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| NUREG-1136 Technical Specifications Wolf Creek Generating Station,
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| ; Unit No.1 Docket No. STN 50-482 Appendix "A" to License No. NPF-42 O :
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| Issued by the U.S. Nuclear Regulatory Commission l
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| Office of Nuclear Reactor Regulation June 1985
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| , ya a'%,,
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| O 6 "Ai'"I88M 8i800282 P pm l
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| NOTICE Availability of Reference Materials Cited in NRC Publications Most documer ts cited in NRC publications will be available from one of the following sources:
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| : 1. The NRC Public Document Room,1717 H Street, N.W.
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| Washington, DC 20555
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| : 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7982
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| : 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
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| Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
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| The following documents in the NUREG series are available for purchase from the NRC/GPO Sales Program: formal NRC staff and contractor reports NRC sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.
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| Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
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| Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
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| Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
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| Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Technical information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555.
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| Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
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| NUREG-1136 Technical Specifications Wolf Creek Generating Station, Unit No.1 Docket No. STN 50-482 Appendix "A" to License No. NPF-42 issued by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1985 s....../
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| m DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1 1.2 ACTUATION LOGIC TEST.......................................... 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1
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| : 1. 4 AXIAL FLUX DIFFERENCE......................................... 1-1 1.5 CHANNEL CALIBRATION........................................... 1-1 1.6 CHANNEL CHECK................................................. 1-1
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| : 1. 7 CONTAINMENT INTEGRITY......................................... 1-2
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| : 1. 8 CONT RO L L E D L EA KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.9 CORE ALTERATION............................................... 1-2 1.10 DOSE EQUIVALENT I-131........................................ 1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-2 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 f ~~s 1.13 FREQUENCY NOTATION........ .................................. 1-3 i n
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| ( _f 1.14 IDENTIFIED LEAKAGE........................................... 1-3 1.15 MASTER RELAY TEST............................................ 1-3 1.16 MEMBER (S) 0F THE PUBLIC...................................... 1-3 1.17 0FFSITE DOSE CALCULATION MANUAL.............................. 1-4 1.18 OPERABLE - OPERABILITY....................................... 1-4 1.19 OPERATIONAL MODE - M0DE...................................... 1-4 1.20 PHYSICS TESTS................................................ 1-4 1.21 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.22 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.23 PURGE - PURGING.............................................. 1-4 1.24 QUADRANT POWER TILT RATI0.................................... 1-5 1.25 RATED THERMAL P0WER.......................................... 1-5 1.26 REACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.27 REPORTABLE EVENT............................................. 1-5 1.28 SHUTDOWN MARGIN.............................................. 1-5 1.29 SITE B0VNDARY................................................ 1-5 s 1.30 SLAVE RELAY TEST............................................. 1-5 1.31 SOLIDIFICATION............................................... 1-5 x
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| WOLF CREEK - UNIT 1 I
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| DEFINITIONS SECTION PAGE DEFINITIONS (Continued) 1.32 SOURCE CHECK................................................. 1-6 1.33 STAGGERED TEST BASIS......................................... 1-6 1.34 THERMAL P0WER................................................ 1-6 1.35 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 1.36 UNIDENTIFIED LEAKAGE......................................... 1-6 1.37 UNRESTRICTED AREA............................................ 1-6 1.38 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-6 1.39 VENTING...................................................... 1-7 1.40 WASTE GAS HOLDUP SYSTEM...................................... 1-7 TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M0 DES....................................... 1-9 O
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| O WOLF CREEK - UNIT 1 II
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| SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-3 i
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| TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION PAGE
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| , 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE....................................... ........ B 2-1 i 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 l
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| 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS............... B 2-3 i
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| WOLF CREEK - UNIT 1 III
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| ~ - . -
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE O
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| E4. 0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T > 200 F...... ... . .............. 3/4 1-1 avg Shutdown Margin - T ,g 5 200 F......... ... .. ..... .... 3/4 1-3 Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality.......... .......... 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating......... ............... . ....... 3/4 1-8 Charging Pump - Shutdown.......... ...................... 3/4 1-9 Charging Pumps - Operating.......... ................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - Operating........................ 3/4 1-12 3/4.1.3 M0VABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D......... ........... ..... ............. 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown............ ...... 3/4 1-18 Rod Drop Time....................................... .... 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits... .......... .. ........... 3/4 1-21 FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER-FOUR LOOP OPERATION...... ..... ..... 3/4 1-22 O
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| WOLF CREEK - UNIT 1 IV
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| \s_ / LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................ 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4
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| ; FIGURE 3.2-2 K(Z)-NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT... 3/4 2-5 9
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| 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R................................................. 3/4 2-8 FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN 0PERATION................................. 3/4 2-9 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-11 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-14 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-15 3/4.3 INSTRUMENTATION
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| ' () 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................
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| TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................
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| 3/4 3-1 3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................ 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................ 3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..................................... 3/4 3-14 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS...................... 3/4 3-22 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-29 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-34 WOLF CREEK - UNIT 1 V
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 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 Movable Incore Detectors................................. 3/4 3-43 Seismic Instrumentation.................................. 3/4 3-44 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-45 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-46 Meteorological Instrumentation........................... 3/4 3-47 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-48 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-49 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 MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-55 Chlorine Detection Systems............................... 3/4 3-56 Fire Detection Instrumentation........................... 3/4 3-57 TABLE 3.3-11 FIRE DETECTION INSTRUMENTS........................... 3/4 3-58 Loose-Part Detection System.............................. 3/4 3-62 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-63 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION............................ ....... 3/4 3-64 O
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| WOLF CREEK - UNIT 1 VI
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| 1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS (s
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| SECTION PAGE INSTRUMENTATION (Continued)
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| TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-66 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-68 TABLE 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION.................................... 3/4 3-69 TABLE 4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-72 3/4.3.4 TURBINE OVERSPEED PROTECTION............................. 3/4 3-76 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 C'ld o Shutdown - Loops Fi11ed............................. 3/4 4-5 Cold Shutdown - Loops Not Fil1ed......................... 3/4 4-6 f'~'s
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| \s , 3/4.4.2 SAFETY VALVES Shutdown............................................... 3/4 4-7 0perating.............................................. 3/4 4-8 3/4.4.3 PRESSURIZER.............................................. 3/4 4-9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-11 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION......................... 3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-18 Ope rati o nal Le akage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-19 l
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| O WOLF CREEK - UNIT 1 VII l
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES........................ ...... ............... 3/4 4-21 3/4.4.7 CHEMISTRY................................................ 3/4 4-22 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...... ........ 3/4 4-23 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS................ ................. ..... 3/4 4-24 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 >
