk;ts** Delete -M- -#.*:

14. E:':ty "; h; 7;;tt h; "id;;.t.. Neutred fl*/ N 4:h: t2II)

15. Containment Water Level ~ M R ,

16. Containment Radiation Ltvel (High R'ange) M RN

17. Thermocouple / Core Cooling Detection System M N

18. J

"..;i 1";..i ";07. i.6 ti k .0;; ";;!t;r M R Au le.- t/me t' level EudMby CrJbm

> b _ ; .:.;. ;; t; ... ;.; L:.,.; .;;.. u :.. Lt. c h;;d p;i tion. l

- ^:.t 4;;k;th 07 tL. LL i .;;.; h ;ntikd h it, ek;;d p;ithn and power is removed.

j (2)***CHAMMEL CALIBRATION may consist-of an electronic calibration of the channel,. not including the detector, 1 for rage decades above 10 R/h and a' one point calibration check of the detector below 10 R/h with an -

-installed or portable gamma source. ^

(1) NueT- cv7nts5 < .g g(,Je e( frem C h4 M Ca A 4 S *- .

_m .. .-_ . ._ . . , - _ .. ,

INSTRUMENTATION

. \'

LOOSE-PART DETECTION SYSTEM 9 '

-LIMITING CONDITION FOR OPERATION

' l.

3.3.3.9 The Loose-Part Detection Syste.a s%11 be OP 8'LE.

APPLICA8ILITY: MODES 1 and 2. '

ACITON: '

s

a. With one or more Loose-Part Detecti n System channels inoperable 'for more than 30 days, prepare and su it a Special' Report to the Commission pursuant to Specifica on 6.9.2 within the next 10 days outlining the cause of the malf nction and the plans for restoring the channel (s) to OPERABLE st us.

4

b. The provisions of Specifica tons 3.0.3 and 3.0.4 are not applicable. ,

a J

SURVEILLANCE REQUIREMENTS- J 4.3.3.9 Each channel of the Loose-Part Detection Systes shall be demonstrate'd- j OPERABLE by performanc of I

a. A CHANNEL ~ HECK at least once per 24' hours,.

H

b. An ANAL CHANNEL OPERATIONAL TEST except for. verification of Setpoint 1 at le t once per 31 days, and

c. AC NNEL CALIBRATION at least once per 18 months. 'i -

/ -J

~ WOLF CREEK - UNIT 1 3/4 3-57 Amendment No..15 b

INSTRUMENTATION h

\"

EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 1

3.3.3.11 The explosive gas monitoring instrumentation chan is shown in Table 3.3-13 shall be OPERABLE with their Alare/ Trip Setpo ts set to ensure that the limits of Specifications 3.11.2.5 are not excee d.

APPLICABILITY: As shown in Table 3.3-13. '

ACTION:

a. With an explosive gas monitoring instr ntation channel Alarm / Trip Setpoint less conservative than requir d by the above specification, declare the channel inoperable and t a the ACTION shown in Table 3.3-13.

b.

With less than the minimum number of explosive gas monitoring instru-mentation channels OPERABLE, ta the ACTION shown in Table 3.3-13.

Restore the inoperable instrum tation to OPERABLE status within 30 days, and, if unsuccessful, p epare and subsit a Special Report to the Commission pursuant to acification 6.9.2 to explain why this inoperability was not corr ted in a timely manner.

c. The provisions of Specif cation 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

/

i 4.3.3.11 Each explosive as monitoring instrumentation channel shall be demonstrated OPERABLE b performance of the CHANNEL CHECX, CHANNEL CALIBRATION and ANALOG CHANNEL OPE TIONAL TEST at the frequencies shown in Table 4.3-9.

I

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WOLF CREEK - UNIT 1 3/4 3-58 Amer.dment No. II, 42

_ _ _ _ _ _ l

i TABLE 3.3 ,

EXPLOSIVE GAS MONITORING INSTRUMENTATION n MINIMUM CHANNELS .

INSTRUMENT OPERABLE APPLICABILITY ACTION

1. (Not Used) b

2. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System-

a. Hydrogen Monitor 1/Recombiner **

44 ;

b. Oxygen Monitor 2/ Rec ner **

42 M

+ .

Y.

u g

4 7 i

a-Y

.U

J b h..

xs -

TABLE 3.3-13 (Continued) g

- During WASTE GAS HOLDUP SYSTEM operation.

ACTION STATEMENTS ACTION 38 - (Not Used)~ !

ACTION 39 - (Not Used)

ACTION 40 - (Not Used)

ACTION 41 - (Not Used) '

ACTION 42 - With the Outlet Oxygen Moni r channel inoperable,

-operation of the system na continue provided grab .;

samples are taken and ana zed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With both oxygen channels or both the' inlet oxygen and inlet hydro n channels inoperable, suspend oxygen ' supply to the r combiner. Addition of waste gas to the system na continue provided' grab samples are taken and analy d at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operatio s and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 4 during other oper tions. ~

ACTION 43 - (Not Used).

ACTION 44 - With.the ni r of channels OPERA 8LE one less than required by the Minimum Channels OPERA 8LE requirement,.

1 suspend o gen supply to.the recombiner. '

(

I WOLF CREEK - UNIT 1 3/4 3-60 Amendment No.15,42

,h

TABLE 4.3-9 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQtilREMENTS r3

$$ CHANNEL MODES FOR WHICH CRANNEL CHANNEL OPERATIONAL SE INSTRUMENT SURVEILLANCE CHECK CALIBRATION TEST

. IS REQUIRED g 1. (Not used)-

" 2. WASTE GAS HOLDUP SYSTEM Explosiv Gas Monitoring System

a. Inlet Hydrogen Monitor Q(4) M **

b. Outlet Hydrogen Monitor D Q(4) M **

c. Inlet Oxygen Monitor D Q(5) M **

,, d. Outlet Oxygen Monitor D Q(6) M **

D Y

D

/

t 5

t

?

t

.st

TABLE 4. 3-9 (Continued) \

l TABLE NOTATIONS

- Ouring WASTE GAS HOLDUP SYSTEM operation.

(1) (Not Used)

(2) (Not Used)

(3) (Not Used)

(4) The CHANNEL CALIEDATION shril include the e of standard gas samples containing a noniinal:

a. One volume percent hydrogan, bala e nitrogen and

b. Four volume percent hydrogen, b ance nitrogen.

(5) The CHANNEL.

containing CALIBRATION shall inc de the use of standard gas samples a nominal:

a. One <olume percent oxyg , balance nitrogen..'and

b. Four volume percent o gen, balance nitrogen.

(6) The CHANNEL containing CALIBRATION hall include the use of standard gas samples a nominal:

a. 10 ppm by vol oxygen, balance nitrogen, and

b. 80 ppm by v use oxygen, balance nitrogen.

i g

WOLF CREEK - UNIT 1 3/4 3-62 Amendment No. 15, 42

~

INSTRUMENTATION g 9

' C

\v 3/4~3.4' TUR81NE OVERSPEED PROTECTION 9 i LIMITING CONDITION FOR OPERATION ,

i 3.3.4 At least one Turbine Overspeed Protection System shall'b OPERA 8LE. I APPLICA8ILITY: MODES 1;.2,*.and 3.*

j ACTION:

a. With one stop valve or one governor valve per h h pressure turbine !

steam line inoperable and/or with one reheat s p valve or'one-reheat intercept valve per low pressure turbi steam line inoper-able, restore the inoperable valve (s) to 0P 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,'or close at least one valve in t affected steam lines or isolate the turbine from the steam sup y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

b. With the above required Turbine 0verspe Protection Systee otherwise 1 inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the urbine from the steam supply.

SURVEILLANCE REQUIREMENTS j

4.3.4.1 The provisions of Specification 4. 4 are not applicable.

4.3.4.2 The above required Turbine Over eed Protection System shall be demonstrated OPERA 8LE:

a. At least once per 7. days by cycling each of the following valves through at least one comp 1 te _ cycle from the running position:

1) Four high pressure urbine stop valves,-

2)- Six low pressure urbine reheat stop valves, and

3) Six low pressu turbine reheat intercept valves.

b. At least once per 1 days by ' cycling each of the four high pressure main turbine gov nor valves through at least one complete cycle ~

from the runni position;.

c. At least once er 31 days by direct observation of the~ movement of- .

each of the ve valves through one complete cycle from the running _ '

position;: '

d. At least nce per 18 months by performance of a CHANNEL' CALIBRATION on the- rbine Overspeed Protection Systems; and-

e. At le t once per 40 months by disassembling'at least one of each of' the ve-valves and performing a visual and surface inspection:of val seats, disks and stems and verifying no unacceptable flaws' or.

co sion.

Not appl cable in M00E 2 or 3 with all main steam line: isolation valves and associ ed bypass valves in the the closed position and all other steam flow paths o the turbine isolated.

WOL CREEK - UNIT l' 3/4 3-63 Amendment'No. II, 42' ,

t

-. , ,. m, . . < - _ , . ,

!ACTC~ 000LANT SYSTEM 3/ 4.2 SAFE 7Y VALVES

\

q

$"UT00WN LIITING CONDITION FOR OPERATION I

3. 4. 2.1 A minimum of one pressurizer Code safety v ve shall be OPERABLE with a lift se;;ing of 2a85 osig : 1% .

A:oLICASILITY: MODES 4 and 5.

ACTION:

With no pressurizer Code safety valve OPERAB E, immediately suspend all crerations involving positive reactivity ch nges and place an OPERABLE RHR

. loco into operation in the shutdown coolin mode.

l SURVEILLANCE REOUIREMENTS i

'l 4.4.2.1 No additional requi ements other than those required by Specification 4.0.5.

I "Tne lift se ting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

WOLF CREEK - UNIT 1 3/4 4-7

m n ,

9 REACTOR COOLANT-SYSTEM' .

~

'3/4.4.4 RELIEF VALVES ' q LIMITING CONDITION FOR OPERATION 3.4.4 Both power-operated relief valves .(PORVs) and their associated block j valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.* -

ACTION:

With one or both PORVs inoperable because of excessive seat leakage, a.

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close-the associated block valve (s) with power maintained to the block .

valve (s); otherwise, be in at least HOT STAND 8Y within the next 6-hours and in HOT SHUTDOWN within the following'6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one PORV inoperable due to causes-other than excessive seat -

leakage,.within I hou'r either restore the PORY to OPERABLE status or close its associated block valve and remove power = from the block "

valve; restore the PORV to OPERABLE status within the following 72-hours or be in HOT STANDBY within the next.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With.both PORVs inoperable due to causes' other than excessive seat.

leakage, within I hour either restore at least one PORV to OPERABLE. '

status or cl';e its associated block valve and remove power f om the - '

block valve:and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months anc in: HOT

' ^ ~

SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one or both block valves inoperable, _ within'1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore. the -

block valve (s) to OPERABLE status or place its associated PORV(s) in manual control. Restore at least one block # valve to OPERABLE status restore any !

within the next hour if both block ~ valves remaining inoperable block valve to OPERABLE status withi are inoperable;'n 72~ hours; otherwise,'be in at least' HOT. STANDBY within-the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Sp4cification.4.0.5, e'ach PORV~ ,

shall be demonstrated OPERABLE at least once per 18 months by performing a. '

CHANNEL CALIBRATION of the actuation instrumentation.

4.4.4.2 Each block valve shall- be demonstrated OPERABLE at least~ once 'per 92 '3 days by operating the valve through one complate' cycle of full travel.unless ' ,

the' block valve is closed in' order to meet the requirements of ACTION b. or c.

lugAT in Specification 3.4.4.

4 -10

i

With all RCS cold leg temperatures above 368'F.

WOLF CREEK - UNIT 1 3/4 4-10 , Amendment No. 63

- ~ _ _ _ _ . _ _ 1.. 1 -._ _ _ , . -

A INSERT 4-10 4.4.4.3 Both PORV position indicators shall be demonstrated OPERABLE at least once per 31 days by performance of a CHANNEL CHECK unless the associated block valve is in the closed position.

4.4.4.4 Both PORV block valve position indicators shall be demonstrated OPERABLE at least once per 31 days by performance of a CHANNEL CHECK unless the block valve is verified in the closed position and power is removed.

RE :T:: :20LANT SYSTEM w

3 2.5 STEA4 GENERATORS p

.:v: :N3 00N0! TION :3R OPERATION

/

3. 4. 5 Each steam generator shall be OPERABLE.

A:oLI:AEILITY: MODES 1, 2, 3 and 4.

ACT!CN:

Witn one or more steam generators inoperable, re ore the inoperable steam generator (s) to OPERABLE status prior to increa ng T,yg aDove 200 F. ,

53VE: LLANCE REQUIREMENTS I

4.4.5.0 Each steam generator shall be d onstrated OPERABLE by performance of tne following augmented inservice inspe ion program and the requirements of Speci fication 4.0.5.

4. 4. 5.1 Steam Generator Samole Sele tion and Inspection - Each steam generator shall ce ceterminec OPERABLE curinp/snutdown Dy selecting and inspecting at least tne minimum number of steam / generators specified in Table 4.4-1.

a.4.5.2 Steam Generator Tube Sadole Selection and Inspection - The steam '

generator tuoe minimum sample ze, inspection result classification, and the corresponding action required hall be as specified in Table 4.4-2. The inservice inspection'of stea. generator tubes shall be performed at the fre-ouencies specified in Speci ication 4.4.5,3 and the inspected tubes shall be verified acceptable per th acceptance criteria of Specification 4.4.5.4. The tubes selected for each i service inspection shall include at least 3% of the total number of tubes i all steam generators; the tubes selected for these inspections shall be s ected on a random basis except:

a. Where expe ence in similar plants with similar water chemistry indicates ritical areas to be inspected, then at least 50% of the tubes in ected shall be from these critical areas;

b. The fi st sample of tubes selected for each inservice inspection' (subs quent.to the preservice inspection) of each steam generator shal include:

WOLF CREEK - UNIT 1 3/4 4-11

l 1

-l

\f 'l

. REACTOR COOLANT SYSTEM 9'v u 1 l

)

l SURVEILLANCE REQUIREMENTS (Continued) j

1) All nonplugged tubes that previously had etectable wall-penetrations (greater than 20%), '

l

2) TubesLin those areas where experience as indicated potential problems, and

3) A tube inspection (pursuant to Spe ification 4.4.5.4a.8)'shall be performed on each selected tub . If any selected tube does not permit the passage of the ed y current probe for a tube inspection, this shall be recor ed and an adjacent tube shall be selected and subjected to a_ tube inspection.

c. The tubes selected as the second nd third samples (if required by

Table 4.4-2) during each inservi e inspection may be subjected ~to a- ,

partial tube inspection provid :

1) The tubes selected for ese samples include the tubes from.

those areas of the tub sheet array where tubes with imperfections'were pr tously found, and ,

2) The inspections inc ude those portions of the tubes where imperfections were previously found.

-The results of each sample insp ction shall be classified into one of'the ;

following three categories: ;

Category Insoection Results .

- J C-1 Less than 5% of the total tubes inspected are degraded tubes'and none of the inspected tuces ;

are defective.

C-2 One or more tubes, but not more than D; of the total tubes inspected are defective,'or between ,

5% and 10% of the. total tubes inspected are ,

degraded tubes. !

C-3 More than~10% of the total tubes inspectec are degraded tubes or more than'L% of the ins:ected tubes are defective.

1 Note: In all inspections, previously degraded'tuces must exrd:! j significant (greater than 10%) further wall penetrations !

to ce included in the above percentage calculatices. i i

~

1 WOLF CREEK - UNIT 1 3/4'4-12' i

b t/

N RE2 :R 2:0LANT SYSTE'4 Q SUr'.E:_ LANCE REOUIREMENTS (Continued)

/

N

. :.5.3 !rscection Frecuencies - The above reouired ins .vice inspections of '

stes generator tuces snail ce performed at the followi g frequencies:

a. The first inservice inspection shall be per ormed after 6 Effective Full Power Months but within 24 calendar tr nths of initial criticality Subsecuent inservice inspections shall b performed at intervals of not less than 12 nor more than 24 calen ar months after the previcus inspection. If two consecutive inspec ions, not in:luding tne pre--

service inspection, result in all ins ection results falling in:c ne C-1 category or if two consecutive i spections demonstrate that pre-vicusly observed degradation has n continued and no additional degradation has occurred, the ins Ction interval may be extendec to a maximum of once per 40 months; , .

5. If the results of the inservic inspection of a steam generator conducted in accordance with able 4.4-2 at 40-month intervals fall in Category C-3, the inspec on frequency shall be increased to at least once per 20 months. he increase in inspection frequency shall apply until the sub equent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 onths; and

c. Additional, unschedu d inservice inspections shall be performed on each steam generato in accordance with the first sample inspection !

specified in Table 4.4-2 during the shutdown subsequent to any of the following co itions:

1) Reactor-t secondary tubes ~ leaks (not including leaks originating from tub to-tube sheet welds) in excess of the limits of Speciff ation 3.4.6.2, or

2) A sei mic occurrence greater than the Operating Basis Earthquake, or

3) A oss-of-coolant accident requiring actuation of the Engineered afety Features, or '

,, A main steam line or feedwater line break.

WOLF CP.EEK - UNIT 1 3/4 4-13

I I

REACTOR COOLANT SYSTEM S 4

. s' V

SURVEILLANCE REOUIREMENTS (Continued)

/

4.4.5.4 Acceptance Criteria

a. As used in this specification:

1) Imoerfection means an exception to th dimensions, finisn or contour of a tube from that require _by fabrication drawings or specifications. Eddy-current test g indications below 20% of the nominal tube wall thickness, f detectable, may be considered as imperfections;

2) Decradation means a service-i duced cracking, wastage, wear or general corrosion occurring n either inside or outsice of a tube;

3) Degraded Tube means 3 tu e containing imperfections greater r than or equal to 20% of the nominal wall thickness causec ey degradation;

4) % Degradation means the percentage of the tube wall tnickness affected or remov by degradation;

5) Defect means an imperfection of such severity that it exceecs the plugging mit. A tube containing a defect is defective:

6) Pluoging Li it means the imperfection depth at or beyond anich tne tuce s all be removed-from service and is equal to 40%

of the n inal tube wall thickness;

7) Unserv ceable describes the condition of a tube if it leaks or cont ns a cefect large enough to affect its structural integ-rit in the event of an Operating Basis Earthquake, a loss of-co lant accident, or a steam line or feedwater line break as ecified in Specification 4.4.5.3c., above:

8) Tube Insoection means an. inspection of the steam generat:r t.:e from the point of entry (hot leg side) completely arourc tne U-bend to the top support of the cold leg; and

\,

WCLF CREEK - UNIT 1 3/4 4-14

k.

RE;: :R COOLANT Svs D*

5JR.E:LLANCE REOL':REMENTS (Continued)

/

9) Preservice Inscection means an inspectio of the full lengin of eacn tupe in eacn steam generator perfo med by eddy current techniques prior to service to establi h a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the e Jipment and teChnioues ex-pected to be used during subsequen inservice inspecticns, o.

The steam generator shall be determin d OPERABLE after completing tne corresponding actions (plug all ubes exceeding the plugging limit and all tuees containirg the gh-wall cracks) required by Table 4.4-2.

4. ' . 5. 5 Reoorts

a. Within 15 days following the empletion of each inservice inspection of steam generator tubes, t number of tubes plugged in eacn steam generator snall be reporte to the Commission in a Special Report-pursuant to Specification 6.9.2;

b. The complete results of the steam generator tube inservice inspection shall be submitted to he Commission in a Special Report pursuant to Specification 6.9.2 ithin 12 months following the completion of tne inspection. This 5 ecial Report shall include:

1) Number and tent of tubes inspected,

2) Location nd percent of wall-tnickness penetration for each indicati n of an imperfection, and

3) Identi ication of tubes plugged.

c. Results f steam generator tube inspections, which fall into Category C-3, sh 11 be reported in a Special Report to the Commission pursuant to Sp ification 6.9.2 within 30 days and prior to resumption of plan operation. - This report shall provide a description of investi-gat ns conducted to determine cause of the tube'cegradation and-co rective measures taken to prevent recurrence.

l I

WOLF CREEK - UNIT 1 3/4 4-15 I

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x ST E AM GE NEll A T O__ll_T ullE__INSPEC I. lON ISkAMPLE INSPECTION 2ND SAMPLE INSPECllON 38tD SAMPLE INSPECllON Sampic Site it Action llequued flesult Action llequoni licsult A4. tion lintunni .

A minimum of C-1\ None N. A. N. A. N. A.

S Tutees per \ N A.

S. G.

C-2 Plug de teve tubes C-1 None and inspect ditional N A. NA Plug defective tubes C-- 1 None 2S tulses in the . G. C-2 and mspect additional C- 2 Plug <leIn teve tulen 4S tutecs in this S. G.

Per form ae.1 son for

!" C3 C-3 n suit of lu si s.uople i \ Perform action for Z C-3 result of first N. A. N. A.

same C-3 Inspect all tubes in All other this S. G., plug de- S. G.s are None N. A. N. A.

fective tubes and C-1 inspect 2S tubes in g each other S. G. Some S. G.s perform action for N C-2 but no N. A. N A-additional C-2 result of second sainple Notification to NI1C S. G. are pursuant to %50.72 C-3 (b)(2) of 10 CFil \

Past 50 Additional inspect all tubes in S. G. is C-3 each S. G. and plug defective tubes.

Notification to NIIC N. A. N A.

pursicant to h50.72 (hi(2) of 10 CF fl Para 50

- /

S 3 %

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l REACTORCOOLANTSYeg

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3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 1

3.4.7 The Reactor Coolant System chemistry shall be main ined within the limits specified in Table 3.4-2.

APPLICABILITY: At all times.

ACTION:

MODES 1, 2, 3, and 4:

a.

With any one or more chemistry par .eter in excess of its Steady-State Limit but within its Transi nt Limit, restore the parameter to within its Steady-State Limit w hin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY following 30 within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the hours.

b. With any one or more chem try parameter in excess of its Transient Limit, be in at least HO STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 'and in COLD SHUTCCWN within the following 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . ,

At All Other Times:

With the concentratio of either chloride or fluoride in the Reactor Coolant System in ex ess of its Steady-State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of ite Transient Limit, reduce the pressurizer pressure to less than or equa to 500 psig, if applicable, and perform an engineering evaluTtion to d ermine the effects of the out of-limit condition on the structural int grity of the Reactor Coolant System; determine that the Reactor Cool t System remains acceptable for continued operation prior to increasi g the pressurizer pressure above 500 psig or prior to proceedin to MODE 4.

SURVEILLAN REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be witnin the imits by analysis of those parameters at the frequencies specified in T 1e 4.4-3.

'aCLF C?.EEK - UNIT 1 3/4 4-22

O TLE'E 3.

. ~2 OE C CD COCLANT SYSTEM CWEMIST9Y LIM!TS STEADY-STATE RANSIENT 0:R;"E*ER LIMIT LIMIT Diss:!vec Cxygen" 1 0.10 ppm i 1.00 ppm Cricrdce 1 0.15 opm i 1.50 ppm

!ueri:e 1 0.15 com 1 1.50 ppm l

l

Licit nct aco icable witn T,yg less than or equal to 250*F.

1

/

WOLF CREEK - UNIT 1 3/4 4-23

e TABLE r.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENT SAMP AND PARAMETER ANALYSI FREQUENCY Dissolved Oxygen

At 1 st once per 72-hours Chloride At east once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Fluoride t least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months "Not required with T,yg less t n or equal to 250*F t

i WCLF CREE.< - UNIT 1 3/4 4-24

F i

REACTOR COOLANT SYSTEM 1(}

PRESSURIZER LIMITING CONDITION FOR OPERATION

/

3.4.9.2 The pressurizer temperature shall be limited o: ,

a, A maximum heatup of 100*F in any 1-hour p tod,

b. A maximum cooldown of 200*F in any 1-h r period, and

c. A maximum spray water temperature di forential of 583*F.

APPLICABILITY: At all times.

ACTION:

With the pressurizer temperature limit in excess of any of the above limits, restore the temperature to within the imits within 30 minutes; perform an engineering evaluation to determine he effects of the out-of-limit condition .

on the structural integrity of the ressurizer; determine that the pressurizer remains acceptable for continued eration or be in at.least HOT STANDBY.within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the ressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

J SURVEILLANCE REQUIREMENT 4.4.9.2 The pressu zer temperatures shall be determined to be within the limits at least on per 30 minutes during system heatup or cooldown. The spray water temp ature differential shall be determined to be within the limit at least ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

i 1

WOLF CREEK - UNIT 1 3/4 4-33

RE; ~:; CCOL*NT SYSTEM NIs

- .:: 3 tuC u;r. :NTEca: <

\J cq)"

_:":~:N3 CONDITION FOR OPERATION

/

3. : .10 The structural integrity of ASME Code Class 1, 2 a d 3 components sr.all be maintained in accordance with Specifd > tion 4.4. 0.

3LICABILITY: All MODES.

_0 4 : ',:

a. With the s;ructural integrity of any A E Code Class 1 component (s) not conforming to the above requireme s, restore the structural integrity of the affected component ) to within its limit or isolate the affected component (s) prior to nereasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NOT considerations.

D. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above equirements, restore the structural integrity of the affected omponent(s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature abov 200*F.

c. With the structural i tegrity of any ASME Code Class 3 component (s) not conforming to t above requirements, restore the structural integrity of the a fected component (s) to within its limit or isolate the affected com. nent(s) from service. _

c. The provision of Specification 3.0.4 ars not applicable. ,

SURVEILLANCE RE0VIR ENTS 4.4.10 In add ion to the requirements of Specification 4.0.5, each reactor coolant pump ywneel shall be inspected per the recommendations of Regulatory Position C.4 b of Regulatory Guide 1.14. Revision 1, August 1975.

WOLF CREEK - UNIT 1 3/4 4-37

i a .

l l

. l REACTOR COOLANT SYSTEM \ .

3/4.4.11 REACTOR CCOLANT SYSTEM VENTS 's)'V l i

LIMITING CONDITION FOR OPERATION 3.4.11 At least one reactor vessel head vent path consisti of at least two valves in series powered from emergency busses shall be OP ABLE and closed.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the above reactor vessel head vent path inoper ble, STARTUP and/or POWER OPERATION may continue provided the inoperable ve path is maintained closed with power removed from the valve actuator of al the valves in the inoperaole vent path; restore the inoperable vent path to ERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in C0 0 SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . '

SURVEILLANCE REOUIREMENTS

/

4.4.11 least onceEach reactor per vessel 18 months head vent ath shall be demonstrated OPERABLE at by: i

a. Verifying all manual iso ation valves in each vent path are locked in the open position,

b. Cycling each valve i the vent path through at least one comolete cycle of full trave from the control room during COLD SHUTOCWN or REFUELING, and

c. Veri fying flow rough the reactor vessel head vent paths during venting durin COLD SHUT 00WN or REFUELING.

i WCLF CREEK - UNIT ; 3/4 4-33

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 ECCS SUBSYSTEMS - TAVG s 200 F LIMITING CONDITION FOR OPERATION 3.5.4 All Safety Injection pumps ard och Otrlh Olayr.p f urrp shall be inoperable.

APPLICABILITY: MODE 5 .ith thc .cctcr iceci Chcec th; top of the nccctor Vcccci ficagc, and Mode 6 with the Reactor Vessel head on.*-etus with th; uctcr iceci ch0ec the tcp of thc Rccctor Vccccl ficngc.

ACTION:

a. With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

$. ibll hs Osrb'ha7d Clage.7Fwp GPCM1.C, wiss ors 4 01 Ce4]updClagr.g funy 6 an irmulls slalus ($$in 4 lan.

EURVEILLANCE REOUIREMENTS 4.5.4.I All Safety Injection pumps shall be demonstrated inoperable **

by verifying that the motor circuit breakers are secured in the open position at least once per 31 days.

4.D,4.2 .y t t<?ggrqU al Ora Cadgugal Cle.7mg Puery stall u drennrelmbd irtopmliz**

01 rrds c&ud ksalxs au su:uted in 01 spn pwlhr, ali had was pr D1 days.

iU1s 01 $0b iYabn lod is [dow Us bsp ) 0111clas tad frrf, [slL br f furry

,y uore m.eep 0,een4y a ,0, dug u apzu ,

- An inoperable pump may he energized for testing or for.. filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in-the closed position.

WOLF CREEK - UNIT 1 5-9 Amendment No. 35

. - .. - --- ~. .

l l

1 1

,1 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY- I LIMITING CONDITION FOR OPERATION 1

3.6.1.1 Primary CONTAINMENT-INTEGRITY shall be maintair.cd.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION: -

Without primary. CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within '

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS ,

4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by_ valves, blind flanges, or- deactivated. automatic: valves -

secured in their positions,.except as provided in Table 3.6-1 of Specification 3.6.3; ;

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c. After each closing of each penetration subject to Type B testing; except the containment air locks, if opened following a Type A or.B.

test, by leak rate testing the' seal with gas at a pressure not less than P,, 48 psig, and verifying that when the measured leakage-' rate-for these' seals is added to the leakage rates determined pursuant to-

~

Sp;;;":..L en 4.C.1.21 for all'other: Type B and C penetrations, the

{ combined-leakage.rateislessthan0.60L,.

' (dM 5ec 6*A ./64 /d "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 1shall be verified closed during each-COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

WOLF CREEK - UNIT 1 3/4 6-1 1

i l

. J l

l CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE Q !

LIMITING CONDITION FOR OPERATION l

3.6.1.2 Containment leakage rates shall be limite to:

a. An overall integrated leakage rate of:

1) Less than or equal to L,, 0.20% y weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P,, 48 ps , or

2) Less than or equal to L t, O. 20% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pt, 24 psig.

b. A combined leakage rate of le s than 0.60 L, for all penetrations and valves subject to Type B an C tests, when pressurized to P,, 48 psig.

APPLICABILITY: MODES 1, 2, 3, and .

ACTION:

a. If Reactor Coolant Sy tem temperature is at or below 200'F, with (',,

either the measured overall i egrated containment leakage rate exceeding O.75 L, or 0.75 Lg , as appli ble, or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L,,

restore the overall integ ated leakage rate to less than 0.75 L, or less than Lt , as applicable, and e combined leakage rate for all penetrations subject l to Type B and C tests o less than 0.60 La prior to increasing the Reactor Coolant System tempe ture above 200 F. 1

.o

b. If the R actor Coolant System temperature is above 200 degrees F, '!

with tb measured combined leakage rate for all oenetrations and l valve = subject to Types B and C test exceeding 0.60 L,, j

1) Restore the combined leakage rate to less than 0.60 L, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by one of the following methods:

a) Repairing the failed containment isolation component, or ]

b) Isolating the penetration containing the failed component by l closing and the deactivating one automatic valve, or ' '

c) Isolating the penetration containing the failed component by closing one manual valve, or d) Isolating the penetration containing the failed component by using a blind flange.

2) If the combined leakage rate is not restored to less than 0.60'L a within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

WOLF CREEK - UNIT 1 3/4 6-2 Amendment No.33

. . . . . . . ~ . . . . - . - . . . . . . -. -. .

-CONTAINMENT SYSTEMS CONTAINMENT ' LEAKAGE SURVEILLANCE REQUIREMENTS

/

't.

4. 6.1. 2 The containment leakage rates shall'be d onstrated'at the following' test schedule and shall be determined in conform ce with the criteria speci-l fled in Appendix J of 10 CFR Part 50 using the thods and provisions of ANSI N45.4-1972:

a. Three Type A~ tests (Overall Integr ed Containment; Leakage Rate)'shall be conducted at-40 2 10 month int vals during shutdown at a pressure ,

not. less than either P,, 48 psig or P t 24 psig, during each 10 year..

service period. The third test of each set shall be conducted during

. the shutdown for the 10 year p ant inserv' ice-inspection; . ,

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l WOLF CREEK - UNIT.1 3/4 6-2a Amendment No.33- _

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N7!NENT $Y$7EM3 g .E:.L3CE RECU:REMENTS (C:ntiruec)

I

c. !* aay periodic Type A test fails to meet ei* er 0.75 L r 0.75 L,.

a tSe test schedule for subsequent Type A tes+s shall be reviewed arc a:: roved by the Commission. If two consec .ive Type A tests fail to-meet either 0.75 L, or 0.75 L g, a Type A st shall be performed at least every 18 months until two consecut' e Type A tests meet eitner 0.75 L a r 0.75 L at which time the ab e test schedule may be res ce:;

t

c. The accuracy of each Type A test shall be verified by a supplemental test wnich;

1) Confirms the accuracy of the t st by verifying.that the supple-mental test result, L ,cminus he sum of the Type A and the supe -

imposed leak,gL , is equal t or less than 0.25 L, or 0.25 Lg :

2) Has a duration sufficient eftablish accurately the change in leakage rate between the T pe A test and the supplemental test; anc

3) Requires that the rate a which gas is injected into the contain-ment or bled from the c ntainment during the supplemental test is between 0.75 L, and 1. L, or 0.75 Lg and 1.25 Lt '

d. Type B and C tests shall b conducted with gas at a pressure not less than P,, 4 psig, at Me vals no greater uan N moms ocen for tests involving:

1) Air locks,

2) Purge supply an exhaust isolation valves with resilient material seals, and

3) Valves press ized with fluid from a seal system.

e. Air locks shall e tasted and demonstrated OPERABLE by the requirements of Specificatio 4.6.1.3;

f. Purge supply d sxhaust isolation valves with resilient material seals shall- tested and demonstrated OPERABLE by the requirements of Specific ion 4.6.1.7.2 and 4.6.1.7.4, as applicable;

g. Leakage fr m isolation valves that are sealed with fluid from a seal system ma be excluded, subject to the provisions of Appendix J,Section II.C 3, when determining the combined leakage rate provided the sea system and valves are pressurized to at least 1.10 P (53 ps ),andthesealsystemcapacityisadequatetomaintain syste, pressure for at least 30 days; and

h. The revisions of Specification 4.0.2 are not applicaole.

WOLF CP.EEK - NIT 1 3/4 6-3

v CONTAINMENT SYSTEMS N CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION I

3. 6.1. 6 The structural integrity of the containeen vessel shall be maintained at a level consistent with the acceptance criteria n Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the abnormal degradation indi ted by the conditions in Specification 4.6.1.6.la.4, restor the tendons to the required level of integrity or verify tha containment integrity .is maintained within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and rform an engineering evaluation of the containment and provide a ecial Report to the Commission within 15 days in accordance th Specification 6.9.2 or be in at least HOT STANDBY within the ext 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With the indicated abnorma degradation of the structural integrity other than ACTION a. at a evel below the acceptance criteria of Specification 4.6.1.6, r store the containment vessel to the ,

required level of integ ity or verify that containment integrity 1s maintained within 15 d'ys and perfann an engineering evaluation of a the containment and p ovide a Special Report to the Commission within 30 days in ac ordance with Specification 6.9.2 or be in at least HOT STANDBY w thin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 h urs. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENT 5 .

4.6.1.6.1 Conta1nment essel Tendons. The structural integrity of the prestressing tendons the containment vessel shall be demonstrated at the end of 1, 3, and 5 y rs following the initial containment vessel structural integrity test and a 5 year intervals thereafter. The structural integrity of the tendons sha be demonstrated by:

a. Determ ing that a random but representative sample of at least 11 tendo (4 inverted U and 7 hoop) each have an observed lift-off fore within the predicted limits established for each tendon. For eac subsequent inspection one. tendon from each grouD (1 inverted U an I hoco) shall be kept unchanged to develop a history and to c rrelate the observed data. The procedure of insoection and the endon acceptance criteria shall be as follows:

WOLF CREEK - UNIT 1 3/4 6-8 Amendment No. 31

\v CONTAINMENT SYSTEMS T

SURVEILLANCE REQUIREMENTS (Continued)

I

1. If the measured prestressin force of t e selected tendon in a groupliesabovetheprescrfbedlower imit, the lift-off test I is considered to be a positive indica on of the sample tendon's acceptability,

2. If the measured prestressing force f the selected tendon in a group lies between the prescribed ower limit and 90% of the prescribed lower limit, two adjac t (accessible) tendons, one on each side of this tendon shal be checked for their prestress-ing forces. If the prestressin forces of these two tendons are above 95% of the prescribed low r limits for the tendons, all three tendons shall be restore to the required level of integrity, and the tendon gro shall be considered as accept-able. If the measured prest ssing force of any two tendons falls below 95% of the pres ibed lower limits of the tendons, additional lift-off testing shall be done to detect the cause and extent of such occurre ce. The condition shall be con-sidered as an indication f abnormal degradation of the contain-ment structure,

3. If the measured prestre sing force of any tendon lies below 90%

of the prescribed lowe limit, the defective tendon shall be completely detensione and additional lift-off testing shall be done so as to determ e the cause and extent of such occurrence. The co ition shall be considered as an indication .

of abnormal degrada ion of the containment structure, '

4. If the average of all measured prestressing forces for each group (corrected or average condition) is found to be less than the minim required prestress level at the anchorage locations for t at group, the condition shall be considered as abnormal degra ation of the containment structure,

5. If from cons utive surveillances the measured prestressing forces for e same tendon or tendons in a group indicate a trend of pr stress loss larger than expected and the resulting '

prestressi g forces will be-less than the minimum required for !

the group efore the next scheduled surveillance, additional ,

lift-off esting shall be done so as to determine the cause and ,1 extent such occurrence. The condition shall be considered

as an i dication of abnormal degradation of the containment struct re, and

6. Unle s there is abnormal' degradation of the containment vessel dur' g the first three inspections, the sample population for su equent inspections shall include at least 6 tendons (3 ho p, 3 inverted U).

WOLF CREEK - U T1 3/4 6-9 Amendment No. 31

CONTAINMENT SYSTEMS 9 SURVEILLANCE REQUIREMENTS (Continued)

I

b. Perfonning tendon detensioning, inspections, and material. tests -on a previously stressed tendon from each group. A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and date ining that over the entire length of the removed wire sample (which shall include the brokee wire if so identified) that:

1. The tendon wires are free of cor sion, cracks, and damage, and

2. A minimum tensile strength of 0 ksi (guaranteed ultimate '

strength of the tendon satori ) exists for at least-three wire samples-(one from each and a one at mid-length) cut fron'each removed wire. -

Failure to mest the requiremen of 4.6.1.6.1.b shall be considered as an indication of abnormal gradation of the containment structure.

c. Performing tendon retension" g of those te'ndons detensioned for.

inspection to at least the 'orce level recorded prior to detensioning-or the predicted value, w chever is-greater, with the tolerance .

within minus zero to plu 6%, but not to exceed 70% of the guaranteed-ultimate tensile.streng of the tendons. During retensioning of.

these tendons-the cha s in load and elongation shall be seasured simultaneously at a a teum of three approximately equally spaced '

levels of force bet n zero and the seating force. If_the.

elongation correspo ing to a specific: load differs by more than 10%

from that recorded uring the installation, an' investigation shall be made-to ensure hat the difference is not related to wire failures or slip f wires in anchorages. This condition shall_be considered as a indication of abnormal. degradation _of the ..

containment-str cture.

d. Verifying th OPERASILITY 'of the sheathing filler grease by assuring: ,

1. _ There re no changes in the presence or physical; appearance of the eathing filler grease including the presence of free.._ ,

wa ,

~

2. unt of grease replaced does not exceed 5% of the net duct v lume,.when injected at 2 10% of the.specified installation- i ressure,

3. Minimum grease coverage exists for the different parts of the '

anchorage system,

. During general visual examination of the containment exter -

surf ace, that grease leakage that could affect containment-integrity is not present, and WOLF. EEK - UNIT 1 3/4 6-10 Amendment No. 31 -

g- py- .l 3 - . n + _ un u _ - - --- ^

V CONTAINMENT SYSTEMS f

t SURVEILLANCE REQUIREMENTS (Continued)

I i S. The chemical properties of the filler ma rial are within the i ,

tolerance limits specified as follows:

Water Content 0 - 10% by d weight Chlorides 0 - 10 ppm Nitrates 0 - 10 ppm Sulfides 0 - 10 pp Reserved Alkalinity >0 Failure to meet the requirements of .6.1.6.1.d shall be considered as an indication of abnormal degrad tion of the containment structure.

4.6.1.6.2 End Anchorages and Adjacent Con rete Surfaces. As an assurance of the structural integrity of the contains t vessel, tendon anchorage assembly hardware (such as bearing plates, stres ng washers, wedges, and buttonheads) of all tendons selected for inspection hall be visually examined. Tendon anchorages selected for inspection sh I be visually examined to the extent

- practical without dismantling the lo $ bearing components of the anchorages.

Bottom grease caps of all vertical endons shall be visually inspected to detect grease leakage or grease ca deformations. The surrounding concrete shall also be checked visually fo indication of any abnormal condition. The frequency of this surveillance s all be in accordance with 4.6.1.6.1.

Significant grease leakage, gre se cap deformation or abnormal concrete condition shall be considered s an indication of abnormal degradation of the containment structure.

4.6.1.6.3 Containment Vess.1 Surfaces. The exterior surface of the containment shall be visu ly examined to detect areas of large spall, severe I

scaling, D-cracking in a area of 25 sq. ft. or more, other surface deterioration or disint ration, or grease leakage, each of which shall be' considered as evidence f abnormal degradation of structural integrity of the l

containment. This in action shall be performed prior to the Type A containment leakage te test.