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| 1 pCi/ GRAM DOSE EQUIVALENT I-131............ .... 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................ .............. . 3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.............. ........... ....... 3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY...... . ... ...... . ..... 3/4 4-30 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY........ ..... .... ....... 3/4 4-31 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE........ . ............ 3/4 4-32 Pressurizer............. ... ...... . ... ... ........... 3/4 4-33 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 3/4.4.11 REACTOR COOLANT SYSTEM VENTS. .. ......... ..... ..... . 3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.. ........ ... ... .. ....... . ... ...... 3/4 5-1 O
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| WOLF CREEK - UNIT 1 VIII
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS J(' SECTION PAGE 3/4.5.2 ECCS SUBSYSTEMS - T avg
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| > 350 F........................... 3/4 5-3 t 3/4.5.3 ECCS SUBSYSTEMS - T avg
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| < 350 F........................... 3/4 5-7 3/4.5.4 ECCS SUBSYSTEMS - T,yg 5 200 F........................... 3/4 5-9 3/4.5.5 REFUELING WATER STORAGE TANK............................. 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.................................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 Containment Air Locks.................................... 3/4 6-4 Internal Pressure........................................ 3/4 6-6 Air Temperature.......................................... 3/4 6-7 Containment Vessel Structural Integrity.................. 3/4 6-8 Containment Ventilation System.................. ........ 3/4 6-11
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| 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.......... ...................... 3/4 6-13 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 TABLE 3.6-1 CONTAINMENT ISOLATION VALVES.......................... 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers....................................... 3/4 6-31 Hydrogen Control Systems................................. 3/4 6-32 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................ 3/4 7-1 G
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| WOLF CREEK - UNIT 1 IX
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE O
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| TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH IN0PERABLE 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 Aux i l i a ry Feedwa te r Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 SAMPLE AND ANALYSIS PR0 GRAM.................. ...... 3/4 7-8 Main Steam Line Isolation Valves......... ............... 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM..................... ..... 3/4 7-11 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK........................... ........... 3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM............... 3/4 7-14 3/4.7.7 EMERGENCY EXHAUST SYSTEM................................. 3/4 7-17 3/4.7.8 SNUBBERS...................... .......................... 3/4 7-19 FIGURE 4.7-1 SAMPLING PLAN 2) FOR SNUBBER FUNCTIONAL TEST......... 3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION.................. ........... 3/4 7-25 O
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| WOLF CREEK - UNIT 1 X
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS O
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| V SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System............................ 3/4 7-27 Spray and/or Sprinkl er Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-30 Halon Systems............................................ 3/4 7-32 Fire Hose Stations....................................... 3/4 7-33 TABLE 3.7-3 FIRE H0SE STATIONS.................................... 3/4 7-34 3/4.7.11 FIRE BARRIER PENETRATIONS................................ 3/4 7-36 3/4.7.12 AREA TEMPERATURE MONITORING.............................. 3/4 7-37 TABLE 3.7-4 AREA TEMPERATURE MONITORING........................... 3/4 7-38 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES i 0perating................................................ 3/4d-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-7 i 's Shutdown................................................. 3/4 8-8 j % 3/4.8.2 D.C. SOURCES 0perating................................................ 3/4 8-9 l
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| TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-11 Shutdown................................................. 3/4 8-12 l
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| l 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-13 Shutdown................................................. 3/4 8-15 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices.......................... .......... 3/4 8-16 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES...................... 3/4 8-18 l
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| WOLF CREEK - UNIT 1 XI
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| LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 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 BUILDING PENETRATIONS. . ... .... ....... .. 3/4 9-4 3/4.9.5 COMMUNICATIONS.. ...... .. ................ . . ........ 3/4 9-5 3/4.9.6 REFUELING MACHINE......... ................ ..... ..... . 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY.. ... ..... . 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level. .................. ...... ........... 3/4 9-9 Low 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................... ......... ......... ... 3/4 9-13 3/4.9.11 WATER LEVEL - STORAGE P0OL ......... ........ ........ . 3/4 9-14 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE........ . . ................. 3/4 9-15 FIGURE 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 1
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| s LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS f
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| SECTION PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................... 3/4 10-1 f
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| 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3 PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS.................................... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-2 Dose....................................... ............. 3/4 11-5
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| ''~s Liquid Radwaste Treatment System......................... 3/4 11-6 s_, Liquid Holdup Tanks...................................... 3/4 11-7
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| ! 3/4.11.2 GASEOUS EFFLUENTS
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| -Dose Rate................................................ 3/4 11-8 TABLE 4.11-2 RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-9 Dose-Noble Gases......................................... 3/4 11-12 Dose-Iodine-131 and 133, Tritium and Radioactive Material in Particulate Form............... 3/4 11-13 Gaseous Radwaste Treatment System........................ 3/4 11-14 Explosive Gas Mixture.................................... 3/4 11-15 i Gas Storage Tanks........................................ 3/4 11-16 d
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| WOLF CREEK - UNIT 1 XIII
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| e LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE O
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| 3/4.11.3 SOLID RADI0 ACTIVE WASTES.... ............................ 3/4 11-17 3/4.11.4 TOTAL 00SE........... .............. .................... 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM.............. ....................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES...... .... ... .......... 3/4 12-9 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS... .. ... . ....... .... . ....... 3/4 12-10 3/4.12.2 LAND USE CENSUS..... ....... .......... ........... ..... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM... . ............... . 3/4 12-14 O
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| O WOLF CREEK - UNIT 1 XIV
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| BASES l
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| \v SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1' B0 RATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE 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
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| '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 DIFFERENCE VERSUS N
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| / 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.0VERSPEED PROTECTION.............................. B 3/4 3-6 i
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| 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 4-1 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 4
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| BASES SECTION PAGE O
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| REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL T0VGHNESS................... ...... B 3/4 4-10 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................ B 3/4 4-11 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RTNDT FOR REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550 F.................... ...... B 3/4 4-12 3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/4 4-16 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................ B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. 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 O
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| WOLF CREEK - UNIT 1 XVI
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| BASES
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| )
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| J 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........... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM. .......................... 8 3/4 7-3 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM............................ B 3/4 7-3 3/4.7.5 U LT I MAT E H E AT S I N '( . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 SNUBBERS.................................................. B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS.................................. B 3/4 7-7 g
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| 3/4.7.11 FIRE BARRIER PENETRATIONS................................. B 3/4 7-7 3/4.7.12 AREA TEMPERATURE MONITORING............................... B 3/4 7-8 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.. ................ B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS.......................................... . B 3/4 9-1 1
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| O i
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| \s WOLF CREEK - UNIT 1 XVII
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| BASES SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING MACHINE.... ................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY. . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............ B 3/4 9-2 3/4.9.9 CONTAINMENT 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......................... ................. B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS. .. B 3/4 10-1 3/4.10.3 PHYSICS TESTS................ .......... ................. B 3/4 10-1 3/4.10.4 REACTOR-COOLANT L00PS................... ................. B 3/4 10-1 3/4.10.5 POSITI0d INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.............. ......................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS............................ ........... B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTES............... ... .......... B 3/4 11-5 3/4.11.4 TOTAL DOSE... ..... .................................... B 3/4 11-5 3/4.12 RADI0 ACTIVE ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM... .. ... . .. ........ .. ... ...... B 3/4 12-1 3/4.12.2 LAND USE CENSUS.. ........................ .............. B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM............ . ........ B 3/4 12-2 O
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| WOLF CREEK - UNIT 1 XVIII
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| l DESIGN FEATURES J
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| ') !
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| SECTION- PAGE 5.1 SITE l
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| ;_ 5.1.' 1 EXCLUSION AREA.............................................. .5-1
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| ' 5.1.2 LOW POPULATION Z0NE......................................... 5-1 '
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| t i
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| 5.1.3 . MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY
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| ! FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS.............. 5-1 j FIGURE 5.1-1 EXCLUSION AREA....................................... 5 j FIGURE 5.1-2 LOW POPULATION Z0NE.................................. 5-3 '
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| ; FIGURE 5.1-3 BOUNDARY FOR GASE0US EFFLUENTS....................... 5-4 1
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| : i L FIGURE 5.1-4 BOUNDARY FOR LIQUID EFFLUENTS........................ 5-5 '
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| ! 5.2 -CONTAINMENT i
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| 5.2.1 CONFIGURATION............................................... '5-1 i
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| i-5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-1 5.3 REACTOR CORE l 5.~ 3.1 FUEL ASSEMBLIES............................................. 5-6 ,
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| l 5.3.2 CONTROL R00 ASSEMBLIES...................................... 5-6
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| : i. 1
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| ;- 5.4 REACTOR COOLANT SYSTEM l- 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-6 i
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| [ 5.4.2 V0LUME...................................................... 5-6 i
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| l 5.5 METEOROLOGICAL TOWER L0 CATION................................. 5-6 i-l 5.6 FUEL STORAGE t i
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| l' 5.6.1 CRITICALITY................................................. 5-7 >
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| 5.6.2 DRAINAGE.................................................... 5-7 L
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| 5.6.3 CAPACITY.................................................... 5-7 ,
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| [ FIGURE 5.6-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE ,
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| . IN REGION 2........................................ 5-8
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| - 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-7 i
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| TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMIT................... 5-9
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| ; ' WOLF CREEK - UNIT 1 XIX y -,~,-,yy e-<=*g----,--r-,,e:.,-*,,.-.s, _sm,,m.-e-..+.4.m- - , . . . .-.--m,-..m. ~ . - . -- - , . . - - . , - - - - - - - - . - - . . - -
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| ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY.............. . ............... ............... 6-1 6.2 ORGANIZATION 6.2.1 0FFSITE...................... .................. ....... ... 6-1 6.2.2 UNIT STAFF.............. ..... .... .... ....... .......... 6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION............................... . 6-3 FIGURE 6.2-2 UNIT ORGANIZATION.......................... ......... 6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION. ...................... 6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)
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| Function.......... ......... ............... ............. 6-6 Composition. ............................................... 6-6 Responsibilities........ ....... .............. ............ 6-6 Records. ............................................. .... 6-6 6.2.4 SHIFT TECHNICAL ADVISOR..... ......... ......... ....... ... 6-6
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| : 6. 3 UNIT STAFF QUALIFICATIONS........................ .......... . 6-6 6.4 TRAINING.................. .. .... ......... ......... ....... 6-7 6.5 REVIEW AND AUDIT
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| ! 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)
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| Function......... . ... .......... ........................ 6-7 Composition................................. .............. 6-7 Alternates....... ... ...... .................... . . ..... 6-7 3 Meeting Frequency...... .. ....... ............. .......... 6-8 Quorum.......................... .... ................. ... 6-8 Responsibilities....................... ......... .......... 6-8 1
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| Records........ ................... ....... ............... 6-9 l
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| WOLF CREEK - UNIT 1 XX
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| ADMINISTRATIVE CONTROLS N
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| SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE (NSRC)
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| Function.................................................... 6-9 Composition................................................. 6-10 Alternates.................................................. 6-10 Consultants................................................. 6-10 Meeting Frequency........................................... 6-10 Quorum...................................................... 6-10 Review...................................................... 6-11 Audits...................................................... 6-11 Records..................................... .............. 6-12 6.6 REPORTABLE EVENT ACTI0N....................................... 6-13 6.7 SAFETY LIMIT VIOLATION........................................ 6-13 6.8 PROCEDURES AND PR0 GRAMS....................................... 6-13 6.9 REPORTING REQUIREMENTS
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| \
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| 6.9.1 ROUTINE REP 0RTS............................................. 6-17 Startup Report.............................................. 6-17 Annual Reports.............................................. 6-17 Annual Radiological Environmental Operating Report.......... 6-18 Semiannual Radioactive Effluent Release Report.............. 6-19 Monthly Operating Report.................................... 6-20 Radial Peaking Factor Limit Report. . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.9.2 SPECIAL REP 0RTS............................................. 6-21 6.10 RECORD RETENTION............................................. 6-21 a
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| +
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| f 1
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| 0
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| .I WOLF CREEK - UNIT 1 XXI
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| ADMINISTRATIVE CONTROLS SECTION PAGE 6.11 RADIATION PROTECTION PROGRAM......... ....................... 6-22 6.12 HIGH RADIATION AREA.......................................... 6-22 6.13 PROCESS CONTROL PROGRAM (PCP)................................ 6-24 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)....................... 6-24 6.15 MAJOR CHANGES TO LIOUID, GASE0US, AND SOLIO RADWASTE TREATMENT SYSTLMS....................... 6-24 9
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| 9
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| O WOLF CREEK - UNIT 1 XXII
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| u,_._,a.m -a _-4 y _--.r . .aL -aw_ .-._mwam._,_m a_.s e -m_. A_e..A4.m*- 4._ww_. m a m.A 4 .c A___-4 _a_,_, 2 _ s a_ . A._ 4h, ,
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| 4 i
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| I l
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| i l
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| l l
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| i SECTION 1.0 DEFINITIONS 1
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| . -- _ . - - . ~ .
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| i e
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| f ,
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| /
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| 1.0 DEFINIJIONS Thedefinedtermsoftf(sse'ctionappearincapitalizedtypeandareapplicable 4- throughout these Technical Specifications.
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| 4 ACTION
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| .1.1 ACTION shall be that part of a Technic'al Specification which prescribes remedial measures required under designated conditions.
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| ACTUATId LOGIC TEST
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| ) 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated
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| ' input combinations in conjunction with each possible interlock logic state and
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| ~
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| verification of the required logic output. The ACTUATION LOGIC TEST shall i- include a continuity check, as a minimum, of output devices.
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| ANALOG CHANNEL OPERATIONAL TES 1.3 An ANALOG CHANNEL OPERATIONAL TEST shah be the injection of a simulated ,
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| signal into the' channel as close to the sensor.as practicable to verify OPERABILITY of ala'rm, interlock and/or trip functkms. 'The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necebsary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required l
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| range and accuracy. r 3 ?
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| lL AXIAL FLUX DIFFERENCE ,
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| /
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| 1.4= AXIAL TLUX DIFFERENCE shall be the difference in normalized f, lux signals ~
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| ; between the top and bottem halves of a two section excore neutron detector.
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| ! CHANNELCALIBRATIOy f f i 1.5- A CHANNEL CALIBRATION shall be the adjestment, as necessary, of the
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| ~
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| channel such that it responds within the required range and accuracy to known t
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| : j. values of input. - The' CHANNEL CALIBRATION shall encompass the entire channel including the sensors and a.larm, interlock and/or trip functions and may be performed by any series of se W wtiel, over, Tapping, or total channel steps such that the entire channd 4,y. :alibrated. ,
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| CHANNEL CHECK.
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| 1.6' A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by' observation. .This determination shall include, where possible, comparison of the channeV indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. . i T
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| i j '
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| l+
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| ^
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| WOLF CREEK - UNIT 1 1-1/
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| . _ , , , , , . . . . . _ . _ , ._m _ _ .m.- . _ . ..., _,,m._ , , _ , , . _ . _ , _ _ , . _ _
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| DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
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| : a. All penetrations required to be closed during accident conditions are either:
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| : 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
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| : 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.
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| : b. All equipment hatches are closed and sealed,
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| : c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
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| : d. The containment leakage rates are within the limits of Specification 3.6.1.2, and
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| : e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
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| CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.
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| 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.
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| Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
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| DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
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| E - AVERAGE DISINTEGP,ATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
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| WOLF CREEK - UNIT 1 1-2 1
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| , DEFINITIONS
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| (/ )
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| v ENGINEERED SAFETY FEATURES RESPONSE TIME
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| , 1.12 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
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| FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
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| IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:
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| 4 a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
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| : b. Leakage into the containment atmosphere from sources that are both-specifically located and known either not to interfere with the v operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or i
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| : c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant-System.
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| > MASTER RELAY TEST 1.15 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
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| MEMBER (S) 0F THE PUBLIC 1.16 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
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| A WOLF CREEK - UNIT 1 1-3
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| DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.17 The 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.
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| OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be ''PERABLE or have OPERABILITY when it is capable of performing its specified function (s),
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| and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
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| OPERATIONAL M0DE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.
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| PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, or (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
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| PRESSURE B0UNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
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| PROCESS CONTROL PROGRAM 1
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| 1.22 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demon:;trated processing of actual or simulated wet solid wastes will be accomplished in such a way as to l assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State l regulations, burial ground requirements, and other requirements governing the disposal of the radioactive waste.
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| 1 PURGE - PURGING l
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| l 1.23 PURGE or PURGING shall be any controlled process of discharging air or l gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
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| WOLF CREEK - UNIT 1 1-4
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| DEFINITIONS O
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| s, ) QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
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| RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3411 MWt.
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| REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.
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| REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 30.73 to 10 CFR Part 50.
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| SHUTDOWN MARGIN Ns/
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| 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
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| SITE B0UNDARY 1.29 The SITE B0UNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
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| SLAVE RELAY TEST 1.30 A SLAVE RELAY TEST shall be the energization of each slave relay and
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| ,. verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include
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| ; a continuity check, as a minimum, of associated testable actuation devices.
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| SOLIDIFICATION l 1.31 SOLIDIFICATION shall be the conversion of wet wastes into a form that l meets shipping and burial ground requirements.