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WO CREEK - UNIT 1 3/4 6-10a Amendment No. 31 l

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l CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 36-inch containment shutdown purge supply and exhaust isolation valve (s)* shall be verified blank flanged and closed at least once per 31 days.

4.6.1.7.2 Each 36-inch containment shutdown purge supply and exhaust isolation valve and its associated blank flange shall be leak tested at least once per 24 months and following each reinstallation of the blank flange when pressurizec to Pa, 48 psig, and verifying that when the measured leakage rate for these valves and flanges, including stem leakage, is added to the leakage rates deter-

- ific-ti;r minedpursuanttoEr$e;akacerata tions, the combined' i.C.1.2d., for 0.60 is less than all other Type B and C penetra-L^.

' Q$.fR S Ecirk IG. fo . l .1 4.6.1.7.3 The cumulative time that all 18-inch containment mini purge suoply and/or exhaust isolation valves have been open during a calendar year shall Oe determined at least once per 7 days.

4.6.1.7.4 At least once per 3 months each 18-inch containment mini-purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.05 Lawhen pressurized to P,.

"Except valves and flanges which are located inside containment. Tnese valves shall be verified to be closed with their blank flanges installeo criar ;a entry into MODE 4 following each COLD SHUTCOWN.

I ft 4

'.C'.F CREE < - UNIT 1

.- 3/4 6-12 u,

CONTAINMENT SYSTEMS 3/4.6.3' CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolat on valves specified in Table 3.6-1 shall be:

OPERABLE with isolation time as shown in Table 3.6-1. !-

APPLICABILITA MODES 1, 2, 3, and 4.

ACTION:

With one or more of the containment 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of'at 'least one closed manual valve or blind flange, or d.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS t

' 4.6.3.1 -The containment isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance,.

repair or replacement work is performed on the valve or its associated actuator,.

control or power circuit by performance of a cycling test, and verification of-isolation time.

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-*Fer "21 va t wi+h e-r:::i;: leakag;, refer te T.d.,,ical Sp::ificati;n 3;s.1.2. -

WOLF CREEK - UNIT 1 3/4 6-16 Amendment No. 33

TABLE 3.6-1 (Continued) -

CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)

9. Other Automatic Valves P-1 AB-HV-11*** Mn. Stm. Isol. A, N.A.

P-2 AB-HV-14*** Mn. Stm. Isol. A N.A.

P-3 AB-HV-17*** Mn. Stm. Isol. A N.A, P-4 AB-HV-20*** Mn. Stm. Isol. A N.A.

P-5 AE-FV-42*** Mn. FW Isol. A N. A.

P-6 AE-FV-39*** Mn. FW Isol. A N.A.

P-7 AE-FV-40*** Mn. FW Isol. A N.A.

P-8 AE-FV-41*** Mn. FW Isol. A N.A.

P-9 BM-HV-4** SG Blowdn. Isol. A 10 P-10 BM HV-1** SG Blowdn. Isol. A 10 P-11 BM-HV-2** SG Blowdn. Isol. A 10 P-12 BM-HV-3** SG Blowdn. Isol. A 10 "The provisions of Specification 3.0.4 are not applicable.

""*'hese valves are included for table completeness. The requirements of Scecification 3.6.3 ce not apply; instead, the requirer.ents of Scacifiestier

3. 7.1. 5, and Sceci fication 3. 3. 2 apoly to the Main Steam !sclation '.'a! .es a: :

Main Ffe -ater Isolation Valves,.re::c;;.<:17 3.u.7

%CLF CREEK - UNIT 1 3/4 6-30

l

stAINMENT SYSTEMS 3'4,f.4 C0!13UST!2LE GAS CONTROL wiCROGEN ANALYZERS

\'Ys LIMITING CONDITION FOR OPERATION

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3.6.4.1 Two independent containment hydrogen an yzers shall be OPERABLE.

APDLICABILITY: MODES 1 and 2.

ACTION:

a. With one containment hydrogen nalyzer inoperable, restore the inoperable analyzer to OPER LE status within 30 days or be in at' least HOT STAN0BY within t next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With both hydrogen anal ers inocerable, restore at least one analyzer to OPERABLE status wi in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 ho s.

SURVEILLANCE RE0VIRE .NTS .

/

4.6.4.1 Each c ntainment hydrogen analyzer shall be demonstrated OPERABLE by the performan of an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 31 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATI0t using sample gas containing ten volume percent hydrogen, balance nitrogen.

P WOLF CREEK - UNIT 1 3/4 6-31

1 PLANT SYSTEMS MAIN FEEDWATER SYSTEM !

LIMITING CONDITION FOR OPERATION .

3,7.1.7 Each main feedwater isolation valve (MFIV).shall be OPERABLE,

- APPLICABILITY; Modes 1, 2, and 3 ACTION:

. MODES 1 and 2: With one MFIV inoperable but open, operation may continue provided the inoperable valve is restored to OPERABLE status within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: otherwise, be in HOT STANDBY within the.next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODE 3: With one MFIV inoperable, subsequent operation in MODE 3 may proceed provided the isolation valve is. maintained closed.

J Otherwise, be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REOUIREMENTS .

f 4.7.1.7 Each MFIV shall:be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. The provisions' .

of sp(cification 4.0.4 are not applicable for entry into MODE 3. .l i

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WOLF CREEK UNIT 1 3/4.7-9b

PLANT SYSTEFS y

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION

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3.7.2 The temperatures of both the reactor and second y coolants in the steam generator shall be greater than 70*F when the p essure of either coolant in the steam generator is greater than 200 psig.

APPLICABILITY: At all times.

ACTION:

With the requirements of the above specifica on not satisfied:

a.

Reduce the steam generator press re of the applicable side to less than or equal to 200 psig withi 30 minutes, and

b. Perform an engineering evalu ion to determine the effect of the overpressurization on the s uctural integrity of the steam generator.

Determine that the steam nerator remains acceptable for continued operation prior to incre ing its temperatures above 200*F.

SURVEILLANCE REQUIREMENTS

/

4.7.2 The pressure in eac side of the steam generator shall be determined to be less than 200 psig at 1 ast once per hour when the temperature of either the reactor or secondary colant is less than 70*F.

/

WO!.F CREEK - UNIT 1 3/4 7-10 '

4 w 1 PLANT SYSTEMS- ]

3/4.7.8 SNUB 8ERS

. LIMITING CONDITION FOR OPERATION j

3.7.8 All snubbers'shall be OPERA 8LE. The only snubber excluded-from the. -

requirement are those installed on nonsafety-related s tems and then only if a their failure or failure of the system or, which they e installed would have -"

no adverse effect on any safety-related system. j APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 an 6 for snubbers' located on 1 systems required OPERA 8LE in those MODES. '

ACTION: f With one or more snubbers inoperable on any ystem, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore' the inoperable snubber (s) to OPERA 8LE stat and perform an engineering evaluation. peri Specification 4.7.8g on the attached comp ent or-declare the attached system inopera ~

ble and follow the appropriate ACTION st asent for that system.

~

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SURVEILLANCE REQUIREMENTS 4.7.8 Each snubber shall be d strated OPERABLE by performance.of the follow- ,

ing augmented inservice inspect n program and the requirements of Specifica- i tion 4.0.5.

a. Inspection Types

{

As used in this specification, type of snubber shall mean snubbers- :;

of 'the same de ign-and menufacturer, irrespective of capacity. 'I

b. Visual'Insp tions- ,

Snubbers re categorized as inaccessible or accessible during ,

, reactor ration.-'Each of these categoriet (inaccessible and .

access le) may be' inspected independently according to the' schedule 1 date ned _by Table 4.7-2.1The visuair inspection interval- for-each !

type snubber shall be determined based upon.the criteria provided :

in able 4.7-2 and the first inspection interval determined using . ,

s criteria shall be based upon the previous.-inspection interval. ,

i established by.the requirements in effect before amendment 44.

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- WOLF CREEK - UNIT 1 -3/4 7-19 Amendment No. 44 L

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PLANT SYSTEMS V ;

SURVEILLANCE REQUIREMENTS (Continued)

c. Visual Inspection Acceptance Criteria 'l. !

Visual. inspections shall verify that: '(1) there are no visible indi-cations of. damage or impaired OPERA 8ILITY,.and 2) attachments to the foundation or' supporting structure are functi al, and (3) fasteners ,

for attachment of' the snubber to the credpone and to the snubber anchorage are functional. Snubbers whichla ar inoperable as a !

result of visual inspections shall be clas ified as unacceptable and ,

may be reclassified acceptable for the p pose of. establishing.the next visual inspection interval, provid that: (1) the cause of the rejection is clearly established and r died.for that particular- 4 snubber:and for other snubbers irresp ctive of type that may be generically susceptible; or (2) the ffected snubber is functionally.

tested in the as-found condition a determined OPERA 8LE'per Speci-fication 4.7.Sf. All snubbers fo connected to'an-inoperable '

common hydraulic fluid reservoir shall be counted as unacceptable for determining the next inspec ion interval. A review and evaluation shall be' performed nd documented to determine system operability with an unaccept le snubber. If operability cannot be '

justified, the system shall e declared inoperable and the ACTION '

requirements shall be met.

d. Transient Event-Inspect ns An inspection shall. performed of all hydraulic and mechanical snub- .

bers attached to se ions of systems that have experienced unexpected, potentially damagi transients as determined from a review of opera-tional data and a isual inspection of the systems within 6 months.

following such a event. In addition to satisfying the visual inspection acc tance criteria, freedoe-of-motion of mechanical i snubbers shal' be verified using at least one~of the following: a

-(1)' manually nduced snubber movement; (2) evaluation of in place

~

snubber pis on setting; or (3) stroking the mechanical snubber through i full range of travel.

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-i WOLF CREEK - UNIT 1 3/4 7-20 Amendment No. 44

r '. W - i<3*E"3 Y

- T . E : c. 'CE :ET. : E'*ECS ( C:" # nue:)

e, ~ .: n c t #ctal Tests Cud g 19e first refueling shutdown and at least once per 13 months thereaf ter curing shutdown, a representat've sample of snuobers of each type shall be tested using one of t following_ sample plans.

The sample plan shall be selected prior o the test period and cannot be changed during the test period. The NRC Regional Admir.istrator shall be notified in writing of the sa ple plan selected for each snuccer type prior to the test period or the sample plan used in the crior test period shall be implement :

5) At least 10% of the total of e ch type of snubber shall be function-ally tested either in place o in a bench te:;t. For each snuober of a type that does not meet th functional test acceptance criteria cf Specification 4. 7.8f. , an a itional 10% of that type of snubbee ,

snall be functionally teste until no more failures are found or until all snubbers of that type have been functionally tested or

2) A representative sample f each type of snubber shall be func-tionally tested in acco ance with Figure 4.7-1. "C" is the total number of snubber of a type found not meeting the accep-tance requirements of peci fication 4. 7. 8f. The cumulative numcer of snubbers of a type tested is denoted by "N".

end of eag_h day's te ing, the new values of "N" and "C" At(pre- the vious day's total p s current day's increments) shall be plotted on the Figure 4.7- If at any time the point plotted falls in !

i the " Reject" regio , all snubbers of that type shall be function- i ally tested. If t any time the point plotted falls in the !

" Accept" region, testing of snubbers of that type may be terminated. Wh n the point plotted lies in the " Continue j Testing" regio , additional snubbers of that type shall be tested until e point falls in the " Accept" region or the

" Reject" regi n, or all the snubbers of that type have been tested; or t

3) An initial epresentative sample of 55 snubbers shall be func-tionally sted. For each snubber type which does not meet -;

the funct onal test acceptance criteria, another sample of at-least on -half the size of the initial sample shall be tested until t e total number tested is equal to the initial sample size m tiplied by the factor, 1 + C/2 where "C" is the number of sn bers found which do not meet the. functional test accept- '

ance riteria. The results from this sample plan shall be plot ed using an " Accept" line which follows the equation N =

55( + C/2). Each snubber point should be plotted as soon as th snubber is tested. If the point plotted falls on or.below t

" Accept" line, testing of that type of snuDber may be

rminated. If the point plotted falls above the " Accept" line, esting must continue until the point falls in the "Acceot" region or all tne snubbers of that type have been tested.

WOLF ; REEK - JNIT 1 3/4 7-21

1 NL/

i l

V l 9

PLANT SYSTEMS i SURVEILLANCE RE0VIREMENTS (Continued)

I

e. Functional Tests (Continued).

Testing equipment failure during functi nal testing may invalidate that day's testing and allow that day' testing to resume anew at a later time provided all snubbers test d with the failed eauipment during the day of equipment failure ,e retested. The represee.at'/e sample selected for the functional st sample plans shall be rancemly selected from the snubbers of each ype and-reviewed before ceginning the testing. The review shall ens re, as far as practicable, tnat they are representative of the va ious configurations, operating environments, range of size, and apacity of snubbers of each type.

Snubbers placed in the same loc ion as snubbers which failed tne previous functional test shall e retested at the time of the next functional test but shall not e included in the sample plan. !f during the functional testing additional sampling is requireo due to failure of only one type of ubber, the functional test results shall be reviewed at that ti e to determine if additional samo!es should be limited to the ty e of snubber which has failed the furc-tional testing.

f. Functionai Tert Acceptan Criteria The snubber functional est shall verify that:

1) Activation (rest aining action) is achieved within tre soeci'iec raftge in both t sion and compression;

2) Snubber bleed ate, or release rate where required, is prese.t in both tensi n and compression, within the specified range; and-

3) For mechani i snubbers, the force required to initiate or main-tain motio of the snubber is within the specified range in totn directions of travel.

Testing methe s may be-used to measure parameters indirectl;. 0-parameters o er than those specified if those results can ce correlated the specifie.d parameters through establisned mete:cs.

g. Service Li e Monitorinc Proc' ram An engin ring evcluation shall be made of each fail,.:re t: eet ta functic 1 test acceotance criteria to determine the ca se :f tre failure The results of this evaluation shall te use:, 'f 1::'i:1: 'e.

in sel cting snubbers to be tested in an effort to dete- me t e OPERA ILITY of other snubbers irresoective of ty;e '..ni cF -' :s subj .t to the same failure mcde.

Ah 3R 7-22 __ _ _ _ . _ . _ _ _ _

Y ,

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

I

g. Service Life Monitorina Program (Con inued)

For the snubbers found inoperable, an engineering evaluation shal_1 be performed on the components to hich the inoperable snubbers are ,

attached. The purpose of this e ineering evaluation shall be to determine if the components to w ich the inoperable snubbers are attached were adversely affecte by the inoperability'of the. snubbers in order to ensure that the co onent remains capable of meeting the-designed service.

If any snubber selected for unctional testing _either fails to lock up or fails to move, i e., frozen-in place, the cause will'be evaluated and, if caused b manufacturer or design deficiency, all snubbers of the same type subject to_the same defect shall be func- .

tionally tested. This ta ting requirement shall be independent of '

the requirements stated n Specification 4.7.8e. for' snubbers 'not meeting the functional est acceptance. criteria.- >

h. Functional Testina of epaired and Replaced Snubbers Snubbers which fail he visual inspection or.the functional test acceptance criteri shall be repaired or replaced. Replacement snubbers and-snubb rs which have repairs which might affect the functional test r suits shall be tested to meet the. functional test criteria before stallation in the unit. Mechanical snubbers shall have met the ac ptance criteria subsequent'to their most recent service, and t freedom-of-motion test must have been performed within-12 mont s before being_ installed in the unit. 1

1. Snubber Serv ce Life Program The servic life of hydraulic and mechanical snubbers shall be moni-tored to sure that the service life is not exceeded between sur-veillance inspections. The maximum expected service life for various-seals, s rings, 'and other critical parts.shall be determined and establi ud based on engineering information and shall be extended-or.she ned based on monitored test results and failure history.

Criti i parts shall be replaced so that the maximum service life willi not exceeded during a period when the: snubber .is required.to be-OPE LE. The parts replacements shall be documented'and the docu-1 ,

men tion shall be retained in accordance with Specification 6.10.2; I

WOLF CREEK - UNIT 1 3/4 7-23 l

+ . -

TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABL SNUBBERS Population Column A Column B Column C per Catergory Extend Interval Repeat Inte al Reduce Interval (Notes 1 and.2) (Extend 3 and 6) (Notes 4 a 6) (Notes 5 and 6) 1 0 0 1 80 0 0 2 100 0 1 4 l

150 0 3 8 200 2 5 13 300 5 12 25 400 8 18 36 500 12 24 48 750 20 40 78 1000 or greater 29 56 109 ;

Note 1: The next visual i spection interval for a snubber category shall- be determined based pon the previous inspection interval and the H number of unace table snubbers found during that interval. !

Snubbers may categorized, based upon their_ accessibility during: i power operati , as accessible or inaccessible. These categories- 1 may be.exami separately or jointly. However, categories must be detemined ' documented before any inspect 1on and that~

deteminati n shalJ.be the basis upon which to detemine the next.

inspection interval.for that category, q Note 2: Interpol ion between population per category a'nd the number of.

unaccep le snubbers is permissible. Use next lower integer for 1 the va of the limit for Columns A, B, and C if that integer _ ~l inclu s a fractional value of unacceptable snubbers as _ determined by i erpolation.

1 WOLF CRE - UNIT 1 3/4 7-24 Amendment No. 44

i

\

9 Table 4.7-2 (Contin d)

SNU88ER VISUAL INSPECTIOY INTERVAL Note 3: If the number of unacceptable snub ers is equal to or less than the number in Column A, the next insp ction' interval may be twice the previous interval but not greate than 48 months.

Note 4: If.the number of. unacceptable ubbers is equal ~to or.less'than the number-in Column B but greater than the. number in Column A, the next.

inspection interval shall be he same as~the previous interval.

Note 5: If the number of unacceptab e snubbers is equal to. or greater than the number in Column C, th next inspection interval shall'be two-thirds of the previout interval. ' However, . i f . the' number of unacceptable snubbers is ess:than the number in Column C:but-greater than.the number n Column 8, the next interval shall be ?

reduced proportionally y interpolation, that 'is, the previous '

interval shall be red ed by a factor that is'one-third of the ratio of-the differen a between the number of unacceptable snubbers' a found during the pre ous interval and the number in Column 8 to the difference in the n ers in Column 8 and C. i Note 6: The provisions of pecification 4.0.2 are applicable for all' inspection interv is up to and including 48 months.

i i

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i WOLF CREEK:- UNIT 1- 3/4 7-251 Anandment' No. 44

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0 10 20 30 40 50 60 70 80 90- 100 N <

FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUB 8ER FUNCTIONAL TEST WOLF CREEK - UNIT 1 3/4 7-26 Amendment No. 44

PLANT SYSTEMS 'N' 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 1

3.7.9 Each sealed source containing radioacti material either in excess of 100 microcuries of beta and/or gama-emitting aterial or 5 microcuries of alpha emitting material shall be free of greater th n or equal to 0.005 microcurie-of removable contamination.

APPLICABILITY: At all times.

ACTION:

a. With a sealed source having r ovable contamination in excess of the above limits, isusediately wi draw the sealed source from use and either:

1. Decontaminate and rep ir the sealed source, or

2. Dispose of the seal source in accordance with Comission Regulations,

b. The provisions of Spec fications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS I

4.7.9.1 Test Requirements Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, r

b. Other person specifically authorized by the Commission or an Agreement S te.

The test method shal have a detection sensitivity of at least 0.005 microcurie per test sample..

4.7.9.2 Test F ncies - Each category of sealed sources (excluding startup sources fission detectors previously subjected to core flux) shall be tJsted at the frwquency described below. -

a. Sour s in use - At least once par 6 months for all sealed sources con ining radioactive materials:

1) With a half-life greater than 30 days (excluding Hydrogen 3),

and

) In any form other than gas.

WOLF.CR EK - UNIT 1 3/4 7-27 Amendment No. 44 -

l PLANT SYSTEMS \ _

SURVEILLANCE REQUIREMENTS (Continued) 9 I

b. Stored sources not in use - Each sealed ource and fission detector i shall be tested prior to use or transfe to another licensee'unless i tested within the previous 6 months, taled sources and fission i detectors transferred without a certi icate indicating the last test l date shall be tested prior to being aced into use; and

c. 5tartup sources and fission detect s - Each sealed startup source and fission detector shall be tes d within 31 days' prior to being subjected to core flux or install d in the core and following repair or maintenance to the source.

4.7.9.3 Reports - A report shall be pr ared and submitted to the Comission on an annual basis if sealed source or ission detector leakage tests. reveal the presance of greater than or equal o 0.005 microcurie of removable contamination.

WOLF CREEK - UNIT 1 3/4 7-28 heendment No. 44

s.

iI PLANT SYSTEMS \

3/4.7.12 AREA TEMPERATURE MONITORING <( ,

LIMITING CONDITION FOR OPERATION I

3.7.12 The temperature limit of each. area given in Table 3.7-4 shall not be exceeded for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or by more than 30'F.

APPLICA81LITY: Whenever the equipment an affected area is required to be OPERABLE. ;

ACTION:

a. With one or more areas ex eding the temperature limit (s) shown in Table 3.7-4 for more tha 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare and submit to the .

Commission within 30 day , pursuant to Specification 6.9.2, a'Special. !

Report that provides a ecord of the cumulative time and the amount by which the temperatu e in the affected area (s) exceeded the limit (s) !

and an analysis to d nstrate the continued OPERASILITY of the-affected equipment, he provisions of Specifications 3.0.3 and 3.0.4 are not appli ble. ,

b. With one or more a eas exceeding the temperature' limit (s) shown in Table 3.7-4 by no e than 30*F, prepare and submit a-Special Report ,

as required by A ION a. above, and within.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the-area (s) to with the temperature limit (s) or declare the equipment in the affected ar a(s) inoperable.

SURVEILLANCE REQUIREME S ,

4.7.12 The tempera'ure in each of the areas shown in Table 3.7-4 shall be ,

determined to be w Lhin its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

J i

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1 WOLF CREEK - UNIT 1 3/4 7-29 Amendment No.'II, 44' I

.. . - - . . . . ~ . .- .

TABLE 3.7-4

\t ,

AREA TEMPERATURE MONITORING

% l MAXIMU TEMPE TURE AREA LIMI (*F) -

1. ESW Pump Room A 11

2. ESW Pump Room B 9

3. Auxiliary Feedwater Pump Room A 119

4. Auxiliary Feedwater Pump Room B 119

5. Turbine Driven Auxiliary Feedwater Pump R m 147-

6. ESF Switchgear Room I 87

7. ESF Switchgear Room II 67 ,

8. RHR Pump Room A 119

9. RHR Pump Room B 119

10. CTMT Spray Pump Room A 119

11. CTMT Spray Pump Room B 119

12. Safety Injection Pump Room' 119

13. Safety Injection Pump Ro B 119

14. Centrifugal Charging P Room A 119

15. Centrifugal Charging ump Room B 119

16. Electrical Penetra on Room A 101

17. Electrical Penet ation Room B 101

18. Component Cool ng Water Room A 119

19. Component ling Water Room B 119

20. Diesel Ge rator Room A 119

21. Diesel nerator Roos 8 119

22. Contr Roos 84 WOLF CREEK 'JNIT 1 3/4 7-30 . Amendment No.15, 44 1

l

W ELECTRICAL POWER SYSTEMS

\

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEV ES LIMITING CONDITION FOR OPERATION

\

3.8,4.1 For each containment penetration provided wi a penetration conductor ,

overcurrent protectivi device (s), each device shall OPERABLE. !

1 APPLICABILITY: HODES 1, 2, 3, and 4. i l i

! ACTION:

With one or more of the above' required contai nt penetration conductor overcurrent protective device (s) inoperable: !

a. Restore the protective device (s) o OPERABLE status or deenergize 4 the circuit (s) by tripping the a sociated backup circuit breaker i or racking out or removing the noperable circuit breaker within '

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the backup circuit reaker to be tripped or the inoperable circuit breaker racked out, r removed, at least once per 7 days thereafter; the provisions f Specification 3.0.4 are not applicable to overcurrent devices in ircuits which have their backup circuit breakers tripped, their i operable circuit breakers racked out, or removed, or 9

b. Be in at least HOT STA OBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the f llowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS I

4.8.4.1 Protective devi es required to be OPERABLE as containment penetration conductor overcurrent p otective devices shall be demonstrated OPERABLE.

a. At least on e per 18 months:

1) By v rifying that the 13.8 kV circuit breakers are OPERABLE by sel eting, on a rotating basis, at least 10% of the circuit br akers, and performing the following:

) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each ,

relay and associated circuit breakers and control circuits function as des'igned, and l!l WOLF EEK - UNIT 1 3/4 8-16 Amendment No. 28,39 1

. . .. . ~,. . . .

$ (

a

. \h ELECTRICAL' POWER 5/ STEMS SURVEILLANCE _REQUIREM NTS (Continued) c) . For each circuit breaker fou inoperable during these.

functional tests, an additi al representative sample lof at least 105 ef=all the cir uit breakers of the inoperable type shall also be functio ally tested until no more-

"?tiures are found or all circuit breakers of that' type have been functionally.t sted.

2) By selecting and functional testing a representative sample of at least 10% of each t of lower _ voltage circuit b,reakers.

Circuit breakers selected or functional. testing shall be selected on a rotating b is. Testing.of these circuit breakers shall consist of inject a current in. excess of.the breakers nominal Setpointrand suring the response time. The measured response time will be onpared to the manufacturer's' data to ensure that.it is le than or equal to a value specified by the manufacturer. Cire t breakers found inoperable during functional.

testing shall be re tored to OPERA 8LE' status ' prior to resuming operation. .For sa h circuit breaker found inoperable during these functional tests, an additional- representative ' sample of at ' '

least 10% of all the circuit breakers of.the inoperable.t shall also be f nctionally tested until no more failures. ype are found or all'c uit breakers of that; type have been functionally tested.

b. At least once pe 60 months by subjecting each circuit breaker to an insoection and ; eventive maintenance in accordance with procedures pre 7ared in co unction with its manufacturer's recommendations.

't WOLF CREEK - UNIT 1 - 3/4 6-17' Amendment No. 5

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9

E:.E.:'.3 ::E:A'::NS

,I_. i . 5 ~ : * ".' *, : :A*:CN3

_:": :N3 ::N:: ::N 00R OPERATICN 1

3.c.5

erso.nnel Direct at thecommunications shall be main ained between the centrol room arc refueling station.

':::E!.:Tv:

During CORE ALTERATIONS

. v. ..

'l en -irect :ommunications between he control room and personnel at the efueling station cannot be maint ned, suspend all CORE ALTERATIONS.

a SURVEILLANCE REQUIREMENTS 4.9.5 Direct communic tions between the control room and personnel at the.

refueling station sha i be demonstrated within I hour prior to the start of and at least on,u pe 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

WOLF CREEK - UNIT 1 3/4 9-5

V REFUELING OPERATIONS _

3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 1

3.9.6 The refueling machine shall be used for movem t of drive rods.or fuel assemblies and shall be OPERABLE with:

a. The refueling machine used for movement of fuel assemolies naving: !

1) A minimum capacity of 4800 pound ,

2) Automatic overload cutoffs wit the following Setpoints:

a) Primary - less than or qual to 250 pounds above the I indicated suspended we ght for wet conditions and less

' than or equal to 350 ounds above the indicated suspencec weight for dry condi ions, and b) Secondary - less an or equal'to 150 pounds above tne primary overload utoff.

3) An automatic load r uction trip with a Setpoint of less tra, or equal to 250 po ds below the suspended weight for we; con- j ditions or dry co itions. ;

4

b. The auxiliary hoist sed for latching and unlatching drive rods anc thimble plug handli g operations having:

{

I

1) A minimum ca acity of 3000 pounds, and

2) A 1000 pou d load indicator which shall be used to moni;ar lif ting I ads for these operation.

APPtICABILITY: Durin movement of drive rods or fuel assemblies witni, :ne reactor vessel.

ACTION:

With the require nts for refueling machine and/or auxiliary hoist OPERABILITY not atisified, suspend use of any inocerable refueling machine crane d/or auxiliary noist from operations involving tro movecent of d ive rods and fuel assemblies witnin the reactor -.es se' 5 J.'E I __A' C E RECUI:Et'ENT 5

.3.i.. .e ef + ; s:..ine ;ec i:r :.e 3 : _e' n 3 e : e. -- :

eac;;r .essei sna': :e ce r strated CPE?AE.E wi:-;n 1.: ;. 3 .'r :-

.C_: C:EE< .N:T'. 3/4 9-5

4 ^,

9 REFUELING OPERATIONS SURVEILLANCE REOUIREMENTS (Continued) to the movement of fuel assemblies in the reacto vessel by performing a load : l.'

test of at least 125% of the secondary automati overload cutoff and demonstra-ting an automatic load . cutoff when the refueli machine load exceeds the Setpoints of Specification 3.9.6a.2) and by nstrating an automatic load reduction trip when the load reduction exce s the Setpoint of Specification 3.9.6a.3).

4.9.6.2 Each auxiliary hoist and associ ed load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the movement of dri e rods within the reactor vessel by' l:

performing a load test of at least 1 0 pounds.

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I WOLF CREEK - UNIT 1 3/4_9-7 Amendment No 3 * -

9 ,

REFUELING OPERATIONS l 1

i 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FAC ITY LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2250 pounds s 11 be prohibited from travel over fuel assemblies in the. spent fuel stor ge facility.

APPLICABILITY: With fuel assemblies in the spent fuel storage facility.

ACTION:

a. With the requirements f the above specification not satisfied, place the crane' load n a safe condition.

b. The provisions of ecifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREME 5 4.9.7 Crane inter eks and physical stops which prevent crane travel with

.: loads in excess o 2250 pounds over fuel assemblies shall be demonstrated OPERABLE within days prior to crane use and at least once per 7 days thereafter duri g crane operation.

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1 WOLF CREEK - UNIT 1 3/4 9-8

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3. 9. '.D . 2 At least 23 feet of water shall be mai tained over the toD of'the ir-sciated fuel assemolies within the reactor p essure vessel.

A3:*:~;EILITY:

Durir.g movement of control r s within the reactor cressure sessei wr.i e ir, MODE 6.

?." TION :

viith the requirerents of the above spe ficatioc not satisfied, suspend all Ocera-ions involving covement of contr 1 rods witcin the pressure vessel.

SUE.'EILLAN"E REOUIREMENTS t%>

4.9.10.2 The water levej hall be determined to be at least its minimum required depth within 2 ours prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter dur'ng movement of control rods within the reactor vessel.

W1..: CREE <,- UNIT 1 3/4 9-13

V 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00VN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Sp ification 3.1.L 1 may be suspended for measurement of control rod wor and SHUTDOW hARGIN provided reactivity equivalent to at least the highe estimated contr:1 rod worth is available for trip insertion from OPERABLE control rod (s).

APPLICABILITY: MODE 2.

ACTION:

a. With any full-length con ol rod not fully inserted and with less than the above reactivi equivalent a_vailable for trip insertion, imediately initiate a d continue boration at greater than or equal to 30 gpm of a soluti n containing greater than or equal to 7000 ppm boron or its equival nt until the SHUTDOWN MARGIN required by Specification 3.1. 1 is restored.,

b. With all full-le th control rods fully inserted and the reactor suberitical by ess than the above reactivity equivalent, imediately initiate and c ntinue boration at greater than or equal to 30 gpm of a solution con ining greater than or equal to 7000 ppm boron or its equivalent til the SHUTOOWN MARGIN required by Specification 3.1.1.1 is restore .

SURVEILLANCEREQU)EMENTS 4.10.1.1 The osition of each full-length control rod either partially or fully withdrawn sh 11 be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 ach full-length control rod not fully inserted shall be demonstrated capable full insertion when tripped from at least the 50% withdrawn position within hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Spe fic4 tion 3.1.1.1.

WOLF CREEK - UNIT 1 3/4 10-1

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. ' 3. 5 CSI : N ?!:2!:AT:CN SYSTEM - ssuTDOWN

.:": :N3 CCN07~:0N c09 CPEr.ATION I

3.10.5 The limitations of Specification 3.1.3. may be suspended during the cerformance of individual full-length shutdow and control rod drop time measurements provided only one shutdown or c trol bank is withdrawn from the fully inserted Desition at a time.

A:DL::AEILITY: MODES 3, 4, and 5 during rformance of rod drop time measurements and during surveillance of gital rod position indicators for OPERASILITY.

ACTION:

With the Position Indication System noperable or with more than one bank of rods withdrawn, immediately open t Reactor trip breakers.

SURVEILLANCE REOUIREMENTS I

4.10.5 The above required P sition Indication Systems shall-be determined to be OPERABLE within 24 ho rs prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by verifying the Demand Position Indication Syste and the Digital Rod Position Indication System agree:

a. Within 12 ste s when the rods are stationary, and

b. Within 24 s eps during rod motion.

KOLF CREEK - UNIT 1 3/4 10-5

3/4.11 RADIDACTIVE EFFLUENTS !

y 3/4.11.1 L1001D EFFLUENTS LIOUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material c , tained in each of the' following unprotected outdoor tanks shall be limited to 1 s than or equal to 150 Curies, excluding tritium and dissolved or entrained n le gases, i

a. Reactor Makeup Water Storage Tank,

b. Refueling Water Storage Tank,

c. Condensate Storage Tank, and I

d. Outside temporary tanks, excludi domineralizer vessels and-liners-being used to solidify radioact a wastes.

R APPLICABILITX: At all times.

ACTION:

a. With the quantity of radio ctive material in any of the above listed tanks exceeding the above imit, immediately suspend all additions of radioactive material to P e tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank l contents to within the .mit, and describe the events leading to this- ..

I condition in'the next nual Radioactive Effluent Release Report, 'l pursuant to Specificat on 6.9.1.7.

b. The provisions of Sp cifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be d ermined to be within-the above limit by analyzing a-representative sample o the tank's contents at least once per 7-days when radioactive meterials re being added and within 7 days following any addition of.

radioactive material o the tank.

~3/4 11-1 . Amendment No. 44, 65 WOLF CREEK - UNIT 1 _ _ _ ___ _

i X

RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION I

3.11.2.5 The concentration of oxygen in the WASTE GAS HOLOUP SYSTEM shall be limited to less than or equal to 3% by vol whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of ygen in the WASTE GAS HOLOUP SYSTEM greater than 3% by volume ut less than or equal to 4% by volume, recute the oxygen concen ation to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

b. With the concentration f oxygen in the WASTE GAS HOLOUP SYSTEM greater than 4% by vol me and the hydrogen concentration greater than 4% by volume, i ediately suspend all additions of waste gases to the system and r uce the concentration of oxygen to less than or equal to 4% by vol e, then take ACTION a. above,

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENT 4,11.2.5 The concent tions of hydrogen and oxygen in the WASTE GAS HOLOUP SYSTEM shall be dete ined to be within the above limits by continuously monitoring the wast gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitor required OPERABLE by Table 3.3-13 of Specification 3.3.3.11.

4 WOLF CREEK - UNIT 1 3/4 11-2 Amendment No. 42 tief g ! SD '

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RADI0 ACTIVE EFFLUENTS T

I l .QAS STORAGE TANKS l

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LIMITING CONDITION FOR OPERATION l

l t L 3.11.2.6 Thequantityofradioactivitycontginedinea gas storage tank shall l be limited to less than or 'aqual to 2.5 x 10 Curies o noble gases (considered as i Xe-133 equivalent). J t

APPLICABILITY: At all times, i ACTION:

a. With the quantity of radioactive ma rial in any gas storage tank exceeding the above limit, immedia ly suspend all additions of radioactive material to the tank, nd within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the limit, an describe the events leading to this condition in the next Annual Ra ioactive Effluent Release Report, pursuant to Specification 6.9. 7.

l

b. The provisions of Specifica ons 3.0.3 and 3.0.4 are not applicable.

SURVElllANCE RE0VIREMENTS !

4.11.2.6 The quantity of radi active material contained in each gas storage tank shall be determined to be wi it the above limit at least once per 7 days when radioactive materials are b ng added and within 7 days following any addition of radioactive material to th tank.

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l WOLF CREEK - UNIT 1 3/4 11-3 Amendment No. 44,65 j

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any MEM8ER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

f.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) repre-sentative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and l radionuclides in the environment in accordance with the methodology and parameters in the ODCM.

2) A Land Use Census to ensure that changes in the use of areas at !

and beyond the SITE BOUNDARY are identified and the modifications to the monitoring program are made if required by the results of this census, and i

3) Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the pMp [p,S-(l ,

measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program

h for environmental monitoring.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable repoYting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.

WOLF CREEK UNIT 1 6-18 Amendment No. 42

1 1

j INSERT 6.8.5 The following programs, relocated from the Technical Specifications to USAR .

Chapter 16, shall be implemented and maintained:

a. Explosive Gas and Storage Tank Radioactivity Monitoring Program

b. Turbine Overspeed Protection Reliability Program

c. Steam Generator Tube Surveillance Program

d. Reactor Coolant Pump Flywheel Inspection Program

e. Snubber-Inspection Program

f. Area Temperature Monitoring Program

g. Primary Water Chendstry Program

h. Containment Tendon Surveillance Program I

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I REACTIVITY CONTROL SYSTEMS 1

BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative HTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting HTC End of Life (EOL) value specified in the CORE OPERATING LIMITS REPORT (COLR). The 300 ppm surveillance limit MTC value specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium baron concentration and is obtained by making these corrections to the limiting EOL MTC value specified in the COLR.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F. This 3

limitation is required to ensure: (1) the moderator temperature coefficient is

\ within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an pF OPERABLE status with a steam bubble, and (4) the reactor vessel is above its l' minimum RT m temperature.

g 3/4.1.2 GUR;J1G LARS - ~

h oration Systems ensures that negative reactivity control is available each mode of facility operation. The component quired to perform this fun on include: (I) borated water sources, (2 ntrifugal charging pumps, (3) rate flow paths, (4) boric acid nsfer pumps, and (5) an emergency power su from OPERABLE diesel rators.

With the RCS average tempe re equal r greater than 350*F a minimum of two boron injection flow paths ar red to ensure single functional capability in the event an assumed ur enders one of the flow paths inoper-able. The boration capability either flow h is sufficient to provide a SHUTDOWN MARGIN from expe operating conditions 1.3% Ak/k after xenon decay and cooldown to F. The maximum expected bor n capability require- ,

ment occurs at E rom full power equilibrium xenon condi and requires '

17,658 gallo f 7000 ppm borated water from the boric acid st e tanks or l 83,754 ons of 2400 ppm borated water from the RWST. With the R verage ature less than 350*, only one boron injection flow path is require l tem WOLF CREEK - UNIT 1 B 3/4 1-2 Amendment No. 23, 61

REACTIVITY CONTROL SYSTEMS 8ASES NORATIONSYSTEMS(Continued) ith the RCS temperature below 200*F, one Boration System is acceptable withou single failure consideration on the basis of the stable reactivity on-dition o the reactor and the additional restrictions prohibiting CORE A ERA-TIONS and itive reactivity changes in the event the single Boron I ction System become inoperable.

The limitati for a maximum of one centrifugal charging p p to be OPERABLE and the Su illance Requirement to verify all char g pumps except the required OPERABLE mp to be inoperable in MODES 4, 5, nd 6 provides assur-ance that a mass additio ressure transient can be reli ed by the operation of a single PORV or an RHR su ion relief valve.

The boron capability req ed below 200*F is ficient to provide a SHUTDOWN MARGIN of 1.3% Ak/k aft xenoa decay a cooldown from 200*F to 140'F. .

This condition requires either 29 allons of 000 ppm borated wa.ter from the boric acid storage tanks or 14,071 g lons o 2400 ppm borated water from the RWST.

The contained water volume limits ne de allowance for water not available because of discharge line location a other hysical characteristics. In the case of the boric acid tanks, all the conta ed volume is considered usable.

The required usable volume may contained in e er or bot 5 of the boric acid tanks.

The limits on contain water volume and boron co entration of the RWST also ensure a pH value between 8.5 and 11.0 for the s tion' recirculated within containment af r a LOCA. This pH band minimizes th evolution of iodine and minimizes the e ect of chloride and caustic stress corro on on mechanical.

systems and compo nts.

The OPE ILITY of one Boration System during REFUELING ensures hat this system is ailable for reactivity control while in MODE 6.

en determining compliance with action statement requirements, addit to the S of borated water with a concentration greater than or equal to the m imum required RWST concentration shall not be considered to be a positive eactivity change.

3/4.1.3 MOVABLE CONTROL AS.SJfiBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at' 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210 and 228 WOLF CREEK - UNIT 1 B 3/4 1-3 Amendment No. M ,61

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fewrter 22,1993

INSERT B1 3/4.1.1.5 CORE REACTIVITY The core is consid eed to be operating within acceptable design limits when measured core react.ivity is within il% Ak/k of the predicted value at steady state thermal conditions. Deviations from the design limit are i normally detected by comparing predicted and measured steady state RCS critical boron concentrations. The difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the design limit is reached. These values are well within the uncertainty.

limits for analysis of boron concentration samples. Therefore,- spurious violations of the design limit due to uncertainty in measuring the RCS boron concentration are unlikely.

The acceptance criteria for core reactivity . (11% Ak/k of the predicted !

value) ensures plant operation is maintained within the assumptions of - the safety analyses.

Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Therefore, every accident evaluation is dependent upon accurate evaluation of core reactivity. SDM and reactivity transients, such as control rod withdrawal accidents or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.