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| I l
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| 1 l
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| i WOLF CREEK - UNIT 1 1-5 l
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| l
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| DEFINITIONS SOURCE CHECK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
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| STAGGERED TEST BASIS 1.33 A STAGGERED TEST BASIS shall consist of:
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| : a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and
| |
| : b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
| |
| THERMAL POWER 1.34 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.
| |
| TRIP ACTUATING DEVICE OPERATIONAL TEST 1.35 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.
| |
| UNIDENTIFIED LEAKAGE 1.36 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.
| |
| UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE B0UNDARY ;
| |
| access to which is not controlled by the licensee for purposes of protection of i individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, com-mercial, institutional, and/or recreational purposes.
| |
| VENTILATION EXHAUST TREATMENT SYSTEM 1.38 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gaseous exhaust stream prior to the release to the environment. l Such a system is not considered to have any effect on noble gas effluents.
| |
| Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered l to be VENTILATION EXHAUST TREATMENT SYSTEM components. l i
| |
| WOLF CREEK - UNIT 1 1-6 l
| |
| | |
| DEFINITIONS VENTING 1.39 VENTING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
| |
| WASTE GAS HOLDUP SYSTEM 1.40 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off- !
| |
| gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
| |
| 1 w.
| |
| WOLF CREEK - UNIT 1 1-7
| |
| | |
| TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours.
| |
| D At least once per 24 hours.
| |
| W At least once per 7 days.
| |
| M At least once per 31 days.
| |
| Q At least once per 92 days.
| |
| SA At least once per 184 days.
| |
| R At least once per 18 months.
| |
| S/U Prior to-each reactor startup.
| |
| N.A. Not applicable.
| |
| P Completed prior to each release.
| |
| O O
| |
| WOLF CREEK - UNIT 1 1-8
| |
| | |
| l 1
| |
| l l
| |
| TABLE 1.2 OPERATIONAL MODES i
| |
| REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER
| |
| * TEMPERATURE l ff
| |
| : 1. POWE! OPERATION -> 0.99 > 5% > 350*F
| |
| : 2. STAR 10P > 0.99 5 5% > 350*F ;
| |
| : 3. HOT STANDBY < 0.99 0 > 350*F
| |
| : 4. HOT SHUTDOWN < 0.99 0 350*F > T
| |
| > 200 F avg
| |
| : 5. COLD SHUTDOWN < 0.99 0 1 200*F
| |
| : 6. REFUELING ** 1 0.95 0 $_140*F
| |
| * Excluding decay heat. ,
| |
| ** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
| |
| (
| |
| I J
| |
| e f
| |
| 4
| |
| ! s l WOLF CREEK - UNIT 1 1-9 i
| |
| | |
| 4 F
| |
| i SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY, SYSTEM SETTINGS r
| |
| i l
| |
| l f
| |
| 4
| |
| , - ,, , - _ . , _ , ,-n,, - , , - - ,wwt,-- w
| |
| | |
| 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS f''N
| |
| - \)
| |
| 2.1 SAFETY LIMITS REACTOR CORE 4
| |
| 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1 for four loop operation.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTION:
| |
| Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY within 1 hour, and comply with the requirements of Specification 6.7.1.
| |
| REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
| |
| APPLICABILITY: MODES 1, 2, 3, 4, and 5.
| |
| O kg_ ,) ACTION:
| |
| MODES 1 and 2:
| |
| Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System p'ressure within its limit within 1 hour, and comply with the requirements of Specification 6.7.1.
| |
| MODES 3, 4, and 5:
| |
| Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.
| |
| . O WOLF CREEK - UNIT 1 2-1 i
| |
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| 560 0 0.2 0.4 0.6
| |
| * 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS N OPERATION O
| |
| WOLF CREEK - UNIT 1 2-2
| |
| | |
| ,_ SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall 4
| |
| be set consistent with the Trip Setpoint values shown in Table 2.2-1.
| |
| APPLICABILITY: As shown for each channel in Table 3.3-1.
| |
| ACTION:
| |
| 4
| |
| : a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.
| |
| With the Reactor Trip System Instrumentation or Interlock Setpoint b.
| |
| i less conservative than the value shown in the Allowable Values
| |
| . column of Table 2.2-1, either:
| |
| . 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 f
| |
| \
| |
| was satisfied for the affected channel, or
| |
| : 2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
| |
| Equation 2.2-1 Z + R + S < TA Where:
| |
| Z= The value from Column Z of Table 2.2-1 for the affected channel, R= The "as measured" value (in percent span) of rack error for the affected channel, S= Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowances) of Table 2.2-1 for the affected channel.
| |
| i
| |
| ;o WOLF CREEK - UNIT 1 2-3 1
| |
| . < - . , , - - - - ~ ~ , _ , . , . - , - - . . . , - . - . - - , . . - _ _ _ ,_
| |
| | |
| TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS o
| |
| SENSOR E TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE
| |
| : 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.
| |
| E U 2. Power Range, Neutron Flux s a. High Setpoint 7.5 4.56 0 $109% of RTP* 1112.3% of RTP*
| |
| : b. Low Setpoint 8.3 4.56 0 $25% of RTP* $28.3% of RTP*
| |
| : 3. Power Range, Neutron F)ux, 2. 4 0.5 0 High Positive Rate 54% of RTP* with 56.3% of RTP* with a time constant a time constant 12 seconds 22 seconds
| |
| : 4. Power Range, Neutron Flux, 2.4 0.5 0 High Negative Rate 54% of RTP* with 56.3% of RTP* with a time constant a time constant y 22 seconds a 22 seconds
| |
| : 5. Intermediate Range, 17.0 8.41 0 Neutron Flux 525% of RTP* $35.3% of RTP*
| |
| : 6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.6 x 105 cps
| |
| : 7. Overtemperature AT 7. 6 3.76 1.73 See Note 1 See Note 2
| |
| + 0.67
| |
| : 8. Overpower AT 5.5 1,43 0.16 See Note 3 See Note 4
| |
| : 9. Pressurizer Pressure-Low 3.7 0.71 2.49 21875 psig 21866 psig
| |
| : 10. Pressurizer Pressure-High 7. 5 0.71 2.49 52385 psig $2400 psig
| |
| : 11. Pressurizer Water Level-High 8.0 2.18 1.96 192% of instrument $93.9% of instrument span span
| |
| *RTP = RATED THERMAL POWER
| |
| ** Loop design flow = 95,700 gpm O O O
| |
| | |
| .. . - ..- . -- - -- . . =- . . . . . .. -.
| |
| TABLE 2.2-1 (Continued) 8 r- REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l 9 SENSOR lE TOTAL ERROR FUNCTIONAL UNIT (S) TRIP SETPOINT
| |
| , 7 ALLOWANCE (TA) Z_ -ALLOWABLE VALUE i c-
| |
| : 12. Reactor Coolant Flow-Low 3.1 2.27 0.6 .>90% of loop >89.1% of loop design flow ** design flow ** -
| |
| I 13. Steam Generator Water 23.5 21.18 2.51 >23.5% of narrow' >22.3% of narrow Level Low-Low range instrument range instrument span span
| |
| ! 14. Undervoltage - Reactor 7.5 1. 3 0 >10578 Volts A.C. >10355 Volts A.C.
| |
| ! Coolant Pumps l
| |
| : 15. Underfrequency - Reactor 3.3 0 0 >57.2 Hz >57.1 Hz j Coolant Pumps i
| |
| : 7 c'
| |
| : 16. Turbine Trip
| |
| ; a. Low Fluid Oil Pressure N.A. N.A. N.A. >590.00 psig >534.20 psig I b. Turbine Stop Valve N.A. N.A. N.A. >1% open >1% open l Closure 1
| |
| l 17. Safety Injection Input N.A. N.A. N.A. N.A. N.A.
| |
| from ESF I
| |
| i u
| |
| l i
| |
| | |
| TABLE 2.2-1 (Continued) 5 q REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS n
| |
| = SENSOR S TOTAL ERROR 7 FUNCTIONAL UNIT ALLOWANCE (TA) Z, (S) TRIP SETPOINT ALLOWABLE VALUE s 18. Reactor Trip System
| |
| $ Interlocks
| |
| : a. Intermediate Range N.A. N.A. N.A. 21 x 10 1 amps 16 x 10 11 amps Neutron Flux, P-6
| |
| : b. Low Power Reactor Trips Block, P-7
| |
| : 1) P-10 input N.A. N.A. N.A. 10% of RTP* >6.7% to < 13.3%
| |
| of RTP* -
| |
| m 2) P-13 input N.A. N.A. N.A.
| |
| cn 510% of RTP* 512.4% of RTP*
| |
| Turbine Impulse Turbine Impulse Pressure Pressure Equivalent Equivalent
| |
| : c. Power Range Neutron N.A. N.A. N.A. 548% of RTP* $51.3% of RTP*
| |
| Flux, P-8
| |
| : d. Power Range Neutron N.A. N.A. N.A. 550% of RTP* $53.3% of RTP*
| |
| Flux, P-9
| |
| : e. Power Range Neutron N.A. N.A. N.A 10% of RTP* >6.7% to < 13.3%
| |
| Flux, P-10 of RTP* -
| |
| : f. Turbine Impulse Chamber N.A. N.A. N.A $10% of RTP* $12.4% of RTP*
| |
| Pressure, P-13 Turbine Impulse Turbine Impulse Pressure Equivalent Pressure Equivalent
| |
| : 19. Reactor Trip Breakers N.A. N.A. N.A N.A. N.A.
| |
| : 20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A.
| |
| Logic
| |
| *RTP = RATED THERMAL POWER O O O
| |
| | |
| ! o o .
| |
| i i TABLE 2.2-1 (Continued)
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| ;9 TABLE NOTATIONS n
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| l 5
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| m NOTE 1: OVERTEMPERATURE AT 4 b
| |
| l AT 3 (y I Tsg) 5 AT, {K -K 2 i
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| [T (7.f 7s 3) - T'] + K (P - P') - f y(AI)}
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| 3 ;
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| E i
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| ; t
| |
| ~ Where: AT = Measured AT by RTD Manifold Instrumentation; ;
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| l
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| = Lead-lag compensator on measured AT; f
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| Ti, T2 = Time constants utilized in lead-lag compensator for AT, T3 =8s, '
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| j T2 = 3 s;
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| ! 1 j y, g
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| = Lag compensator on measured AT;
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| ! O
| |
| = Time constant utilized in the lag compensator for AT, T3 = 0 s; Ta j AT, = Indicated AT at RATED THERMAL POWER; K3 = 1.10; i i K2 = 0.0137/*F; i
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| 1 T* =
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| The function generated by the lead-lag compensator for T,yg dynamic compensation;
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| { .
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| i
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| = l l T4, T3 Time constants utilized in the lead-lag compensator for T,yg, 14 = 28 s,
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| ] 13 = 4 s; I
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| j T = Average temperature, "F; 1 =
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| : y. 3 Lag compensator on measured T,yg;
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| =
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| 4 Ts Time constant utilized in the measured T,yg lag compensator, is = 0 s; l
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| 4
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| | |
| TABLE 2.2-1 (Continued)
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| :c y TABLE NOTATIONS (Continued) 9 NOTE 1: (Continued)
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| A T'
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| 7 5 588.5 F (Nominal T,yg at RATED THED. MAL POWER);
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| E K3 = 0.000671; Z
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| ~ P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);
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| S = Laplace transform operator, s 2; and f (AI) is a function of the indicated difference between top and bottom detectors of the 3
| |
| power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
| |
| 7 (i) for qt - 9 between b -35% and + 7%, f (AI) = 0, where qtand q b 3 re percent O'
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| RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt *9 IS b
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| total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of g t
| |
| 9bexceeds -35%, the AT Trip Setpoint shall be automatically reduced by 1.26% of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of q t
| |
| 4bexceeds +7%, the AT Trip Setpoint shall be automatically reduced by 1.05% of its value at RATED THERMAL POWER.
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| NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.3% of AT span.
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| O O O
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| | |
| C
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| ( /
| |
| { TABLE 2.2-1 (Continued) 8 q TABLE NOTATIONS (Continued)
| |
| $ NOTE 3: OVERPOWER AT 1 + T S) ( 1 i ) (b ) ( 1 1 AT( (1 +2 T S) (1 3+ T 5) < ATo 4 -K 5{K j 1+157 1 + TsS ) T - Ks [T (1 + TsS) - T"]2
| |
| - f (AI)}
| |
| e i C
| |
| ) 5.
| |
| w 3
| |
| w Where: AT = Measured AT by RTD manifold instrumentation; 1
| |
| 7
| |
| = Lead-lag compensator on measured AT; It. T2 =
| |
| Time constants utilized in lead-lag compensator for AT, I =8s,T2 = 3 s; t
| |
| 1 y, 3
| |
| = Lag compensator on measured AT;
| |
| ] T3 =
| |
| Time constant utilized in the lag compensator for AT, r3 = 0 s;
| |
| , AT = Indicated AT at RATED THERMAL POWER; i
| |
| K4 = 1.08;
| |
| ! r
| |
| =
| |
| , K3 0.02/*F for increasing average temperature and 0 for decreasing average j temperature;
| |
| ; y}'b 3
| |
| =
| |
| The function generated by the rate-lag compensator for T,yg dynamic j compensation; i
| |
| r2 =
| |
| Time constant utilized in the rate-lag compensator for T,yg, T7 = 10 s; 1 =
| |
| 1, 3 Lag compensator on measured T,yg;
| |
| =
| |
| Ts Time constant utilized in the measured T,yg lag compensator, Ts = 0 s;
| |
| | |
| TABLE 2.2-1 (Continued)
| |
| :c
| |
| { TABLE NOTATIONS (Continued)
| |
| O NOTE 3: (Continued) m 7 Ks = 0.00128/ F for T > T" and Ks = 0 for T $ T";
| |
| E T = Average temperature, *F;
| |
| - T" = Indicated T avg at RATED THERMAL POWER (Calibration temperature for AT instrumentation, 5 588.5 F);
| |
| 5 = Laplace transform operator, s 2; and f 2(AI) = 0 for all al.
| |
| I NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than m 4.1% of AT span.
| |
| O O O
| |
| | |
| I s
| |
| 4
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| ?
| |
| i i
| |
| I BASES FOR SECTION 2.0 J
| |
| SAFETY LIMITS
| |
| ; AND l LIMITING SAFETY SYSTEM SETTINGS 1
| |
| i
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| [
| |
| i t
| |
| i a
| |
| r f
| |
| l t
| |
| :l ll
| |
| )
| |
| f I
| |
| l
| |
| | |
| O NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.
| |
| O 1
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| I I
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| l l
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| l 1
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| 0 ,
| |
| | |
| 2.. SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
| |
| Operation above the upper boundary of the nucleace boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation (R-GRID). The W-3 DNB correlation (R-GRID) has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.
| |
| The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated traasients is limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
| |
| The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
| |
| These curves are based on an enthalpy hot channel factor, F H, f 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F H at reduced power based on the expression:
| |
| F g = 1.55 [1+ 0.2 (1-P)]
| |
| Where P is the fraction of RATED THERMAL POWER.
| |
| These limiting heat flux conditions are higher than those calculated for i the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f t(AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent O)
| |
| \ with core Safety Limits.
| |
| WOLF CREEK - UNIT 1 8 2-1
| |
| | |
| SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
| |
| The reactor vessel, pressurizer, and the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which permits a naximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.
| |
| The entire RCS is hydrotested at greater than or equal to 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation.
| |
| O O
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| WOLF CREEK - UNIT 1 8 2-2
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| | |
| 2.2 LIMITING SAFETY SYSTEM SETTINGS V BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.
| |
| To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this n option utilizes the "as measured" deviation from the specified calibration
| |
| / ) point for rack and sensor components in conjunction with a statistical combin-(,/ ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Setpoint.
| |
| : S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from
| |
| , the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift i factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
| |
| The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and i other instrumentation utilized in these channels are expected to be capable of l operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, p in excess of the allowance that is more than occasional, may be indicative of more serious problems and shoul m rrant further investigation.