Design calculations and saiety analyses are performed for each fuel cycle. These are used to predetermine reactivity behavior and RCS boron concentration requirements for reactivity control during fuel depletion.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core ,

reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of life (BOL) do not agree, then-the l assumptions used in the reload cycle design analysis or_ requirements may not -4 be accurate. If reasonable agreement between measured and predicted . core i reactivity exists at BOL, then the prediction may be normalized to- the l measured boron concentration. Thereafter, any significant deviations in the !

measured boron concentration from the predicted boron letdown curve that i develop during fuel depletion may be an indication that the calculational !

model is not adequate for core burnups beyond BOL, or that an unexpected change in core conditions has occurred.

The normalization of predicted RCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the. control rods in their normal positions for power l operation. The normalization is performed at BOL conditions, so that ' core reactivity relative to predicted values can be continually monitored and evaluated'as core conditions change during the cycle.

Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core

~3 conditions. The required completion time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on the low

~ probability of a DBA occurring'during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.

Following evaluations of the core design and safety analysis, the cause' of the reactivity anomaly may be resolved. If the cause of the reactivity an'omal y is a mismatch in core conditions at the time of RCS boron i concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that_ core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause :

of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the' reactor core is. acceptable for-continued operation, then the boron letdown curve may be renormalized and power. operation may continue. If operational restriction or additional surveillance requirements are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.

The required completion time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is adequate for preparing whatever operating restrictions or surveillances that may.be required to allow continued operation.

I

REACTIVITY CONTROL SYSTEMS -

BASES HOVABLE CONTROL ASSEMBLIES (Continued) .l steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.

Since the Digital Rod Position System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges'are 4 picked for verification of agreement with demanded position. j l

For purposes of determining compliance with Specification 3.1.3.1, any ;

immovability of a control rod invokes ACTION Statement 3.1.3.1.a. Before utili- '

zing ACTION Statement 3.1.3.1.c, the rod control urgent failure alarm must be illuminated or an electrical problem must be detected in the rod control system. ,

The rod is considered trippable if the rod was demonstrated OPERABLE during the '

last performance of Surveillance Requirement 4.1.3.1.2 and met the rod drop time.

criteria of Sp=ifinti= 3.1.3.t during the last performance of Surveillance Requirement 4.1. Q 4,0 3,l.L The ACTION statements which permit limited variations from the basic' !

requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of ,

peaking factors and a restriction in THERMAL POWER. These restrictions provide '

assurance of fuel rod integrity during continued operation. In addition,: those safety analyses affected by a misaligned rod are reevaluated to confirm that the !

results remain valid during future operation.

The power reduction and shutdown time limits given in ACTION statements 1 3.1.3.2.a.2, 3.1.3.2.b.2, and 3.1.3.2.c.2, respectively, are initiated at the ;

time of discovery that the compensatory actions required for POWER OPERATION can no longer be met.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or equal to 551*F and with all reactor coolant' pumps operatin g ensures that.the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more fre-quent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable -

LCOs are satisfied.

WOLF CREEK - UNIT 1 B 3/4 1-4 Amendment No N ,46 tbwrt)er 22.1993 ,.

INSTRUMENTATION ,

BASES' 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated ACTION.will.be initiated when the radiation level monitored;by each channel or combination thereof. reaches its Setpoint, (2) the specified coincidence logic is maintained,' and (3)- -+

sufficient redundancy is maintained to permit a channel. to.be out-of-service for testing or maintenance. 'The radiation monitors for-plant operations senses radiation levels in selected plant systems'and locations and determines-whether or not predetermined limits are being exceeded. 'If the/ are, the ,

signals are combined into logic matrices sensitive to combinations: indicative 1 of various accidents and abnormal conditions. Once the required logic . 3 D

combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or . 's Control Room Emergency Ventilation Systems.

3/4.3.3.2 m MST@ ,

OPERASILITY of the movable.incore detectors with the specift minimum c nt of equipment' ensures that the measurements obt from l use of this sys urately represent the spatial neutro distribution of'the-core. .The OPE Y of this system is demo ed by irradiating. i each detector used and deters the accept y of its voltage curve.

For the purpose.of measuri , ,, - F "(X,Y) 'a full - incore - flux --

map is used. Quarter-cor maps, as defin P-8648, June 1976, may be used in recalibr of the Excore Neutron Flux De System, and full .,

incore flux r symmetric incore thimbles may be used for oring the. q QUAD R TILT RATIO when one Power Range Neutron Flux channel-rable.

CX*lETf 0 OPERA 8ILITY of the. seismic instrumentation ensures that suffi H capability able to promptly determine the magnitude o e event and evaluate the res those features.importa ety. This ~

capability is ~ required 'to pem ris . measured response to that-used in the design basis for the armine'if plant shutdown is. ;

required pursuant to of 10 CFR Part 1 . instrumentation is -)

consistent wi recommendations of Regulatory Guide 1. ,

trumentation i for akes," April 1974.

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WOLF CREEX - UNIT 1 B 3/4 3-4 Amendment No. 61 u .- ., ,-e , , _ v _,r. ,

6 s

INSTRUMENTATION BASES 4/4.3.3.4 DWTW !

ILITY of the meteorological instrumentation ensu I sufficient' mete 1 data-is aval able for esti ential radiation i doses to the public as a routine ntal re ease of .

radioactive materials to the ateo . capability is required to ,

evaluate the need for 1 protective sea rotect the health and safety of th and is consistent with the recomme a e ulatory ~ .

Su . , Onsite Meteorological Programs," February 1972. _

3/4.3.3.5 REMDTE SHUThnW INSTRtMENTATION

~

The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the facility from locations outside of the control room' and that a fire will >

not preclude achieving safe: shutdown.- The Remote Shutdown System transfer .

switches, power circuits, and control circuits'are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is required in the event control room habitability isLlost and is ;

consistent.with General Design Criteria 3 and 1.9 and Appendix R of 10 CFR Part

50. ,

1 3/4.3.3.6 ACCIDENT MDNITORING INSTRUMENTATION '

The OPERABILITY of the accident monitoring instrumentation ensures that: ,

sufficient-information is available on selected plant-parameters to monitor and' assess these variables _following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2,

" Instrumentation for Light-Water-Cooled Nuclear Power Plants ~ to Assess Plant Conditions During and Following an Accident," December 1980 and NUREG-0737, "

" Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.3.7 DELETED 3/4.3.3.8 DELETED b

9 h

WOLF CREEK - UNIT 1 B 3/4 3-5 Amendment No. M, 66~

-- , ,. ~ -- - ..a. -

7 INSTRUMENTATION BASES ^

t 3/4.3.3.9 40966-PART 00:C 20% I;372Et.TI@ $EIED sufficient capa ILITY of the loose part detection instrumentation e at Reactor Coolant System andvailable a to detect loose se itiget s in the components. The allowable out-o - Reactor Coolant System-are consistent with th Surveillance Requirements -)

Detection P ndationsofRejulato .133, " Loose-Part or the Primary Systees of L,ght-Water-Cooled " May 3/4.3.3.10- DELETED 3/4.3.3.11 DELETED 3/4.3.4 -TU 0!N: OV:::P:00 PROTCOTibN DC-lf]ED tection instrume ificationais provided to ensure' that the turbine ove aa :d pro ~~~~ 7 d the turbine speed control will protect the turbine frow ov PERABLE ano

. 1though the orientation of the turbine is such that th o damaging missiles which could impact and d minimal ety related components equ structures.is-on from excessive turbine overspe,ed is required.

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.1 WOLF CREEK - UNIT 1 B 3/4 3-6 Amendment No. 25, 42

3/4.4 REACTOR COOLANT SYSTEM 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 the safety analysis limit DNBR {1.32) during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within G hours.

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident; however, single failure considerations require that three loops be OPERABLE. A single reactor coolant loop provides sufficient heat removal if a bank withdrawal accident can be prevented; i.e., by opening the Reactor Trip System breakers.

In MODE 4, and in MODE 5 with t eactor coolant loops filled, a single reactor coolant loop or RHR loop pro vides sufficient. heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) te OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHa loop provides sufficient heat removsi capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing componenc, require that at least two RHR 1 cops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

Addition of borated water with a concentration greater than or equal to the minimum required RWST concentration but less than the actual RCS boron concentration shall not be considered a reduction in boron concentration.

The restrictions on starting a reactor coolant pump in MODES 4 and S are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam. Ne reHef'

ity of ; ;ir.gi

ftty valv; i; ;d quate te relieve my everprMs~ -

WOLF CREEK - UNIT 1 B 3/4 4-1 Amendment No. 51 teveter 22,1993

REACTOR COOLANT SYSTEM BASES SAFETY VALVES (Continued) l

- "ica which could occur during shutdown. In the ev=nt thet in.,

h1 L4r.JL" o erating RHR loop, connec , provides overpressure relief capab CS overpressurization. In addition, the y0vas o ection System pro means of nrMmT~ a6~a.nst RCS overnrossuri7ation at inw tomnoraturae - ~

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from ~a complete loss-of-load assuming no Reactor trip 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.

Addition to the RCS of borated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered a positive reactivity change. Cooldown of the RCS for restoration of operability of a pressurizer code safety valve, with a negative moderator temperature coeffi t ~ shall not be considered a positive reactivity change provided the RCS s bora ed to the COLD SHUTDOWN, xenon-free conditions per specification 3. 1./.

I 3/4.4.3 PRESSURI ER The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient opera- ,

tion. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of .

the plant to control Reactor Coolant System pressure and establish natural l circulation.

3/4.4.4 RELIEF VALVES ;

The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the 1 design step load decrease with steam dump. Operation of the PORVs minimizes ,

the undesirable opening of the spring-loaded pressurizer code safety valves. '

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. j The PORVs are equipped with automatic actuation circuitry and manual I control capability. Because no credit for automatic PORV operation is taken in the USAR analyses for MODE 1, 2 and 3 transients, the PORVs are considered OPERABLE in either the manual or automatic mode. The automatic mode is the preferred configuration, as this provides pressure relieving capability without reliance on operator action.

WOLF CREEK - UNIT 1 8 3/4 4-2 Amendment No. 63 tbwmber 22, 1993

REACTOR COOLANT SYSTEM BASES

}]4.4.5 STEA" OE C AT00 DM760 l The Surveillance Requirements for inspection of the steam generator tubes e ure that the structural integrity of this portion of the RCS will be main-tai d. The program for inservice inspection of steam generator tubes is base on a modification of Regulatory Guide 1.83, Ravision 1. Inservice inspec on of steam generator tubing is essential in order to maintain surveill ce of the conditions of the tubes in the event that there is evidence o mechanical damage or progressive degradation due to design manufacturi errors, or inservice conditions that lead to corrosion Inservice ins ction of steam generator tubing also provides a mea of characterizing he nature and cause of any tube degradation so t corrective measures can be ken.

Unscheduled in evice inspections are performed on eac steam generator following: 1) reacto to secondary tube leaks; 2) seismic ccurrence greater than the Operating Bas Earthquake; 3) a loss-of-coolan accident requiring actuation of the Enginee d Safety Features, which for is specification is defined to be a break grea r than that equivalent t he severance of-a 1" inside diameter pipe, or, fo a main steamline or f dline, a break greater than that equivalent to a ste generator safety lve failing open; to ensure that steam generator tubes reta sufficient in grity for continued operation. Transients less seve than these o not require inspections because the resulting stresses are ell with the stress criteria established by Regulatory Guide 1.121 which unp1 ged eam generator tubes must be capable of withstanding.

The plant is expected to be oper e in a manner such that the secondary coolant will be maintained within t se ch istry limits found to result in-negligible corrosion of the steam enerator bes. If the secondary coolant chemistry is not maintained with these limit localized corrosion may likely result in stress corros n cracking. Th extent of cracking during plant operation would be lim ed by the limitatio of steam generator tube leakage between the reactor oolant System and the econdary Coolant System (reactor-to-secondary lea ge - 500 gallons per day r steam generator).

Cracks having a reactor o-secondary leakage less than his limit during operation will have a adequate margin of safety to wit tand the loads imposed during norm operation and by postulated acciden . Operating plants have demonstrated at reactor-to-secondary leakage of 500 lions per day per steam generator n readily be detected by radiation monitor of steam l generator blo . Leakage in excess of this limit will requ e plant l shutdown and unscheduled inspection, during which the leaking tubes will be '

located and lugged.

Was age-type defects are unlikely with proper chemistry treatmen of the second y coolant. However, even if a defect should develop in servic it will e found during scheduled inservice steam generator tube examinatio .

P1 ging will be required for all tubes with imperfections exceeding the p) gging limit of 40% of the tube nominal wall thickness. Steam generator

/ubeinspectionsofoperatingplantshavedemonstratedthecapabilityto _

WOLF CREEK - UNIT 1 B 3/4 4-3 tbvetoer 22,1993

REACTOR COOLANT SYSTEM jASES

T = c = Tc = (Continu w) _

l re ' detect degradation that has penetrated 20% of the original . wall thickness.

Whenever the resu ny steam generat ng inservice inspection fall into Category C-3, these r w reported to the Commission pur-suant to Specification 6.9.2 pri r ion of plant operation. Such cases will be considered Commission on -by-case basis and may result in a requir or analysis, laboratory exam s, tests, addi-tional edd - ent inspection, and revision of the Technica fications, ifJ ary.

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 provided to monitor and detect leakage from the reactor coolant pressure bordary. These. Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of.an impending gross failure of the pressure boundary. There-fore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

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 I gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the I event of a main steam line rupture or under LOCA condtions.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

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WOLF CREEK - UNIT 1 B 3/4 4-4 teveter 22,1993

REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued) l The CONTROLLED LEAKAGE limitation restricts operation when the total flow from the reactor coolant pump seals exceeds 8 gpm per RC pump at a nominal RCS pressure of 2235 psig. This limitation ensures adequate performance of the RC pump seals.

The 1 gpm leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a sub-stantial length of time, verification of valve integrity is required. Since those valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7 C:l = ST Ri DOCET60 The limitations on Reactor Coolant System chemistry ensure that corros of the actor Coo.lant System is minimized and reduces the potential fo Reactor Co t System leakage or failure due to stress corrosion.

Maintaining th emistry within the Steady-State Limits provi adequate corrosion protecti o ensure the structural integrity of Reactor Coolant System over the life o e plant. The associated eff s of exceeding the-oxygen, chloride, and fluor limits are time an aperature dependent.

Corrosion studies show that oper on may be inued with contaminant concentration levels in excess of th t -State Limits, up to the Transient Limits, for the specified limited ti vals without having a significant effect on the structural integri of the Rea r Coolant System. The time interval pemitting continu peration within in estrictions of the Transient Limits provid ime for taking corrective ons to restore the contaminant concent ons to within the Steady-State limi The S tilance Requirements provide adequate assurance that cen-tratio n excess of the limits will be detected in sufficient time to ke c ctive action. _

3/4.4.8 SPECIFIC ACTIVITY, The limitations on the specific activity of the reactor coolant e are !

that the resulting 2-hour doses at the SITE BOUNDARY will not exceed ;

appropriately small fraction of 10 CFR Part 100 dose guideline values l following a steam generator tube rupture accident in conjunction with an WOLF CREEK - UNIT 1 B 3/4 4-5 tbveter 22,1993

i l

REACTOR COOLANT SYSTEM i

BASES l

PRESSURE / TEMPERATURE LIMITS (Continued)

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided telow.

h idary side of the steam generator must not be pr ed abc if the temperature of the steam or is below 70'r.

4. The pressurizer heatu col o es shall not exceed 100*F/h and 200*F/h, u ively. The spray sha . be used if the tem e difference between the pressurizer an ray fluid greater than 583*F.

J f. Syctem preservice hydrotests and in-service leak and hydrotests s be performed at pressures in accordance with the requirements at AE Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT at the end of 13.6 effective full power years (EFPY) of service life. Ne,13.6EFPYservicelife period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT o The selection of such a limitingassures ko,f the that limEing unirradiated all components material.

in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RT,o,; the. results ci these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT,o,. Therefore, an adjusted reference temperature, based upon the fluence and copper content and nickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART , computed by Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement l of Eeactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT end of 13.6 EFPY as well as adjustments for possible errors in the,o, at the pressure l and temperature sensing instruments.

WOLF CREEK - UNIT 1 B 3/4 4-7 Amendment No. 40,71

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Although the pressurf: r operate: in tempee:ture r:nge: :b:ve th::: for which there i: reason for : ncern of nonductil; failur;, sperating limits are prcvided to assur : mpatibility of operation with the fetigue analysis

-performed 'n :::crdane: with the ASME Code r quir : nt:.

The OPERABILITY of two PORVs, or two RHR suction relief valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 368'F.

Either PORV or either RHR suction relief valve 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 temperature of the steam generator less than or equal to 50*F-above the RCS cold leg tempera-tures, or (2) the start of a centrifugal charging pump and its injection into a water solid RCS.

I.

In addition to opening RCS vents to meet the requirement of Specifica-tion 3.4.9.3c., it is acceptable to remove a pressurizer Code safety valve, open a PORV block valve and remove power from the valve operator in conjunction with disassembly of a PORV and removal of its internals, or otherwise open the RCS.

WOLF CREEK - UNIT 1 B 3/4 4-13 Amendment No. 6, 49

.- . - = . - .- . .-- _ . - - - ,

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'Io BASES i COLD OVERPRESSURE (Coritinued)

RCP eliminates the possibility of a 50*F difference existing between indicated 0 and ' actual RCS temperature as a result of heat transport effects. Considering-instrument uncertainties only, an indicated RCS temperature.ofs350*F is.

sufficiently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditionst the. -

pressurizer safety valves provide acceptable and redundant overpressure '

protection.

c The Maximum Allowed PORY Setpoint for the Cold Overpressure: Mitigation:

System will be updated based on the results of examinations of reactor vessel ,

material irradiation surveillance specimens performed as required by 10 CFR j Part 50, Appendix H.

3/4.4.10 GTRUCTURAL INTECRITY OfMI6 7 1

inservice inspection and testing programs for ASME Code Clar !

and 3 comp s ensure that the structural integrity.and oper a..

readiness of thes onents will be' maintained at an able~ level '

throughout the life of ant. These program n accordance with.

Section'XI of the ASME Boiler assur el Code and applicable Addenda ,

as required by 10 CFR Part 50.55a where specific written relief has been granted by the Commiss rsuant to Part 50.55a(g)(6)(1).

Componen e Reactor Coolant System were design rovide access:

to pe service inspections in accordance with'Section XI.o ME ,

r and-Pressure Vessel Code, 1974 Edition and Addenda through Summer 3/4.4.li "EACT0" COOLANT 0"0 TEM YENTE D6LMcD actor Coolant System vents are provided to exhaust .noncondensibl ses and/or s from the Reactor. Coolant ' System that could . inhibit ' al-circulation co oling. The OPERABILITY'of'a reactor ves ead vent path ensures the capabili ists to perform this-functio s 1

The valve redundancy of th ctor nt: System vent paths serves)to-minimize.the probability of'inadver irreversible actuation while '

ensuring that a single failur - valve p ~ supply or control system does not p? event isolation of vent path.

lhe fun , capabilities, and testing requirements o . Reactor _-

Cool t.n en vents are consistent with the requirements of -

Item 1. of p' -0737, " Clarification of TMI Action Plan Re<viirements," . November 1 ,

J

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i WOLF CREEK - UNIT 1 B 3/4 4-15 Amendment No. 40, 57 l

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3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

juyAT*

bd'I 3/4.6.1.2 CONTAIN":NT LEAXAO: DEL 67FD The limitations on containment leakage rates ensure that the total cont ' ent leakage volume will not exceed the value assumed in the sa analyses the peak accident pressure, P,. As an added conserva , the measured overal tegrated leakage rate is further limite less than or equal to 0.75 L, or . Lt , as applicable, during per nance of the periodic test to account for possib radation of the ainment leakage barriers between leakage tests.

For reduced pressure tests, t ea characteristics yielded by measurements L g and L,, shall ablish the mum allowable test leakage rate Ltof not more tha , (Lg/L,,). In the event /L,,_is greater than 0.7, L shall be t cified as equal to L, (Pt/P,)1/2 Th rveillance testing for measuring leakage rates are consis with th quirements of Appendix J of 10 CFR Part 50.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are

, required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

WOLF CREEK - UNIT 1 B 3/4 6-1

~ . . _ - . _ . _ _ _ _ - . _ _ _ _ _ _ _ . _ _ . _ _____-- _ _ _ _ _ _ - = . . . _ _ _ . ..

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[ INSERT B6-1 1

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k' i Containment leakage rates shall be within the following. limits: ,

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1) An overall integrated leakage rate of less than or equal'to L a, 0.20% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at.P a, 48 psig- .

j' 2) A combined leakage rate of less than 0.60 La for all penetrations and valves subject to Type B and C tests, when pressurized to P a, 48 psig, ,

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i CONTAINMENT SYSTEMS BASES l

3/4.6.1.4 INTERNAL _ PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented frem exceeding its design negative pressure differential with respect to the outside atmosphere of 3.0 psig, and (2) the containment peak pressure does not exceed the design pressure of 60 psig during steam line break conditions.

The maximum peak pressure expected to be obtained from a steam line break event is 48.9 psig. The limit of 1.5 psig for initial positive containment pressure will limit the total pressure to 50.4 psig, which is less than design pressure and is consistent with the safety analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a steam line break accident. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature.

3/4.6.1.6 -CONTAIN" NT "CSSEL STRUCTURAL INTEGRITY WO ,

This limitation ensures that the structural integrity of the containmen will maintained in accordance with safety analysis requirements for t ife of the e lity. Structural integrity is required to ensure that th ntain-ment will wi tand the maximum pressure of 50.4 psig in the even a steam line break acci l

t. The measurement of containment tendon l' -off force, the tensile tests of th endon wires or strands, the visual e ination of tendons, anchorages an xposed interior and exterior aces of the contain-ment, and the Type A leaka test are sufficient to monstrate this capability.

The Surveillance Requirements fo nstrating the containment's structural integrity are in complian w the recommendations of proposed Regulatory Guide 1.35, "Inservic rveillan of Ungrouted Tendons-in Prestressed Concrete Containe Structures," A 1 1979, and proposed Regulatory Guide 1.35.1, " ermining Prestressing ces for Inspection of Prestressed Concrete C ainments," April 1979.

The requir Special Reports from any engineering evalua ' of containment normalities shall include a description of the ten condition, the con

1on of the concrete (especially at tendon anchorages), the ~ pection l pro re, the tolerance on cracking, the results of the engineering eva tion '

the corrective actions taken.

WOLF CREEK - UNIT 1 B 3/4 6-2 Amendment No. 50

l CONTAINMENT SY5TEMS BASES 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 36-inch containment purge supply and exhaust isolation valves are requirec to be closed and blank flanged during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Maintaining these valves closed and blank flanged during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 36-inch containment valves cannot be inadvertently opened, the valves are blank flanged.

The use of the containment mini purge lines is restricted to the 18-inch purgt supply and exhaust isolation valves since, unlike the 36-inch valves, the 18-inch valves are capable of closing during a LOCA or steam line break accident. There-fore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not be exceeded in the event of an accident during containment purging operation. Opera-tion will be limited to 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months during a calendar year. The total time the Containment Purge (vent) System isolation valves may be open during MODES 1, 2, 3, and 4 in a calendar year is a function of anticipated need and operating experience. Only safety-related reasons, e.g. , containment pressure control or the reduction of airborne radioactivity to facilitate personnel access for surveillance and maintenance activities, should be used to support the additional time requests. Only safety-related reasons should be used to justify the opening of these isolation valves during MODES 1. 2, 3 and 4, in any calendar year regardless of the allowable hours.

Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust "tpply valves will provide early indication of recilient material seal degradation and will allow opportunity for repair before gro;s leak-age failures could develop. The 0.60 L, leakage limi c' !;::i ficeL 3.o. l.2.o.

shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves' and penetrations subject to Type B and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures _that containment ;

depressurization and cooling capability will be available in the event of a l LOCA or steam line break. The pressure reduction and resultant lower containment l 1eakatn rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Cooling System are redundant to each other in providing post-accident cooling of the containment- .

I atmosphere. However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 8.5 and 11.0 for the WOLF CREEK - UNIT 1 B 3/4 6-3

BASES SPRAY ADDITIVE SYSTEM (Continued) solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. The educator flow test of 52 gpm with RWST water is equivalent to 40 gpm NaOH solution. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses. ,

3/4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Cooling System ensures that: (1) the containment air temperature will be maintained within limits during normal operation, and (2) adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post-LOCA conditions.

The required design cooling water flow to the Containment Cooling System is verified by the surveillance testing requirements of Specification 4.6.2.](b) i which is performed at 18 month intervals. The testing requirements of Specification 4.6.2.3(a), performed at 31 day intervals, ensure that the fan units and the cooling water flow paths (supply and return) from the Essential Service Water System headers are OPERABLE.

The Containment Cooling System and the Containment Spray System are redundant to each other in providing post accident cooling of the containment atmosphere. As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the Containment Cooling System havebeenappropriatelyadlssted. However, the allowable out-of-service tir-requirements for the Containment Spray System have been maintained consistent with that assigned other inoperable ESF equipment since the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC54 thru 57 of Appendix A to 10 CFR Part 50. Containment. isolation within the time limits specified for those isolation valves designed to close auto-matically ensures that the release of radicactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the 6 t::t b ~

--a& 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 condtions. Either recombiner unit is capable of con-trolling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. Operation of the Emergency Exhaust System with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. These Hydro-WOLF CREEK - UNIT 1 B 3/4 6-4 Amendment No. 50

PLANT SYSTEMS BASES 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 blowdown in the event of a steam line rupture.

This restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Requirements are consistent with the assumptions used in the safety analyses. ,

3.4.7.1.6 STEAM GENERATOR ATMOSPHERIC RELIEF VALVES The operability of the main steamline atmospheric relief valves (ARV's) ensures that reactor decay heat can be dissipated to the atmosphere in the event of a steam generator tube rupture and loss of offsite power and that the Reactor Coolant System can be cooled down for Residual Heat Removal System operation. The number of required ARV's assures that the subcooling can be achieved, consistent with the assumptions used in the steam generator tube rupture analysis, to facilitate equalizing pressures between the Reactor Coolant System and the faulted steam generator. For cooling the plant to RHR "

initiation conditions, only one ARV is required. In this case, with three ARV's operable, if the single failure of one ARV occurs and another ARV is assumed to be associated required with the faulted steam generator, one ARV remains available for heat removal.

i Each ARV is equipped with a manual block valve (in the auxiliary building) i to provide a positive shutoff capability should an ARV develop leakage. Closure !

of the block valves of all ARV's because of excessive seat leakage does not endanger the reactor core; consistent with plant accident and transient analyses, decay heat can be dissipated with the main steamline safety valves or a block valve can be opened manually in the auxiliary building and the ARV can be used to control release of steam to the astmosphere. For the steam generator tube rupture event, primary to secondary leakage can be terminated by depressurizing

[i the Reactor Coolant System with the pressurizer power operated relief valves. I

/ '

~

4 3/4.7.2 STEAH GEWERATOR FRE35URE/TEMFERATURE i. IMITATION htf7@

the pressure-indu n steam generator pressure and temperatura mies tnat allowable fracture toughness str w ' unit steam oenardesi2 ao not exceed the maximum are based on a " ass generator RT imitations of 70*F and 200 psi; b:1t+hrTrac t u re. NDT of 60 F and are su + to prevent WOLF CREEK - UNIT 1 B 3/4 7-3 Amenoment No. I'.

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1 i

INSERT B2 .

3/4.7.1.7 ' MAIN FEEDWATER ISOLATION VALVES The OPERABILITY of the main feedwater isolation valves (1) provides a pressure boundary to permit auxiliary feedwater addition in the event of'a-main steam or feedwater line break; (2) limits the RCS cooldown and-mass and energy releases for secondary line breaks inside containments and (3) mitigates steam generator overfill events such as a feedwater malfunction, with protection provided by feedwater isolation via the steam generator high-high level trip signal. The OPERABILITY of the main feedwater isolation valves within the closure times of the surveillance-requirements is consistent with the assumptions ~used in the safety analysis.

Insert to Bases Page B3/4 7-3 1

a.m--,,- _ - - - - - - - . - - . - - _ _ . , - - _ - - . - . -

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PLANT SYSTEMS BASES 3/A.7.9 JE ::G D04 @

All snubbers are required OPERABLE to ensure that the structural integrity o the Reactor Coolant System and all other safety-related systems is main-tai ad during and following a seismic or other event initiating dynamic loads.

S bbers are classified and grouped by design and manufacturer but no* by si:e. Fe example, mechanical snubbers utilizing the same design featur .of the 2-kip, 0-kip, and 100-kip capacity manufactured by Company "A" ar of the same type. e same design mechanical snubbers manufactured by Comp y "B" for the purpos of this Technical Specification would be of a diff rent type, as would hydrau 'c snubbers from either manufacturer. Snubbers also be classified and gr ed by inaccessible or accessible for visual inspection purposes. Therefor each snubber type may be grouped for in ection in accorcance with acces 'bility.

A list of individua nubbers with detailed inform ion of snubber loca-tion and size" and of system affected shall be availa e at the plant in accordance with Section 50.7 c) of 10 CFR Part 50. he accessibility of each snubber shall be determined an approved by the P1 t Safety Review Committee.

The determination shall be based pon the existi radiation levels and the expected time to perform a visual spection i each snubber location as well as otner factors associated with acc ssibili during plant operations (e.g.,

temperatura, atmosphere, location etc. a the recommendations of Regulatory Guides 8.8 and 8.10. The addition or de tion of any hydraulic or mechanical snubber shall be made in accordance wi ction 50.59 of 10 CFR Part 50. p The visual inspection frequene is based pon maintaining a constant-level of snubber protection durin an earthquak or severe transient. Therefore, the required inspection interva varies inversely ith the observed snuboer failures and is determined by he number of inopera le snubbers found during an inspection of each type. I order to establish _the 1 spection frequency for each type of snubber on a afety-related system, it wa assumed that the fre-quency of snubber failur s and initiating events is cons nt with time and that the failure of any sn ei could cause the system to be un otected and to result in failure during an ssumed initiating event. Inspections rformed before that interval has apsed may be used as a new reference point to determine the next inspect n. However, the results of such early inspect' ns performed before the ori nal required time interval has elapsed (nominal t1 e less 250 may not be u d to lengthen the required inspection interval. Any 1 spection whose resul s require a shorter in'spection interval will override the revious schedule.

' e acceptance criteria are to be used in the visual inspection to det r-min OPERABILITY of the snubbers. For example, if a fluid port of a.hydraul1 s ober is found to be uncovered, the snubber shall be declared inoperaole and hall not be determined OPERABLE via functional testing. Since the visual 1

\' l WCLF CREEK - UNIT 1 B 3/4 7-5 i

PLANT SYSTEMS BASES

\

88ERS (Continued) ins ions are augmented by functional testing program, the visual inspec* on need no be a hands on inspection, but shall require visual scrutiny suf' cient to assure hat fasteners or mountings for connecting the snubbers to s ports

~

or foundati s shall have no visible bolts, pins or fasteners'missin , or otner visible signs f physical damage such as cracking or loosening.

To provide as rance of snubber functional reliability, o .of three functional testing m hods are used with the stated accepta- e ' criteria: ,

1. Functionally 'st 10% of a type of snubber wit an additional 10%-

tested for each unctional testing failure,

2. Functionally test a ample size and deter,ine sample acceptance or rejection using Figur 4.7-1, or 3 Functionally test a repres tative ample size and determine sample '

acceptance or rejection usin th stated equation.

Figure 4.7-1 was developed using ' al 's Sequential Probability Ratio Plan" as described in " Quality Control an ndustr 1 Statistics" by Acheson J. Duncan.

Permanent or other exempti s from the surv llance program for individual snubbers may be granted by th Commission if a jus 'fiable basis for ex'emption is presented and, if applic le, snubber life destru ive testing was performed to qualify the snubber fo the applicable design condit' ns at either the com-pletion of their fabric ion or'at a subsequent date. Sn bers so exempted shall be listed in th list of individual snubbers indicati the extent of the exemptions.

The servi life of a snubber is established via manufacture input and information rough. consideration of the snubber service conditions nd associated nstallation and maintenance records (newly installed _ snub r, seal' replaced spring replaced, in high radiation area, in high temperature a es, etc.). The requirement to monitor'the snubber se,vice life is included to ensu that the snubbers periodically undergo a performance. evaluation in v1 of heir age and operating conditions. These' records will-provide statistical ses_for future consideration of snubber service life.

1 3/4.7.9 SEALED SOURCE C0" IPT!O" D6C fd- .

'I l

h _* inns on removable was testing, including h :, contamination it based for sources on '0 recui-4 M Jy.49(a)(3) ; for limits plutonium. This limitatinn ud w arr that 4 eor;: 'rnm -Byproduct, Source, and Soeri =1 S.icer Material sources will not exceed allowable in a * % j l

WOLF CREEK - UNIT 1 B 3/4 7-6

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l PLANT SYSTEMS 1 1

1 BASES h 00RCE CONTAMINATION (Continued) ,

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Sealed sources sified into three groups according t use, with'-

Surveillance Requirement urate with the pr y.of damage to a source in that group. :Those sourc requently handled are required to be tested more often'than th are ealed sources which are.

continuously enclosed a. shielded mechanism ( . ., ' led sources within radiation mo or boron measuring devices) are considere stored -1 an ot be tested unless they are removed from the shielded mechan . !

3/4.7.10 DELETED 3/4.7.11- DELETED 3/4.7.12 ^ RE". ?!P:n."T'J0: "0"!T0"I t4 DELEI6@

erature limitations e'nsure that safety-related ;'-[ I will not be subjecte tures in excess af +h:'. w&vTronmental qualification temperatures. Ex au .; e % e temperatures may degrade-equipment.and can of its OPERA 81LI . ture limits in wance for instrument error of 13*F.

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WOLF CREEK - UNIT 1 B 3/4 7-7 Amendment No. 15 i l

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, a l ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 CLECTRICAL COUII"CNT PCOTECTIVE OCVICCC- M N Containment electrical penetrations and penetration conductors _are prot ed by e er deenergizing circuits not required during reactor. operation or demonstr ng the OPERABILITY of primary and backup overcurrent' prote on circuit bre s during periodic surveillance.

The Surveillan Requirements applicable to lower vol e circuit breakers-provide assurance of b ker reliability by testing at st one representative 1 sample of each manufacture ' brand of circuit brea . Each manufacturer's molded case and metal case.ci it breakers are ouped into representative samples which are then tested on otating is to ensure that all breakers are tested. If a wide variety exists n any manufacturer's brand of cir-cuit breakers, it is necessary to div at manufacturer's breakers.into groups and treat each group as a arate ty of breaker for surveillance purposes.

A list of contain penetration conductor overcu nt protective devices j whose circuit limit fault current exceeds the penetrati rating, with 1 information of 1 tion and size and equipment powered by the tected circuit, is available the plant site in accordance with Section 50.71(c f i 10 CFR P 50. The addition or deletion of any containment penetrat condu r overcurrent protective device would be made in accordance with Se 1on 50.59 of 10 CFR Part 50.

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WOLF CREEK - UNIT 1 B 3/4 8-3 Amendment No '28. 1 l

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REFUELING OPERATIONS BASES l

3/4.9.5 -C0 Z NICATIONS DEU!@

The requiremeni T r e -"nications capabil4+y r.:a => cnat refueling station personnel can be n n--tl,, inf5rmea of ;gw4fic ut chances in the fac41W--st.ius or core reactivity conditions durino CORE ALTERAnum.

3/4.9.6 -RETUCLING MAC;;INC DREI60 ILITY requirements for the refueling machine and hoist ensure tha : ulator cranes will be u vement of drive rods and fuel assemblies, (2 icient load capacity to lift a drive rod or fuel assemb the c als and reactor vessel are protected ssive lifting force in the even inadvertently p ring lifting operations.

3/4.9.7 CRANE TRAVEL - STENT TUEL STORA0 TACILITY CEU T6D iction on movement of loads in excess of the nominal a fuel and contro eembly and associated handling to er fuel assemblies in the storage p s ensures t e event this load is dropped: (1) the activity release w . ted to that contained in a single fuel assembly, an possible disto fuel in the storage racks will no n a critical array. This assumption istent with l th y y release assumed in the. safety analyses. l 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING MODE, and (2) sufficient. coolant circulation is maintained through the core to minimize the effect of a boron dilution i incident and prevent boron stratification. The minimum of 1000 gpm allows flow rates which provide additional margin against vortexing at the RHR pump /

suction while in a reduced RCS inventory condition.

Addition of borated water with a concentration greater than or equal to the minimum required RWST concentration but less than the actual RCS boron-concentration shall not be considered a reduction in boron concentration.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of RHR capability. With the reactor vessel head removed and at least 23 feet of water above the reactor 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.

WOLF CREEK - UNIT 1 B 3/4 9-2 Amendmer,t No. 35 Noveter 22,1993

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3/4.10 SPECIAL TEST EXCEPTIONS s

BASES 3/4.10.1 ShJ~:0L; "" *,C:f' DELEIG D

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.This sp -

exception provides that a minimum rod worth is immediately aval a etivity_gnn* :. woen tests are performed ;

for control rod worth measureme '5.= spew' ' t exception is required to permit the perio '" ":. . . i ation of the actual versus pr . re reactivity?

c d4+' vucurring as a result'of fuel burnup or fuel cycling ooerati 3/4.-10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS -

. l This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance -[

of such PHYSICS TESTS as those required to: (1) malsure control rod worth, and (2) determine the reactor stability index and damping factor under xenon '

oscillation conditions.

3/4.10.3 PHYSICS TESTS-This soecial test except' ion permits PHYSICS TESTS to be performed at- less s.

than or equal to 5% of RATED THERMAL POWER with the RCS T,yg sligntly lower than normally allowed so that the fundamental nuclear characteristics _of the core and related instrumentation can be verified. In order for-various

, characteristics to be accurately measured,-it_is at times necessary to operate-outside the normal restrictions of these Technical Specifications. For instantei 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 which in turn may cause the RCS T to fall slightly below the minimum temperature of Specification 3.1.1.4. avg 3/4.10.4 REACTOR COOLANT LOOPS' '

This special test' exception permits reactor criticality uncer no. flow - .

conditions and is required to perform certain STARTUP and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.5 ^:2:'::N :.w::AT:CN T/ST:.1 ShJTOC u . DCLEIES

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WOLF CREEK - UNIT 1 B 3/4 10-1 ,

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3/4.11 RADI0 ACTIVE EFFLUENTS '

8ASES 3/4.11.1 LIQUID EFFLUENTS-3/4.11.1.1 DELETED 3/4.11.1.2 OELETED 3/4.11.1.3 DELETED 3/4.11.1.4 L:00:0 ll0 LOOP '^."r! LDE/N a

nks listed in this specification include all those outdoor r e tanks that a surrounded by liners, dikes, or walls capabl ding.

the tank contents a t do not have tank overflows an ounding area drains connected to the L q dweste Treatmen .

Restricting the quantity of ct erial contained in 'the specified tanks provides assurance n the event of an rolled release _of the tanks' contents, ulting concentrations would be e n the limits of 10 CFR Par , ppendix B, Table II, Column 2, at the nearest p water '

su nd the nearest surface water supply in an UNRESTRICTED AREA.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DELETED 3/4.11.2.2 DELETED 3/4.11.2.3 DELETED t 3/4.11.2.4 DELETED 3/4.11.2.5 EXFLOO!"C 0's "IT"" D6MI6D is specification:is provided to ensure that the concentration potentia losive gas mixtures contained in the WASTE GAS SYSTEM is maintained below lammability limits of hydrogen a gen. -Automatic-control features are inc in the system to -the hydrogen and oxygen:

concentrations from reaching the y limits. 'These automatic control features include isolat ce of hydrogen and/or. oxygen.

Maintaining the concente of hydrogen and o low their flammability limits provides nce that'the releases of radioac terials will be contro11 conformance with the requirements of General Des iterion 60 oJ App 6 dix A to 10 CFR Part 50.

I WOLF CREEK - UNIT 1 ~ 8 3/4 11-1 Amendment No. 42 l

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RADIOACTIVE EFFLUENTS

-BASES 3/4.11.2.6 OAO OT00AOC TA=0 Wif0 ks included in this specification are.those tanks for wh a quantity of r ivity contained is not limited direct 1 rectly by-another Technical Spec on. Restricting the y of radioactivity >

contained in each gas storage vide ance that in the event of an uncontrolled release of the tank's he resulting whole body exposure to'a MEMBER OF THE PUBLIC nearest SITE 8 will not exceed 0.5 rem.

This is consisten tandard Review Plan 11.3.-_Branc cal. Position -

ET!B 11-5 " ulated Radioactive Releases Due to a Waste Gas Sys k:or F , ' in NUREG-0800, July 1981.