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| WOLF CREEK - UNIT 1 B 2-3 i
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| | |
| LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
| |
| The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
| |
| Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.
| |
| Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
| |
| The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
| |
| Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
| |
| Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power.
| |
| The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.
| |
| WOLF CREEK - UNIT 1 B 2-4
| |
| | |
| }
| |
| LIMITING SAFETY SYSTEM SETTINGS n
| |
| BASES i
| |
| l Intermediate and Source Range, Neutron Flux .
| |
| ! I The Intermediate and Source Range, Neutron' Flux trips provide core protection during reactor startup to mi.tigate the consequences of an un-controlled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip rof the Power Range, Neutron Flux channels. The Source Range channels will j initiate a Reactor trip at about 105 counts per second unless manually blocked l when P-6 becomes active. The Intermediate Range channels will initiate a ,
| |
| ; Reactor trip at a current level equivalent to approximately 25% of RATED '
| |
| i THERMAL POWER unless manually blocked when P-10 becomes active.
| |
| ! Overtemperature AT g i
| |
| 4 The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power i distribution, provided that the transient is slow with respect to piping i transit delays from the core to the temperature detectors (about 4 seconds),
| |
| l and pressure is within the range between the Pressurizer High and Low Pressure
| |
| ; trips. The Setpoint is automatically varied with: (1) coolant temperature to j correct for temperature induced changes in density and heat capacity of water
| |
| : , and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) presst rizer pressure, and (3) axial power distribution.
| |
| I With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater i than design, as indicated by the difference between top and bottom power range l nuclear detectors, the Reactor trip is automatically reduced according to the j notations in Table 2.2-1. l 4
| |
| l Overpower AT l The Overpower AT Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under_all
| |
| :' possible overpower' conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint
| |
| ; is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change
| |
| '. of temperature for dynamic compensation for piping delays from the core to the i loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpcwer AT trip provides protection to mitigate i the consequences of various size steam breaks as reported in WCAP-9226, " Reactor .
| |
| Core Response to Excessive Secondary Steam Releases."
| |
| I l
| |
| f WOLF CREEK - UNIT 1 B 2-5 i
| |
| i w a,+*e vm.-o-,,-mem-,-v,w<--.---o-+-r- + _me J - 1, m m a s-r,e. ~ - - - -w e~r w --r ~-~~w r v r m w a m nwew - --
| |
| | |
| LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own Trip Setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.
| |
| On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
| |
| The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
| |
| Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pres-sure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
| |
| Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
| |
| On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%
| |
| of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.
| |
| O WOLF CREEK - UNIT 1 B 2-6
| |
| | |
| o
| |
| ,a LIMITING SAFETY SYSTEM SETTINGS ,
| |
| ,,~,
| |
| BASES Steam Generator Water Level -
| |
| TheSteamGeneratorWaterLevelLow-Lowtripprotects'tNereactorlfrom loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specified Setpoint prevides allowances for starting delays of the Auxiliary Feedwater System. ),
| |
| Undervoltage and Underfrequency - Reactor Coolant Pump Busses s
| |
| .. . L> c The Undervoltage and Underfrequency Reactor Coolagt PumpvBus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time, delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips -
| |
| from momentary electrical power transients. For undervoit. age, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two ar more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.
| |
| On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%
| |
| [s of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately
| |
| \ 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.
| |
| Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9.
| |
| Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.
| |
| The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.
| |
| l 2'. '
| |
| b f
| |
| Q 1
| |
| > =
| |
| WOLF CREEK - UNIT 1 B 2-7
| |
| | |
| LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Trip System Interlocks The Reactor Trip System Interlocks perform the following functions:
| |
| P-6 On increasing power, P-6 allows the manual block of the Source Range Reactor trip (i.e., prevents premature block of Source Range trip),
| |
| provides a backup block for Source Range Neutron Flux doubling, and allows deenergization of the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.
| |
| P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked.
| |
| P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the single loop Low Flow trip.
| |
| P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.
| |
| P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and de-energizes the Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated.
| |
| Provides input to P-7.
| |
| P-13 Provides input to P-7.
| |
| O WOLF CREEK - UNIT 1 B 2-8
| |
| | |
| 4 i i i ,
| |
| l !
| |
| e 1
| |
| l i
| |
| i I
| |
| i I SECTIONS 3.0 AND 4.0 l
| |
| i LIMITING CONDITIONS FOR OPERATION f i
| |
| < AND SURVEILLANCE REQUIREMENTS i
| |
| b i
| |
| i 4
| |
| 4 I
| |
| (
| |
| 4 4
| |
| Y P
| |
| 1
| |
| | |
| l 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
| |
| 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
| |
| 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
| |
| : a. At least HOT STANDBY within the next 6 hours,
| |
| [(",)\ b.
| |
| c.
| |
| At least HOT SHUTDOWN within the following 6 hours, and At least COLD SHUTOOWN within the subsequent 24 hours.
| |
| Where corrective measures ara completed that permit operation under the ACTION requirements, the action may oe taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.
| |
| This Specification is not applicable in MODE 5 or 6.
| |
| 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met l without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.
| |
| i l
| |
| [T
| |
| \v)
| |
| WOLF CREEK - UNIT 1 3/4 0-1
| |
| | |
| APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
| |
| 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
| |
| : a. A maximum allowable extension not to exceed 25% of the surveillance interval, but
| |
| : b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
| |
| 4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements ara stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment.
| |
| 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.
| |
| 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
| |
| : a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i);
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| O WOLF CREEK - UNIT 1 3/4 0-2
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| | |
| APPLICABILITY
| |
| [
| |
| SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
| |
| ASME BOILER AND PRESSURE VESSrl REQUIRED FREQUENCIES FOR CODE AND APPLICABLE ADDENDA PERFORMING INSERVICE TERMIN0 LOGY FOR INSERVICE INSPECTION AND TESTING INSPECTION AND TESTING ACTIVITIES ACTIVITIES Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days
| |
| : c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities; i d. Performance of the above inservice inspection and testing activities V shall be in addition to other specified Surveillance Requirements; and
| |
| : e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
| |
| {
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| x WOLF CREEK - UNIT 1 3/4 0-3
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| | |
| 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,yg >200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k for four loop operation.
| |
| APPLICABILITY: MODES 1, 2*, 3, and 4.
| |
| ACTION:
| |
| With the SHUTDOWN MARGIN less than 1.3% Ak/k, 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 equivalent until the required SHUTDOWN MARGIN is restored.
| |
| SURVEILLANCE REQUIREMENTS
| |
| /m
| |
| / T 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal i V to 1.3% Ak/k:
| |
| : a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
| |
| If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be 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,ff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; l
| |
| l
| |
| : c. When in MODE 2 with K,ff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical l control rod position is within the limits of Specification 3.1.3.6;
| |
| : d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification l 4.1.1.1.le. below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and
| |
| *See Special Test Exception Specification 3.10.1.
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| WOLF CREEK - UNIT 1 3/4 1-1 l
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| l i
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| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : e. When in MODE 3 or 4, at least once per 24 hours by consideration of the following factors:
| |
| : 1) Reactor Coolant System boron concentration,
| |
| : 2) Control rod position,
| |
| : 3) Reactor Coolant System average temperature,
| |
| : 4) Fuel burnup based on gross thermal energy generation,
| |
| : 5) Xenon concentration, and
| |
| : 6) Samarium concentration.
| |
| 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% 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.le. above. The predicted reactivity values shall he adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.
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| l l
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| l l
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| l l
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| O WOLF CREEK - UNIT 1 3/4 1-2 l
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| l
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| | |
| __s REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T < 200*F ava LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1% Ak/k.
| |
| APPLICABILITY: MODE 5.
| |
| ACTION:
| |
| With the SHUTDOWN MARGIN less than 1% Ak/k, 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 equivalent until the required SHUTDOWN MARGIN is restored.
| |
| SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% Ak/k:
| |
| p.
| |
| \ a. Within I hour after detection of an inoperable control rod (s) and at
| |
| (\s_-/ least once per 12 hours thereafter while the rod (s) is inoperable.
| |
| If the inoperable control rod is immovable or untrippable, the SHUTOOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or.untrippable control rod (s); and
| |
| : b. At least once per 24 hours by consideration of the following factors:
| |
| : 1) Reactor Coolant System boron concentration,
| |
| : 2) Control rod position,
| |
| : 3) Reactor Coolant System average temperature,
| |
| : 4) Fuel burnup based on gross thermal energy generation,
| |
| : 5) Xenon concentration, and
| |
| : 6) Samarium concentration.
| |
| \.
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| WOLF CREEK - UNIT 1 3/4 1-3
| |
| | |
| REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:
| |
| : a. Less positive than 0 Ak/k/ F for the all rods withdrawn, beginning of cycle life (80L), hot zero THERMAL POWER condition, or
| |
| : b. Less negative than -4.1 x 10 4 Ak/k/ F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.
| |
| APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2#* .
| |
| Specification 3.1.1.3b. - MODES 1, 2, and 3#.
| |
| ACTION:
| |
| : a. With the MTC more positive than the limit of Specification 3.1.1.3a.
| |
| above, operation in MODES 1 and 2 may proceed provided:
| |
| : 1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 Ak/k/ F within 24 hours or be in HOT STANDBY within the next 6 hours.
| |
| These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;
| |
| : 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and
| |
| : 3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
| |
| : b. With the MTC more negative than the limit of Specification 3.1.1.3b.
| |
| above, be in HOT SHUTDOWN within 12 hours.
| |
| *With Keff greater than or equal to 1.
| |
| #See Special Test Exception Specification 3.10.3.
| |
| O WOLF CREEK - UNIT 1 3/4 1-4
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1. 3 The MTC shall be determined to be within its limits during each fuel cycle.as follows:
| |
| : a. The MTC shall be measured and compared to the BOL limit of Specifica-tion 3.1.1.3a., above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and
| |
| : b. The MTC shall be measured at any THERMAL POWER and compared to
| |
| -3.2 x 10 ' Ak/k/ F (all rods withdrawn, RATED THERMAL POWER condi-tion) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates the MTC is more negative than.-3.2 x 10 4 Ak/k/ F, the MTC shall be remeasured, and compared to the E0L MTC limit of Specification 3.1.1.3b. , at least once per 14 EFPD during the remainder of the fuel cycle.
| |
| \
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| v l
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| i i
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| r I
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| WOLF CREEK - UNIT 1 3/4 1-5
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| | |
| REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LINITING CONDITION FOR OPERATION 3.1.1.4 The Reactor shall be greater than Coolant or equalSystem to 551lowest F. operating loop temperature (Tavg)
| |
| APPLICABILITY: MODES 1 and 2#*.
| |
| ACTION:
| |
| With a Reactor Coolant System operating loop temperature (Tavg) less than 551 F, restore T to within its limit within 15 minutes or be in HOT avg STAND 8Y within the next 15 minutes.
| |
| SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (T "9) shall be determined to 8
| |
| be greater than or equal to 551 F:
| |
| : a. Within 15 minutes prior to achieving reactor criticality, and
| |
| : b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T avg is less than 561 F with the T,yg-T ref Deviation Alarm not reset.
| |
| #With Keff greater than or equal to 1.
| |
| *See Special Test Exception Specification 3.10.3.
| |
| O WOLF CREEK - UNIT 1 3/4 1-6
| |
| | |
| REACTIVITY CONTROL SYSTEMS 3/4.1.2 B0 RATION SYSTEMS
| |
| 'Q FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION i
| |
| 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 centrifugal charging pump to the Reactor Coolant System if the Boric Acid Storage System is OPERABLE as given in Specification 3.1.2.5a. for MODES 5 and 6 or as given in Specification 3.1.2.6a. 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 Specification 3.1.2.5b. for MODES 5 and 6 or as given in Specification 3.1.2.6b. for MODE 4.
| |
| i 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 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, 4
| |
| sealed, or otherwise secured in position, is in its correct position.
| |
| i n
| |
| WOLF CREEK - UNIT 1 3/4 1-7
| |
| | |
| REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 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 centrifugal charging pump to the Reactor Coolant System, and
| |
| : b. Two flow paths from the refueling water storage 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 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 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200 F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 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 path that is not locked, sealed, or otherwise secured in position, is in its correct position;
| |
| : b. At least once per 18 months during shutdown by verifying that 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 verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the Reactor Coolant System.
| |
| *The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifagal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first.
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| WOLF CREEK - UNIT 1 3/4 1-8
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| | |
| _ REACTIVITY CONTROL SYSTEMS
| |
| .,m CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One centrifugal charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powerec from an OPERABLE emergency power source.
| |
| APPLICABILITY: MODES 4, 5, and 6.
| |
| ACTION:
| |
| With no centrifugal charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
| |
| ,-_ SURVEILLANCE REQUIREMENTS
| |
| ( )
| |
| 'O 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 dif-ferential pressure of greater than or equal to 2400 psid when tested pursuant to Specification 4.0.5.
| |
| 4.1.2.3.2 All centrifugal charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable
| |
| * at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.
| |
| *An inoperable pump may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
| |
| O b
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| WOLF CREEK - UNIT 1 3/4 1-9
| |
| | |
| REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.*
| |
| ACTION:
| |
| With only one centrifugal charging pump OPERABLE, restore at least two cen-trifugal charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and boratea to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200 F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in HOT SHUTOOWN within the next 6 hours.
| |
| SURVEILLANCE RE0UIREMENTS 4.1.2.4 At least two centrifugal charging pumps 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 Specification 4.0.5.
| |
| t
| |
| *The provisions of 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 Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first.
| |
| O WOLF CREEK - UNIT 1 3/4 1-10
| |
| | |
| REACTIVITY CONTROL SYSTEMS
| |
| . ,9
| |
| ( BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 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) with:
| |
| : 1) A minimum contained borated water volume of 55,416 gallons,
| |
| : 2) A minimum boron concentration of 2000 pp1, 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 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) Verifying the contained borated water volume, and
| |
| : 3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.
| |
| : b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 37 F.
| |
| O V
| |
| WOLF CREEK - UNIT 1- 3/4 1-11
| |
| | |
| REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2, and 3 and one of the following borated water sources shall be OPERABLE as required by Specifica-tion 3.1.2.1 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 boron, 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 2000 and 2100 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 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200 F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
| |
| : b. With the RWST inoperable in MODE 1, 2, or 3, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
| |
| : c. With no borated water source OPERABLE in MODE 4, restore one borated water source to OPERABLE status within 6 hours or be in COLD SHUTDOWN within the following 30 hours.
| |
| O WOLF CREEK - UNIT 1 3/4 1-12
| |
| | |
| REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each required borated water source shall be demonstrated OPERABLE:
| |
| : a. At least once per 7 days by:
| |
| : 1) Verifying the boron concentration in the water,
| |
| : 2) Verifying the contained borated water volume of the water source, and
| |
| : 3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.
| |
| .b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 37*F or greater than 100 F.
| |
| O O
| |
| WOLF CREfK - UNIT 1 3/4 1-13
| |
| | |
| 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 (indicated position) of their group step counter demand position.
| |
| APPLICABILITY: MODES 1* and 2*.
| |
| ACTION:
| |
| : a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours.
| |
| : b. With more than one full-length rod inoperable or misaligned from the group step counter demand position by more than 12 steps (indicated position), be in HOT STANDBY within 6 hours.
| |
| : c. With one full-length rod trippable but inoperable due to causes other than addressed t.y ACTION a. above, or misaligned from its group step counter den:and height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour:
| |
| : 1. The rod is restored to OPERABLE status within the above alignment requirements, or
| |
| : 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2. The THERMAL POWER level shall be restricted pursuant to Specifica-tion 3.1.3.6 during subsequent operation, or
| |
| : 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. 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; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours;
| |
| *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
| |
| O WOLF CREEK - UNIT 1 3/4 1-14
| |
| | |
| REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore detectors and F (Z) and F H are verified to be 9
| |
| within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.
| |
| O'u 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 hours except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours.
| |
| 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.
| |
| O WOLF CREEK - UNIT 1 3/4 1-15
| |
| | |
| TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)
| |
| Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 1-16
| |
| | |
| REACTIVITY CONTROL SYSTEMS
| |
| ()
| |
| O POSITION INDICATION SYSTEMS-0PERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within i 12 steps.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTION: ,
| |
| : a. With a maximum of one digital rod position indicator per bank inoper-able either:
| |
| : 1. Determine the position of the nonindicating rod (s) indirectly by the movable incore detectors at least once per 8 hours and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
| |
| : 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER s within 8 hours.
| |
| As _s b. With a maximum of one demand position indicator per bank inoperable either:
| |
| : 1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
| |
| : 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation monitor is inoper-able, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours.
| |
| l WOLF CREEK - UNIT 1 3/4 1-17
| |
| | |
| REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluding demand position indica-tion) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted.