3/4.11.3 DELETED 3/4.11.4 OELETED l

WOLF CREEK - UNIT 1 B 3/4 11-2 Amendment No. 42

i Attachment V to NA 94-0089 Page 1 of 4 ATTACHMENT V RESULTS OF APPLICATION OF THE NRC FINAL POLICY STATEMENT ON ,

TECHNICAL SPECIFICATION IMPROVEMENTS i

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1 Attachment V to NA 94-0089 j Page 2 of 4 l 1

l Introduction The NRC's Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, 58 FR 39132, July 22, 1993 (the Policy Statement). ,

urges licensees to upgrade plant Technical Specifications by' focusing the Technical Specifications on those requirements that are of. controlling importance to operational safety. To identify those requirements, the Policy-Statement includes four criteria to be used in screening the Technical Specifications. Technical Specifications that satisfy one or more of the' criteria must be retained. Specifications that do not satisfy . any of the criteria may be removed from the Technical Specifications. The Policy Statement states that removed requirements must be relocated into a licensee-controlled program or procedure. This attachment. provides the results of applying the Policy Statement screening criteria to the WCGS Technical Specifications,

Background

The NRC issued an Interim Policy Statement on Technical Specification Improvement, 52 FR 3788, February 6, 1987. In accordance with the Interim Policy Statement, the purpose of Technical Specifications is to. impose those-conditions or limitations upon reactor operation necessary to ' obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health _and safety by establishing those conditions of operation which cannot be changed without prior Commission approval . :and by identifying those features which are of controlling importance to safety.

The criteria contained in the Interim Policy Statement were applied to the Westinghouse Standard Technical Specifications (STS), NUREG-0452,. Revision 4 and Draft Revision 5, and submitted to the NRC in WCAP-11618. The results of

.the NRC review were issued by letter to the Westinghouse Owners Group dated May 9, 1988.

In . July 1993, the NRC issued the Final.. Policy Statement on ' Technical ;

Specification Improvements. The Final Policy. Statement incorporates .the information obtained from public comments and from the . experience gained in applying the interim policy criteria 'during development of new,~. vendor-specific STS. The new STS'for Westinghouse. plants are contained in NUREG-1431 issued in September 1992. :

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I Attachment V to NA 94-0089 Page 3 of 4 Application of the Screening Criteria Application of the criteria'from the Final Policy Statement to the Technical ,

Specifications was begun by preparing a screening form similar to that used in WCAP-11618 except that a separate screening criterion for risk-significant structures, systems, and components was added as required by the Final' Policy Statement. Each of the 115 Technical Specifications was evaluated using the ,

screening criteria and the clarifications included in the. . Policy Statement I discussion of each criteria. I During ' the Technical Specification evaluations, . reference was made .to the current Westinghouse STS and bases (Ref. 2), the screening forms in WCAP-11618 (Ref. 3), the NRC evaluation of WCAP-11618 (Ref . 4) , the results 'of an NRC test application of screening criteria to the WCGS Technical . Specifications (Ref. 5), and the results of applying the interim selection criteria to the-North Anna Plant.

Table 1 provides a summary of the results of applying the Final Policy Statement criteria. Table 1 also provides, for comparison, the results of the NRC review of previous Westinghouse STS in Ref. 4. The notes to Table 1 include information regarding the disposition of Technical Specifications and provide supporting - information justifying some of the proposed Technical- ,

Specification changes.

The screening forms for those Technical Specifications that ' did not satisfy any of the criteria and,-therefore proposed for relocation, are included in '

Table 2.

I Appendix A is the Probabilistic Safety Assessment evaluation that was used to identify structures, systems, and components that satisfied Criterion 4 on the ,

screening forms.

References In this attachment and on the screening forms, the following references have ,

been used:

1. WCGS Technical Specifications and Bases (NUREG-1136) as amended.

2. Standard Technical Specifications, Westinghouse Plants, NUREG-~ 1 1431, September 1992

3. J. D. Andrachek, et. al., Methodically Engineered, Restructured, and Improved Technical Specifications, MERITS Program - Phase.II Task 5, Criteria Application, WCAP-11618, November 1987. -

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4. NRC letter to Westinghouse Owners Group (T. Murley to R. . Newton) ,

"NRC Staff Review of Nuclear Steam _ Supply System Vendor Owners Groups' Application of the Commission's Interim Policy _ Statement Criteria to Standard Technical Specifications," May 9, 1988.

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J E Attachment V to NA 94-0089 Page 4 of 4 i-' . 5. NRC memorandum (V. Stello to NRC. Commissioners), " Test Application.

l of TSIP : Technical Specification Selection Criteria," February 7, 1986.

6. 'NRC Generic Letter 85-05,

" Inadvertent Boron Dilution Events ,." -

January 31, 1985.

7. NSAC-183, ." Risk of PWR Reactivity Accident During ' Shutdown ~ and Refueling."

8. TU Electric Letter to'NRC, TXX-93098, dated April'30, 1993, and: $

NRC Approval and SER dated November 3, 1993.

9. TR-92-0063, " Wolf Creek Generating Station Individual Plant .

Examination Summary Report," September 1992.

10. WASH-1400, " Reactor Safety Study,'An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," October 1975.

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TABLEI Summary of Criteria Application Results Reactivity Control Systems Tech STS Rev. Technical Specification NRC WCGS Results Note Spec 5 Number Title Results Number 3.1.1.1 3.1.1.1 Shutdown Margin Retain See Note 1 1 3.1.1.2 3.1.1.2 Shutdown Margin s200 F Retain N/A See Note 1 1 3.1.1.3 3.1.1.3 Moderator Temp. Retain Retain Coefficient 3.1.1.4 3.1.1.4 Min. Temperature for Retain Retain

~ Criticality 3.1.2.1 3.1.2.1 Boration Path Shutdown Relocate Relocate 3.1.2.2 3.1.2.2 Boration Path Operating Relocate Relocate 3.1.2.3 3.1.2.3 Charging Pumps Relocate See Note 2 2 Shutdown 3.1.2.4 3.1.2.4 Charging Pumps Relocate Relocate Operating 3.1.2.5 3.1.2.5 Borated Water Sources Relocate Relocate Shutdown 3.1.2.6 3.1.2.6 Borated Water Sources Relocate Relocate Operating 3.1.3,1 3.1.3.1 Movable Control Retain Retain 1 Assemblies - Group Height 3.1.3.2 3.1.3.2 Position Indication - Relocate Retain 3 Operating 3.1.3.3 3.1.3.3 Position Indication - Relocate Relocate 3 Shutdown 3.1.3.4 3.1.3.4 Rod Drop Time Relocate Relocate 4 3.1.3.5 3.1.3.5 Shutdown Rod Insertion Retain Retain Limits 3.1.3.6 3.1.3.6 Control Rod Insertion Retain Retain 1 Limits

i TABLE I (Cont.) l Summary of Criteria Application Results Power Distribution Limits Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.2.1 3.2.1 Axial Flux Differ. Retain Retain 3.2.2 3.2.2 Heat Flux Hot Channel Retain Retain Factor 3.2.3 3.2.3 Nuclear Enthalpy Rise Retain Retain Hot Channel Factor 3.2.4 3.2.4 Quadrant Power Tilt Retain Retain Ratio 3.2.5 3.2.5 DNB Parameters Retain Retain TABLE I (Cont.)

Summary of Criteria Application Results Instrumentation Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.3.1 3.3.1 Reactor Trip System Retain Retain Instrumentation 3.3.2 3.3.2 Eng. Safety Feature Retain Retain Actuation System Instrumentation 3.3.3.1 3.3.3.1 Radiation Monitoring Retain Retain Instrumentation 3.3.3.2 3.3.3.2 hiovable Incore Detectors Relocate Relocate 3.3.3.3 3.3.3.3 Seismic Instrumentation Relocate Relocate 3.3.3.4 3.3.3.4 hieteorological Relocate Relocate Instrumentation 3.3.3.5 3.3.3.5 Remote Shutdown Retain Retain Instrumentation 3.3.3.6 3.3.3.6 Accident hionitoring Retain Retain 5 Instrumentation 3.3.3.9 3.3.3.9 Loose Parts Detection Relocate Relocate System 3.3.3. I 1 Explosive Gas hionitoring Not Relocate 6 Instrumentation Reviewed 3.3.4 3.3.4 Turbine Overspeed Relocate Relocate 7 Protection

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Summary of Criteria Application Results Reactor Coolant System Tech STS Rev. 5 Technical Speci6 cation NRC WCGS Note Spec Number Title Results Results Number 3.4.1.1 3.4.1.1 Reactor Coolant Loops Retain Retain and Coolant Circulation 3.4.1.2 3.4.1.2 RCS Hot Standby Retain Retain 3.4.1.3 3.4.1.3 RCS Hot Shutdown Retain Retain 3.4.1.4.1 3.4.1.4.1 Cold Shutdown Loops Retain Retain :

Filled 3.4.1.4.2 3.4.1.4.2 Cold Shutdown Loops Retain Retam {

Not Filled 3.4.2.1 3.4.2.1 Safety Valves -Shutdown Relocate Relocate 3.4.2.2 3.4.2.2 Safety Valves -Operating Retain Retain i 3.4.3 3.4.3 Pressurizer Retain Retain 3.4.4 3.4.4 Relief Valves Retain Retain 3.4.5 3.4.5 Steam Generators Relocate Relocate 8 3.4.6.1 3.4.6.1 Leakage Detection Retain Retain Systems 3.4.6.2 3.4.6.2 Operational Leakage Retain Retain l 3.4.7 3.4.7 Chemistry Relocate Relocate 9 !

3.4.8 3.4.8 Speci&c Activity Retain Retain l 3.4.9.1 3.4.9.1 Pressure / Temperature Retain Retain Limits 3.4.9.2 3.4.9.2 Pressurizer Relocate Relocate Pressure / Temperature 3.4.9.3 3.4.9.3 Overpressure Protection Retain Retain System 3.4.10 3.4.10 Structural Integrity Relocate Relocate 10 3.4. I 1 3.4.11 RCS Vents Relocate Relocate l

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TABLE 1 (Cont.)

Summary of Criteria Application Results Emergency Core Cooling Systems Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.5.1 3.5.1 Accumulators Retain Retain 3.5.2 3.5.2 ECCS Subsystems Tavg Retain Retain 2 350 F 3.5.3 3.5.3 ECCS Subsystems Tavg Retain Retain

< 350 F 3.5,4 ECCS Subsystems Tavg Not Retain 2, 11 -j 5;200 F Reviewed 3.5.5 3.5.5 RWST Retain Retain TABLE 1 (Cont.)

Summary of Criteria Application Results Containment Systems Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.6.1.1 3.6.1.1 Containment Integrity Retain Retain 12 3.6.1.2 3.6.1.2 Containment Leakage See Note 12 See Note 12 12 1 3.6.1.3 3.6.1.3 Containment Airlocks Retain Retain 3.6.1.4 3,6.1.' Internal Pressure Retain Retain 3.6.1.5 3.6.1.6 Air Temperature Retain Retain 3.6.1.6 3.6.1.7 Contain. Vessel Structural Relocate Relocate 13 Integrity 3.6.1.7 3.6.1.8 Containment Ventilation Retain Retain 14 System 3.6.2.1 3.6.2.1 Containment Spray Retain Retain System 3.6.2.2 3.6.2.2 Spray Additive System Retain Retain 3.6.2.3 Containment Cooling Retain Retain System 3.6.3 3.6.3 Containment Isolation Retain Retain Valves 3.6.4.1 3.6.4.1 Hydrogen Analyzers Retain Delete 15 3.6.4.2 3.6.4.2 Hydrogen Control System Retain Retain i

TABLE 1 (Cont.)

Summary of Criteria Application Results Plant Systems Tech STS Rev. 5 Technical Specification Title NRC WCGS Note Spec Number Results Results Number 3.7.1.1 3.7.1.1 Safety Valves Retain Retain 3.7.1.2 3.7.1.2 Auxiliary Feedwater System Retain Retain 3.7.1.3 3.7..l.3 Condensate Storage Tank Retain Retain 3.7.1.4 3.7.1.4 Specific Activity Retain Retain 3.7.1. 5 3.7.1.5 Main Steam Isolation Valves Retain Retain 3.7.1.6 Steam Generator Not Retain Atmospheric Relief Valves Reviewed 3.7.1.7 Main FeedwaterIsolation Not Add 16 Valves Reviewed 3.7.2 3.7.2 Steam Generator Relocate Relocate Pressure / Temperature Limits 3.7.3 3.7.3 Component Cooling Water Retain Retain 3.7.4 3.7.4 Essential Service Water Retain Retain System 3.7.5 3.7.5 Ultimate IIcat Sink Retain Retain 3.7.6 Control Room Emerg. Retain Retain Ventilation System 3.7.7 3.7.8 Emerg. Exhaust System - Retain Retain Auxiliary Building 3.7.8 3.7.9 Snubbers Relocate Relocate 17 3.7.9 3.7.10 Scaled Source Contamination Relocate Relocate 3.7.12 3.7.13 Area Temperature Relocate Relocate 18 Monitoring l

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Summary of Criteria Application Restilts Electrical Power Systems Tech STS Rev. 5 Technical Specification Title NRC WCGS Note Spec Number Results Results Number

3. 8.1.1 3.8.1.1 AC Sources Operating Retain ,

Retain 3.8.1.2 3.8.1.2 AC Sources Shutdown Retain Retain 3.8 2.1 DC Sources Operating Retain Retain 3.8.2.2 DC Sources Shutdown Retain Retain 3.8.3.1 3.8.3.I Onsite Power Distrib. - Operating Retain Retain 3.8.3.2 3.8.3.2 Onsite Power Distrib. - Shutdown Retain Retain 3.8.4.1 3.8.4.I Containment Penetration Relocate - Relocate Conductor Overcurrent Protection Devices

j TABLE 1 (Cont.)

Summary of Criteria Application Results Refueling Operations Tech STS Rev. 5 Technical Specification Title NRC WCGS Note Spec Number Results Results Number 3.9.1 3.9.1 Boron Concentration Retain Retain 3.9.2 3.9.2 Instrumentation Retain Retain 3.9.3 3.9.3 Decay Time Retain Retain 3.9.4 3.9.4 Containment Building Retain Retain Penetrations 3.9.5 3.9.5 Communications Relocate Relocate 3.9.6 3.9.6 Refueling Machine Relocate Relocate 3.9.7 3.9.7 Crane Travel - Spent Fuel Stor. Relocate Relocate Facility 3.9.8.1 3.9.8.1 RHR and Coolant Recirculation - Retain Retain High Water Level 3.9.8.2 3.9.8.2 RHR and Coolant Recirculation - Retain Retain Low Water Level 3.9.9 3.9.9 Containment Ventilation System Retain Retain 3.9.10.1 Water Level Reactor Vessel - Retain Retain Fuel Assemblies 3.9.10.2 Water Level Reactor Vessel - Not Relocate 19 Control Rods Reviewed 3.9. I 1 3.9.11 Water Level -Storage Pool Retain Petain 3.9.12 Spent Fuel Assembly Storage Not Retain Reviewed 3.9.13 3.9.12 Emergency Exhaust System Fuel Retain Retain Building i

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TABLE I (Cont) l Summary of Criteria Application Results Special Test Exceptions ;

Tech STS Rev. 5 Technical Specification NRC WCGS Note i Spec Number Title Results Results Number 3.10.1 3.10.1 Shutdown Margin Relocate Delete 20 3.10.2 3.10.2 Group Height, Insertion, Retain Retain and Power Distribution Limits 3 10.3 3.10.3 Physics Tests Retain Retain 3.10.4 3.10.4 Reactor Coolant Loops Retain Retain 3.10.5 3.10.5 Position Indication Relocate Relocate 20 System Shutdown TABLE I (Cont.)

Summary of Criteria Application Results Radioactive Effluents Tech STS Rev. 5 Technical Specification NRC WCGS Note .

Spec Number Title Results Results Number 3.11.1.4 3.11.1.4 Liquid Holdup Tanks Relocate Relocate 21 3.11.2.5 3.11.2.5 Explosive Gas Mixture Relocate Relocate 6 3.I1.2.6 3.11.2.6 Gas Storage Tanks Relocate Relocate 21

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.i Notes to Table 1. ]

NOTES: 1. SDM -in Modes 1.and '2 ' is ensured by the control. rods. ;

maintained at or.' above their insertion limits and, for certain ll events which add positive reactivity, the'boration-capability of- i the.ECCS-is credited. The NRC review issued to the WOG. dated May. ,

h 9, 1988, concluded that the SDM TS could be. relocated for Modes 1 and 2 and retained for Modes 3, 4, and 5. However,. Wolf Creek has determined that.the SDM requirements'for Modes 1 and 2 should be retained in theiTechnical Specifications under other : Reactivity Control- Systems . The changes to, Technical Specification 3.1.1'.1' consist of' deleting Modes 1 and 2.from the LCO applicability and-incorporating the Modes 1 and 2 requirements under new TechnicalL Specification 3.1.1.5 and . existing Technical Specifications-3.1.3.1 and 3.1.3.6.

Action a of LCO 3.1.3.6 has been added - to address .the'l required actions for a loss of SDM in Modes 1 and 2. New'LCO 3.1.3.6' Action a provides one hour to verify SDM - or. initiate - boration, consistent with the timing for Actions 3.1.3.la and 3.1. 3. lc .. of LCO 3.1.3.1. The Actions for LCO 3.1.1.1 in Modes 3 and 4 and'LCO

3.1.1.2 in Mode 5 have been revised-'to replace "immediately" with "within 15 minutes" to implement - boration, . per. . the - ' STS . SR-4.1.1.1.la for Modes 1 and 2 has.been incorporated into . Action - 4 3.1,3. la and 3.1. 3.1c of LCO 3.1.3.1. -SR 4.1'.1.1.1b and ' Action 3 .1. 3 .1. c . 3 . b) of LCO 3.1.3.1' have been deleted since they ' are redundant to renumbered SR - 4.1;3. 6.1. .' SR 4.1.1.1.1c, ' regarding estimated critical position, has been. moved to. Technical-Specification -3.1.3.6, Control Rod . Insertion . ' Limits, . as.-SR. ,;

4.1.3 A.2. . Moving - these SDM . requirements . for Modes .1 and 2 to Technical Specification 3.1.3.1 and L. 3.1.3.6- improves the- .

specifications by placing actions and surveillances for inoperable rods and insertion limits with their appropriate LCOs.

SR 4.1.1,1.1d and SR 4.1.1.1.2, regarding measuring SDM prior to ' '

5% ratec thermal power (RTP) with' rods fully inserted and ;

maintaining core reactivity within predicted . values,. have . been convertea into new Technical Specification- 3;1.1.5, Core Reactivity.

2. SR 4.1.2.3.2 limits the number.'of- operable centrifugal charging pumps to one in Modes 4 , ~ 5, and ' 6 '(except when ;the y reactor vessel head is removed) . _ This is ' an operating restriction.

of the reactor vessel cold overpressure analysis. This SR.will.be retained under LCO-3.5.4, ECCS Subsystems - Tavg s200'F for Modes 5 and 6. .The footnote' .to 3.1 2.3 is deleted because it is' v redundant to the footnote for. Specification 3 . 5 . 4 '. SR - 4.5. 3. 2 '

addresses Mode 4. ,

3. The NRC review of LCO 3.1.3.2 and LCO 3.1.3.3' concluded that they could be relocated. However, if an associated SR is necessary to meet the operability requirements for a retained LCO,1 the SR should be relocated to the ' retained LCO. Our: evaluation found that LCO 3.1.3.2 is associated with a transient ' analysis 4

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initial condition and supports LCO 3.1.3.1. As such, LCO 3.1.3.2 will be retained as ' is. The surveillance associated with LCO 3.1.3.3 is not required for any retained LCO and, therefore, SR 4.1.3.3 will be relocated.

4. Th3 NRC review of this LCO concluded that- it -could be relocated. However, if an associated SR is necessary to meet the' operability requirements for a retained LCO, the SR should be relocated to the retained LCO. SR 4.1.3.4 'is required to ensure the operability of control rods under LCO 3.1.3.1 and will be retained under that LCO with the rod drop time limit given in new USAR Section 16.1.3.2. This is consistent with STS.

5. The Regulatory Guide 1.97, Rev. 2, Type A variables identified in USAR Appendix 7A are retained. The neutron flux (Gamma-Metrics) and RVLIS instrumentation will be added. The non-Type A variables are identified and evaluated on the screening form. The relocated instruments are:

Containment Pressure - Extended Range ,

PZR Safety Valve Position Indication Unit Vent High Range Nc'*le Gas Monitor PORV and PORV Block Valve position indicators have been deleted from Technical Specification 3 . 3 . 3 . 6 ' and monthly channel checks have been added to LCO 3.4.4 as discussed in the Safety Evaluation, Attachment I.

6. This specification will be relocated and an Explosive Gas .

I Monitoring Program statement will be incorporated into new Section 6.8.5.

7. This specification will be relocated and a Turbine Overspeed Protection Reliability Program statement will be incorporated into new Section 6.8.5.

8. This specification will be relocated and a Steam Generator Tube Surveillance Program statement will be included in new Section 6.8.5.

9. This specification will be relocated and a Primary Water Chemistry Program statement will be included in new Section 6.8.5.

10. The LCO will be relocated and the associated SR regarding-RCP flywheel integrity will be retained in new Section 6.8.5 as a programmatic requirement. ]

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11. This LCO is intended to prevent a loss of the decay heat-removal function in Mode 5 and Mode 6 with the reactor vessel head "

installed by allowing the safety injection pumps to be operable when the water level is below the vessel flange. The LCO will be retained. Consideration was given to incorporating the -

p a

restrictions on- pump - operation into LCO 3.4.9.3, Overpressure Protection,- which ' would have been in conformance with the STS approach. However, the Modes and RCS. temperatures:for which these-specifications apply prevented combining. them .into one :

specification.

12. Containment testing is a requirement imposed by Appendix J.of. ;

10 CFR.50. This LCO will be relocated - however, the values , of parameters defining leakage limits from 3.6.1.2. will - be retained under the Containment Integrity Bases. SR 4. 6.1. lc . will be modified to eliminate reference- to a specification that was relocated, and instead reference corresponding USAR Section 16.6.1.1.

13. This specification will be relocated and a Containme.nt Tendon Surveillance Program statement will be incorporated 'i nto new ,

Section 6.8.5.

14. SR 4.6.1.7.2 will be modified to eliminate reference to a- !

specification that was relocated, and instead reference corresponding USAR Section 16.6.1.1.

15. LCO 3.6.4.1 is deleted since it 'is. redundant t o LCO 3 . 3 . 3 . 6 and is obsolete per the STS.

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16. A new Technical Specification for operability L of .the Main Feedwater Isolation Valves (MFIVs)- will be added ito the Technical Specifications for Wolf Creek. The requirements will be identical >

to those in the Callaway Technical Specifications. Inclusion of a specification for the -MFIVs is consistent. 'with NRC Policy-Statement criterion 3 regarding accident mitigating components. j

17. This specification will'be relocated and a Snubber Inspection .;

Program statement will be included in New Section 6.8.5. i

18. This specification will be relocated and.an Area Temperature Monitoring Program statement will be included in 'new Section 6.8.5.

19. This specification places a lower limit on the amount of water above the top of the fuel assemblies in the reactor vessel during.

movement of control rods. The Bases state-that this ensures the water-removes 99% of the assumed 10% iodine gap activity released i from the rupture of an irradiated fuel assembly in the event ofLa; fuel handling accident during core alterations. However, the movement of - control rods is not associated with the' initial.

conditions of a fuel handling accident, and the Bases . do not-address any concerns regarding inadvertent criticality which could.

lead. to a breach of the fuel rod cladding. Inadvertent' 1 criticality during Mode 6 is prevented by maintaining proper boron- l concentration in the coolant in accordance with LCO' 3 . 9 '.1. -

Therefore, this LCO will be relocated. ]j

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.20. The NRC review concluded that: (1) . special ' test . exceptions-3.10.1- through 3 .10. 4 . may be included.' with ; corresponding LCOs -

which are remaining in Technical Specifications,.and (2)- special test. exception 3.10.5 may be relocated along: with LCO - 3 .1. 3 . 3 .

LCO 3.10.1 is only applicable in~ Mode 2. As-discussed'in Note 1 above the SHUTDOWN MARGIN requirements for Modes 1 and ' 2 are retained in other Reactivity Control. System . Technical.

Specifications. Retained Special . Test- Exceptions- 3.10.2 and-3.10.3 address Special Test Exception 3.10.1 for'LCOs'3.1.3.1 and 3.1.3.6. Therefore, LCO 3.10.1 will be deleted. Also, per the stated NRC conclusion, LCO 3.10.5 will be relocated. LCOs 3.10.2.

through 3.10.4 will be retained as they are.

21. This , specification will be relocated and a . Storage -Tank'-

Radioactivity Monitoring , Program . statement will be included in Section 6.8.5.

TABLE 2 SCREENING FORMS FOR SPECIFICATIONS TO BE RELOCATED Screening Forms for the following Technical Specifications are attached:

REACTIVITY CONTROL SYSTEMS 3.1.1.1 SHUTDOWN MARGIN Shutdown Margin requirements for Modes 1 and 2 will be ,

incorporated under other Reactivity Control System Technical Specifications.

3.1.2.1 FLOW PATHS - SHUTDOWN 3.1.2.2 FLOW PATHS - OPERATING 3.1.2.3 CHARGING PUMPS - SHUTDOWN 3.1.2.4 CHARGING PUMPS - OPERATING 3.1.2.5 BORATED WATER SOURCES - SHUTDOWN 3.1.2.6 BORATED WATER SOURCES - OPERATING 3.1.3.3 POSITION INDICATION SYSTEM - SHUTDOWN 3.1.3.4 ROD DROP TIME POWER DISTRIBUTION LIMITS NONE INSTRUMENTATION 3.3.3.2 MOVABLE INCORE DETECTORS 3.3.3.3 SEISMIC INSTRUMENTATION 3.3.3.4 METEOROLOGICAL INSTRUMENTATION 3.3.3.6 ACCIDENT MONITORING INSTRUMENTATION 3.3.3.9 LOOSE-PART MONITORING INSTRUMENTATION 3.3.3.11 EXPLOSIVE GAS MONITORING INSTRUMENTATION 3.3.4 TURBINE OVERSPEED PROTECTION

_ ..e

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l c q REACTOR COOLANT SYSTEM' I

SAFETY VALVES --SHUTDOWN 3.4.2.1 3.4.5 STEAM GENERATORS a

3.4.7 CHEMISTRY ,

3.4.9.2 PRESSURIZER P/T LIMITS 3.4.1'O STRUCTURAL INTEGRITY 3.4.11 REACTOR COOLANT SYSTEM VENTS )

EMERGENCY CORE COOLING SYSTEMS

4 4

NONE CONTAINMENT SYSTEMS

, .i 3.6.1.2 CONTAINMENT LEAKAGE f

3.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY PLANT SYSTEMS ,

3.7.2 STEAM' GENERATOR PRESSURE / TEMPERATURE LIMITATION 3.7.8 SNUBBERS-1 3.7.9 SEALED SOURCE CONTAMINATION 3.7.12 AREA TEMPERATURE MONITORING ELECTRICAL POWER SYSTEMS l 3.8.4.1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE' DEVICES REFUELING OPERATIONS j 3.9.5 ' COMMUNICATIONS' 3.9.6 REFUELING MACHINE.

3.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY 3.9-10.2 WATER LEVEL -~-REACTOR' VESSEL / CONTROL RODS-9 6

. + . , , ,

+ ., .y . ., a., . , . , - , ,,-q.v

SPECIAL TEST EXCEPTIONS 3.10.1 SHUTDOWN MARGIN 3.10.5 POSITION INDICATION SYSTEM - SHUTDOWN RADIOACTIVE EFFLUENTS 3.11.1.4 LIQUID HOLDUP TANKS 3.11.2.5 EXPLOSIVE GAS MIXTURE 3.11.2.6 GAS STORAGE TANKS

TEC11NICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.1.1.1 SHtrTDOWN MARGIN Applicable Modes: 1, 2, 3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

1 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Based on the discussion below, this LCO satisfies criterion 2 for Modes 3,4, and 5. For Modes I and 2, the criterion is not satistied.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design 13 asis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES" then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

j (3) DISCUSSION Shutdown Margin (SDM) requirements provide sufTicient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for nonnat shutdown and anticipated operational occurrences. 'Ihe SDM defines the degree of suberiticality that would be obtained immediately following the insenian or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn During power operation, SDM control is ensured by operating with q the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion.

When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron Concentration. .j The 11ases for 'his 13 state that sufficient SDM ensures (1) the reactor can be made suberitical from all operating conditions, (2) reactivity trarcients associated with postulated accidents are controllable within acceptable limits , and (3) the reactor will be maintained si fliciently subcritical to preclude inadvertent criticality in the shutdown condition. The most restrictive =

condition is EOI. at no load operating Tavg associated with a MSLB. A minimum SDM of 1.3% Delta-k/k is required to contrd the react 9ity added by the cooldown. 'the SDM requirements must also protect against; i

a. Inadvertent boron dilution.

b. An uncontrolled rod withdrawal fmm subcritical or low power condition,

! c. Startup of an inactive reactor coolant pump, and

d. Rod ejection.

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~ In Modes 1 and 2,~ SDM is verified by obserymg that the requirements for rod insertion limits are met. In Modes 3,4, and 5,. )]

SDM is verified by performing a reactivity balance calculation. )

The SDM (boration control) TS is not applicable to a process variable indicating in the control room a significant degradation .]'

of the RCPB. Therefore, SDM does not satisfy criterion 1. -l I

SDM is an initial condition of accident and transient analyses, flowever, during operation in Modes 1 and 2, the available . ^!

SDM is determined by the rod insertion limits. Derefore, SDM (boration control) requirements are not applicable to a process variable, design feature, or operating restriction that either assumes the failure of or presents a challenge to the integrity of a ;

fission product barrier and, thus, do not satisfy criterion 2 for these operating Modes. ' liowever, this TS Au satisfy criterion 2 ,;

for Modes 3 and 4. j i

%e TS requirements for SDM are not applicable to an SSC that is part of the primary success path and which functions or .

actuates to mitigate a DBA or transient; these requirements, therefore, do not satisfy criterion 3. ]

Ref 4 concluded that the LCO could be relocated for Modes 1 and 2 but must be retained for Modes 3,4, and 5? Ilowever, Wolf Creek has determined that the SDM requirements for Modes 1 and 2 should be retained in the. Technical SpeciGcations under other Reactivity Control Systems. The changes to Tecimical Specification 3.1.1.1 consist of deleting Modes 1 and 2 from - -

the LCO applicability and incorporating the Modes I and 2 requirements under new Technical Specification 3.1.1.5 and' existing Technical Specifications 3.1.3.1 and 3.1.3.6.

Action a of LCO 3.1.3.6 has been added to address the required actions for a loss of SDM in Modes I and 2. New LCO 3.1.3.6 Action a provides one hour to verify SDM or initiate boration, consistent with the timing for Actions 3.1.3.la and 3.1.3.le of LCO 3.L3.1. He Actions for LCO 3.1.1.1 in Modes 3 and 4 and LCO 3.1.1.2 in Mode 5 have been revised to replace "immediately" with "within 15 minutes' to implement boration per the STS. SR 4.1.1.1.la for Modes 1.and 2 has been incorporated into Actions 3.1.3.la and 3.1.3.lc of LCO 3.1.3.1. SR 4.1.1.1.lb and Action 3.1.3.1.c.3.b) of LCO 3.1.3.1 have '

been deleted since they are redundant to renumbered SR 4.13.6.1. SR 4.1.1.1 lc, regarding es9 mated critical position, has ;jj been moved to Technical Specification 3.L3.6, Control Rod insertion Limits, as SR 4.1.3.6.2. Moving these SDM- ;

requirements for Modes I and 2 to Technical Specifications 3.1.3.1 and 3.1.3.6 improves the specifications by placing actions !

and requirements for inoperable rods and insertion limits with their appropriate LCOs.

SR 4.1.LI.ld and SR 4.1.1.1.2, regarding measuring SDM prior to 5% RTP with rods fully inserted and maintaining core reactivity within predicted values, have been converted into new Technical Speciti.,ation 3.1,1.5, Core Reactivity.

From Reference 2, shutdown margins during power operation have not been shown to be tisk significant to public health and' ,!

safety by either operating experience or PSA. Baron dilution accidents, rod ejection accidents and retum to power from plant ~

transients are the major scenarios for which the shutdown margin is needed. Ilowever, none of the three are dominant risk - I contributors. Shutdown margins are not modeled in the Wolf Creek IPE. Therefore, this TS does not satisfy Criterion 4J I Based on the above, the SDM requirements for Modes 1 and 2 will be retained under other Reactivity Control System Technical Specifications. He LCO for Modes 3,4, and 5 should be retained because SDM in these modes is not verified by.

the rod insertion limits.

(4) CONCLUSION

.X. His Technical Specification is retained. <

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_ ne Technical Specification may be relocated to the following controlled document (s):

j Hit die f.'

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L1 TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 31.2.1 HORATION FLOW PAlilS - SilUTDOWN Applicable Modes: 4,5, and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO 2, (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to ,

the integrity of a tission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the faihire of or presents a challenge to the integrity of a tission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "'(ES", then the Technical Specification shall be retained in the Tecimical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION 1he Bases for this LCO state that the purpose is to assure negative reactivity control is available during each Mode of facility operation.

The purpose of the boration subsystem of the CVCS is to provide the means to control the boron concentration to maintain SDM. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of an assumed boron dilution event, the automatic response of the BDMS, or that required of the operator, is to close the appropriate valves in the reactor makeup system before the shutdown margin is lost. Automatic actuation of the boration subsystem is not required to mitigate the event. Shutdown Margin (SDM) requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for nonnal shutdown and anticipated operational occurrences.

1hc SDM defines the degree of suberiticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn.' During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

Ref. 5 notes that the normal capability to control reactivity with baron is not credited in the accident analysis.

The boration subsy stem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The boration subsystem TS is not associated with a process variable that is an initial condition of an event that assumes failure of or challenges the integrity of a tission product barrier. As stated in the analyscs of boron dilution events, the BDMS perfonns automatic acticns in response to detecting an assumed boron dilution event. These actions are credited for events

.+. . . .. - . _.. .

occurring in Modes 3,4, and 5.' The actions include providing an alarm, automatically isolating the dilution flow path, and automatically initiating boration of the RCS fmm the RWST via the charging pumps. For these events, the primary success -

path' for mitigation includes isolating the dilution floupath. The subsequent actuation of equipment to establish a boron - >

injection flowpath is intended to regain the required SDM; This is desirable, but beyond the scope of a primary success path-action. Therefore, tids TS does not satisfy criterion 2.

The boration subsystem TS does not apply to any SSC that is a part of the primary success path and which functions to mitigate a Dl3A or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrict; therefore, this TS does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the RWST, and associated flowpaths is required as part of the ECCS TS.

The shutdown 110w paths, used to inject borated water to maintain SDM, have not been shown to be significant to public health and safety by either operating experience or PSA. The shutdown flow paths are modeled in the WCGS IPE for the ATWS event, in the LTS fault tree. Ilowever, the core damage values for the ATWS event sequences are extremely low, well below the NRC screening value of 1.0E-06. Therefore, this TS does not satisfy Criterion 4.

(4) CONCLUSION This Technical Specification is retained.

1 1he Tecludcal Specilication may be relocated to the following controlled document (s):

USAR Chapter 16.

3121d%

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 31.2.2 BORA110N FLOW PATIIS - OPERA 1TNG I Applicable Modes: 1,2, and 3 ]

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnonnal degradation of the reactor coolant pressure boundary,

_K, (2) A process variable, design feature, or operating restriction that is an initial condition of a Design -

Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A stmeture, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or ,

presents a challenge to the integrity of a tission product barrier.

X (4) A structure, system, or component which operating experienc~ or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each Mode of facility operation.

The purpose of the boration subsystem of the CVCS is to provide the means to control the boron concentration to maintain SDM. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of an assumed baron dilution event in Mode 3, the automatic response of the BDMS, or that required of the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Automatic actuation of the boration subsystem is not required to mitigate the event. In Modes 1 and 2, a dilution event is initially mitigated by the RTS and the reactor is shut dow n by insenion of the control rods. Continued dilution will tend to take the reactor critical; however, the operator has more than 30 minutes to stop the dilution flow. Ref. 5 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. Shutdown Margin (SDM) requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM detines the degree of suberiticality that would be obtained immediately following the insertion or scram of all shutdown and -

control rods, assuming that the single rad assembly of highest worth is fully withdrawn. During power operation, SDM control is ensured by operating with the Autdown banks fully withdrawn and the control banks within the limits of LCOs -

3.1.3.5 and 3.1.3.6, for rod insertion. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

Based on the foregoing, the boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion 1.-

The boration subsystem TS is not associated with a process variable that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. As stated in the analyses of boron dilution events, the DDMS

performs automatic actions in response to detecting an assumed boron dilution event. Rese actions are credited for events occurring in Mode' 3, - ne actions include providing an alarm, automatically isolating the dilution flow. path, and automatically initiating boration of the RCS from the RWST via the charging pumps for these events, the primary success path for mitigation includes isolating the dilution flowpath. He subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. His is desirable, but beyond the scope of a primary success path' action. In Modes I and 2, the operator is required to isolate the dilution flow path subsequent to a reactor trip. Herefore, the boration subsystem is not a design feature required to be operable to~ mitigate these events, and this TS does not satisfy.- 't criterion 2.

The boration subsystem TS does not apply to an SSC that is part of the primary success path and which functions or actuates .

i to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the RWST, and -

associated flowpaths is required as part of the ECCS TS.

For the MSLB event, the sequence of events takes the plant to cold shutdown conditions and, therefore, boration of the RCS is .

necessary. Ilowever, the boration 110wpath in this case is required as part of the ECCS function.

From Reference 2, the operating flow paths, used to inject borated water to maintain SDM, have not been shown to 'be significant to public health and safety by either operating experience or PSA. While the operating flow paths are not modeled in the Wolf Creek IPE, they are similar to the shutdown flow paths, which have been modeled in the Wolf Creek IPE and which have extremely low risk values. Therefore, this TS does not satisfy Criterion 4. ,

(4) CONCLUSION His Technical Specification is retained.

,X. He Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

r 3120 dos i

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1 TECIINICAL SPECIFICATION SCREENING FORM (1) TECllNICAL SPECIFICATION 31.2.3 CfIARGING PUMPS - SHUTDOWN l Applicable Modes: 4,5, and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO*, the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION ne Bases for this 1.C0 state that the purpose is to assure negative reactivity control is available during each Mode of facility operation. Equipment required to perform this function includes: (1) borated water sources, (2) CCPs, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power source from the EDGs.

The purpose of the boration subsystem of the CVCS is to provide the means to control the boron concentration to maintain SDM. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of an assumed boron dilution event, the automatic response of the BDMS, or that required of the operator, is to close the appropriate valves in the reactor makeup system before the shutdown margin is lost. Automatic actuation of the boration subsystem is not assumed to mitigate the event. Ref. 5 notes that the normal capability to control reactivity with boron is not credited in the accident analysis. Shutdown Margin (SDM) requirements provide suflicient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. He SDM .

defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. 'During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion. When the unit is in the shutdown and refueling modes, the SDM requirements are met -

by means of adjustments to the RCS boron concentration.

The boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB; therefore, this TS does not satisfy criterion L

He boration subsystem TS is not associated with a process variable that-is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. As stated in the analyses of boron dilution events, the BDMS

' performs automatic actions in response to detecting an assumed boron dilution event. Rese actions are credited for events occurring in Modes 3,4, and 5. He actions include providing an alarm, automatically isolating the dilution flow path, and .

automatically initiating boration of the RCS from the RWST via the charging pumps.' As stated in Ref. 3 for these events, the primary success path for mitigation includes isolating the dilution tiowpath. He subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. His is desirable, but beyond the scope of a primary success path action. Herefore, this TS does not satisfy criterion 2.

The boration subsystem TS does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the 13 does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the RWST, and associated flowpaths is required as part of the ECCS TS.

This TS also senes to prevent a cold overpressure event from occurring by limiting the number of operable CCPs to one in Modes 4,5, and 6 except when the reactor vessel head is removed. His restriction is part of an initial condition for the cold overpressure analysis. The specific SR that imposes this restriction will be incorporated into TS 3/4.5,4 for Modes 5 and 6. '

TS 4.5.3.2 addresses Mode 4..

r From Reference 2 the charging pumps, used to inject borated water into the RCS to maintain SDM, have not been shown to be .

significant to public health and safety by either operating experience or PSA. He charging pumps have been modeled in the 1,TS fault tree which is a top event for the ATWS initiating event. The core damage values for the ATWS event are extremely low, well below the NRC screening value. In addition, WCOS IPE comparison with NUMARC 93 01 Section 9.3.1, using both the Risk Reduction and Risk Achievement methods, demonstrated that the CCPs are not risk significant.

Herefore, this TS does not satisfy Criterion 4.

(4) CONCLUSION 1 His Technical Specification is retained.

SR 4.1.2.3.2 will be retained and incorporated into TS 3/4.5.4.

1 The Technical Specification may be relocated to the following controlled document (s): .

USAR Chapter 16.

>w u.

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i TECIINICAL SPECIFICATION SCREENLNG FORM (1) TECllNICAL SPECIFICATION 3.12 4 ClIARGING PITMPS - OPERNITNG Applicable Modes: 1,2, and 3 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant -

abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the faihtte of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical' Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled doounent.

(3) DISCUSSION 1he Bases for this LCO state that the purpose is to assure negative reactivity control is available during each Mode of facility operation. The equipment required to perform this function includes: (1) barated water sources, (2) CCPs, (3) separate Dow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from operable EDGs.

Ref, 3 states that the purpose of the boration subsystem of the CVCS is to provide the means to control the boron concentration to maintain SDM. The boration subsystem is not assmned to operate to mitigate the consequences of a DBA or transient, in the case of an assumed boron dilution event in Mode 3, the automatic response of the BDMS, or that required of the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Automatic actuation of the horation subsystem is not assumed to mitigate the event. In Modes 1 and 2, a dilution event is initially mitigated by the RTS and the reactor is shut down by insertion of the control rods. Continued dilution will tend to take the reactor critical; however, the operator has more than 30 minutes to stop the dilution Dow. Ref. $ notes that the normal capability to control reactivity with boron is not credited in the accident analysis. Shutdown Margin (SDM) requirements provide sunicient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM deGnes the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6. for rod insertion. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

13ased on the foregoing, the boration subsystem TS is not associated with installed instrumentation used to detect or indicate a significant degradation of the RCPil; therefore, this TS does not satisfy criterion 1.

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The boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an in iial condition of an event that assumes failure of or challenges the iotegrity of a fission product barrier. As stated in the analyses of boron dilution events, the BDMS performs automatic actions in response to detecting an assumed boron dilution event.

These actions are credited for events occurring in Mode 3.1he actions include providing an alann, automatically isolating the dilution flow path, and automatically initiating boration of the RCS from the RWST via the charging pumps. For these events, the primary success path for mitigation includes isolating the dilution flowpath. .The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM. Although desirable, this is beyond the scope of a primary success path action. In Modes 1 and 2, the operator is required to isolate the ddution flow path

' subsequent to a reactor trip. Therefore, the boration subsystem is not required to be operable to mitigate these events, and the -

TS does not satisfy criterion 2.

The boration subsystem is not a system that is part of the primary success path and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; therefore, the 15 does not satisfy criterion 3. Ref. 3 also notes that operability of the charging pumps, the RWST, and associated flowpaths is required as part of the ECCS TS.

For the MSLB event, the sequence of events takes the plant to cold shutdowTt conditions and, therefore, boration of the RCS is necessary. Ilowever, the boration flowpath in this case is required as part of the ECCS function.

From Reference 2, the charging pumps, used to inject borated water into the RCS to maintain SDM, have not been shown to ,

be significant to public health and safety by either operating experience or PSA. While the charging pumps have not been ,

modeled for this application in the WCGS IPE, the application is similar to the shutdown mode application which is included in the NIWS model. The core damage values for the ATWS event are extremely low, well below the NRC cutoff values. In addition, the WCGS IPE comparison with NUMARC 93-01 Section 9.3.1, using both the Risk Reduction and Risk- '

Achievement methods, demonstrated that the CCPs are not risk significant. Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION

_. This Technical Specification is retained.

2. The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

1824 duc l

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 3.1.2.5 UORATTD WATER SotrRCE. SiltrfDOWN Applicable Modes: 5 and 6 (2) EVALUATION OF POLICY STATEMENT CRITERIA ,

is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature. or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment .

has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Tecimical Specification may be relocated to a controlled document.

(3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each Mode of facility operation. Equipment required to perform this function includes, depending on operating conditions, a combination of: (i) borated water sources, (2) CCPs, (3) separate flow paths, (4) bonc acid transfer pumps, and (5) an emergency power source from the EDGs.

The purpose of the boration subsystem of the CVCS is to provide the means to control the boron concentration to maintain SDM.1he boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of -

an assumed boron dilution event, the automatic response of the BDMS, or that required of the operator, is to close the appropriate valves in the reactor makeup system before the shutdown margin is lost. Automatic actuation of the boration subsystem is not assumed to mitigate the event. Ref. 5 notes that the normal capability to control reactivity with boron is not credited in the accident analysis.

Shutdown Margin (SDM) requirements provide suflicient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences. The SDM defines the degree of suberiticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. During power operation, SDM control is ensured by operating with I the shutdown banks fully withdrawn and the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion. ;

When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS ;

boron concentration.

1he boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of j the RCPH; therefore, this TS does not satisfy criterion 1.

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s He boration subsystem TS is not applicable to a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure of or challenges the integrity of a fission product barrier. Rus, this TS does not satisfy criterion 2.

The boration subsystem TS does not apply to an SGC that is part of the primary success path and functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. As stated in the analyses of boron dilution events, the BDMS performs automatic actions in response to detecting an assumed boron ,

dilution event. Rese actions are credited for events occurring in Modes 3,4, and 5 He actions include providing an alarm, .

automatically isolating the dilution tiow path, and automatically initiating boration of RCS from the RWST via the charging -

pumps. As stated in Ref. 3 for these events, the primary success path for mitigation includes isolating the dilution flowpath, ne subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM.

Although desirable, this is beyond the scope of a primary success path action. Therefore, this LCO does not satisfy criterion !

3. ;

Ref. 3 also notes that operability of the charging pumps, the RWST, and associated flowpaths is required as part of the ECCS .-'

TS. ,

from Reference 2, the borated water sources have not been shown to be significant to public health and safety by either operational experience or PSA. LER 482/90-025-00 was written to report that a common return line for the Safety injection pumps to the refueling water storage tank (RWST) had frozen. His could have caused the loss of the RWST, liowever, design changes have been implemented which will preclude this from occurring in the future. ne boric acid tank (BAT), due _ :i' to it's generally rugged, simple design, has a low failure probability. It was not included in the WCGS IPE, which in general, does not model shutdown. He RWST is not included in the WCGS IPE model for the boron dilution accident during shutdown, but is modeled for accident mitigation while operating. Due to it's simple design and passive function, the RWST has a low failure probability, well below the NRC screening criterion. Herefore, this TS does not satisfy Criterion 4.

(4) CONCLUSION nis Technical Specification is retained.

X, ne Technical Specification may be relocated to the following controlled document (s): ;

USAR Chapter 16.

3 M a15

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y TECllNICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 31.2 6 BORATED WATER SOtTRCES - OPERATING Applicable Modes: 1,2,3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in tl.e control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ _X_ (3) A structure, system, or component that is part of the primary success path and which functions or uctuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four nf the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The Bases for this LCO state that the purpose is to assure negative reactivity control is available during each Mode of facility operation. The equipment required to perform this function includes, depending upon operating conditions, combinations of; (1) borated water sources, (2) CCPs, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from operable EDGs. ,

The purpose of the boration subsystem of the CVCS is to provide the means to control the boron concentration to maintain SDM. The boration subsystem is not assumed to operate to mitigate the consequences of a DBA or transient. In the case of an assumed boron dilution event in Mode 3 or 4, the automatic response of the BDMS, or that required of the operator, is to close the appropriate valves in the reactor makeup system before the SDM is lost. Automatic actuation of the boration subsystem is not assumed to mitigate the event. In Modes 1 and 2, a dilution event is initially mitigated by the RTS and the reactor is shut down by insertion of the control rods. Continued dilution will tend to take the reactor critical; however, the operator has more than 30 minutes to stop the dilution 11ow and maintain SDM Shutdown Margin (SDM) requirements provide sullicient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for nonnal shutdown and anticipated operational occurrences. The SDM defines the degree of suberiticality that would be obtained inunediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn. During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and ~

the control banks within the limits of LCOs 3.1.3.5 and 3.1.3.6, for rod insertion. When the unit is in the shutdown and -

refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

Bef. $ notes that the normal capability to control reactivity with boron is not credited in the accident analysis. 'i i

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Based on the foregoing, the boration subsystem TS is not applicable to installed instrumentation used to detect or indicate a significant degradation of the RCPB, therefore, this TS does not satisfy criterion 1.

He boration subsystem TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of an event that assumes failure cf or challenges the integrity of a fission product barrier. Hus, the TS does not satisfy criterion 2.

He boration subsystem TS does not apply to a system that is part of the primary success path and which functions to mitigate a DDA or transient that either assumes the failure of or presents a challenge to the integrity of a tission product barrier. As stated in the analyses of boron dilution events, the DDMS performs automatic actions in response to detecting an assumed boron dilution event. Rese actions are credited for events occurring in Modes 3 and 4. He actions include providing an alarm, automatically isolating the dilution tiow path, and automatically initiating boration of RCS from the RWST via the charging pumps. For these events, the primary success path for mitigation includes isolating the dilution 11owpath. He subsequent actuation of equipment to establish a boron injection floupath is intended to regain the required SDM. Although desirable, this is beyond the scope of a primary success path action. In Modes 1 and 2, the operator is required to isolate the dilution flow path subsequent to a reactor trip. Derefore, the boration subsystem is not required to be operable to mitigate these events, and the TS does not satisfy criterion 3. ,

Ref. 3 also notes that operability of the charging pumps, the RWST, and associated flowpaths is required as part of the ECCS

'I S.

For the MSLB event, the sequence of events takes the plant to cold shutdown conditions and, therefore, boration of the RCS is I

necessary. Ilowever, the boration floupath in this case is required as part of the ECCS function.

From Reference 2, the borated water sources have not been shown to be significant to public health and safety _by either operational experience or PSA. LER 482/90-025-00 was written to report that a common return line from the Safety Injection pumps to the refueling water storage tank (RWST) had frozen. His could have caused the loss of the RWST. However, design changes have been implemented which will preclude this from occurring in the future. He boric acid tank (BAT), due to it's generally rugged, simple design, was not modeled in the WCGS IPE. The RWST is included in the WCGS IPE model but, due to it's simple design and passive function, has a low failure probability, well below NRC screening criteria. ,

Herefore, this TS does not satisfy Criterion 4.

(4) CONCLUSION uis Technical Specification is retained.

_X_ he Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16 3126 dec .,

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TECliNICAL SPECIFICATION SCREENING FORM (1) TEcliNICAL SPECIFICATION 3.1.3.3 POSITION INDICATING SYSlTMS - SilUTDOWN Applicable Modes: 3,4, and 5 with the Reactor Trip Dreakers Closed (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

I YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assmnes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ J_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is 'NO, the Tedmical Specification may be relocated to a controlled document.

(3) DISCUSSION Control rod position is used by the operator to verify that the rods are correctly positioned and to verify that the rods are inserted into the core following a reactor trip . Rod pt 4 tion is also use<l during a reactor startup.

Operability of the control rod position indicators is required te je mine control rod positions and thereby ensure compliance with the rod alignment and insertion limits. Rese red ai.gna s requirements are applicable during power operation to maintain power distribution limits. Rod insertion limits are requireu to maintain SDM during Modes 1 and 2. The Bases do not address the shutdown condition. The LCO requires that one position indicator be operable to determine the position of any rod not fully inserted. Rod position indication may be used duriy a control rod withdrawal event from shutdown condition, but it is not required to be operable as an initial condition (,1. aitigating signal.

He position indication system TS is not applicable to installed instrumentation used to detect and indicate in the control room significant abnormal degradation of the RCPB. Herefore, this TS does not satisfy criterion L The position indication system TS, for shutdown conditions, is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, the TS does not satisfy criterion 2.

Finally, the position indication system TS does not apply to an SSC that is part of the primary success path and which functions to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, the TS does not satisfy criterion 3.

From Reference 2, the control rod position indicating systems have not been shown to be significant to public health and safety by either operational experience or PSA. The Zion PRA study (Reference 5) included this system, and it was shown to not be risk significant for their plant. He system is not modeled in the WCGS IPE, but there is no indication that it would be identified us risk significant ifit were included in the WCGS IPE model. Herefore, this TS does not satisfy Criterion 4.

(4) CONCLUSION This Technical Ft ;cification is retained.

1 The Technical Specitication may be relocated to the following controlled document (s):

USAR Chtq)ter 16.

313) 6 4 J

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TECilNICALSPECIFICATION SCREENING FORM

- (1) TECilNICAL SPECIFICATION 113 4 ROD DROP TiMl!

Applicable Modes: I and 2 (2) EVALUATION OF POLICY STATEMENT CRITERIA la the Technical Specification applicable to:

YES NO X (1) installed instnanentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_. 1 (3) A structure, system, or component th t is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity er a tission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION 1he Bases state that this TS ensures the control rod drop times are consistent with the assumptions of the safety analyses.

Therefore, the drop time may be considered a variable that is an initial condition of several events that could present a challenge to a fission product ba: Tier. Ilowever, this parameter cannot be monitored, controlled, or maintained within the bounds of the safety analysis by the plant operators. Also, this parameter is one that contributes to the definition of an operable control rod, however, rod drop time is not used to define an operable control rod during plant operation in Modes 1 and 2.

Ref. 3 determined that this specification is not installed instrumentation used to detect, and indicate in the control room, a significant abnonnal degradation of the RCPB. Nor is it an SSC that is part of the primary success path and which functions to mitigate any event. Ref. 3 also stated that Rod Drop Time is a variable that is an initial condition of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Ref. 4 concluded that this LCO may be relocated but the associated SR should be relocated to a retained LCO if the SR is necessary to meet the operability requirements of an LCO. Ref. 2 has relocated this LCO but included the rod drop time limit and conditions required for measuring it as an SR under an LCO for rod alignment.

'the rod drop time TS is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a l significant abnormal degradation of the RCPB. . lherefore, this TS does not satisfy criterion 1.

The rod drop time TS is associated with a design feature (rod insertion time) that is an initial condition of a DBA or transient : ,

analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. IIowever, rod J drop time is not a parameter that is maintained, during plant operations, within the bounds assumed in the accident analyses.

Therefore, this TS does not satisfy criterion 2.

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The rod drop time TS does apply to an SSC (operable control rod) that is part of the primary success path and which ftmetions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Ilowever, this parameter is not used to define an operable control rod during plant operation. Therefore, this 13 does not satisfy criterion 3.

From Reference 2, the rod drop time TS has not been shown to be significant to public health and safety by either operational experience or PSA. While not modeled directly in the WCGS IPE, the A1WS event which is a part of the WCGS IPE model, would provide a conservative approximation of the excessive rod drop time scenario. The NIWS event has a worst case coremelt sequence frequency well below the NRC cutoff value. Therefore, this TS does not satisfy Criterion 4. !

(4) CONCLUSION 1 1his Technical Specification is retained.

Rod drop time and plant conditions for measurement will be relocated as an SR under LCO 3.1.3.1.

1 1he Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

nu s.

TECIINICAL SPECIFICATION SCREENING FORM 4

k

. (1) TECllNICAL SPECIFICATION 3 3.3.2 MOVABI.E INCORE DETECTORS Applicable Modes: Refer to new USAR Section 16.3.1.1 (2) EVALUATION OF POLIC'l STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instnunentation that is used to detect, and indicate in the control room, a significant abnonnal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design llasis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product banrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component w hich operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES" then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Techtical Specification may be relocated to a controlled document.

(3) DISCUSSION His LCO requires the movable incore detectors to be operable, within dermed conditions, whenever the system is used for recalibration of excore detectors, monitoring the quadrant power tilt ratio, or measurement of Fq and F-Delta !!. If the system is not operable, the required action is not to use the system for these purposes. He requirements for maintaining Fa and F- ,

Delta 11 w ithin limits are addressed in the TS for power distribution limits.

Ref. I states that the operability of the movable incore detectors ensures the accurate measurement of spatial neutron flux distribution of the core.

Ref. 3 notes that the movable incore detector system is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB. Also, the system is not a process variable, design feature, or - ,

operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Further, the movable incore detector system is not an SSC that is pan of ;;

the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a cnallenge to the integrity of a fission product barrier.

The movable incore detector TS is not applicable to installed instrumentation that is used to detect and indicate in the control :

room, a significant abnormal degiadation of the RCPB.

ne movable incore detector TS is associated indirectly with an operating restriction (fiux distribution limits) that is an initial ;

condition of a DD/. or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barries. How ever, this operating restriction is maintained by other 13 requirements.

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The movable incore detector TS does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DIIA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

From Reference 2 the movable incore detectors have not been shown to be significant to public health and safety by either operational experience or PSA. The detectors are used only for periodic surveillance of the core power distribution and for calibration of the excore detectors and do not initiate any automatic protection action. The detectors are not modeled in the WCOS IPE.

Ilased on the above, the 1,C0 does not satisfy criteria 1,2,3 or 4.

(4) CONCLUSION Tids Technical Specification is retained.

1 The Technical Specification may be relocated to the following controlled document (s):

l'l-USAR Chapter 16.

D12 doc

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 3.3.3.3 SDSMIC INSTRUMENTNflON Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO

_ .X. (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant '

abnormal degradation of the reactor coolant pressure boundary.

_, 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

,X. (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or ,

presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown 'o be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled '

document.

(3) DISCUSSION lhe TS Bases state that the seismic monitoring instruments are to determine the magnitude of a seismic event so that the measured response of the plant can be compared to the response used in the design basis and determine if a shutdown is required in accordance with 10 CFR 100. The occurrence of a scismic event would represent a challenge to fission product barriers. Ilowever, the ability of the plant to withstand an SSE is a design requirement. The seismic monitoring instrumentation performs no role in mitigating a seismic event or in achieving a safe shutdown condition after a seismic event has occurred.

Ref. 3 determined that the seismic instrumentation is not installed instrumentation that is used to detect degradation of the RCPB. Seismic instrumentation is not assumed to function in the safety analysis and is not an SSC that is part of the primary success path and which function or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The seismic instrumentation TS is not applicable to installed instrumentation that is used to detect and indicate in the control room, a significant abnonnal degradation of the RCPB.

The seismic instrumentation is not associated with a process variable, design feature, or operating restriction that is an initial -

condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The seismic instrumentation TS does not apply to an SSC that is part cf the primary success path and which functions or -

actuates to mitigate a DBA or transient that either assumes the faihe. of or presents a challenge to the integrity of a fission product barrier.

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1 From Reference 2, the seismic instrumentation has not been shown to be significant to public health and safety by either operational experience or PSA.~ The seismic instrumentation is not designed to monitor seismic events that are of sufficient severity to be a dominant plant risk. The seismic instrumentation is not modeled in the WCGS IPE.

Based on the above, the seismic instrumentation requirements do not satisfy criteria 1,2,3, or 4.

. (4)- CONCLUSION

_ This Technical Specification is retained.

_X_ The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

331) dos

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e TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.3 3.4 METEOROLOGICAL INSTRUMENTATION Applicable Modes: At all times (1) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instmmentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initist condition of a Design .

Basis Accident or Transient analysis that either assumes the failure of on presents a challenge to the integrity of a fission product barrier.

,X_ (3) .A structure, system, or component that is part of the primary success path and which functions or '

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or

. presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION ne meteorological instrumentation ensures that data is available to estimate potential radiological doses to the public from accidental or routine releases of radioactive materials to the atmosphere. He instmmentation is used to assess the need for recommending protective measures following an accident. He meteorological instrumentation is not used to mitigate a DBA or transient.

Ref. 3 evaluated this instrumentation and concluded that it is not installed instmmentation that is used to detect degradation of the RCPB. Neither is it assumed to function in the safety analysis and is not an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either nsumes the failure of or presents a challenge to the integrity of a fission product barrier.

He meteorological instrumentation TS is not applicabic to installed instrumentation that is used to detect and indicate in the control room, a significant abnormal degradation of the RCPB. Therefore, this TS does not satisfy criterion L ne meteorological instrumentation TS is also not associated with a process variable, design feature or operating restriction '

that is an initial condition of a DBA or transient amdysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, this TS does not satisfy criterion 2.

He meteorological instrumentation TS does not apply to an SSC that is part of the primary success path and which functions I or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission l product barrier. Herefore, this TS does not satisfy criterion 3.

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From Reference 2, the meteorological instntmentation has not been shown to be significant to public health and safety by either operational experience or PSA. Offsite dose calculations for large accidental releases of radioactive materials rely'on conservative meteorological and evacuation assumptions and do not take credit for the meteorological instruments cited in this TS to guide emergency measures to protect the public. Finally, per Reference 1, no severe radioactive releases per the PWR Release Category 4, as defined in WASil 1400 (Reference 6) were found to exist at WCGS.

13ased on the above, the meteorological instrumentation does not satisfy criteria 1,2, 3, or 4. This is consistent with the NRC's conclusion in Ref. 4.

(4) CONCLUSION i

, _. 'This Technical Specification is retained.

1 1he Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

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l TECIINICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 3 3 3 6 ACCIDENT MONITORINGINSTRIPMENTA110N Applicable Modes: 1,2, and 3 ;

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

A structure, system, or component that is part of the primary success path and which functions or -

(3) actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ,

t (4) A stnicture, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Re instrumentation that satisfies criteria 3 or 4 are the Type A variables in USAR Appendix 7A, as well as the a risk-significant variables listed in the discussion below. Some of the TS 3/4.3.3.6 instruments may be relocated to USAR Chapter 16. Others must be retained in the Technical Specifications. He Neutron Flux Monitors and the Reactor Vessel Water Level Indicating System (RVLIS) will be added to the Tecimical -

Specifications.

If the answer to any one of the above questions is "YES" then the Technical Specification shall be retained in the Technical Specifications. l If the answer to all four of the above questions is "NO*, the Technical Specification may be relocated to a controlled [

document.

(3) DISCUSSION Operability of the accident monitoring instrumentation ensures that sufficient infbrmation is available on selected plant parameters to monitor and assess these variables following an accident. He instrumentation allows the operator to verify the response of automatic safety systems and to take preplanned manual actions to accomplish a safe plant shutdown.

The accident monitoring instrumentation is not intended to be a leading indicator of RCS leakage Although accident monitoring instruments respond to the consequences of a LOCA, the instruments captured by criterion 1 are those that are intended to prevent a LOCA from occurring and to give some indication of RCS leakage prior to the LOCA. Therefore, accident monitoring instrumentation TS is not applicable to installed instrumentation that is used to detect, and indicate in the ~

control room, a significant abnormal degradation of the RCPB and does not satisfy criterion 1. ,

l Accident monitoring instrumentation is not associated with a process variable, design feature, or operating restriction that is -

an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Although some variables that are accident monitoring instruments may also establish initial conditions at the time of a DBA or transient (for example, pressurizer level), the post-event function is separate and distinct from the pre-event function. Herefore. the accident monitoring instrumentation TS does not satisfy criterion 2.

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Specific accident monitoring instmmentation provides the operator with information needed to perform the required manual actions to bring the plant to a stable condition following an accident. His instrumentation is a component of the primary success path and functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, specific accident monitoring instrumentation satisfies criterion 3.

Ref. 4 states that accident monitoring instrumentation that satisfies the definition of Type A variables in Regulatory Guide 1.97 satisfies criterion 3 and should be retained in the TS. Ref. 4 also states that non-Type A, Category 1 instruments are to be evaluated for inclusion in the TS based on results of risk analyses. In accordance with USAR Table 7A-2, the instruments .

that are either (1) Type A or (2) non-Type A, Category 1 are:

INSTRUMENT TS 1YPE A CA1TGORY Neutron Flux !

Core Exit Temperature Yes 1 Reactor Vessel Level 1 RCS T-Cold Yes Yes 1 RCS T-Ilot Yes Yes 1 RCS Pressure Yes Yes 1 Pressurizer Level Yes Yes 1 RWST Level Yes Yes 2 Steam Generator Level-Wide Range Yes 1 Steam Generator Level-Narrow Range Yes Yes 1 Steam Line Pressure Yes Yes 1 Condensate Storage Tank Level (Pressure) 1 Containment Pressure Yes Yes 1 Containment Pressure. Extended Range Yes 1 Containment Normal Sump Level Yes Yes 1 Containment RilR Sump Level 1 Containment Isolation Valve Position 1 Containment Hydrogen Concentration Yes 1 Containment Area Radiation Yes Yes 1 Radiation Level in RCS (Sampling System) 1 Auxiliary Feedwater Flow Rate Yes 2 PORV Position Indicator Yes 2 PORV Block Valve Position Indicator Yes N.A.

Safety Valve Position Indicator Yes 2 Unit Vent-Iligh Range Noble Gas Monitor Yes 2 he Type A variables satisfy criterion 3 and will not be relocated. The Non-Type A variables are evaluated in the following paragraphs.

1. Neutron Flux-Source Range Neutron flux is a R.G.1.97 Category 1, Type B variable. In the emergency operating procedures (EOPs), neutron fiux is the specified means to verify reactor subcriticality and is to be monitored during EOP usage. Indication of significant post-trip power generation results in entry into a function Restoration Procedure (IRP) designed to ensure adequate shutdown reactivity. Based on the significance of this variable in the EOPs, neutron ilux will be incorporated into the TS.

2. Core Exit Temperature his indication is important for determining inadequate core cooling and will be retained.

I

3. Reactor Vessel level ne EOPs make use of a number of reactor vessel level indicating system (RVLIS) setpoints related to RCS inventory control and indication ofinadequate core cooling. Rese include

a. Indication ofinadequate core cooling

b. An alternate to RCS subcooling and pressurizer level as a safety injection initiation criterion.

c. A means of controlling charging flow if pressurizer level indication is not available.

d. A means of determining if RIIR operation will be effective based on collapsed liquid level, ne detection ofinadequate core cooling represents a potential near term breach of the fuel cladding integrity. Use of reactor vessel level indication in the EOPs includes events that are DBAs. Based on this information, reactor vessel level will be incorporated into the TS.

4. Steam Generator Level-Wide Range Steam generator level-wide range is used in the EOPs as an indicator of steam generator (SG) dryout and as a criterion for establishing feed and bleed cooling of the RCS. Loss of SG level does not, in and of itself, represent an approach to a breach of a fission product barner. His instrumentation does provide information required to .

perform a manual action which preserves a critical safety function (heat sink). He steam generator wide range level monitors are modeled in the operator actions OPA-OFB and OPA-OFC, operator feed and bleed action, which -

are used in the small LOCA, SGTR, secondary break and the TRA, TRO and LSP event trees. He most probable core damage sequence containing this operator action has a frequency of 1.4E-09/ year, which is below the NRC signiticance criteria. Ilowever, due to the above discussion, these indicators will be retained

5. Condensate Storage Tank Level (Pressure)

This variable is R.G.1.97 Category 1 if needed to ensure water supply for the ABV system. Ilowevers it may be Category 3 if the CST is not the primary source of supply. He primary source of ARV supply is the ESW system, and the ARV pump suction lines are automatically transferred to the ESW system upon loss of CST level as l

indicated by low pressure in the pump suction lines. Since there is no manual action required for switchover to the alternate source of auxiliary feedwater (ESW system), the CST levd measurement is not a Type A variable. He condensate storage tank level indicator has not been shown to be significant to public health and safety by either operational experience or PSA. It is not included in the WCGS IPE model. Herefore, the CST level (suction pressure) need not be added to the TS.

6. Containment Prenure-Extended Range R.G.1.97 defines the purpose of this variable as " detection of potential for or actual breach; accomplishment of mitigation". He EOPs do not base any decisions or actions on this variable. All actions related to containment pressure are based on the normal range containment pressure indication which is a Type A variable. Extended range pressure is not required to take appropriate actions to ensure the integrity of any fission product barrier. He extended range containment pressure indicators have not been shown to be significant to public health and safety by '-

either operational experience or PSA; Acy are not included in the WCGS IPE model. His TS does not satisfy criterion 3 or 4 and will be relocated.

i i

7. Containment RHR Sump level nis parameter is not a Type A variable. It is a Type B, Category I variable. The Containment Normal Sump Level is a Type A, Category 1 variable that will remain in the TS. Although the RIIR smnp level could be used for event identification, it is not required and would not be flooded with water immediately following an event since there is a curb around it. Also, since switchover to sump recirculation is automatic, verification of water level is not required nor part of a preplanned manual safety function. He containment RHR sump level indicators have not been shown to be significant to public health and safety by either operational experience or PSA. Hey are not included in the WCGS IPE model. Herefore, the RilR sump level instrumentation will not be added to the TS.

8. Containment holation Valve Position

~

he EOPs make use of this indication as part of an immediate response to a reactor trip with safety injection -

actuated. He operator is directed to confirm containment isolation, for both Phase A and B signals, as an immediate response to any safety injection. If the position indication shows any valves to be open, then the operator is directed to close them. Failure of this indication or failure of the operator to manually close an open valve could result in a release path for radioactive materials to the environment. Ilowever, a double failure would have to occur which is not a DDA requirement. For DBAs, there are Type A variables (cc.ntainment pressure, normal sump level, containment radiation) which provide the operator with information required to perform actions which ensure the containment integrity critical safety function during a DBA. While containment isolation is an important aspect of the containment analysis, indication ofisolation valve position has not been shown to b- eignificant to p % nealth .

and safety by either operational experience or PSA. Containment isolation valve position indication is not included 6 in the WCGS IPE. Therefore, this instrumentation will not be added to the TS.

9. Containment Ilydrogen Concentration Level TS 3/4.6.4, Combustible Gas Control, which requires the operability of the containment hydrogen analyzers, is .

evaluated on the TS Screening Form for LCO 3.6.4.1. In accordance with that form and the Safety' Evaluation, Attachment 1, LCO 3.6.4.1 will be deleted, since it is redundant to LCO 3.3.3.6 and is obsolete per the STS.

10. Radiation Levelin RCS R.G.1.97 defines the purpose of monitoring this variable as detection of breach (of the fuel cladding). He EOPs do not base any decisions or actions on this variable. His variable is not required to assure the integrity of any fission product barrier. His TS does not satisfy criterion 3. He radiation level in the RCS is an important indicator of a :

fuel cladding breach; however, it has not been shomt to be significant to public health and safety by either operational experience or PSA, and thua does not satisfy criterion 4. He RCS radiation level indication has not -

been modeled in the WCGS IPE. For these reasons, this variable need not be incorporated into TS.

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11. Auxillary Feedwater Flow Rate ne auxiliary feedwater flow rate indicator has not been shown to be significant to public health and safety by either operational experience or PSA. It is not included in the WCOS IPE model. However, the AFW Ilow rate indication should be retained for several reasons. First, it is included in NUREG-1431 Table 3.3.31. Second, SR 4.7.1.2.1 requires AFW flow rate indication. nird, AFW flow rate indication is being retained in TS Table 3.3 9 for the auxiliary shutdown panel (ASP). If an LCO and SR for the ASP AFW flow rate indication are being retained, then they should also be retained in TS 3.3.3.6.

12. PORY and PORV Block Valve Position Indicators PORV and PORV block valve position indicators have been deleted from Technical Specification 3.3.3.6. Loss of position indication requires that the Actions associated with LCO 3.4.4 be entered; therefore, there is no need to also have these indicators under LCO 3.3.3.6. It is further noted that these indicators are not Type A variables at Wolf Creek, nor are they RG 1,97 Category 1. Monthly channel checks for these indicators have been added as SR 4.4.4.3 and SR 4.4.4.4.

13. Safety Valve Position Indicator his instnunent is not a Type A or Category 1 indication. He safety valve position indicators have not been shown to be significant to public health and safety by either operational experience or PSA, and they are not included in the WCOS IPE model. This instrument is a Type D, Category 2 variable and will be relocated.

14. Unit Vent -Illgh Range Noble Gas Monitor nis instrument is not a Type A or Category 1 indication. De unit vent - high range noble gas monitor has not been shown to be significant to public health and safety by either operational experience or PSA. It is amt included in the WCGS IPE model. The WCOS Emergency Plan also has provisions to issue olisite Pn>tective Action Recommendations based on plant condition. Indication from this monitor would not be required to make those determinations. His is a Type D, Category 2 variable and will be relocated.

(4) CONCLUSION ,

This Technical Specification is retained.

'As indicated above; Neutron Flux and RVLIS will be added.

" ne Technical Specification may be relocated to the following controlled document (s):

"USAR Chapter 16 (Containment pressure Extended-range;' Safety valve position indicator, Unit vent wide-range nobic gas monitor).

JORV and PORV Block Valve position indicators have been deleted from LCO 3.3.3.6 as discussed above.

3316 dne

i TECilNICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.3.3.9 LOOSE.PART MONITORINGINSTRUMENTATION Applicable Modes: 1 and 2 (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: ,

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design -

Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which fimetions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION ne Loose Part Detection Instrumentation provides the capability to detect loose parts in the RCS which could cause damage to some component in the RCS. Loose parts are not assumed to initiate any DBA, and the detection of a loose part is : lot required for mitigation of any DBA.

The Loose Part Detection System TS is not associated with installed instrumentation used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He Loose Part Detection Instrumentation TS is not applicable to a process variable, design feature, or operating restriction that is an initial condition of any DBA or transient analysis. nus, this TS does not satisfy criterion 2.

The Loose Part Detection Instrumentation TS does not apply to any SSC assumed to function in the safety analysis. It is not part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this TS does not satisfy criterion 3.

From Reference 2, the Loose-Part Monitoring Instrumentation has not been shown to be significant to public health and safety by either operational experience or PSA. Loose parta would not be expected to damage the RCS pressure boundary or afTect initiating event frequencies or PRA results. He Loose.Part Monitoring System is not modeled in the WCGS IPE. Herefore, this TS does not satisfy criterion 4. 4 1

-i (4) CONCLUSION nis Technical Specification is retained.

X He Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16 3319 dos e

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(1) TEC11NICAL SPECIFICATION 3 3.3.11 EXPLOSIVE GAS MONITORINGINSTRUMENTATION' I Applicable Modes: During Waste Gis IIoldup System operation l i

1 (2) EVALUATION OF POLICY STATEMENT CRITERI A i Is the Technical Specification applicable to: )

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

._ 1 (3) A structure, system, or component that is part of the primary success path and which ftmetions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Tecimical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION ne explosive gas monitoring instrumentation provides the capability to detect the concentration of oxygen and hydrogen in the waste gas holdup system (at the hydrogen recombiners) and provide an alarm if the concentrations exceed prescribed limits. According to LCO 3.3.3.11, this TS assures the operability of the instrumentation required for LCO 3.11.2.5, Explosive Gas Mixture of the Radioactive EfDuents TS. According to the Bases of LCO 3.11.2.5, the purpose of the limits on explosive gas concentrations and the monitoring instrumentation is to prevent an explosion in the waste gas holdup system.

(The Bases for 3.3.3.1I were deleted in Operating License Amendment No. 42.) An explosion could result in a release of radioactive materials contained in the gaseous waste holdup system. Although release of the contents of a waste gas decay tank is an analyzed DBA, the analysis assumes that the tank ruptures non-mechanistically and not as the result of a hydrogen explosion. Herefore, the explosive gas limits are not an initial condition of a DBA.

He explosive gas monitoring instrumentation is not applicable to installed instrumentation used to detect, and indicate in the control room, a significant abnonnal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He explosive gas monitoring instrumentation is not applicable to a process variable, design feature, or operating restriction that is an initial condition of any DBA or transient analysis. Hus, this TS does not satisfy criterion 2.

He explosive gas monitoring instrumentation is not assumed to function in the safety analysis. It is not a part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Rus, this TS does not satisfy criterion 3.

From Reference 2, the Explosive Gas Monitoring Instrumentation has not been shown to be sigruficant to public health and safety by either operational experience or PSA. He function of this instrumentation is to preclude inadvertent radioactivity releases fram the Waste Gas lloldup System due to a tank failure from a waste gas explosion. Severe accidents dominate public risk, not inadvertent releases. His system is not modeled in the WCGS IPE. Hus, this TS does not satisfy criterion 4.

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(4) CONCLUSION lids Tecimical SpeciScation is retained.'

.X_ The Technical SpeciDeation may be relocated to the following controlled document (s):-

USAR Chapter 16. (The LCO will be relocated but a program statement will be added to new.

TS Section 615).

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.3.4 TURIIINE OVERSPEED PROTECTION Applicable Modes: 1,2, and 3 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Tecimical Specification applicable to:

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnonnat degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shon to be significant to public health and safety.

If the answer to any one of the above questions is "YES*, then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION He Turbine Overspeed Protection System actuates to mitigate a potential turbine overspeed event. His prevents the generation of potentially damaging missiles from the turbine. The tmbine overspeed event is not a DBA. His event is evaluated to determine the probability of damage to equipment needed for safe shutdown. He turbine has a favorable -

orientation from the standpoint of low trajectory missiles; however, the combination of overspeed probability with high -

trajectory strike probability must meet the NRC's requirements for overall probability, i.e., less than 1E-7 per year.

He Turbine Overspeed Protection System is not applicable to installed instrumentation used to detect, and indicate in the control room, a significant abnonnal degradation of the RCPB; therefore, this TS does not satisfy criterion L o

%e Turbine Overspeed Protection System is not associated with a process variable, design feature, or operating restriction-that is an initial condition of any DBA or transient analysis. Bus, this TS does not satisfy criterion 2.

He Turbine Overspeed Protection System is not assumed to function in the safety analysis. It does not apply to any SSC that - l is a part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the !

failure of or presents a challenge to the integrity of a fission product barrier, Rus, this TS does not satisfy criterion 3. -

From Reference 2, the turbine overspeed protection has not been shown to be significant to public health and safety by either operational experience or PSA. PRA studies discussed in Reference 2 indicate that the probability of turbine missile ejection and resultant damage to safety-related structures are so low that they have little or no impact on the quantification of core --

damage frequency. Due to the physical location and orientation of the turbine at WCGS in relation to the majority of safety. 1 I

related equipment, these low probability values will be valid. Turbine overspeed protection is not included in the WCOS IPE model. Bus, this TS does not satisfy criterion 4.

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his Technical Specification is retained.

.. 1 he Technical Specification may be relocated to the following controlled document (s)

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{> USAR Chapter 16. (1he LCO may be relocated but a program statement will be added to new TS Section 6.8.5).

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.4 2.1 SAFE 1Y VALVES - SillTmoWN Applicable Modes: 4 and 5 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO

_ X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

,,,, 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a chauenge to the integrity of a lission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or

, actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be signiticant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical t

Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION This TS is applicable to Modes 4 and 5. The safety valves, together with the reactor protection system, protect the RCS from being pressurized above its Safety Limit of 2735 psig. The pressurizer safety valves provide overpressure protection during both power operation and hot standby. Ilowever, the safety valves are not assumed *.o function to mitigate a DBA or transient in Modes 4 and 5. According to the Bases, only one safety valve is required to relieve any overpressure condition which could occur during shutdown. In the event no safety valves are operable during shutdown there are several other means to >

provide the required protection. for example, the RIIR relief valves in an operating RIIR loop connected to the RCS or the -

Overpressure trotection System, which relies on the Pressurizer PORVs, can provide the needed protection. Ref. 2 Bases note that overpressure protection during shutdown is provided by operating procedures and by meeting the requirements of the LCO for low temperature overpressure protection (LCO 3.4.9.3). LCO 3.4.9.3 is applicable when in Mode 3 and any RCS -

cold leg temperature is less than or equal to 368 *F and in Modes 4,5, and 6 with the vessel head installed.

The safety valve TS is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB. This TS does not satisfy criterion 1.

The safety valve TS is not associated with a process variable, design feature or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure or presents a challenge to the integrity of a fission product barrier. Thus, this TS does not satisfy criterion 2.

The safety valves are not assumed to function in the safety analy sis to mitigate overpressure transients in Modes 4 and 5. The pressurizer safety valve TS is not applicable to components that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assmnes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

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-1 From Reference 2, the operation of the pressurizer safety valves during shutdown has not been shown to be significant to public health and safety by either operational experience or PSA. A recent probabilistic study indicated that the frequency of experiencing an event leading to RCS pressure above 2485 psig while at shutdown conditions is less than 1.0E-10 per year.-

- As the WCOS IPE model does not include shutdown conditions, the pressurizer safety valves are not in the model.1herefore, this TS does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

J_ The Technical Specification may be relocated to the following controlled document (s):

f USAR Chapter 16.

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(1) TECilNICAL SPECIFICATION 3.4 5 STTAM GENERATORS Applicable Modes: 1,2,3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a Rssion product barrier.

_. 1 (3) A structure, system, or component that is part of the primary success path and which functions or -

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or .

presents a challenge to the integrity of a fissior product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment.

has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION nis TS establishes the inservice inspection requirements for the steam generator (SG) tubes which are part of the RCPU. It is intended to maintain the structural integrity of this portion of the RCPB, ne LCO requires the SGs to be operable in Modes I,2,3, and 4; operability in this case refers to the structural integnty of the SG tubes by means of an augmented inservice inspection (ISI) program that is performed periodically during plant outages.

His specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnonnal degradation of the RCPB; and, therefore, this TS does not satisfy criterion 1.

His specification is not applicable to a process variable or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. He -

specification is applicable to the design feature of SG tube strength which comes into play, for example, during a LOCA or MSLB to avoid a combined LOCNSGfR or MSLB/SGTR event. Ilowever, tube integrity is neither an active design feature i

nor monitored or controlled during plant operation, rather during shutdown conditions under the SG ISI program. Rus, the structural integrity and assumed passive post-accident performance of the SG tubes is maintained by periodic inspection. ,

nerefore, this TS does not satisfy criterion 2.

He SG tubes are components of the RCS that are part of the primary success path and which function to mitigate a DBA or transient that either assumes the failure of or presettts a challenge to the integrity of a fission product barrier. ne post-uccident or post-transient perfbrmance of the SGs, which is a passive function, is maintained by the periodic inspection and ')

repair of the SG tubes specified in this LCO. Ilowever, the operability of the SG tubes is not maintained during operation of. i the phmt through any actions performed or parameters monitored by the operating staff. Also, the SG tubes do not perfonn ;

any active function or actuation required for DBA or transient mitigation. Herefore, this TS does not satisfy criterion 3, I

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'l Reference 2 states that the steam generators have not been shown to be significant to public health and safety by either operational experience or PSA. Also, for WCGS, the SGTR initiated event centributes .less than 1.5% to the core damage

. frequency. WCGS does have three dominant (i.e. above 1.0E-07/ year) containment bypass sequences resulting from the SGTR initiated event. They are: 1. SGTR event AFW and cooldown fail; 2. SGTR event, failure to stabilize RCS and ruptured SG pressure, secondary side relief valve (RV) closes; and 3. SGTR event, failure to stabilize RCS and ruptured SG .

pressure, secondary side _ RV sticks open, cooldown fails. Ilowever, steam generator bypass is controlled by Tecimical Specification 3.4.6, " Leakage Detection Systems", which limits the leakage from all steam generators not isolated from the .,

RCS to 1 gallon per minute. This limitation assures that dosage contribution from the tube leakage will be limited to a small -

fraction of the 10CFR Part 100 dose guideline values in the event of a SGTR event. . Therefore, this 13 does not satisfy j Criterion 4.

Ref. 4 concluded that this LCO could be relocated out of TS but that the SRs must be retained. -

(4) CONCLUSION i _

This Technical Specification is retained.

.2;,, lhe Technical Specification may be relocated to the following controlled document (s):

The LCO may be relocated to USAR Chapter 16; however, a SG tube surveillance program statement will be added to new TS Section 6.8.5.

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 3 4.7 CilEMISITtY Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design -i Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or ,

presents a challenge to the integrity of a fission product barrier. !

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION !

His specification places limits on the oxygen, chloride, and fluoride content of the RCS to minimize corrosion of the RCPB.

He RCS chemistry TS is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnonnal degradation of the RCPU. He RCS chemistry TS does not satisfy criterion 1.

Chemistry restrictions are not used as initial conditions for safety analysis. liowever, the chemistry requirements are applicable, albeit indirectly, to a design feature (RCS integrity) that is an initial condition of a DBA or transient analysis that either assumes the failure or presents a challenge to the integrity of a fission product barrier. But RCS integrity is a passive rather than an active design feature, nus, the RCS chemistry TS does not satisfy criterion 2. ,

, t ne chemistry requirements for the RCS are applicable to the integrity of the RCS which is a system that is part of the primary success path and which functions nr actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Ilowever, the chemistry requirements do not directly assure the RCS integrity, but provide an indication of a concern. RCS integrity is assured through ISI and engineering evaluations of structural integrity. Herefore, the RCS chemistry TS does not satisfy criterion 3.

From Reference 2, RCS chemistry has not been shown to be significant to public health and safety by either operational experience or PSA. Primary system corrosion is a slow process which would be detected in inservice inspections or small leakages before it caused a rupture. Undetected corrosion would not be expected to have a significant affect on LOCA -

frequencies. RCS chemistry is not modeled in the WCGS IPE. Herefore, the RCS chemistry TS does not satisfy criterion 4.

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(4) CONCLUSION

. This Technical Specification is retained.

.X The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

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TECilNICAL SPECIFICATION SCREENING FORM (1) TECIINICAL. SPECIFICATION 3 4.9.2 P/r LIMITS - PRESSURIZER Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Teclutical Specification applicable to:

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be signincant to public health and safety, if the answer to any one of the abose questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO*, the Technical Speification may be relocated to a controlled document.

(3) DISCUSSION Pressure and temperature (P/T) limits are placed on the pressurizer (PZR) to be consistent with the requirements of the ASME Code. In accordance with the Bases, although the PZR operates in temperature ranges above those for which there is reason for concem of non-ductile failure, operational limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

He P/r limits are not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB. Herefore, this TS does not satisfy criterion 1.

De P/r limits are not applicable to a process variable or design feature that is an initial condition of a DBA or transient analysis that either assurnes the failure of or presents a challenge to the integrity of a fission product barrier. While the TS imposes operating restrictions, they are not associated with a DBA or transient analysis or with precluding the occurrence of an unanalyzed event but, rather, with maintaining fatigue cycles within approved limits. Herefore, this TS does not satisfy criterion 2.

He Pff limits are associated with an SSC that is part of the primary success path and which functions or actuates to mitigate .

a DilA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.- For example, the PZR must maintain its structural integrity following a MSLB or SDLOCA to maintain RCS circulation and cooling capability, llowever, the passive functional integrity of the PZR is not maintained by any activities of the plant stalT l during plant operation. Pressurizer integrity is a design feature maintained by ASME Code design and component cyclic / transient limit requirements imposed outside of this TS. nus, this TS does not satisfy criterion 3.

Prom Reference 2, the Pressurizer P/T limit technical specitication has not been shown to be significant to public health and safety by either operational experience or PSA. The consequences of a pressurizer failure due to operation outside the specified pressure / temperature limits are expected to be much less severe that of a reactor vessel failure, which has been shown in risk studies to not be a dominant risk. While the pressurizer P/T limits are not modeled in the WCGS IPE, they are included in the Modular Accident Analysis Program (MAAP) computer code which provides Level 1 success criteria and Level 2 releases. Thus, the TS does not satisfy criterion 4, (4) CONCLUSION

'lhis Technical Specification is retained.

X, - The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.410 SillUCT1 TRAL IN1T.GRilY Applicable Modes: All Modes

l (2) EVALUATION OF POLICY STATEMENT CRITERIA 1 Is the Technical Specification applicable to

YES NO X (1) Installed instnunentation that is used to detect, and indicate in the control room, a signiticant abnonaal degradation of the reactor coolant pressure boundary, X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

,_Xu (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION This specitication provides the inspection requirements for the ASME Code Class 1,2, and 3 components to ensure their -

structural integrity, 1his specitication is not applicable to installed instrumentation that is used to detect, and indicate in the control room,'a significant abnonnat degradation of the RCPD. Therefore, the structuralintegrity requirements do not satbfy criterion 1.

This specification is not applicable to a process variable, design feature, or operating restriction that is an initial condition of DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

While the TS imposes an operating restriction regarding pressure boundary integrity, it is not monitored or controlled during plant operation. The assumed integrity of Class 1,2, and 3 components is assured by means of periodic inspections.

Therefore, this TS does not satisfy criterion 2.

ASME Code Class 1,2, and 3 components are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of or present a challenge to the integrity of a fission product barrier. Individual ASME Code Class 1,2, and 3 components may satisfy criterion 3 and the requirements that ensure the integrity / operability of these components are included in the individual specifications that cover these components. Ilowever, as stated above, this specification addresses the passive, pressure boundary function of these components Therefore, this 13 does not satisfy .

criterion 3.

Ref. 4 concluded that the LCO for this specification could be relocated out of TS; however, the associated SR must be relocated to the TS programmatic requirements.

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1 from Reference 2, the structural integrity of ASME Code 1,2 and 3 components has not I,een shown to be significant to ,

public health and safety by either operational experience or PSA. Failure modes of these components would not be identified ,

from the requirements of this technical specification. The structural integrity of ASME Code 1,2 and 3 components is not 1 modeled in the WCOS IPE. Therefore, this TS does not satisfy criterion 4. )

(4) CONCLUSION This Technical Specification is retained.

1 1he Technical Specification may be relocated to the following controlled document (s)!

USAR Chapter 16. (The 1.C0 may be relocated but a program statement will be added to new TS Section 6.8.5).

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l TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.4.11 REACTOR COOL ANT SYSTEM VENTS Applicable Modes: 1,2, 3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specitication applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, sy stem, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component u hich operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION ne RCS vents are provided to exhaust, from the reactor vessel, noncondensable gases and/or steam from the RCS which could inhibit natural circulation core cooling following any event involving a loss of otTsite power and requiring long tenn cooling, such as a LOCA. Heir function, capabilities, and testing requirements are consistent v.ith NUREG-0737, Item II.D.1, which assumes a severely damaged core. Ilowever, the vents are not required to operats to mitigate any DBA or transient. Operation of the vents is not assumed in the safety analysis. His is because operation of the vents is not part of the prunary success path. Operation of the vents is an assumed operator action aller an event has occun si and is required only if '

there is indication that natural circulation is not occurring, ne TS requirements for RCS vents are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He RCS vents TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Aus, this TS does not satisfy criterion 2.

He TS for the RCS vents does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, the RCS vent requirements do not satisfy criterion 3.

From Reference 2, the reactor coolant system vents have not been shown to be significant to public health and safety by either operational experience or PSA. On Westinghouse designed PWRs, buildup of suflicient non-condensable gases or steam within the primary system to inhibit natural circulation is unlikely. Also, the contribution of inadvertent opening of the head -

vent valves to the small LOCA initiating event frequency is not a primary contributor to risk. He reactor coolant system vents are not modeled in the WCGS IPE. Herefore, this TS does not satisfy criterion 4.

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He Technical Specification may be relocated to the following controlled document (s): -1 X

USAR Chapter 16.

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TECHNICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3 6.1.2 CONTAINMENT LEAKAGE Applicable Modes: 1,2,3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: ,

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or ,

presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or con.ponent uhich operating experience or probabilistic safety' assessment has shown to be significant te public health and safety.

If the ansuer to any one of the above questions is "YES", then the Tecnnical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Tecimical Specification may be relocated to a controlled document.

(3) DISCUSSION His TS identifies the allowable leakage rates for the containment structure which are established to meet 10 CFR 50, Appendix J. Rese requirements ensure that the leakage rates from containment will not exceed the value assumed in the safety analyses at the peak accident pressure.

His specification is not applicable to installed instrumentation that is used to detect, and indicute in the control room, a significant abnormal degradation of a the RCPB; and, therefore, the TS does not satisfy criterion 1.

His specification is applicable '.o parameters that are an initial condition of a DBA or transient analysis that either assumes ,

the failure of or presents a challenge to the integrity of a fission product barrier, llowever, the process variables for which the requirements are applicable (containment design pressure and allowable leakage rates) are not variables that are monitored and controlled during power operation such that process values remain within the analysis bounds. Containment integrity is assured by periodic inupection and testing. Therefore, this specification does not satisfy criterion 2.

ne specification applies to containment leakage rate limits. Thus, it is applicable to a structure that is part of the primary success path and w hich function to mitigate a DBA or transient that either assumes the faihire of or presents a challenge to the integrity of a fission product barrier, llowever, the intent of criterion 3 is to capture only those SSC (and supporting systems) that are part of the primary success path of a safety sequence analysis. Operability of the containment is assured by a separate LCO (3.6.1.1), and the limits imposed by the leakage rate requirements are neither monitored or controlled during operation nor part of the primary success path of the containment function. A efore, this TS does not satisfy criterion 3.

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_. _ m 2t From Reference 2, containment leakage has not been shown to be significant to public health and safdy by either operational experience or PSA. PRAs indicate that risk is dominated by events in which the containment is b>pt ssed, unisolated, or fails ,

structurally. The technical specification value for overall containment leakage is included in the WCG5 MAAP model, but contributes only a small fraction of the total release in the Level 2 IPE.1herefore, this TS does not satisfy criterion 4.

Ref. 4 concluded that this LCO could be relocated out of TS but that the limiting values of Paand La must be retained in TS.

(4) CONCLUSION This Tecimical Specification is retained.

- The Technical Specification may be relocated to the following controlled document (s): *

' 1he LCO may be relocated to USAR Chapter 16, but the . limiting values of Pa and La will be retained in the Containment integrity Dases. Relocation of the LCO requires that revisions be made to SR 4.6.1.lc and SR 4.6.1.7.2.

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(1) TECllNICAL SPECIFICATION 16.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY Applicable Modes: 1,2. 3, and 4 (2) EVALUATION OF POLICY STATDIENT CRITERIA Is the Technical Specification applicable to:

YES NO

_ JL (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to -

the integrity of a fission product barrier.

r X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barder.

X A structure, system, or component which operating experience or probabilistic safety assessment i (4) has shown to be significant to public health and safety. ,

if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical ,

Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION ne containment series as a barrier to prevent the release oflission products following a LOCA or MSLB inside containment.

To mitigate the poential consequences of a DilA, it is necessary that the contaimnent structure meet its structural ,

requirements. His specification is intended to detect abnormal degradation of the containment structural elements. His TS outlines an appropriate inspection and testing program to demonstrate this capability, ne program consists of the measurement of tendon linoff force, tensile tests of tendon wires, and visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment.

This specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnomial degradation of a the RCPII; and, therefore, this TS does not satisfy criterion 1.

His specification is applicable to a design feature (the containment) that is an initial condition of a DBA or transient analysis ,

that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Containment structural integrity is assumed to be available for many DBAs. Ilowever, containment structural integrity is not monitored or controlled during plant operation but, rather, via periodic inspections and tests. Herefore, this TS does not satisfy criterion 2.

He specification upplies to the detection of abnormal degradation of containment structures and therefore to containment ,

structural integrity. nus, it is applicable to a structure that is part of the primary success path which functions to mitigate a ,

DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier Ilowever, the functional mode addressed by the 13 is maintaining the passive, pressure boundary integrity. His TS does not -

address the capability of the containment to function or actuate in order to mitigate the consequences of a DBA or transient.

Herefore, this 13 is not required to ensure the operability of containment and, thus, does not satisfy criterion 3.

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- Ref. 4 concluded that this LCO could be relocated out of TS but that the associated SRs should be retained to meet the operability requirements for a retained LCO, in this case LCO 3.6.1.1. ' Ref. 2 incorporated the SRs regarding tendon surveillance into Section 6 of the TS.

From Reference 2 containment leakage has not been shown to be significant to public health and safety by either operational .

experience or PSA. FRAs indicate that risk is dominated by events in which the containment is bypassed, unisolated, or fails structurally. None of the sequences addressed in the containment and' source term analysis could realistically threaten containment due to hydrogen combustion. No WCGS contaimnent vulnerabilities uere identified as a result of Supplement 3 - ,

to Generic Letter 88-20.

The WCGS IPE also showed that the overall release frequency per year is dominated by releases due to containment bypass sequences of which interfacing system LOCAs make up the vast majority, i

The best estimate containment failure mode will occur at 2.13 times the design pressura due to membrane stresses in the ;

containment mid-height region which exceed the pre-stress and cause through-concrete cracking and yielding of the liner, reinforcing steel, and pre-stressing tendons, lhe material properties used in these calculations do not change rapidly, so testing and inspection requirements of this technical specification are not critical. Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION 1his Technical Specification is retained. i

,1 The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (The LCO may be relocated; but a program statement will be added to new TS Section 6.8.5).

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TECHNICAL SPECIFICATION SCREENING FORM (1) TEcliNICAL SPECIFICATION 3.6.4.1 IIYDROGEN ANALYZERS (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ 1 (2) .A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to ,

the integrity of a fission product barrier X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes 'he failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A stmeture, system or component which operating experience or probabilistic safety assessment has shown to be sigrdficant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION He operability of the systems and equipment 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. He hydrogen monitoria system is initiated after the occurrence of a LOCA to detect the buildup of hydrogen. He purpose of the hydrogen control features is to maintain containment integrity by assuring that a hydrogen burn or explosion would not overpressurize containment.

He TS requirements for hydrogen analyzers are not applicable to installed instrumentation used to detect a significant abnonnal degradation of the RCPB; therefore, this 'IS does not satisfy criterion 1.

He hydrogen analyzers TS are not associated with a process variable, design feature, or operating restriction that is an initial ;

condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Rus, this requirement does not meet criterion 2.

He TS for hydrogen analyzers are applicable to an SSC (containment) that is part of the primary success path and which functions or actuates to mitigate a DBA or transien'. that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, these requirements satisfy criterion 3.

LCO 3.6.4.1 is deleted since it is redundant to LCO 3.3.3.6 and is obsolete per the STS.

(4) CONCLUSION 1 he requirements of this Technical Specification are retained under LCO 3.3.3.6..

The Technical Specification may be relocated to the following controlled document (s).

1 1 His Technical Specification is deleted. 1 1

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E TECliNICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 312 STEAM GENERATOR PTT LIMITATION Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis tlat either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment -

has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer 'o all four of the above questions is "NO", the Tecimical Specification may be relocated to a controlled document.

(3) DISCUSSION Pressure and temperature (P/T) limits are placed on the steam generators (SG) to prevent a non-ductile failure of either the -

RCPD or the secondary side pressure boundary. The specification places limits on the SG Prf to ensure that the pressure induced stresses are within the maximum allowable fracture toughness stress limits. The Pff limits are based on a SG RTm sufficient to prevent brittle fracture.

He SG Pil limits are not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPD. Herefore, the SG Pff limits do not satisfy criterion 1.

He Pff limits are not applicable to a process variable, design feature, or operating restriction that is an initial condition of DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

While the TS imposes an operating restriction, it is not employed to prevent unanalyzed accidents and transients. . Under the conditions when this TS could be required, an unanalyzed event of any significance from a safety function standpoint (decay heat removal, accident mitigation, and reactor shutdown) is unlikely to result. Herefore, this TS does not satisfy criterion 2.

He Pff limits are associated with an SSC that is part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. For -

example, the SG must maintain its structural integrity following a MSLD or SBLOCA to maintain'RCS circulation'and cooling capability. Ilowever, the TS limitations apply only to shutdown conditions when RCS temperature is imusually low (less than 70 *F). Under these conditions, the SG is not required to function to mitigate any DBAs or transients. .nerefore, this TS does not satisfy criterion 3.

From Reference 2, the steam generator pressure / temperature limitation has not been shown to be significant to public health and safety by either operational experience or PSA. His technical specification is intended to prevent brittle fracture of a SG when at low pressures and temperatures, something which is not likely during plant operation, which is the analyzed condition for the WCOS IPE study. This condition, then, is not modeled in the WCGS IPE. Herefore, the TS does not satisfy criterion 4.

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(4). ' CONCLUSION -

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_ This Technical Specincation is retained.

_X 1he Techniesi Specincation may be relocated to the following controlled document (s):

USAR Chapter 16.

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TECitNICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.7.8 SNUBBERS Applicable Modes: 1,2,3, and 4. Also Modes 5 and 6 for those systems required to be operable in these Modes.

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design .

Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a Hssion product barrier.

X (3) A structure, system, or component that is part of the primary success path and w hich functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presen'.s a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical -

Specifications.

If the answer to all four of the above questions is "NO", the Tecimical Specification may be relocated to a controlled document.

(3) DISCUSSION ne snubbers are required to be operable to ensure that the structural integrity of the RCS and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. He restraining action of the snubbers ensures that the initiating event failure does not propagate to other parts of the failed system or to other safety systems.

Snubbers also allow normal thermal expansion of piping and nozzles to eliminate excessive thermal stresses during heatup or ,

cooldown. Snubber surveillance is conducted under the requirements of the Wolf Creek Snubber Surveillance Program.

He TS requirements for snubbers are not applicable to installed instnunentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He snubber TS is associated with a design feature or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Ilowever, the snubber requirements are not explicitly considered in the accident analysis. The availability of the snubbers is assumed based -'

on the performance of a program of periodic augmented inspection and testing. Snubber operability is not required to be monitored and controlled during plant operation. Some snubbers (inaccessible) can only be inspected during plant outages.

Hus, this 'lS does not satisfy criterion 2.

Rose snubbers that are required to function during DBAs or transients to prevent the initiating event fn m propagating to other systems or components that are part of the primary success path may be considered components that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or ,

presents a challenge to the integrity of a fission product barrier. liowever. snubbers are not explicitly considered in DBA or i transient analyses but are a structural / design feature whose operability is assured by an inspection program Herefore, this !

TS does not satisfy criterion 3. j I

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From Reference 2. for non-RCS and other high energy systems such as the feedwater and main steam systems inside the containment building, the snubber technical specification has not been shown to be significant to public health and safety by either operational experience or PSA. Reference 2 reviewed the Zion and Millstone PRAs and determined that the snubbers,'

which ensure the operability of certain safety-related equipment during a seismic event, are not risk dominant for this scenario. While the seismic portion of the WCGS IPEEE is not complete at this time, there is no reason to believe the results will be different. Thus, this 13 does not satisfy criterion 4.

- For snubbers which are not part of the RCS or other high energy systems, then, this technical specification can be relocated.

(4) CONCLUSION Als Technical Specitication is retained.

I

.X._ The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (The LCO may be relocated but a program statement will be added to new TS Section 6.8.5).

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 17M SEAL.ED SOURCE CONTAMINA110N Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Tectuucal Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnonnal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient aralysis that :ither assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the pnmary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a tission product barrier.

X (4) A stmeture, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specificationt if the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The TS limitations ensures that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. The hmitations on removable contamination for sources requiring leak testing, including alpha emitters, is bas 4on 10 CI R Part 70.39(a)(3) limits for plutonium.

The TS requirements for sealed source contamination are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB, therefore, this TS does not satisfy criterion 1.

The scaled source contamination 15 is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DI A or transient analysis that either assumes the failure of or presents a ciudlenge to the integrity of a tim n product barrier. Thus, this TS does not satisfy cnterion 2.

1he 1S for se sicd source contamination does not apply to an SSC that is part of the primary success path and which ftmetions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, this TS does not satisfy cnterion 3.

I' rom Reference 2, sealed source contamination has not been shown to be significant to public health and safety by either operational experience or PSA. Scaled sources are used for cabbration and other purposes which have no impact on plant risk. 'Ihis technical specification is not included in the WCGS IPE. Therefore, this TS does not satisfy criterion 4.

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-(4) CONCLUSION -

'this Technical Specification is retained.

1.- The Technical Specification rnay be relocated to the following controlled document (s):

USAR Chapter 16.

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i TECIINICAL SPECIFICATION SCREENING FORM (1) TEC11NICAL SPECIFICATION 3.7.12 AREA TTiMPERATURE MONITORING Applicable Modes: Whenever equipment in the area is required to be OPERABLE.

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specitication applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment .

has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION

~Dris specification places a limit on the temperature of the areas of the plant which contain safety-related equipment. This is required to ensure that the temperature of the equipment does not exceed its environmental qualification temperature during normal operation. Exposure to excessively high temperatures may degrade the equipment and cause a loss ofits operability.

The TS requirements for area temperature monitoring are not applicable to installed instrumentation used to detect a-significant abnonnal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The area temperature monitoring TS is associated with the variable of room temperature which is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier Thus, this TS does not satisfy criterion 2.

1he TS for area temperature monitoring does apply to the operability of SSCs that are part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity ,

of a fission product barrier. Ilowever, the TS is only indirectly applicable to the operability of these systems and components. ;

Therefore, this TS does not satisfy criterion 3.

From Reference 2, the area temperature monitors have not been shown to be significant to public health and safety by either i operational experience or PSA. The area temperature monitors have not been included in the WCOS IPE. Therefore, this TS !

does not satisfy criterion 4. j j

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L (4) CONCLUSION

_ uis Technical Specification is retained. ,

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.X., The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (The LCO may be relocated but a program statement will be added to new -

TS Section 6.8.5).. _

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TECliNICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 38.4.1 CONTAINMENT PENETRATION CONDUCTOR OVERCtTRRENT PROTECTIVE DEVICES Applicable Modes: 1,2, 3, and 4 (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specincation applicable to:

YES NO X (1) Installed instrumentation that is useo to detect, and indicate in the control room, a significant abnormal degradation of the reactor coo' ant pressure boundary.

_ X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A stmeture, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The containment penetration conductor overcurrent protective devices are installed to minimize the potential for a fault in a component inside contairunent, or in cabling which penetrates containment. Tids prevents an electrical penetration from being damaged in such a w ay that the containment structure is breached.

the TS requirements for these devices are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion i.

The containment penetration conductor overcurrent protective devices do help to preserve the assumptions of the accident analysis by enhancing proper equipment operation. Ilowever, they are not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this TS does not satisfy criterion 2.

The contaimnent penetration conductor overcurrent protective devices provide equipment and distribution system protection from faults or improper operation of other protective devices in addition to that provided by the design of the distribution system. The TS for containment penetration conductor overcurrent protective devices does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

The containment penetration conductor overcurrent protective devices are installed to minimize the potential for a fault in a component inside containment or in cabling which penetrates the containment from damaging the electrical penetration in such a way that the containment stmeture is breached. From Reference 2, the protective devices have not been shown to be significant to public health and safety by either operational experience or PSA. The overcurrent protective devices have not been included in the WCGS IPE. Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION nis Technical Specification is retained.

X ne Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

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TECl!NICAL SPECIFICATION SCREENING FORM (1) TECllNICAL SPECIFICATION 3.9.5 COMMUNICATIONS Applicable Modes: During Core Alterations (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO

,1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant O abnormal degradation of the reactor coolant pressure boundary, X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION nis specification requires communication between the control room and the refueling station to ensure that any abnormal change in the facility status or core reactivity observed on the control room instrumentation can be communicated to the refueling station personnel during core alterations.

He TS requirements for communications are not applicable to installed instrumentation used to detect a significant abnonnal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He communications TS are not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrict. Thus, this TS does not satisfy criterion 2.

The TS for refueling communications does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

From Reference 2, communications between the control room and the refueling station during core modifications has not been shown to be significant to public health and safety by either operational experience or PSA. The WCGS IPE does not model the plant during refueling operations. Herefore, this TS does not satisfy criterion 4.

(4) CONCLUSION 1his Technical Specification is retained.

X_ 'Ihe Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

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(1) TECIINICAL SPECIFICATION 3.9.6 REFUEIING MACIIINE Applicable Modes: During movement of drive rods or fuel assemblies within the Reactor Vessel.

(2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO', the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION His specification assures that: (1) manipulator cranes will be used for movement of drive rods and fuel assemblies, (2) each crane has sutlicient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

The TS requirements for the refueling machine are not applicable to mstalled instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He refueling machine TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or inmsient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Rus, this TS does not satisfy criterion 2.

He TS for the refueling machine does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DDA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Herefore, this 'IS does not satisfy criterion 3.

He requirements of this technical specification are not significant to public health and safety by either operational experience or PSA. He refueling machine is used to transpon fuel assemblies during refueling operations. He WCGS IPE models the -

phmt during power operations, and therefore does not include the refueling machine in any risk quantifications. Ilowever, if the refueling machine were included in the model, it's significance would negligible. Herefore, this TS does not satisfy criterion 4. l

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(4) CONCLUSION

, l lhis Technical Specification is retained.

J_ 1he Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16, 3% dos e +

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TECliNICAL SPECIFICATION SCREENING FORM (1) TECllNICAL SPECIFICATION 3 9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY Applicable Modes! With fuel assemblies in the spent fuel storage facility.

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a signjncant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design llasis Accident or Transient anal) sis that either assumes the failure of or presents a challenge to the integrity of a fission product burier.

X (3) A structure, system, or component that is part of the primary success path cnd which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical .;-- .

Specifications. . ~ "J If the answer to all four of the above r;uestiona i: "NO", the Techmeal Specification may be relocated to a controlled document.

(3) DISCUSSION His specification ensures that loads in excess of one fuel assembly containing a control rod, plus the weight of the fuel handling tool, will not be moved over other fuel assemblies stored on the spent fuel storage racks. Therefore, in the event of a drop of this load, the activity released is limited to that contained in one fuel assembly. His also prevents any possible distortion of fuel assemblies in the storage racks from achieving a critical configuration. nis specification applies to prevention of a heavy load drop accident and assures that the damage caused by the load is limited to the equivalent of one spent fuel assembly. His assumption is consistent with the activity release assumed in the DBA accident analyses for a fuel handling accident; how ever, the load drop event is not a DBA.

We TS requirements for crane travel are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He spent fuel facility crane travel TS is associated with an operating restriction for a heavy load _ drop event. This specification is not applicable to a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a lission product bairier.

nus, this TS does not satisfy criterion 2.

The TS for crane travel does not apply to an SSC that is part c: ...e pimary success path and which functions or actuates to mitigate a DDA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barner. Herefore, this TS does not satisfy criterion 3.

From Peference 2, the spent fuel storage facility crane has not been shown to be significant to public health and safety by either og erational expenence or PSA. Reference 2 reviewed several environmental reports related to these cranes, and found their risk significance to be minimal. He spent fuel storage facility crane is not modeled in the WCGS IPE. Therefore, this TS does not satisfy criterion 4.

(4). CONCLUSION This Technical Specification is retained.

X The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

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TECilNICAL SPECIFICATION SCREENING FORM (1) TECilNICAL SPECIFICATION 3.9.10.2 WNITR LEVEL - REACTOR VESSEL / CONTROL RODS Applicable Modes: 6, during movement of control rods within the Reactor Vessel.

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO

_, J (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant 1 abnormal degradation of the reactor coolant pressure boundary.

_ J (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

3,, (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ J. (4) A structure, system or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specificat, ions.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION This specification places a lower limit on the amount of water above the top of the fuel assemblies in the reactor vessel during movement of control rods. The Bases state that this ensures the water removes 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly in the event of a fuel handling accident (FIIA). Ilowever, the j movement of control rods is not ussociated with the initial conditions of an Fila, and the Bases do not address any concerns regarding inadvertent criticality which could lead to a breach of ae fuel rod cladding. Inadvettent criticality during Mode 6 is prevented by maintaining proper boron concentration in the coolant in accordance with LCO 3.9.L The TS requirements for water level reactor vessel are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The watu .. 'el - reactor vessel TS is not associated with a process variable or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product ;

barrier. Thus, this TS does not satisfy criterion 2.

'the TS for water level- reactor vessel do not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

The reactor water level dunng refueling operations has not been shown to be significant to public health and safety by either operational experience or PSA. While refueling operations, including the reactor water level, are not modeled in the WCGS IPE, but they would not be important in any of the dominant accident sequences at WCGS.

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(4)- CONCLUSION This Technical Specification is retained.-

X The Technical Specification may be relocated to the following controlled document (sli USAR Chapter 16.

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECIINICAL SPECIFICATION 3.10.1 SPECIAL TEST EXCEFUON - SifUTDOWN MARGIN Applicable Mode: 2 (2) EVALUATIOli OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable , design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (3) A structure, system, or component that is part of the primary success path and which functions or -

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ,

X (4) A stmeture, system, or component u hich operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION Ref. 4 states that "Special Test Exceptions 3.10.1 through 3.10.4 may be included with corresponding LCOs which are remaining in Technical Specifications. Special Test Exception 3.10.5 may be relocated outside of Technical Specifications along uith LCO 3.1.3.3."

LCO 3.10.1 is only applicable in Mode 2. As discussed in the Screening Form for TS 3.1.1.1, the SDM requirements for Modes I and 2 are retained in other Reactivity Control System Technical Specifications. Retained Special Test Exceptions 3.10.2 and 3.10.3 address Special Test Exception 3.10.1 for LCOs 3.1.3.1 and 3.1.3.6. Herefore, Technical Specifications 3.10.1 will be deleted.

Shutdown margin has been shown to not be a dominant risk contributor. See TS 3.1.1.1. Therefore, these requirements do not satisfy enterion 4.

(4) CONCLUSION Btis Technical Specification is retained.

De Technical Specification may be relocated to the following controlled document (s):

X This Technical Specification is deleted.

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I (1) TECIINICAL SPECIFICATION } 10.5 SPECIAL 'IEST EXCEPTION - POSIRON INDICABON l SYSTEM - SillJfDOWN l Applicable Modes: 3,4, and 5 during performance of rod drop time measurements and during surveillance of DRPI I for Operability.

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to.

YES NO

_ 1 (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable , design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, systern, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety, if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION Ref. 4 states that "Special Test Exceptions 3.10.1 through 3.10.4 may be included with corresponding LCOs which are .

remaining in Technical Specifications. Special Test Exception 3.10.5 may be relocated outside of Technical Specifications along w ith LCO 3.1.3.3."

In accordance with its Screening Form LCO 3.1.3.3 may be relocated from TS. Therefore, LCO 3.10.5 may be relocated.

(4) CONCLUSION This Technical Specification is retained.

X De Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16.

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TECl!NICAL SPECIFICATION SCREENING FORM (1) TECIINICAL' SPECIFICATION 311.1.4 LIOtTID IIOI. DUP TANKS Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Tecludcal Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4)- A structure, system, or component uhich operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3). DISCUSSION ne liquid holdup tank specifications impose limits on the quantity of radioactive material contained in specific outdoor tanks that may contain radwaste. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentration would be less than the limits of 10 CFR 20, Appendix B, Table 11, Column 2, at the m.arest potable water supply and the nearest surface water supply in an unrestricted area. He tanks addressed by this specincation are:

a. Reactor Makeup Water Storage Tank

b. Refueling Water Storage Tank

c. Condensate Storage Tank

d. Outside temporary tanks, excluding demineralizer vessels and liners being used to solidify radioactive wastes.

Rese tanks are not addressed by the safety analysis of radioactive release from a subsystem or component.

De TS requirements for liquid holdup tanks are not applicable to installed instrumentation used to detect a significant abnonnal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He liquid holdup tanks E is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Hus, this TS does not satisfy criterion 2.

, , . . . = = - - - .. ..- ....

.~ . _ .-.

The TS for liquid holdup do not apply to an SSC that is part of the primary success path and which functions or actuates to .

mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

From Reference 2, the liquid holdup tanks, which hold radwaste, have not been shown to be significant to public health and '

safety by either operational experience or PSA. Risk of radioactivity release is dominated by severe accidente, not releases of radionuclides generated from normal operations. For this reason, the liquid holdup tanks are not modeled in the WCGS IPE.

'lherefore, this TS do not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

1 The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (The I.CO may be relocated but a program statement will be added to new -

TS Section 6.8.5).

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1 TECIINICAL SPECIFICATION SCREENING FORM "l (1) TECIINICAL SPECIFICATION 3112.5 EXPLOSIVE GAS MIXT 11RE Applicable Modes; At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO l

X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_ X (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission pro <'act barrier.

_ 1 (3) A structure, synem, or component that is part of the primary success path and which functions or >

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

I X (4) A structure, system, or component which operatmg experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be included in the new Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION nis specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the fiaminability limits of hydrogen and oxygen. . Maintaining these limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirernents of GDC 60 of Appendix A to 10 CFR 50. The accident analysis concerning the gaseous radwaste system assumes that a storage tank ruptures, from unspecified causes, and releases its contents without mitigation.

-i

%e TS requirements for explosive gas mixture are not applicable to installed instrumentation used to detect a significant ';

abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1. ;

l The explosive gas mixture TS is associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission - .j product barrier. Rus, this 'IS does not satisfy ciiterion 2. 'l 1

The TS for explosive gas mixture does not a' pply to an SSC that is part of the primary success path and which functions or actuates to miti Fate a DBA ,'r transient that either assumes the failure of or presents a chadenge to the integrity of a fission product barrier. Therefore, this TS do not satisfy criterion 3.

l

, .. .. - ~ . - -. -. . - .

i lhe explosive gas mixture of the waste gas holdup tanks has not been shown to be significant to public health and safety by either operational experience or PSA. Risk of radioactivity release is dominated by severe accidents, not releases of

' radionuclides generated from normal eperations. In addition. from Reference 2 the quantity of radioactivity contained in each pressurized gas storage tank in the waste gas holdup system is limited to assure a release would be substantially below the dose guideline values of 10 CFR Part 100, The waste gas holdup tanks are not modeled in the WCGS IPE. Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained. ,

X 1he Technical Specification may be relocated to the following controlled document (s):-

USAR Chapter 16. (The LCO may be relocated but a program statement will be added to new TS Section 6.8.5).

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TECIINICAL SPECIFICATION SCREENING FORM (1) TECliNICAL SPECIFICATION 311.2 6 GAS STORAGE TANKi Applicable Modes: At all times (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: ;

YES ' NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant

abnonnat degradation of the reactor coolant pressure boundary.

_ 1 (2) A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_ 1 (3) A structure, system, or component that is part of the primary success path and which functions or .

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION '!

The gas storage tank specifications impose limits on the quantity of radioactive material contained in those tanks for which ;

the quantity of radioactivity contained is not limited directly or indirectly by another TS. Restricting the quantity of ,

radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem. 1 This is consistent with Branch Technical Position ETED 11-5, " Postulated Radioactive Releases Due to a Waste Gas System . ,

1.cak or Failure." The accident analysis concerning the gaseous radwaste system assumes a rupture of a storage tank without mitigation.

The TS requirements for gas storage tanks are not applicable to installed instmmentation used to detect a significant abnormal' .

degradation of the RCPB; therefore, this TS doer not satisfy criterion 1.

The gas storage tank TS is associated with a process variable or operating restriction (quantity of contained radioactivity) that .j is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity = .j of a fission product barrier. Ilowever, the barrier in this case is the tank itself which is not a barrier that is monitored and -J controlled during power operation of the plant. Therefore, this TS does not satisfy criterion 2. ']

The 13 for gas storage tanks does not apply to an SSC that is part of the primary success path and which functions or actuates )

to mitigate a DBA or transient that either assumes the faihire of or presents a challenge to the integrity of a fission product . ;

harrier.1herefore, this TS does not satisfy criterion 3. i

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1' rom Reference 2, the waste gas holdup tanks, which hold radwaste, have not been shown to be significant to public health and safety by either operational experience or PSA. Risk of radioactivity release is dominated by severe accidents, not releases of radionuclides generated from normal operations. In addition, from Reference 2 the quantity of radioactivity contained in each pressurized gas storage tank in the waste gas holdup system is Imated to assure.a release would be substantially below the dose guideline values of 10 CFR Part 100. He waste gas holdup tanks are not modeled in the WCGS IPE. Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

2 The Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (He LCO may be relocated but a program statement will be added to new -

TS Section 6.8.5).

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4 APPENDIX A 1.0 OBJECTIVE Four criteria are included in the NRC's Final Policy Statement for determining the requirements to be included in the Technical Specifications. The Wolf

Creek Technical Specifications have been evaluated based on those four-criteria. The purpose of this document is to determine if the parameters, components, or systems addressed by the Technical Specifications are-significant from an operating experience or probabilistic safety assessment.

(PSA) perspective (i.e. the fourth criterion).

2.0 EVALUATION BASES The evaluation of the risk impact of the Technical Specifications in regards to the fourth criterion is based on the following:

A. The Technical Specifications that are relocated will be transferred to USAR Chapter 16.3 and will be implemented by-programs and procedures-subject to' control by WCNOC, within the constraints of 10CFR50.59.

B. The risk criteria used in determining the disposition of a Technical Specification are the following:

1. If the Technical Specification contains constraints of, prime importance in limiting the likelihood or severity of the accident sequences that are found to dominate risk, it will_be retained, a

2. If the Technical Specification includes items involved in one of these dominant sequences but has an insignificant impact on the '

probability or severity of that sequence and is not significant based on operating experience, it will be relocated to USAR Chapter 16.3.

3. If the Technical Specification is not involved in risk dominant sequences and is not significant based on operating experience, it will be relocated to USAR Chapter 16.3.

C. The measures related to risA used in this evaluation are core damage frequency (CDF) and offsite health effects. These measures are e consistent with the Final Policy Statement on Technical Specification Improvements and the Safety Goal and Severe Accident Policy Statements.

D. The criteria used to determine if'a sequence is risk dominant is the q following: For core melt, any sequence whose frequency was found to be greater than 1.0E-06 per reactor year in the Wolf Creek IPE (Reference 9) is considered to be a dominant sequence. This is slightly over 2% of the ;

total Wolf Creek core damage frequency of-4.2815E-05 per reactor year. ,

These sequences, 14 in all, are identified in Table 3.4-1 of Reference 9 and repeated here in Table 4.1. In addition, any sequence whose '

frequency of containment bypass was found to be greater than 1.0E-07 per reactor year in the Wolf Creek IPE is also considered to be a dominant

'I

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sequence. Three sequences met this criterion and are identified in' Table 3.4-4 of Reference 9 and also repeated below in Table 4.2 For offsite health effects, any sequence whose frequency.of severe radioactive release was found to be greater' than 1.0E-07 per reactor: year'

, in the Wolf Creek IPE is considered to be a dominant sequence. -Severe

. radioactive release for WCGS is defined as a WASH-1400-(Reference 10) PWR-release category 4 accident. As noted in Section 4.3.3.1 of Reference 9, ,

while sequences with frequencies greatsr than 1.0E-07 per reactor year.

exist at WCGS, none of these sequences meet the.PWR release category 4

. inventory release criteria. Offsite health effects, then,,are not a factor in determining whether to relocate the WCGS Technical Specifications. ,

.i E. Wolf Creek systems and functions that are important from a PSA or operating experience perspective are listed in Table 4.3. These >

identified systems, as well as .the risk domiriant sequences'as determined in paragraph D, were.used to screen the requirements of.the Technical j Specifications reviewed. If the requirements of a Technical l Specification were not found to be modeled in the Wolf' Creek IPELand no '

significant risk issues were identified from a review of the. risk.

insights or operating experience, that Technical. Specification would be relocated to USAR Chapter 16,3 unless the other three criteria mandated that it be retained. l t

3.0 METHOD USED Screening forms were developed which formalized the' review of each Technical Specification under the four criteria of the Final Policy Statement. These screening forms contain:

1. The number and title of the Technical Specification; 't

2. Im evaluation of the Technical Specification against'the. Final Policy Statement's four criteria;

~

3. A discussion of the information used in arriving at the conclusions for the- [

four criteria; and 4 A conclusion as to whether the Technical Specification should be retained.

or relocated. ;

This methodology is based on the approach presented in WCAP-11618 (Reference 3).

4.0 RISK DOMINANT SEQUENCES :!

a The tables that follow contain the dominant sequences in regards to risk and-containment bypass. Recall that significant offsite health effects are not'a :

concern at WCGS. '

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TABLE 4.1 LIST OF DCMINANT SEQUENCES SEQUENCE PERCENT SEQUENCE SEQUENCE WLNBER PROBABILITY CONTRIB DESCRIPTION IDENTIFIER 1 5.89E-06 13.77 STATION BLACKOUT INITIATING EVENT OCCURS IEV-SB0 SBC CUTSETS - CIA 4PONENTS FAIL AFTER LSP, SB0 EVENT RESULTS SYS-CCWS AC POWER IS NOT RECOVERED WITHIN 8 HOURS AFTER AN SB0 8HR-FAILS 2 4.47E-06 10.44 CONTROL BUILDING SWITCHGEAR ROOMS FLOCOING IEV OCCURS IEV-FL4 3 2.87E-06 6.70 STATION BLACKOUT INITIATING EVENT OCCURS IEV-SB0 i

$80 CUTSETS - COMPONENTS FAIL AFTER LSP, SB0 EVENT RESULTS SYS-CCWS AC POWER IS RECOVERED WITHIN 8 HOURS AFTER AN SB0 BHR-SUCCESSFUL CORE UNCOVERY OCCURS WITHIN 8 HOURS AFTER AN SB0 W/ RCD CNUS-FAILS 4 2.86E-06 6.67 LOSS OF OFFSITE POWER . INITIATING EVENT OCCURS IEV-LSP AUXILIARY FEEDWATER SYSTEM FAILS - 2/4 SG'S FROM 1/3 PUMPS SYS-AF2WO OPERATOR BLEED AND FEED COOLING FAILS SYS-OFC BOTH COMPONENT COOLING WATER TRAINS DO NOT FAIL DEL-CCW 5 2.77E-06 6.48 STATION BLACKOUT INITIATING EVENT OCCURS IEV-SBO S80 CUTSETS - COMPONENTS FAIL AFTER LSP, SB0 EVENT RESULTS SYS-CCWS AC POWER IS RECOVERED WITHIN 8 HOURS AFTER AN SB0 BHR-SUCCESSFUL' CORE DOES SOT UNCOVER WITHIN 8 HOURS AFTER AN 580 W/ RCD CNUS-SUCCESSFUL FRACTION OF AC RECOVERY AT 8 HR nFTER 5B0 FROM OFFSITE W/ RCD OP-08 HIGH PRESSURE RECIRCULATION FAILS - 1/4 PMPS FROM 1/2 TRAINS SYS-HPR12 6 2.20E-06 5.14 STATION BLACKOUT INITIATING EVENT OCCURS IEV-SBO COMPONENT Fall. AFTER LSP, SB0 EVENT RESULTS OTH-CCWS AUXILIARY FEEDWATER SYSTEM FAILS - 2/4 SG'S WITH TDAFW PUMP' SYS-AFT A! 20WER IS RECOVERED WITHIN 2 HOURS AFTER- AN SB0 2HR-SUCCESSFUL CORE DOES NOT UNCOVER WITHIN 2 HOURS AFTER AN SB0 W/O RCD CNU2F-SUCCESSFUL

' FRACTION OF AC RECOVERY AT 2 HR AFTER SBO FROM EDG W/0 RCD EDF-02 HIGH PRESSURE St RESTORATION AFTER.SBO, SMS OR CCW FAILS 2/2 OTH-RRI22 7 2.20E-06 5.13 RECOVERABLE CONTROL BUILDING BASEMENT FLOOD- IEV OCCURS IEV-FL38 FAILURE TO MITIGATE THE CONTROL BUILDING FLOOD EVENT SYS-FL38 8 2.19E-06 5.11 LOSS OF THE OPERATING CCW TRAIN IEV OCCURS IEV-CCWA

'HRA FAILURE TO PROVIDE RCP SEAL COOL IN TIMELY MANNER OPA-RCPSEAL HRA SUCCESS TO TRIP RUNNING RCP DN LOSS OF SEAL COOL B4 MLO RCPTRIP-SUC RCP SEAL LOCA (SLO) OCCURS AFTER CCWA LOSS (SEAL COOL LOSS) SYS-SLOCCW .

COMPONENT COOLING WATER SYSTEM TRAIN B DOES NOT FAIL DEL-CCWBO 9 1.86E-06 4.35 MEDIUM LOCA . INITIATING EVENT OCCURS IEV-MLO HIGH PRESSURE RECIRCULATION FAILS - 1/4 PMPS FROM 1/2 TRAINS SYS-LC2

-10 1.65E-06 3.84 STATION BLACKOUT INITIATING EVENT OCCURS IEV-SB0 F%9 - -'-' * ' ' ' -r' ---a- --- __. ----- ------- ---------- -

.q SBC CUTSETS -' COMPONENTS FAIL AFTER LSP, SB0 EVENT RESULTS SYS-CCWS AUXILIARY FEEDWATER SYSTEM FAILS - 2/4 SG'S WITH TDAFW PUMP STS-AFT AC POWER IS NOT RECOVERED WITHIN 2 HOURS AFTER AN $80 2HR-FAILS 11 1.64E-06 3.83 STATION- BLACKOUT INITIATING EVENT OCCURS IEV-S80 COMPONENT FAIL AFTER LSP, SB0 EVENT RESULTS OTH-CCWS AC POWER IS RECOVERED WITHIN 8 HOURS AFTER AN SB0 BHR-SUCCESSFUL CORE DOES NOT UNCOVER WITHIN 8 HOURS AFTER AN S80 W/ RCD CNU8-SUCCESSFUL FRACTION OF AC RECOVERY AT 8 HR AFTER 580 FROM D G W/ RCD ED-08 HIG*1 PRESSURE RECIRCULATION FAILS - 1/2 PMPS FROM 1/1 TRAIN SYS-HPR11 12 1.53E-06 3.56 STATION BLACKOUT INITIATING EVENT OCCURS IEV-SB0

$80 CUTSETS - COMPONENTS FAIL AFTER LSP, SBO EVENT RESUL15 SYS-CCWS AUXILIARY FEEDWATER SYSTEM FAILS - 2/4 SG'S WITH TDAFW PUMP SYS-AFT AC POWER IS RECOVERED WITHIN 2 HOURS AFTER AN SB0 2HR-SUCCESSFUL CORE DOES NOT UNCOVER WITHIN 2 HOURS AFTER AN SBC W/0 RCD CNU2F-SUCCESSFUL FRACTION OF AC RECOVERY AT 2 HR AFTER SB0 FROM OFFSITE W/0 RCD OPF-02 HIGH PRESSURE RECIRCULATION FAILS - 2/2 CCPS FROM 2/2 TRAINS SYS-HPR22 13 1.31E-06 3.06 LOSS OF OFFSITE POWER INITIATING EVENT OCCURS IEV-LSP COMPONENT COOLING WATER SYSTEM TRAIN A FAILS SYS-CCWA COMPONENT COOLING WATER SYS1EM TRAIN B DOES NOT FAIL DEL-CCW8 HRA FAILURE TO PROVIDE RCP SEAL COOL IN TIMELY MANNER .

OPA-RCPSEAL RCP SEAL LOCA (SLO) OCCURS - LSP AND CCWA FLS (SEAL COOL LOSS) SYS-SLOLSP 14 1.15E-06 2.68 LARGE LOCA INITIATING EVENT OCCURS IEV-LLO LOW PRESSURE RECIRCULATION SYSTEM FAILS SYS-LC1

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TABLE 4.2 i DOMINANT CONTAINMENT BYPASS SEQUENCES -

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' Percent l Sequence Cont *ibution Number Frecuency (to core melt) Secuence Descriotion i 26 2.44E-07 0.57 SGTR event, AFW and cooldown fail 27 2.41E-07 0.56 SGTR event, failure to l 4

stabilize.RCS and~

ruptured SG pressure, -'['

. secondary side RV closes

.i 1- 32 1 45E-07 0.34 SGTR event, failurc to .!

i stabilize RCS and '[

ruptured SG pressure,

[ secondary side RV sticks open, cooldown fails l

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i TABLE 4.3 SYSTEMS AND FUNCTIONS THAT ARE IMPORTANT FROM A PSA OR OPERATING EXPERIENCE !

PERSPECTIVE !

j Systems /eauinment cenerally the same as Callaway Class 1E Lower Medium Voltage System - 4.16Kv (NB)

Standby Generation System (NE)

Class 1E Low Voltage System - 480V (NG)

Class 1E '125VDC System (NK)

Class 1E Instrument AC System (NN) i 1

Reactor Coolant Pump Seals (BB)

Auxiliary Feedwater System (AL) i Component Cooling Water System (EG)

Safety Injection System (RM)

Chemical Volume and Control System-(BG)

. Essential- Service Water System (EF) .

Residual Heat Removal System (EJ)

Pressurizer PORV's (BB)

Refueling Water Storage System _(BN)

Steam Generator Atmospheric Relief Valves (??)

Containment Spray System (EN)

Solid State Protection System (SB) 7300 Process Protection System (SB)

Engineered Safety Features Actuation System (SA)

LOCA and Shutdown Sequencers (NF) >

Offsite Power (NA)

Service Water System (EA) ,

Main Feedwater System (AE)

Auxiliary-Building HVAC (for RHR) (GL)

Miscellaneous Buildings HVAC (for MDAFW) (GF)

ESW Pumphouse Ventilation System (GD)

Diesel Generator HVAC System (GM)

Emergency Fuel Oil System (JE)

Auxiliary Turbine System (FC)

Instrument Air System (KA) i i

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i Attachment VI to NA 94-0089 l l

Page 1 of 81 ATTACHMENT VI P

PROPOSED UPDATED SAFETY ANALYSIS REPORT REVISIONS s

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4 Attachment _VI to NA 94-0089 ,

Page 2 of 81 )

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NOTE: The following USAR mark-~up pages are provided as-information-only pages.

to show WCNOC's intent to relocate the applicable portions.of the . Technical- '

. Specifications to the USAR. The actual page format, content, and pagination of the USAR Revision may' differ slightly from the following pages. This . is. 3 based on WCNOC's intent to make-the new USAR Section 16 format, content, and pagination consistent with-the rest-of the USAR.

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I 161 f 3/41) REACTIVITY CONTROL SYSTEMS 161.1 INTENTIONAL l.Y BLANK 16 1.2(3/4.1.2) BORATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 16.1.2.1(3.1.2.1) As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:

a. A flow path from the Boric Acid Storage System via a boric acid transfer pump and a centnfugal charging pump to the Reactor Coolant System if the Boric Acid Storage System is OPERABLE as given in '

Section 16.1.2.5a. for MODES 5 and 6 or as given in Section 16.1.2.6a for MODE 4; or

b. The flow path from the refueling water storage tank via a centrifugal charging pump to the Reactor Coolant System if the refueling water storage tank is OPERABLE as given in Section 16.1.2.5b for MODES 5 and 6 or as given in Section 16.1.2.6b for MODE 4.

APPLICABILITY MODES 4. 5 and 6.

ACTION-With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 16.1.2.1.1 (4.1.2.1) At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual.

power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

BASES 16.1.2.1.2 The Boration Systems ensures that negative reactivity controlis available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) centrifugal charging pumps, (3) separate flow paths. (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature equal to or greater than 350*F a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The boration capability of either flow path is sufficient to provide a ?

SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capabildy requirement 18.1-1 l

I

occurs at EOL from full power equilibrium xenon conditions and requires 17,658 gallons of 7000 ppm borated water from the boric acid storage tanks or 83,754 gallons of 2400 ppm borated water from the RWST. With the RCS average temperature less than 350*, only one boron injection flow path is required.

With the RCS temperature below 200'F, one Boration System 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 changes in the event the single Boron injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4,5, and 6 provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR suction relief valve.

The boron capability required below 200'F is sufficient to provide a SHUTDOWN MARGIN of 1.3% Ak/k after xenon decay and cooldown from 200*F to 140'F.

This condition requires either 2968 gallons of 7000 ppm borated water from the boric acid storage tanks or 14.071 gallons of 2400 ppm borated water from the RWST.

The contained water volume limits include allowance for water not available because of discharge line location and other physical charactenstics. In the case of the boric acid tanks, all of the contained volume is considered usable.

The required usable volume may be contained in either or both of the boric acid tanks.

The limits on contained water volume and boron concentration of the RWST -

also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boration System during REFUELING ensures that this system is available for reactivrty control while in MODE 6.

When determining compliance with action statement requirements, addition to the RCS of borated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered to be a positive reactivity change.

16.1-2 1

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REACTIVITY CONTROL SYSTEMS ELOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 16.1.2.2(3.1.2.2) At least two of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from the Boric Acid Storage System via a boric acid transfer pump and a centnfugal charging pump to the Reactor Coolant System, and

b. Two flow paths from the refueling water storage tank via centnfugal charging pumps to the Reactor Coolant System.

APPLICABILITY: MODES 1. 2. and 3.

ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE. restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow Paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVElLLANCE REQUIREMENTS 16.1.2.2.1 (4 1.2.2) At least two of the above required flow paths shall be demonstrated OPERABLE.

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow oath that is not locked.

sealed, or otherwise secured in position, is in its correct position:

b. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety injection test signal and

c. At least once per 18 months by venfying that the flow path required by Section 16.12.2a delivers at least 30 gpm to the Reactor Coolant System.

BASES 16.1.2.2.2 See Section 16.1.2.1.2

The provisions of Technical Specifications 3.0.4 and 4.0.4 are not appheable for entry into MODE 3 for the centnfugal charging pump declared inoperable pursuant to Technical Specification 4.5.3.2 provided that the centnfugal charging pump is restored to OPERABLE status wdhin 4 hoars or prior to the ternperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

16 1-3

BJACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 16.1.2.3 (3 1.2.3) One centnfugal charging pump in the boron injection flow path required by Section 16.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 4 5. and 6.

ACTION:

With no centnfugal charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positrve reactivdy changes.

SURVEILLANCE REQUIREMENTS 16.1.2.3.1 (4.1.2.3.1) The above required centrifugal charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a differential pressure of greater than or equal to 2400 psid when tested pursuant to Technical Specification 4 0.5.

BASES 16.12.32 See Section 16.1.2.1.2 16.1-4

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 16.1.2.4 (3.1.2.4) At least two centrifugal charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1.2 and 3

ACTION' With only one centnfugal charging pump OPERABLE, restore at least two centnfugal charging pumps to OPERABLE status wdhin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: restore at least two charging pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 16.1.2.4.1 (4.1.2.4) At least two centnfugal charging pumps shall be demonstrated OPERABLE by venfying, on recirculation flow, that the pump develops a differential pressure of greater than or equal to 2400 psid when tested pursuant to Technical Specification 4.0.5.

BASES 16,1 '.4.2 See Section 16.1.2.1.2

The provisions of Technical Specifications 3.0.4 and 4.0 4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Technical Specification 4.5.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

l

'l 16.1-5 1

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 16.1.2.5 (3.1.2.5) As a minimum, one of the following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 2968 gallons,

2) Between 7000 and 7700 ppm of boron, and

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) wrth:

1) A minimum contained borated water volume of 55,416 gallons,

2) A minimum boron concentration of 2400 ppm, and -

3) A minimum solution temperature of 37'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 16.1.2.5.1 (4.1.2.5) The above required borated water source shall be demonstrated OPERABLE:

a At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Venfying the contained borated water volume, and

3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.

l

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 37'F.

BASES 16.1.2.5 2 See Section 16.1.2.1.2 16.1-6

REACTIVITY CONTROL SYSTEMS SORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 16.1.2.6 (3.1.2.6) As a minimum, the following borated water sources shall be OPERABLE -

as required by USAR Section 16.1.2.2 for MODES 1,2, and 3 and one of the following borated water sources shall be OPERABLE as required by USAR Section 16.1.2.1 for MODE 4:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 17,658 gallons,

2) Between 7000 and 7700 ppm of baron, and

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) with.

1) A minimum contained borated water volume of 394,000 gallons

2) Between 2400 and 2500 ppm of boron,

3) A minimum solution temperature of 37'F, and

4) A maximum solution temperature of 100*F.

APPLICABILITY: MODES 1. 2. 3. and 4.

ACTION:

a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources in MODE 1,2 or 3, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN w. thin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With the RWST inoperable in MODE 1,2, or 3, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With no borated water source OPERABLE in MODE 4, restore one borated water source to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , SURVEILLANCE REQUIREMENTS 16.1.2.6.1(4.1.2.6) Each required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Venfying the boron concentration in the water, 16.1-7

1 l

, 1 h

2) Verifying the contained borated water volume of the water j source, and , l

3) Venfying the Boric Acid Storage System solution temperature when it is the source of borated water,

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by venfying the RWST temperature when the outside air temperature is either less than 37'F or greater than 100'F.

BASES 16.1.2.6.2 See Section 16.1.2.1.2

16.1-8 l

16.13(3/41.3.1 MOVABLE CONTROL ASSEMBUES

- EgSITioN INDICATION SYSTEM-SPUTDOWN LIMITING CONDITION FOR OPERATION 16.1.3.1(31.33) One digital rod position Indicator (exct; ding demand position indication) shall be OPERABLE and capable of determining the control rod position wrthin + 12 steps for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3*#. 4*#. and 5*#

ACTION:

With less than the above required posttion indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.

SURVEILLANCE REQUIREMENTS 16.1.3.1.1 (4.1.3 3) Each of the above reauired digital rod position indicator (s) shall be determined to be OPERABLE by venfying that the digital rod position indicator agrees with the demand position indicator wthin 12 steps whan exercised over the full range of rod travel at least once per 18 months.

BASES -

16.1.3.1.2 See Technical Spec:fication Bases 3/41.3.

With the Reactor Trip System breakers in the closed position.

See Special Test Exception in Section 16.10.2.

1 i

16.1-9

c.

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 10.1.3.2 (3.1.3 4) The indrvidual fulLlength shutdown and control rod drop time from the physical fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T avg greater than or equal to 551*F, and

b. All reactor coolant pumps operating APPLICABILITY: MODES 1 and 2.

ACTION' a, With the rod drop time of any fulLlength rod determined to ex.ceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times wdhin limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 16,1.3.2.1(4.1.3.4) See Technical Specification 4.1.3.1.3 BASES 16 1.3.2.2 See Technical Specification Bases 3/4.1.3 I

I i

i 16.1-10

f

(:

16.2 INTENTIONALLY BLANK l

l 1.

4 4

i 16.2 1

l 16.3 (3/4 3) INSTRUMENTATION 16.3.1 (3/4 3.3) MONITORING INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 16.3.1.1 (3.3.3.2) The Movable incore Detection System shall be OPERABLE with:

a. At least 75% of the detector thimbles,

b. A minimum of two detector thimbles per core quadrant, and c Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Movable incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System,

b. Monitonng the QUADRANT POWER TILT RATIO, or

c. Measurement of Fo (X,Y,Z) and F,(X,Y)

ACTION

a. With the Movable Incore Detection System inoperable, do not use the system for the above applicab!e monitoring or cahbration functions.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 163.1.1.1(43.3.2) The Movable incore Detection System shall be demonstrated OPERABLE at ,

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

a. Recahbration of the Excore Neutron Flux Detection System, or

b. Mondoring the QUADRANT POWER TILT RATIO, or

c. Measurement of Fa (X,Y,Z) and F,(X,Y)

BASES

.16.3.1.1.2 The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

16.3-1

x.

1 4

M For the purpose of measuring Fo "(X,Y,Z) or F,g (X,Y) a fullincore flux map is used Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the i QUADRANT POWER TILT RATIO when one Power Range Neutron Flux channelis inoperable.

-i f

w 3

6

-16.3-2

'l l

i l

, , ,, - , . , . - , . - - n - , - - , - - .,-v--- - . - - - --- - --- -

1 I

INSTRUMENTATION j SEISMIC INSTRUMENTATION l

LIMITING CONDITION FOR OPERATION 16.3.1.2(3.3.3.3) The seismic monitoring instrumentation shown in Table 16.3-1 shall be i OPERABLE APPLICABILITY: At all times.

ACTION:

a. Wi'h one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submd a Special Report to the Commission pursuant to Technical Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.3.1.2.1.a (4.3.3.3.1) Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.3 2.

16 3.1.2.1.b (4.3.3 3.2) Each of the above required seismic moni'oring instruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed wrthin 10 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnrtude of the vibratory ground motion. A Special Report shall be prepared and submrtted to the Commission pursuant to Technical Specification 6.9.2 within 14 days describing the magnitude, frequency spectrum, and resultant effect upon facildy features important to safety.

BASES 16.3.1.2.2 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 capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974.

16.3-3

TABLE 16.31 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Peak Recording Accelerographs

a. Radwaste Base Slab + 1.0g i

b. Control Room + 1.0g 1

c. ESW Pump Facility + 1.0g 1-

d. Ctmt Structure + 2.0g . -1

e. Auxiliary Bldg. S1 Pump Suetions - + 1.0g 1

f. SGB Piping + 5.0g 1

g. SGC Support . + 1.0g i

2. Triaxial Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers (Active)

a. Ctmt. Base Stab + 1.0g 1

b. Ctmt. Oper. Floor + 1.0g 1 .

c. Reactor Support + 1.0g i

d. Aux. Bldg. Base Slab + 1.0g 1

e. Aux. Bldg. Control Room Air Filter + 1.0g i

f. Free Field + 0.5g 1

3. Triaxial Response-Spectrum Recorder (Passive)

Ctmt. Base Slab + 1.0g 1

4. Triaxial Seismic Switches ACCELERATION LEVEL North - s East Vertical -

a. OBE Ctmt. Base Slab 0.06g 0.06g 0.06g 1

b. SSE Ctmt. Base Slab 0.159 0.15g 0.16g 1

c. OBE Ctmt. Oper. Fl. 0,07g 0.07g 0.07g - 1

d. SSE Ctmt. Oper. Fl. 0.16g 0.17g - 0.16g i '

e. System Trigger 0.01g 0.01g 0.01g 1 l

16.3-4 i l

I l

i

TABLE 16 3-2 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ;

ANALOG CHANNEL-CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CAllBRATION TEST

1. Triaxial Peak Recording Accelerographs

a. Radwaste Base Slab N.A. R N.A.

b. Control Room NA R N.A.

c. ESW Pump Facilrty N A. R N.A.

d. Ctmt Structure N.A, R N.A.

e. Auxiliary Bldg Sl Pump Suction N.A. R N.A.

f. SGB Piping N.A. R NA g SGC Support N A. R N.A.

2. Triaxial Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers (Active)

a. Ctme. Base Slab M R SA

b. Ctme. Oper, Floor M R SA

c. Reactor Support M R SA"

d. Aux. Bldg. Base Slab M R SA"

e. Aux. Bldg. Control Room Filters M R SA"

f. Free Field M R SA"

3. Triaxial Response-Spectrum Recorder (Passive)

Ctmt Base Slab NA R N. A.*

4. Triaxial Seismic Switches

a. OBE Ctmt. Base Stab M R SA

b. SSE Ctmt. Base Slab M R SA

c. OBE Ctmt. Oper. Fl. M R SA

d. SSE Ctmt. Oper Fl. M R SA

e. System Trigger M R SA

' Checking at the Main Control Board Annunciation for contact closure output in the Control Room shall be performed at least once per 184 days.

"The Bi-stable Trip Setpoint need not be determined during the performance of an ANALOG CHANNEL OPERATIONAL TEST.

16.3-5

l 1

INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION .l l

')

16.3.1.3(3.3.3.4) The meteorological mondoring instrumentation channels in Table 16.3-3 '

shall be OPERABLE.

APPLICABILITX At all times.

ACTION. -

a. With one or more required meteorological monitoring channe's inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Technical Specification 6 9 * .vithin the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b. T., jlovisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.3.1.3.1(43.34) Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CAllBRATION at the frequencies shown in Table 16.3-4.

BASES 16.3.1.3.2 The OPERABILITY of the meteorologicalinstrumentation ensures that sufficient meteorological data is available for et o.iting potential radiation doses to the public as a result of routine or a: Wental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent wrth the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

16.3-6 6

g.

- TABLE 16.3-3 I

METEOROLOGICAL MONITORING INSTRUMENTATIOJJ MINIMUM INSTRUMENT LOCATION OPERABLE' .

1. Wind Speed Nominal Elev.10m 1 Nominal Elev. 60m 1

2. Wind Direction NominalElev 10m 1 Nominal Elev. 60m .1 .

3. Air Temperature - AT Nominal Elev.10m-60m 1 t

t I

s 16.3-7

TABLE 16.3-4 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMqtJ_TS T

CHANNEL CHANNEL INSTRUMENT CHECK- CALIBRATION T

1. Wind Speed

a. Nominal Elev.10m D SA b Nominal Elev. 60m D SA

^

2. Wind Direction

a. Nominal Elev.10m D SA

b. Nominal Elev 60m D SA ,

3. Air Temperature - AT

a. Nominal Elev.10-60m D SA P

t f

4 4

16.3-8 ,

i

'J

~

.i 1

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 16.31.4(3.33.6) The accident monitoring instrumentation channels shown in Table 16.3-5 shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTION:

a.With the number of OPERABLE accident monitoring instrumentation channels less than the Total Nurnber of Channels shown in Table 16.3-5, restore the inoperable channel (s) to OPERABLE status within 30 days or prepare and submit a Special Report to the Commission pursuant to Technical Specification 6.9.2 withiit the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

b.With the number of OPERABLE accident monitoring instrumentation channels, except the unit vent-high range noble gas monitor, less than the Minimum Channels OPERABLE requirements of Table 16.3-5, restore one channel to OPERABLE status within 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.With the number of OPERABLE channels for the unrt vent-high range noble gas monitor less than the Minimum Channels OPERABLE requirements of Table 16.3-5, initiate an alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and either restore +!,e inoperable channel to OPERABLE status within 7 days, or prepare and submit a Special Report to the Commission pursuant to Technical Specification 6.9.2 within the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperabikty, and the plans and schedule for restoring the channels to OPERABLE status.

d.The provisions of Technical Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.3.1.4.1(4.3.3 6) Each accident monttering inseumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 16.3-6.

BASES 16.3 1.4.2 See Technical Specification Bases 3/4.3 3 '

16.3-9

i TABLE 16.3-5 ACCIDENT MONITORING INSTRUMENTATION TOTAL MINIMUM ~ '

NO.OF CHANNELS INSTRUMENT CHANNELS OPERABLE

1. Containment Pressure - Extended Range 2 1

2. Safety Valve Position Indicator 1Nalve 1Nalve

3. Unit Vent - High Range Noble Gas Monitor N.A. 1 t

I 1

16.3-10

i i

TABLE 16 34 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ;

(NSTRUMENT CHANNEL CHANNEL CHECK CALIBRATION

1. Containment Pressure - Extended Range M R ,

2. Safety Valve Position Indicator M N.A

3. Unit Vent - High Range Noble Gas Monitor M R f

16.3-11

m- .

-i INSTRUMENTATION LQOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION l

1 163.1.5(33.39) The Loose-Part Detection System shall be OPERABLE. l l

APPLICABILITY: MODES 1 and 2.

ACTION: ,

a. With one or more Loose-Part Detection System channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Technical Specification 6 9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b. The provisions of Technical Specifications 3 0.3 and 3.0 4 are not applicable.

SURVEILLANCE REQUIREMENTS ,

a 16.3,1.51(4.3.39) Each channel of the Loose-Part Detection System shall be demonstrated OPERABLE by performance of.

a. A CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b An ANALOG CHANNEL OPERATIONAL TEST except for venfication of Setpoint at least once per 31 days, and

c. A CHANNEL CAllBRATION at least once per 18 months.

BASES 16.3 1.5.2 The OPERABILITY of the loose-part detection instrumentation ensures that sufficient capability is availabie to detect loose metallic parts in the Reactor Coolant System and avoid or mdigate damage to Reactor Coolant System components The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133," Loose-Part Detection Program for the Pnmary System of Light-Water-Cooled Reactors." May,1981.

16.3-12

INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 163.1.6(3.33.11) The explosive gas monitoring instrumentation channels shown in Table 16.3 7 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Section 16.11.2 are not exceeded.

APPLICABillT1 As shown in Table 16.3-7.

ACTION.

a. With an explosive gas monitoring instrumentation channel Alarm / Trip '

Setpoint less conservative than required by the above specification.

declare the channelinoperable and take tho ACTION shown in Table 16.3-7.

b. With less than the minimum number of explosive gas monitoring instru-mentation channels OPERABLE, take the ACTION shown in Table 16.3-7.

Restore the inoperable instrumentation to OPERABLE status wdhin 30 days, and, if unsuccessful, prepare and submrt a Special Report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.

c. The provisions of Technical Specification 3.0.3 and 3.0,4 are not applicable.

SURVEILLANCE REQUIREMENTS -

16 3.1.6.1 (4.3.3.11) Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CAllBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.3-8.

BASES 16 3.1.6.2 Intentionally Blank 16.3-13 l

TABLE 16 3-7 EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System

a. Hydrogen Monitor 1/Recombiner 1

b. Oxygen Monitor 2/Recombiner 2 ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner.

ACTION 2 - With the Outlet Oxygen Monitor channel inoperable, operation of the system may continue provided grab samples are taken and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With both oxygen channels or both the inlet oxygen and inlet hydrogen channels inoperable, suspend oxygen supply to the recombiner. Addition of waste gas to the system may continue provided grab samples are taken and analyzed at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations.

f 1

I J 16.3-14

[-

1 TABLE 16 3-8 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CAllBRATION TEST IS REQUIRED

1. WASTE GAS HOLDUP SYSTEM Explosive Gas Mondoring System a Inlet Hydrogen Monitor D Q(1) M ,

b. Outlet Hydrogen Monitor D O(1) M

c. Inlet Oxygen Monitor D Q(2) M d Outlet Oxygen Monitor D Q(3) M TABLE NOTATIONS

" Dunng WASTE GAS HOLDUP SYSTEM operation.

(1) The CHANNEL CAllBRATION shallinclude the use of standard gas samples containing a nominal:.

a. One volume percent hydrogen, balance nitrogen and

b. Four volume percent hydrogen, balance nitrogen.

(2) The CHANNEL CAllBRATION shallinclude the use of standard gas samples containing a nominal:

a. One volume per :ent oxygen, balance nitrogen, and

b. Four volume percent oxygen, balance nitrogen-(3) The CHA.NNEL CALIBRATION shallinclude the use of standard gas samples containing a nominal '

a. 10 ppm by volume oxygen, balance nitrogen, and

b. 80 ppm by volume oxygen, balance nitrogen 16 3-15

4 INSTRUMENTATION .j 16.32(3/4.34) TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION .

16.3.2.1 (3.3.4) At least one Turbine Overspeed Protection System shall be OPERABLE.

i . APPLICABILITY: MODES 1,2? and 3?

ACTION?

a.. With one stop valve or one governor vane per high pressure turbine steam kne inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoper-able, restc,re the inoperable valve (s)to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or close at least one valve in the affected steam lines ~

or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. Wrth the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply.

SURVEILLANCE REQUIREMENTS 16.3.2.1.1a (4.3.4.1) The provisions of Technical Specification 4.0.4 are not applicable.

16.3.2.1.ib (4.3.4.2) The above required Turbine Overspeed Protection System shall be -

demonstrated OPERABLE:

a. At least once per 7 days by cychng each of the following valves through at least one complete cycle from the running position:

1) Four high pressure turbine stop valves,

'2) Six low pressure turbine reheat stop valves, and ;

3) Six low pressure turbine reheat intercept valves.

b. At least once per 31 days by cycling each of the four high pressure main turbine governor valves through at least one complete cycle -

from the running position;

c. At least once per 31 days by direct observation of the movement of each of the above valves through one complete cycle from the running position;-

d. At least once per 18 months by performance of a CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems; and

e. At least once per 40 months by disassembhng at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and venfying no unacceptable flaws or corrosion.

Not applicable in MODE 2 or 3 with all main steam kne isolation valves and I associated bypass valves in the closed position and all other steam flow paths to the turbine isolated.

16.3-16

BASES 16.3.2.1.2 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. Although the orientation of the turbine is such that the number of potentially damaging missiles which could impact and damage safety-related components, equipment, or structures is minimal, protection frorn excessive turbine overspeed is required.

l-P 9

16.3-17

I 16.4 (3/4.4) REACTOR COOLANT SYSTEM l i

16 41 (3/4 4 2) SAFETY VALVES j i

SHUTDOWN LIMITING CONDITION FOR OPERATION !

16.4.1.1(3.4.2.1) A minimum of one pressurtzer Code safety valve shall be OPERABLE with a hft setting of 2485 psig 11%.*

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactMty changes and place an OPERABLE RHR loop into operation in the shutdown coohng mode.

SURVEILLANCE REQUIREMENTS 16.4.1.1.1(4.4.2.1) No additional requirements other than those required by Technical Specification 4.0.5.

BASES 16.4.1.1.2 The pressunzer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to reheve 420,000 lbs per hour of saturated steam. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capabihty and will prevent RCS overpressurization, in addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

The hft setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. >

16.4-1 l

t-REACTOR COOLANT SYSTEM 16.4 2 (3/4.4.5) STEAM GENERATORS LIMITING CONDITION FOR OPERATION 16.4.2.1(3 4.5) Each steam generator shall be OPERABLE.

APPLICASILITY: MODES 1,2,3 and 4.

A_CTLOff With one or more steam generators inoperable, restore the snoperable steam generator (s) to OPERABLE status pnor to increasing T,y above 200*F.

SURVElLLANCE REQUIREMENTS 16.4 2.1.1.a (41.5 0) Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Technical Specification 4.0 5.

16.4 2.1.1.b (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 16 4-1.

16.4 2.1.1.c (4.4.5 2) Steam Generatoj Tube Sample Selection Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 16 4-2. The inservice inspection of steam generator tubes shall be performed at the Fre-quencies specified in Section 16.4,2.1.1.d (4.4.5 3) and the inspected tubes shall be venfied acceptable per the acceptance entena of Section 16.4.2.1.1.e (4.4.5.4). The tubes selected for each inservice inspection shallinclude at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

(a) Where experience in similar plants with similar water chemistry indicates entical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas; (b) The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include.

16.4 2

f REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) ;

1) All nonplugged tu' bes that previously had detectable wall penetratens (greater than 20%),

2) Tubes in those areas where experience has indicated potential problems, and

3) A tube inspection (pursuant to Section 16 4.2.1.5a.8 (4.4.5.4a.8)) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube t inspection, this shall be recorded and an adjacent tube shall ,

be selected and subjected to a tube inspection (c) The tubes selected as the second and third samples (if required by Table 16.4-2) during each inservice inspection may be subjected to a partial tube inspection provided.

1)' The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and

2) The inspections include those portions of the tubes where imperfections were previously found. l The results of each sample inspection shall be classified into one of the ;

.following three categories: .;

Cateaory Inspection Results C-1 Less than 5% of the totaltubes inspected are "

degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the totaltubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are I degraded tubes or more than 1% of the inspected

tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit -

significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. ,

4 16.4-3

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REACTOR COOLANT SYSTEM

' SURVEILLANCE REQUIREMENTS (Continued)

' 16.4.2.1.1.d (4.4.5.3) Inspection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

1 (a) The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial enticality. 'i Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous -

inspection. l_f two consecutive inspections, not including the pre '

service inspection, result in allinspection results falling into the -

C-1 category or if two consecutive inspections demonstrate that pre-viously 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 results of the inservice inspection of a steam generator conducted in accordance with Table 16.4-2 at 40-month intervals fall :

in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency -

i.

shall apply until the subsequent inspections satisfy the errteria of Section 16.4.2.1.1.d (a)(4.4.5.3a).; the interval may then be extended to a maximum of once per 40 months; and ,

(c) Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 16.4 2 during the shutdown subsequent _to any of the following conditions:

1) Reactor-to-secondary tubes leaks (not including leaks originating .

from tube-to-tube sheet welds)in excess of the limrts of Technical Specification 3.4.6.2, or 'i

2) A seismic occurrence greater than the Operating Basis Earthquake, 5 or-

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam hne or feedwater line break. ,

l D

1 16.44 i

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l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i 16.4.2.1.1.e (4 4.5.4) Acceptance Lnteria J (a) As used in this specification:

1) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabncation drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;

2) O_emdation means a service-induced cracking, wastage, wear or general corrosion occurring on ether inside or outside of a tube;

3) Dearaded Tube means a tube containing imperfections greater than or equal to 20% of the nominalwallthickness caused by degradation;

4) % Dearadation means the percentage of the tube wallthickness affected or removed by degradation:

5) Defect means an imperfection of such seventy that it exceeds the plugging limit. A tube containing a defect is defective:

6) Pluacina Limd means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40%

of the nominal tube wall thickness;

7) Unserviceable describes the condition of a tube if it leaks or ,

contains a defect large enough to affect its structuralintegnty in the event of an Operating Basis Earthquake, a lossaf-coolant accident, or a steam line or feedwater line break as specified in Section 16.4.2.1.1.d (c) (4.4.5.3c), above;

8) lube inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and 9

16 4-5 8

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REACTOR COOLANT SYSTEM i SURVEILLANCE REQUIREMENTS (Continued)

9) Preservice inspection means an inspection of the fulllength of each tube in each steam generator performed by eddy current techniques prior to service to estabhsh a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used dunng subsequent inservice inspections.

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 16.4-2.

16.4.2.1.f (4.4.5,5) Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Technical Specification 6 9.2:

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Technical Specification 6 9.2 within 12 months following the completion of the inspection. This Special Report shallinclude:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each 1 indication of an imperfection, and

3) Identification of tubes plugged c Results of steam generator tube inspections, which fallinto Category C-3, shall be reported in a Special Report to the Commission pursuant to Technical Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

BASES 16.4 2.1.2 The Surveillance Requirements for inspection of the steam generator tubes ensure that the structuralintegrity 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 conddions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, ,_

manufactunng errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken

^

16.4-6

y

' Unscheduled inservice inspections are performed on each steam generator following 1) reactor to secondary tube leaks; 2) seismic occurrence gre9ter than the Operating Basis Earthquake. 3) a loss-of-coolant accident requiring actuation of the Engineered Safety Features, which for this specification is defined to be a break greater than that equivalent to the severance of a 1" inside diameter pipe, or, for a main steamline or feedline, a break greater than that equivaient to a steam generator safety valve failing open: to ensure that steam generator tubes retain sufficient integnty for continued operation Transients less severe than these do not require inspections because the resulting stresses are well within the stress enteria established by Regulatory Guide 1.121 which unplugged steam generator tubes must be capable of withstanding The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these lim:ts, localized corrosion may '

likely result in stress corrosion cracking: The extent of cracking during plant operation would be limrted by the limitation of steam generator tube leakage between the reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator).

Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. 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 proper chemistry treatment of the '

secondary coolant. However, even if a defect should develop in service it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominalwallthickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the originaltube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fallinto Category C-3, these results will be reported to the Commission pursuant to Specification 6.9 2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations. tests, addi-tional eddy-current inspection, and revision of the Technical Specifications, if necessary.

I 16.4-7 i I

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TABLE 4 4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION I

Preservice inspection No Yes

-i

!' No. of Steam Generators per Unit Two Three Four Two Three Four First inservice Inspection All One Two Two- ,

Second & Subsequent inservice Inspections One1 One l , One2 One3 TABLE NOTATIONS

1. The Inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant)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 conditions in one or more steam generators may be found to be more severe than ,

those in other steam generators Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.

2. The other steam generator not inspected during the first inservice inspection shall be inspected. The third and ,

subsequent actions should follow the instructions described in 1 above.

3. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions desenbed in 1 obove.

I i

i 16,4-8 i

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION I 2ND SAMPLE INSPECTION l 3RD SAMPLE INSPECTION I l

Sampic Size Result Action Required Result Action Required Result Action Required A minimum C-1 None N. A. N. A. N. A. N. A.

of S Tubes per C-2 Plug defective tubes C-1 None N. A. N. A.

Steam and inspect Generator additional 2S tubes C-2 Plug defective tubes C-1 None in this S.G. and inspect additional 4S tubes C-2 Ptag defective tubes in this S. G.

C-3 Perform Action for E C-3 result of first sample 8

C-3 Perform action for N. A. N. A.

C-3 result of first sample C-3 Inspect all tubes in A:! *.er S. G.s are None N. A. N. A.

this S G plug C-1 defective tubes and inspect 2S tubes in Some S. G.s C-2 but' Perform action for N. A. N. A.

each other S.G. no additional S. G. is C-2 result of second C-3 sample Notification to NRC pursuant to 50.72(b)(2) of 10 CFR Part 50 Additional S. G. is inspect all tubes in N. A. N. A.

C-3 each S. G and plug defective tubes.

Notification to NRC pursuant to 50.72(b)(2) of 10 CFR l Part 50 S=3 % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during inspection

i l

l REACTOR COOLANT SYSTEM I 1643(3/44.7) CHEMISTRY i I

i LIMITING CONDITION FOR OPERATION 16 4.3.1 (3.4.7) The Reactor Coolant Systern chemistry shall be maintained within the limits specified in Table 16.4-3.

APPLICABILITY: At all times.

ACTION:

MODES 1, 2. 3, and 4:

a. With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any one or more chernistry parameter in excess of its Transient Limrt, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN .

within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

At All Other Times:

With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressunzer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limrt condition on the structuralintegrity .

of the Reactor Coolant System; determine that the Reactor Coolant System [

remains acceptable for continued operation prior to increasing the pressurizer ,

pressure above 500 psig or prior to proceeding to MODE 4.

SURVEILLANCE REQUIREMENTS 16 4.3.1.1 (4 4.7) The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 16.44.

BASES 16.4.3.1.2 The limrtations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage of failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structuralintegnty of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect of the structural integnty of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provv,'es time for taking corrective actions to restore 1.he contaminant concentrations to within the Steady-State Limitt..

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

16 4-10

TABLE 16.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen

s 0.10 ppm s 1.00 ppm Chloride s 0.15 ppm s 1.50 ppm Fluoride s 0.15 ppm s 1.50 ppm i

Limd not applicable with T avg less than or equal to 250*F. ;

i 16.4-11 I

TABLE 16.4-4 REACTOR COOLANT SYSTEM j CHEMISTRY SURVEILLANCE REQUIREMENTS

. SAMPLE AND PARAMETER ANALYSIS FREQUENCY Dissoked Oxygen * - At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Chloride At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Fluoride At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months t

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'Not required with T avg less than or equal to 250*F. I 16.4-12 l

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16,4 4 f 3/4 4 9) PRESSURFJTEMPERATURE UMITS PRESSURIZJR, LIMITING CONDITION FOR OPERATION 1644.1(34.92) The pressurizer temperature shall be limited to.

a. A maximum heatup of 100*F in any 1-hour penod,

b. A maximum cooldown of 200'F in any 1-hour penod, and

e. A maximum spray water temperature differential of 583'F.

APPUCABluTY: At all times ACTION.

Wrth the pressunzer temperature lim *s in excess of any of the above limits restore the temperature to within the hmrts within 30 minutes; perform an engineenng evaluation ti determine the effects of the out-of-limit cordttion on the structuralintegrity of the pressunzer; determine that the pressurizer remains acceptable for enntinued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce ibe pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 16.4.4 1.1 (4.4.9.2) The pressunzer temperatures shall be determined to be within the limits at least once per 30 minutes dunng system heatup or cooldown. The spray water teroperature differential shall be determined to be within the ljmit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

BASES 16441.2 The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section lil, Appendix G.

The pressurizer heatup and cool rates shall not exceed 100*F/h and 200'/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583*F.

Although the pressurtzer operates in temperature ranges above those for which there is reason for concern of ;

nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis i performed in accordance with the ASME Code requirements. j Also see Technical Specification Bases 3/4.4 9.

16.4-13 i

1 I

l 85 ACTOR COOLANT SYSTgff 16 4.5 (3/4 4.10) STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 16.4.5.1 (3.4.10) The structuralintegrity of ASME Code Class 1,2 and 3 components shall be maintained in accordance with Section 16 4.5.1.1 (4.4.10).

APPLICABILITY: All MODES.

ACTION:

a. With the structuralintegrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integnty of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.

b. With the structuralintegnty 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 structuralintegnty of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integnty of the affected component (s) to within its limit or isolate the affected component (s) from service.

d. The provisions of Technical Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.4 5.1.1 (4.4.10) in addition to the requirements of Technical Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4 b of Regulatory Guide 1.14, Revision 1, August 1975. (See Technical Specification 6.8.5)

BASES 164.5.12 The inservice inspection and testing programs for ASME Code Class 1,2, and 3 components ensure that the structuralintegrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific wrrtten relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspection sin accordance with Section XI of the ASME Boiler and Pressure Vessel Code,1974 Eddion and Addenda through Summer 1975.

16.4 14

16 4 6 (3/4 4.11) REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 16.4.6 (3 4.11) At Mast one reactor vessel head vent path consisting of at least two valves in series powered from emergency busses shall be OPERABLE and closed.

MfklCABILITY: MODES 1,2, 3, and 4.

ACTION:

With the above reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 16.461.1(4.4.11) Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 rnonths by:

a. Verifying all manualisolation valves in each vent path are locked in the open position,

b. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING, and

c. Venfying flow through the reactor vessel head vent paths during venting during COLD SHUTDOWN or REFUELING.

BASES 164.6.12 Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability exists to perform this function, The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure vent valve power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of item II.B.1 of NUREG-0737,"Clanfication of TMI Action Plan Requirements," November 1980.'

16.4-15

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l' 16.5 INTENTIONALLY BLANK 4

1.

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16.5-1 5

ms . ., , ,, ..-e . . . - , ,.e- - -&

CONTAINMENT SYSTEMS 16.6(3/4 6) PRIMARY CONTAINMENT 16.6.1 (3/4.6.1) CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 16.6.1.1 (3.6.1.2) Containment leakage rates shall be limited to'

a. An overallintegrated leakage rate of.

1) Less than or equal to L a,0.20% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pa ,48 psig, or

2) Less than or equal to L t,0.020% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pt ,24 psig.

L A combined leakage rate of lesa : San 0.60 La for all penetrations and valves subject to Type B and C tests, when pressurized to Pa ,48 psig.

APPLICABILITY: MODES 1,2, 3, and 40 ACTION:

a. If Reactor Coolant System temperature is at or below 200*F, with either the measured overallintegrated containment leakage rate exceeding 0.75 Laor 0.75 L as t applicable, or the measured combined leakage rate for all penetrations and vai.es subject to Types B and C tests exceeding 0.60 La '

restore the overall integrated leakage rate to less than 0.75 La or less than Lt . as applicable, and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 La prior to increasing the Reactor Coolant System temperature above 200*F.

b. It the Reactor Coolant System temperature is above 200 degrees F, with the measured combined leakage rate for all penetrations and valves subject to Types B and C test exceeding 0.60 La-

1) Restore the combined leakage rate to less than 0.60 La within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by one of the following methods:

a) Repairing the failed containment isolation component, or b) Isolating the penetration containing the failed component by closing and the deactivating one automatic valve, or c) Isolating the penetration containing the failed component by closing one manual valve, or d) Isolating the penetration containing the failed component by using a blind flange.

2) If the combined leakage rate is not restored to less than 0.60 La within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

16.6-1 J

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CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE 1

SURVEILLANCE REQUIREMENTS 16.6.1.1.1 (4 6.1.2) The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the enteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972: ,

a. Three Type A tests (Overallintegrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during shutdown at a pressure not less than either Pa,48 psig, or P .t 24 psig, dunng each 10-year service period. The third test of cach set shall be conducted during the shutdown for the 10-year plant inservice inspection; i

b. If any periodic Type A test fails to meet either 0.75 L a or 0.75 Lt the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet erther 0.75 La or 0.75 L t, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet erther 0.75 Laor 0.75 L att which time the above test schedule may be resumed;

c. The accuracy of cach Type A test shall be venfied by a supplemental test which:

1) Confirms the accuracy of the test by venfying that the supple-mental test result, Lc, minus the sum of the Type A and the super.

imposed leak, Lo, is equal to or less than 0.25 La or 0.25 Lt :

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplementaltest; and

3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment dunng the supplemental test is between 0.75 La and 1.25 La or 0.75 Lt and 1.25 Lt '

d, Type B and C tests shall be conducted with gas at a pressure not less than Pa,48 psig, at intervals no greater than 24 months except for i tests involving- l

1) Air locks,

2) Purge supply and exhaust isolation valves with resilient material seals, and l

3) Valves pressurized with fluid from a seal system.

e, Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3;

f. Purge supply and exhaust isolation valves with resilient material j seals shall be tested and demonstrated OPERABLE by the requirements j of Technical Specification 4.6.1.7.2 and 4.6.1.7.4, as applicable; I

16.6-2

g Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J, Section Ill.C.3, when determining the combined leakage rate provided the seal system and valves are pressurtzed to at least 1.10 Pa (53 psig), and the seal system capacrty is adequate to maintain system pressure for at least 30 days; and

h. The provisions of Technical Specification 4.0.2 are not applicable.

BASES 16.6.1.1.2 The limrtations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overallintegrated leakage rate is further limrted to less than or equal to 0.75 Laor 0.75 L ,t as applicable, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

For reduced pressure tests, the leakage characteristics yielded by measurements Ltm and L am shall establish the maximum allowable test leakage rate L of not more than La(Ltm/ lam). In the event Ltm/ lam is greater than t

0.7, Lt shall be specified as equalto La (Pt /Pa N The surveillance testing for measuring leakage rates are consistent with the requirement, of Appendix J of 10 CFR Part 50.

16.6-3 1

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1 1

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY

. LIMITING CONDITION FOR OPERATION 16.6.1.2(3.6.1.0) The structuralintegnty of the containment vessel shall be maintained at a level consistent with tho acceptance cnteria in Section 16.6.1.2.1 (4.6.1.6).

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION

a. Wdh the abnormal degradation indicated by the conditions in Section 16.6.1.2.1a.4 (4.6.1.6.1a.4), restore the tendons to the required level of integnty or verify that containment integnty is maintained within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Technical Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the indicated abnormal degradation of the structuralintegnty other than ACTION a. at a level below the acceptance enteria of Section 16.6.1.2.1 (4.6.1.6), restore the containment vessel to the required level of integnty or verify that containment integnty is maintained within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Technical Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN wrthin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Technical Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.6.1.2.1.a (4.6.1.61) Containment Vessel Tendons. The structuralintegrity of the prestressing tendons of the containment vessel shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integnty test and at 5-year intervals thereafter.. The structuralintegnty of the tendons shall be demonstrated by; (a) Determining that a random but representative sample of at least 11 tendons (4 inverted U and 7 hoop) each have an observed lift-off force within the predicted hmits established for each tendon. For each subsequent inspection one tendon from each group (1 inverted U and 1 hoop) shall be kept unchanged to develop a history and to corrvate the observed data. The procedure of inspection and the tendon acceptance entena shall be as follows:

16.6-4

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1, if the measured prestressing force of the selected tendon in a group lies above the presenbed lower limrt, the hft-off test is considered to be a posrtive indication of the sample tendon's acceptability,

2. If the measured prestressing force of the selected tendon in a group lies between the prescribed lower limit and 90% of the prescribed lower limit, two adjacent (accessible) tendons, one on each side of this tendon shall be checked for their prestress-ing forces. If the prestressing forces of these two tendons are above 95% of the presenbed lower limits for the tendons, all three tendons shall be restored to the required level of integrity, and the tendon group shall be considered as accept-able. If the measured prestressing force of any twc andons falls below 95% of the prescribed lower limits of the tendons, additional hft-off testing shall be done to detect the cause and extent of such occurrence The condition shall be con-sidered as an indication of abnormal degradation of the contain-ment structure,

3. If the measured prestressing force of any tendon lies below 90%

of the prescribed lower limit, the defective tendon shall be completely detensioned and additional hft-off testing shall be done so as to determine the cause and extent of such occurrence. The condition shall be considered as an indication of abnormal degradation of the containment structure,

4. If the average of all measured prestressing forces for each .-

group (corrected for average cond! tion) is found to be less than the minimum required prestress level at the anchorage locations for that group, the condition shall be considered as abnormal degradation of the containment structure,

5. If from consecutive surveillances the measured prestressing forces for the same tendon or tendons in a group indicate a trend of prestress loss larger than expected and the resulting prestressing forces will be less than the minimum required for the group before the next scheduled surveillance, additional lift-off testing shall be done so as to determine the cause and extent of such occurrence. The condition shall be considered as an indication of abnormal degradation of the containment structure, and

6. Unless there is abnormal degradation of the containment vessel

' during the first three inspections, the sample population for ;

subsequent inspections shall include at least 6 tendons (3 hoop, 3 inverted U).

16.6-5

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I CONIAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i (b) Performing tendon detensioning, inspections, and matenal tests on a previously stressed tendon from each group. A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and determining that over the entire length of the removed wire sample (which shallinclude the broken wire if so identified) that: ,

1. The tendon wires are free of corrosion, cracks, and damage, and

2. A minimum tensile strength of 240 ksi(guaranteed ultimate strength of the tendon material) exists for at least three wire samples (one from each end and one at mid-length) cut from each removed wire.

Failure to meet the requirements of Section 16.6.1.2.1.a (b) (4.6.1.6.1.b) shall be considered as an indication of abnormal degradation of the containment structure.

(c) Performing tendon retensioning of those tendons detensioned for inspection to at least the force level recorded prior to detensioning or the predicted value, whichever is greater, with the tolerance within minus zero to plus 6%, but not to exceed 70% of the guaranteed ultimate tensile strength of the tendons. During retensioning of these tendons the changes in load and elongation shall be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating force, if the elongation corresponding to a specific load differs by more than 10%

from that recorded during the installation, an investigation shall be made to ensure that the difference is not related to wire failures or slip of wires in anchorages. This condition shall be considered as an indication of abnormal degradation of the containment structure.

(d) Venfying the OPERABILITY of the sheathing filler grease by assuring:

1. There are no changes in the presence or physical appearance of the sheathing filler-grease including the presence of free water,

2. Amount of grease replaced does not exceed 5% of the net duct volume, when injected at 110% of the specified installation pressure,

3. Minimum grease coverage exists for the different parts of the anchorage system,

4. During general visual examination of the containment external surface, that grease leakage that could affect containment integrity is not present, and 16 6-6

CONTAINMENT SYSTEMS

~

SURVEILLANCE REQUIREMENTS (Continued)

5. The chemical properties of the filler material are within the tolerance limrts specified as follows:

Water Content 0 - 10% by dry weight Chlorides 0 - 10 ppm Ndrates 0 - 10 ppm Sulfides 0 - 10 ppm Reserved Alkalinity >0 Failure to meet the requirements of Section 16.6.1.2.1.a (d)(4.6.1.6.1.d) shall be considered as an indication of abnormal degradation of the containment structure.

16.6.1.2.1.b (4.6.1.6.2) End Anchorages and Adjacent Concrete Surfaces. As an assurance of the structuralintegnty of the containment vessel, tendon anchorage assembly hardware (such as bearing plates, stressing washers, wedges, and buttonheads) of all tendons selected for inspection shall be visually examined. Tendon anchorages selected for inspection shall be visually examined to the extent ,

practical without dismantkng the load bearing components of the anchorages.

Bottom grease caps of all vertical tendons shall be visually inspected to detect grease leakage or grease cap deformations. The surrounding concrete shall also be checked visually for indication of any abnormal condition. The frequency of this surveillance shall be in accordance with Section 16.6.1.2.1 (4.6.1.6.1).

Significant grease leakage, grease cap deformation or abnormal concrete condition shall be considered as an indication of abnormal degradation of the containment structure.

16.6,1.2.1.c (4.6.1.6 3) Containment Vessel Surfaces. The exterior surface of the containment shall be visually examined to detect areas of large spall. severe scaling, D-cracking in an area of 25 sq. ft. or more, other surface deterioration or disintegration, or grease leakage, each of which shall be considered as evidence of abnormal degradation of structuralintegnty of the containment. This inspection shall be performed prior to the Type A containment leakage rats test (See Technical Specification 6.8.5)

BASES 16.6 1.2.2 This limitation ensures that the structuralintegrity of the containment will be maintained in accordance with safety analysis requirements for the life of the facilfty. Structurat integrrty is required to ensure that the contain.

ment will withstand the maximum pressure of 50.4 psig in the event of a steam line break accident. The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and rsxterior surfaces of the contain-ment, and the Type A leakage test are sufficient to demonstrate this capabihty.

16.6-7 -

p The Surveillance Requirements for demonstrating the containment's structuralintegrity are in compliance with the recommendations of proposed Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures." April 1979, and proposed .

Regulatory Guide 1.35.1, " Determining Prestressing Forces for inspection of '

Prestressed Concrete Containments," April 1979.

The required Special Reports from any engineenng evaluation of ,

containment abnormalities shallinclude a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerance on cracking, the results of the engineering evaluation and the corrective actions taken.

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4 1668

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16 7 (3/4.7) PLANT SYSTEMS 16.7.1 (3/4.7.2)_ STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 16.7,1.1(3,7 2) The temperatures of both the reactor and secondary coolants in the steam generator shall be greater than 70"F when the pressure of erther coolant in the steam generator is greater than 200 psig APPLICABILITY: At all times.

ACTION-With the requirements of the above section not satisfied;

a. Reduce the steam ger.erator pressure of the applicable side to less than or equal to 200 psig wrthin 30 minutes, and

b. Perform an engineering evaluation to determine the effect of the overpressunzation on the structuralintegrrty of the steam generator.

Determine that the steam generator remains acceptable for continued operation pnor to increasing its temperatures above 200'F.

SURVEILLANCE REQUIREMENTS 16.7.1.1.1 (4.7.2) The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the reactcr or secondary coolant is less than 70*F.

BASES 16 7.1.1.2 The hmrtation 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 hmits The limitations of 70 *F and 200 psia are based on a steam generator RTNDT of 60*F and are sufficient to prevent brittle fracture, 16.7-1

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-i PLANT SYSTEMS

. 16 7.2(3/4.7.8) SNUBBERS LIMITING CONDITION FOR OPERATION 16.7.2.1 (3.7.8) All snubbers shall be OPERABLE. The only snubbers excluded frem the requirement are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

APPLICABILITY: MODES 1. 2,3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.

ACTION:

Wth one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Section 16.7.2.1.1g (4.7.8g) on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 16.7.2.1.1(4.7.8) Each snubber shall be demonstrated OPERABLE by performance of the follow-ing augmented inservice inspection program and the requirements of Technical Specification 4.0.5.

j- a. Inspection Types As used in this specification type of snubber shall mean snubbers of the same design and manufacturor, irrespectrve of capacity,

b. Visual inspections Snubbers are categorized as inaccessible or accessible dunng reactor operation. Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 16.7-1. The visualinspection interval for each type snubber shall be determined based upon the criteria provided in Table 16.7-1 and the first inspection interval determined using this enteria shall be based upon the previous inspection interval as established by the requirements in effect betore Operating License Amendment 44.

4 16.7-2 a

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. VigualInspection Acceptance Cnteria Visual inspections shall venfy that: (1) there are no visible indi-cations of damage or impaired OPERABILITY, and (2) attachments to the foundation or supporting structure are functional, and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visualinspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that. (1) the cause of the rejection is clearly estabhshed and remedied for that particular snubber and for other snubbers irrespective of type that may be genencally susceptible; or (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Section 16.7.2.1.1f (4.7.8f). All snubbers found connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable for determining the next inspection interval. A review and evaluation shall be performed and documented to determine system operabihty with an unacceptable snubber. If operability cannot be justified, the system shall be declared inoperable and the ACTION requirements shall be met.

d Transient Event inspection An inspection shall be performed of all hydraulic and mechanical snub-bers attached to sections of systems that have expenenced unexpected potentially damaging transients as determined from a review of opera-tional data and a visualinspection of the systems within 6 months following such an event. In addition to satisfying the visual inspection acceptance entena, freedom-of-motion of mechanical snubbers shall be venfied using at least one of the following:

(1) manually induced snubber movement, (2) evaluation of in-place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.

16.7-3

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. E_unctional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers of each type shall be tested using one of the following sample plans.

The sample plan shall be selected prior to the test period and cannot be changed dunng the test period. The NRC Regional Administrator shall be notified in wnting of the sample plan selected for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:

1) At least 10% of the total of each type of snubber shall be function.

ally tested either in-place or in a bench test. For each snubber of a type that does not meet the functional test acceptance enteria of Section 16.7.2.1.1f (4.7.8f), an additional 10% of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested or

2) A representative sample of each type of snubber shall be func-tionally tested in accordance with Figure 16.7-1 (4.7-1). "C" is the ,

total number of snubbers of a type found not meeting the accep-

,f tance requirements of Section 16.7.2.1.1f(4.7.8f). The cumulative number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of"N" and "C"(pre-vious day's total plus current day's increments) shall be plotted on the Figure 16.7-1 (4.7-1). If at any time the point plotted falls in the " Reject" region, all snubbers of that type shali be function-alty tested. If at any time the point plotted falls in the

" Accept" region, testing of snubbers of that type may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the

" Reject" region or allthe Sr;ubbers of that type have been tested; or

3) An initial representative sample of 55 snubbers shall be func-tionally tested. For each snubber type which does not meet the functiond test acceptance enteria, another sample of at least one-half the s!ze of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the factor,1 + C/2 where "C"is the number of snubbers found which do not meet the functional test accept-ance criteria The results frorn this sample plan shall be plotted using an " Accept"line which follows the equation N =

55(1 + C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept"line, testing of that type of snubber may be terminated. If the point plotted falls above the " Accept"line testing must cor,tinue until the point falls in the " Accept" region or all the snubbers of that type have been tested.

16.7-4

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Pt. ANT SYSTEMS ;

SURVEILLANCE REQUIREMENTS (Continued) !

e. Functional Tests (Continued)

Testing equipment failure during functional testing may invalidate i that day's testing and allow that day's testing to resume anew at a later time provided ali snubbers tested with the failed squipment during the day of equipment failure are retested. The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers of each type.

Snubbers placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing. additional sampling is required due to failure of only one type of snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the func.

tional testing

f. Functional Test Acceptance Criteria The snubber functional test shall verify that:

1) Activation (restraining action)is achieved within the specified range in both tension and compression;

2) Snubber bleed rate, or release rate where required, is present in both tension and compression, within the specified range; and

3) For mechanical snubbers, the force required to initiate or main-tain motion of the snubber is within the specified range in both directions of travel Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.

9. Service Life Monitorina Proaram An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.

7 16.7-5

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) J f

g. Service Life Monitorina Proaram (Continued) for the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service.

If any snubber selected for functional testing either fails to lock up or fails to move, i e., frozen-in-place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of the requirements stated in Section 16.7.2.1.1e (4 7.8e). for snubbers not meeting the functional test acceptance criteria.

h Functional Testina of Repaired and Replaced Snubbers Snubbers which fail the visualinspection or the functional test acceptance enteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test enteria before installation in the unit. Mechanical snubbers shall have met the acceptance enteria subsequent to their most recent servica and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.

i Snubber Service Life Proaram The service life of hydraulic and mechanical snubbers shall be moni-tored to ensure that the service life is not exceeded between sur-veillance inspections. The maximum expected service life for various seals, springs, and other entical parts shall be determined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history.

Cntical parts shall be replaced so that the maximum service life will not be exceeded dunng a period wh6n the snubber is required to be OPERABLE, The parts replacements shall be documented and the docu-mentation shall be retained in accordance with Technical Specification 6.10.2.

(See Technical Specification 6.8 5) 16.7-6

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4 BASES >

16.7.2.1.2 All snubbers are required OPERABLE to ensure that the structuralintegrity of the Reactor Coolant System and all other safety-related systems is main-4 tained dunng and following a seismic or other event initiating dynamic loads.

Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip, and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" <

for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer. Snubbers may also be classified and grouped by inaccessible or accessible for visualinspection purposes Therefore, each snubber type may be grouped for inspection in accordance with accessibildy.

A list of individual snubbers with detailed information of snubber location and size and of systems affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Plant Safety Review Commrttee.

The determination shau be based upon the existing radiation levels and the expected time to perform a visualinspection in each snubber location as well as other factors associated w;th accessibility during plant operations (e g.,

temperature, atmosphere, location etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

The visualinspection frequency is based upon maintaining a constant level of snubber protection during an earthquake or severe transient. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number ofinoperable snubbers found during an inspection of each type. In order to establish the inspection frequency for each type of snubber on a safety-related system, it was assumed that the fre-quency of snubber failures and initiating events is constant with time and that the failure of any snubber could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%)

may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

The acceptance enteria are to be used in the visualinspection to deter-mine OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic

' snubber is found to be uncovered, the snubber shall be declared inoperable and i shall not be determined OPERABLE via functional testing Since the visual -)

inspections are augmented by functional testing program, the visualinspection l need not be a hands on inspection, but shall require visual scrutiny sufficient J' to assure that fasteners or mountings for connecting the snubbers to supports or foundations shall have no visible bolts, pins or fasteners missing, or other visible signs of physical damage such as cracking or loosening.

16.7-7 I

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1 To provide assurance of snubber functional reliabikty, one of three

. functionaltesting methods are used with the stated acceptance entena:

1. Functionally test 10% of a type of snubber with an additional 10%

tested for each functional testing failure, or -

2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or

3. Functionally test a representatrve sample size and determine sample acceptance or rejection using the stated equation.

Figure 4.7-1 was developed using "Wald's Sequential Probabildy Ratio Plan" as desenbed in "Quakty Control and Industrial Statistics" by Acheson J. Duncan Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber hfe destructive testing was performed to quahfy the snubber for the applicable design conddions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the kst of individual snubbers indicating the extent of the exemptions. -

The service kfe of a snubber is established via manufacturer input hnd information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spnng replaced, in high radiation area, in high temperature area, etc.), The requirement to monitor the snubber service hfe 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 hfe.

a s

l' 16.7-8 1

~

q H

r

\ ,,, b + . - , - - - -- ,

TABLE 16.7-1 SNU' BBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABLE SNUBBERS Population Column A Column B Column C per Category Extend Interval Repeat Interval Reduce Interval (Notes 1 and 2) (Extend 3 and 6) (Notes 4 and 6) (Notes 5 and 6) 1 0 0 -1 80 0 0 2 100 0 1 4 150 0 3 8 200 2 5 13 300 5 12 25 -

400 8 18 36 500 12 24 48 750 20 40 78 1000 or greater 29 56 109 Note 1: The next visual inspect'on interval for a snubber category shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval.

Snubbers may be categortzed, based upon their accessibility during power operatic,n, as accessible or inaccessible. These categories may be examined separately or jointly..However, categories must be determined and documented before any inspection and that determination shall be the basis upon which to determine the next '

inspection interval for that category.

Note 2: Interpolation between population per category and the number of unacceptable snubbers is permissible. Use next lower integer for the value of the limit for Columns A, B, and C if that integer includes a fractional value of unacceptable snubbers as determined by interpolation.

16.7 9 a

l T

.l Note 3: If the number of unacceptable snubbers is equal to or less than the number in Column A, the next inspection interval may be twice the previous interval but not greater than 48 months.

Note 4: If the number of unacceptable snubbers is equal to or less thari the number in Column 8 but greater than the number in Column A, the next inspection interval shall be the same as the previous interval.

E Note 5: If the number of unacceptable snubbers is equalto or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shat! be reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one-third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Column B and C.

Note 6: The provislor.s of Technical Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months.

i t-16.7 10

._ . . _ . - ._ _ , _n ._ ._ _. . ~ ,

10 9'

8

~

7 REJECT j 3V O

CONTINUE-2 TESTING ,

3 2 -

ACCEPT 1 e 0 10 20 30 40 50 60 70. 80 90 '100 N

d e

/Z ,7- ' / l

' FIGURE =4.7-1 SAMPLE PLAN 2) FOR SNUB 8ER FUNCTIONAL TEST j I4 , 7 - 11 l

. _ ___ m

f PLANT SYSTEMS 1673_(3/479) SEALED SOURCE CF "INATION LIMITING CONDITION FOR OPERATION 16.7.3.1 (3.7.9) Each sealed source containing radioactive matenal either in excess of 100 microCuries of beta and/or gamma-emitting material or 5 microCtries of alpha emittirig material shall be free of greater than or equal to 0.005 microcurie of removable contamination.

APPLICABILITY: At all times.

ACTION:

a. With a sealed source having removable contamination in excess of the above limits, irnmediately withdraw the sealed source from use and either;

1. Decontaminate and repair the sealed source, or

2. Dispose of the sealed suurce in accordance with Commission Regulations.

b. The provisions of Technical Specifications 3.0.3 and 3 0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.7.3.1.1.a (4.7.9.1) Test Requirements Each sealed source shall be tested for leakage and/or contamination by; a The licensee, or b Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample.

16.7.3.1.1.b (4.7.9 2) Test Frequencies - Each category of sealed sources (excludlng startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency desenbed below.

a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:

1) With a half-life greater than 30 days (excluding Hydrogen 3) and

2) in any form other than gas. i l

16.7-12

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and

c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair ,

or maintenance to the source.

16.7.3.1.1.c (4.7.9.3) Reports - A report shall be prepared and submrtted to the Comrnission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination BASES 16.7.3.1.2 The limitations on removable contamination for sources requinng leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3)limrts for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate wrth the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are cons'dered to be stored and need not be tested unless they are removed from the shielded mechanism.

16.7-13

l l

l

.1 PLANT SYSTEMS ;

i 16 7 4l3/4 712) AREA TEMPERATURE MONITORING -l LIMITING CONDITION FOR OPERATION 16.7.4.1(3.7.12) The temperature limit of each area grven in Table 16.7-2 shall not be exceeded for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or by more than 30'F.

APPLICABAIE Whenever the equipment in an affected area is required to be OPERABLE.

ACTION-.

a. With one or more areas exceeding the temperature limit (s) shown in -

Table 16.7 2 for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , prepare and submit to the Commission wrthin 30 days, pursuant to Technical Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Technical Specifications 3.0.3 and 3.0 4 are not applicable. .

b. With one or more areas exceeding the temperature limit (s) shown in Table 16.7-2 by more than 30'F, prepare and submrt a Special Report as required by ACTION a above, and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either restore the area (s) to within the temperature limit (s) or declare the equipment in the affected area (s) inoperable, SURVEILLANCE REQUIREMENTS 16.7.4.1.1 (4.7.12) The temperature in each of the areas shown in Table 16.7-2 shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(See Technical Specification 6.8.5)

BASES 16.7 4.1.2 The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of +3*F.

I 16.7-14 ,

1 1

l l l

)

TABLE 16.7-2 AREA TEMPERATURE MONITORING

_ MAXIM UM TEMPERATURE AREA LIMIT ( *F )

1. ESW Pump Room A 119

2. ESW Pump Room B 119

3. Auxiliary Feedwater Pump Room A 119

4. Auxiliary Feedwater Pump Room B 119

5. Turbine Driven Auxiliary Feedwater Pump Room 147

6. ESF Switchgear Room I 87

7. ESF Switchgear Room 11 87 8 RHR Pump Room A 119 e

9. RHR Pump Room B 119

10. CTMT Spray Pump Room A 119

11. CTMT Spray Pump Room B 119 12 Safety injection Pump Room A 119

13. Safety injection Pump Room B 119 14 Centnfugal Charging Pump Room A 119 15 Centrifugal Charging Pump Room B i 119

16. Electrical Penetration Room A 101

17. Electncal Penetration Room B 101

18. Component Cooling Water Room A 119

19. Component Cooling Water Room B 119 20 Diesel Generator Room A 119

21. Diesel Generator Room B 119

22. Control Room 84 16.7-15

168(3/48) ELECTRICAL POWER SYSTEMS 168.1(3/484) ELECTRICAL EQUIPMENT PROTECTIVE OfVICES

, CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 16.8.1.1 (3.8.4.1) For each containment penetration provided with a penetration conductor overcurrent protective device (s), each device shall be OPERABLE.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION With one or more of the above required containment penetration conductor overcurrent protective device (s) inoperable:

a. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated backup circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoperable circuit breaker racked out, or removed, at least once per 7 days thereafter; the provisions of Technical Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, theit inoperable circurf breakers racked out, or removed, or

b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 16.8.1.1.1 (4.8.4 1) Protective devices required to be OPERABLE as containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE.

a. At least once per 18 months:

1) By venfying that the 13.8 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers, and performing the following-a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and venfying that each relay and associated circuit breakers and control circuits function as designed, and 16.8-1

FJf.CTRICAL POWER SYSTEM _S SURVEILLANCE REQUIREMENTS (Continued) c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

2) By selecting and functionally testing a repr asentative sample of at least 10% of each type of lower volta 3 e circuit breakers.

Circuit breakers selected for functional tes' ng shall be selected on a rotating basis. Testing of tt se circurt breakers shall consist of injecting a current in exce s of the breakers nominal Setpoint and measuring the response time. The measured response time will be compared to the manufacturer's data to ensure that it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sarnple of at least 10% of allthe circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

BASES 16.8.1.1.2 Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent r:rotection circuit breakers during periodic surveillance.

The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker. Each manufacturer's molded case and metal case circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufacturer's brand of circuit breakers, it la necessary to divide that manufacturer's breakers into groups and treat each group as a separate type of breaker for surveillance purposes.

A list of containment penetration conductor overcurrent protective devices whose circuit limiting fault current exceeds the penetration rating, with information of location and size and equipment powered by the protected circurt, is available at the plant site in accordance with Section 50.71(c) of 10 CFR Part 50. The addition or deletion of any containment penetration conductor overcurrent protective device would be made in accordance with Section 50.59 of 10 CFR Part 50.

16.8-2 1

16.9 (3/4.9) BfFUElitLG OPERATIONS 1Rt(3/4 9.5) COMMUNICATIONS LIMITING CONDITION FOR OPERATION 16.91.1 (3.9 5) Direct communications shall be maintained between the control room and personnel at the refuehng station.

APPLICABILITY: During CORE ALTERATIONS.

ACTION.

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.

SURVEILLANCE REQUIREMENTS 16.91.1.1 (4.9 5) Direct communications between the control room and personnel at the refueling station shall be demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months dunng CORE ALTERATIONS.

BASES 16 9.1.1.2 The requirement for communications capabihty ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conddions during CORE ALTERATIONS.

16.9-1

REFUELING OPERATIONS 16 9 2 (3/4 9 6) REFUELING MACHING LIMITING CONDITION FOR OPERATION 16.92.1(39.6) The refueling machine shall be used for movement of dnve rods or fuel assemblies and shall be OPERABLE with:

a. The refueling machine used for movement of fuel assemblies having:

1) A minimum capacity of 4800 pounds,

2) Automatic overload cutoffs wdh the following Setpoints:

a) Primary -less than or equal to 250 pounds above the indicated suspended weight for wet conditions and less than or equal to 350 pounds above the indicated suspended weight for dry condrtions, and b) Secondary - less than or equal to 150 pounds above the primary overload cutoff. ,

3) An automatic load reduction trip wdh a Setpoint of less than or equal to 250 pounds below the suspended weight for wet conddions or dry conditions.

b. The auxiliary hoist used for latching W " Jehing drive rods and thimble plug handling operations having:

1) A minimum capacdy of 3000 pounds, and

2) A 1000-pound load indicator which shall be used to monitor lifting loads for these operation.

APPLICABILITY: During movement of drive rods or fuel assemblies wdhin the reactor vessel ACTION.

With the requirements for refueling machine and/or auxiliary hoist OPERABILITY not satisfied, suspend use of any inoperable refueling machine crane and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor vessel SURVEILLANCE REQUIREMENTS _

16 9.2.1.1.a (4 9.6.1) The refueling machine used for movement of fuel assemblies wrthin the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior 16.9-2

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) to the movement of fuel assemblies in the reactor vessel by performing a load test of at least 125% of the secondary automatic overload cutoff and demonstra.

ting an automatic load cutoff when the refueling machine load exceeds the Setpoints of Section 16.9.2.1a.2 ) and by demonstrating an automatic load reduction trip when the load reduction exceeds the Setpoint of Section 16.9.2.1a 3.

16.9.2.1.1.b (4 9.6.2) Each auxihary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the movement of drive rods within the reactor vessel by performing a load test of at least 1250 pounds.

BASES 16.9 2.1.2 The OPERABILITY requirements for the refueling machine and auxiliary hoist ensure that: (1) manipulator cranes will be used for movement of drive rods and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from .

excessive lifting force in the event they are inadvertently engaged during lifting operations.

16.9-3

!~

. * ~ .

j i

REFUELING OPERATIONS j 16.9 3 (3/4.9 7) CRANE TRAVEL - SPENT FUEL STORAGE FACILITY LIMITING CONDITION FOR OPERATION j 16.9 3.1 (3.9.7) Loads in excess of 2250 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage facility.

APPLICABILITY: With fuel assemblies in the spent fuel storage facihty.

ACTION.

a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.9.3 1.1 (4.9 /) Crane interlocks and physical stops which prevent crane travel with loads in excess of 2250 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least on e per 7 days thereafter during crane operation.

BASES

.- = =

169.3.12 The restriction on movement of loads in excess of the nominal weight of a fuel and contro' rod assembly and associated handling tool over other fuel assemblies in the storage r ool area 3 ensures that in the event this load is dropped (1) the activity rele9se will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuelin the storage racks will not result in a entical array. This assumption is consistent with the activity release assumed in the safety analyses t

t 16.9-4 l

l l

16.9 4 (3/4 910) WATER LEVEL - REACTOR VESSEL CONTROL RODS LIMITING CONDITION FOR OPERATION 16.9.4.1 (3.9.10 2) At least 23 feet of water:ihall be maintained over the top of the irradiated fuel assemblies within the reactor pressure vessel.

APPLICABILITY; Dunng movement of control rods ,vF.hin the reactor pressure vessel while in MODE 6.

ACTIO_N:

With the requirements of the above specification not satisfied, suspend all cperations involving movement of control rods within the pressure vessel.

SURVEILLANCE REQUIREMENTS 16.9.4.1.1(4.9.10.2) The water level shall be determined to be at least its minimum required death within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months pnor t, le start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter dunng movement of control rods within the reactor vessel.

BASES 16.9 4.1.2 See Technical Specification Bases 3/4.910 16.9-5 i

16.10(3/4.10) SPECI AL TEST EXCEPTIONS 16 10.1(3/4 10.11 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 16.10.1.1(3.10.1) The SHUTDOWN MARGIN requirement of Section 16.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactrvity equivalent to at least the highest estimated control rod worth is available for tnp insertion from OPERABLE control rod (s).

APPLICABILITY: MODE 2.

ACTION:

a. With any full-length control rod not fully inserted and with less than the above reactrvrty equivaient available for tnp insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equalto 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Section 16.1.1.1 is restored.

b. With all full-length control rods fully inserted and the reactor subcntical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Section 16.1,1.1 is restored.

SURVEILLANCE REQUIREMENTS 1610.1.1.1 a (4.10.1.1) The position of each full-length control rod either partially or fully wrthdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

16.101.1.1.b (4.10.1.2) Each full-length control rod not fully inserted shall be demonstrated capable of fullinsertion when tripped from at least the 50% withdrawn position withir 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limds of Section 16.1.1.1.

BASES 16.10.1.1.2 This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to l permit the periodic venfication of the actual versus predicted core reactivity ,

condition occurring as a result of fuel burnup or fuel cycling operations. 'l l

'l l

16.10-1 l

l

SPECIAL TELT EXCEPTIONS 16.10.2 (3/410.5) POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 16.10.2.1 (3.10.5) The limitations of Section 16.1.3.1 may be suspended during the performance of individu^' full-length shutdown and control rod drop time measurements providtd . y one shutdown cr control bank is withdrawn from the fully inserted posttion at a time.

APPLICABILITY. MODES 3,4, and 5 during performance of rod drop time measurements and during surveillanc e of digital rod position indicators for OPERABILITY.

ACTION.

With the Position Indication System inoperable or wrth more than one bank of rods withdrawn, immediately open the Reactor tnp breakers.

SURVEILLANCE REQUIREMENTS 9

16.10 2.1.1(4.10 5) The above required Position Indication Systems shall be determined to be OPERABLE wdhin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by venfying the Demand Position Indication System and the Digdal Rod Position Indication System agree:

a. Within 12 steps when the rods are stationary, and

b. Within 24 steps during rod motion. i BASES 16.10.2.1.2 This special test ereption permrts the Position Indication Systems to be inoperable during rod drop time measurements.

16.10-2 i-

=!

L!

l j

1611 (3/411) RADIOACTIVE _Jf_fl,UENTS 16.11.1 LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 16.11.1.1 (3.11.1.4) The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limrted to less than or equal m 60 Curies, excluding tritium and dissolved or entrained noble gases. ,

a. Reactor Makeup Water Storage Tank,

b. Refueling Water Storage Tank,

c. Condensate Storage Tank, and

d. Outside temporary tanks, excluding demineralizer vessels and liners being used to solidify radioactive wastes.

APPLICABydy; A" p times.

eCTION:

a With the quantity of r 1 adivo materialin any of the above listed tanks exceeding tr e above limit, immediately suspend all additions of radioactive mate ial to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduca the tank contents to within tr e limit, and desenbe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Technical Specification 6 9.1.7.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVE!LLANCE REQUIREMENTS 16.11.1.1.1 (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 limrt by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added and within 7 days following any addition of radioactive material to the tank. (See Technical Specification 6.8.5)

BASES 16.11.1.1.2 The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restncting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 11, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA 16.11-1 n

16.11.2 EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 16.11.2.1 (3.11.2.5) The concentration of cxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 3% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABLl,EY_: At all times.

ACTIO3 a With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 3% by volumo but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTFA greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, then take ACTION a. above

c. The provisions of Technical Specifications 3.0.3 and 3 0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 16.11.2.1.1 (4.11.2.5) The concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by continuously monttenng the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 16 3-6 of Section 16.3.1.6.

(See Technical Specification 6.8 5)

BASES 16.11.2.1.2 This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits These automatic control features include isolation of the source of hydrogen and/or oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of .adioactive materials will be controlled in conformance with the requirements of General Design Cnterion 60 of Appendix A to 10 CFR Part 50.

16.11 2 l

BAQlOACTIVE EFFLUENTS i

1611.3 GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION

')

16 11.3.1 (3.11.2.6) The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 2.5 x 105 Cunes of noble gases '

(considered as Xe-133 equivalent).

APPLICABILITY: At all times.

ACTION-

a. With the quantity of radioactive matenalin any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the limit, and desenbe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Technical Specification 6.9.1.7.

b. The provisions of Technical Specifications 3 0 3 and 3.0.4 are not apphcable SURVEILLANCE REQUIREMENTS 16.11.3.1.1 (4.11.2.6) The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above hmit at least once per 7 days when radioactive materials are being added and within 7 days following any addition of radioactive material to the tank. (See Technical Specification 6.8.5)

BASES 16.11,3.1.2 The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not hmited directly or indirectly by another Technical Specification. Restricting the quantity of radioactiv1y contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem.

This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5,. " Postulated Radioactive Re' eases Due to a Waste Gas System Leak or Failure,"in NUREG-0800, July 1981.

16.11-3

.

ML20069B568

Text

c.

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