| |
| APPLICABILITY: MODES 3*#, 4*#, and 5*#.
| |
| ACTION:
| |
| With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.
| |
| SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicator agrees with the demand position indicator within 12 steps when exercised over the full range of rod travel at least once per 18 months.
| |
| *With the Reactor Trip System breakers in the closed position.
| |
| #See Special Test Exception Specification 3.10.5.
| |
| i l
| |
| O' WOLF CREEK - UNIT 1 3/4 1-18 l
| |
| | |
| d REACTIVITY CONTROL SYSTEMS
| |
| /7 R00 DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage 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 full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
| |
| : b. With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided
| |
| ['~'N THERMAL POWER is restricted to less than or equal to 66% of RATED t, THERMAL POWER.
| |
| SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
| |
| : a. For all rods following each removal of the reactor vessel head,
| |
| : b. For specifically affected individual rods following any maintenance r on or modification to the Control Rod Drive System which could l affect the drop time of those specific rods, and I
| |
| : c. At least once per 18 months.
| |
| l
| |
| ,V WOLF CREEK - UNIT 1 3/4 1-19
| |
| | |
| REACTIVITY CONTROL SYSTEMS SHUTOOWN R0D INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.
| |
| APPLICABILITY: MODES 1* and 2*#.
| |
| ACTION:
| |
| With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either:
| |
| : a. Fully withdraw the rod, or
| |
| : b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
| |
| SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined fully withdrawn:
| |
| : a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
| |
| : b. At least once per 12 hours thereaf ter.
| |
| *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
| |
| #With Keff greater than or equal to 1.
| |
| O WOLF CREEK - UNIT 1 3/4 1-20 l
| |
| | |
| REACTIVITY CONTROL SYSTEMS l Q CONTROL R0D INSERTION LIMITS T
| |
| - LIMITING CONDITION FOR OPERATION 4
| |
| 3.1.3.6 The control banks shall be limited in physical insertion as shown in
| |
| - Figure-3.1-1.
| |
| APPLICABILITY: MODES 1* and 2*#.
| |
| 2 ACTION:
| |
| With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant'to Specification 4.1.3.1.2:
| |
| E a. Restore the control banks to within the limits within 2 hours, or
| |
| : b. Reduce THERMAL POWEt within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figure, or '
| |
| : c. Be in at least HOT STANDBY within 6 hours.
| |
| i Q SbRVEILLANCEREQUIREMENTS
| |
| )
| |
| 1
| |
| , _ 4.1.3.6 The position of each controlcbank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hou s, f
| |
| j' *See Special Test Exceptions Specifications'3.10.2 and 3.10.3.
| |
| #With K,ff greater than or equal to 1.
| |
| I WOLF CREEK - UNIT 1 3/4 1-21 Y ..
| |
| i
| |
| _ ---_. - -- ~ _ . _ _
| |
| | |
| (FULLY WITHDRAWN) 228 "
| |
| F~-
| |
| ^
| |
| (29.3%, 228);. '
| |
| -(79.3%, 228)L .
| |
| 220 .o .- - . . .
| |
| .. x .:_. t J r. :2.n _ 2cc q=_._ _ ;r - 4 n ra- + = _
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| nr. . u- : _ _ _ -l,m._..__,_. __ __ m_+_
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| __. .._= .. m..
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| r_ t,. _._ ._._2 =___, : . _ n_ . r- ---
| |
| 4 _ __7. _.__ J .___ J-- ,;__ _. . r
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| :=. = .: }: r +2 = :-~; = t r _ .+ r ::r r n- - tr-- - -j._ _ - _- t -- - : : -
| |
| _:= 2n=ub- -- __.=_ p -
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| _ _ _;-- , = : =;c
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| _._._. - B AN K B .___.___ L _
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| . _ . . . _ . _ _ - 4 rr =_. =__ =. :=._n.=_4. .
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| c _ tra
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| :::r.}., =: = _== = t= tar. = r =.; r =rr-gg ,.._.t-- :- -a g-- : = f _:t=n== =n d' =--(100%,161) 160 (0%,161).::=" ._=. - - c . : ;: ---- '
| |
| 2 . . - : =--- ----- _.: r i:. ... _-_ j=_ .;. .- I: -dn=,
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| o x r u . : : == r _ , ._ _ :x = u:;_. c . a : =xr , _:
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| .c :___.__. .
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| .t g =. c:i__=...2_n__.=_...
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| .x._... zr._ : 2; . :
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| . . _ _ . .c-. =. .--- . .._.arr- - -_
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| _.4.._..___=._ = F _,- 1.___ _.
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| , - : = . f === t = . . ---:.= +. . -
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| --L a... .
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| g a_ _ ~ _E5I T-'
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| , -= . . J.r. n ar't=~2K.
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| = : _ u.;_ _ . _ _..
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| :=
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| _ . _ . - - =_{'_=_=_.... _,
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| 2
| |
| =_=. _ =_ . n_ =7 ..__ : ~r rM n
| |
| _~_ .;, 7, m .
| |
| _ _ _ .___=. _ _r_. g' . =7Iqr ' ". ~ ~ ~_~_..__._i._....._
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| +
| |
| ;- =nar cr+ _; _: ._2= n- . =n: . 2.
| |
| - ;-.x- .- .-- ,-- r ---p.r--
| |
| 2
| |
| = r.2 2 m 100
| |
| . - - - - -d lx x.=
| |
| x:- jrux mt _u._ __ .-
| |
| o -~. =-
| |
| -:-- l = =-
| |
| - =:=,
| |
| * +
| |
| . : ==_=_ =! ===
| |
| : n. c. = 7 r -- --- 1- .2 = ~ . = { r=-l=: = =;= =+_= _ =n . _
| |
| g -.-.- -- _. ;. .r_--. . q=' =~ =~ h_=~ _r. i . . . .
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| . _ . _ . . . _ . .1.,_....___.__.. . _ . _ _
| |
| g -t=" j =-. _-n= -.. -_. } -.<._._=r .. .. .l.a .:;~. t = t _ _;.
| |
| ' -- ' ~
| |
| ~ - ' - - ~ ' ~ ~
| |
| +--
| |
| 80 -
| |
| $ _. 2? c . c _::i Mii:r[p."ijgiriityg- g"- ~''dj{ gBANK DEj~Q.,_ _{-g' , _
| |
| O - = - .
| |
| . : t =- m i rd--- - -- :f _ -r .u.,- - , _ _ . t- :=1- --
| |
| O ; a . _: ; __u :. . x m._ - -- y 2= :r , n ,; ;y--
| |
| _ _ , . _.=
| |
| x %
| |
| t
| |
| _7_: n im7r y,,--- . 21 c;=;=r_c
| |
| _3 ._ :m m l,-r -;rnj::r:T: , . :+--
| |
| l .1_-. - : t - rx a_ . .
| |
| .+.-
| |
| -- - d=-Ez i
| |
| ~ '~
| |
| (0%,46)__ '
| |
| .r -
| |
| . - .i- E 3..._._..
| |
| .__g._. _
| |
| ;; . __m - _._ ,b; . _..
| |
| QQ - .:. - - .
| |
| _..____.._~_ __ _ _ . . . _ _
| |
| (30.2%, 0) .
| |
| 0 0 20 40 60 80 100 l RELATIVE POWER (Percent)
| |
| (FULLY INSERTED) 1 l
| |
| FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER-FOUR LOOP OPERATION WOLF CREEK - UNIT 1 3/4 1-22
| |
| | |
| b' .
| |
| p 3/4.2 POWER DISTRIBUTION LIMITS e '
| |
| j
| |
| (
| |
| i 4 3/4.2.1 AXIAL FLUX DIFFGRENCE y /
| |
| LIMITING CONDITION FOR OPERATION .
| |
| 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:
| |
| : a. 1 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU, and
| |
| : b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTV. r The indicated AFD may devia't'e outside the above i required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indicated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumulative penalty' deviation time does not exceed 1 hour during the
| |
| ~~
| |
| previous 24 hours. .
| |
| 1 The indicated AFD may deviate outside the above required target band at greater than 15% but less .than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time doesi not exceed 1 hour during the previous 24 hours.
| |
| APPLICABILITY: MODE 1 above 15% of RATED THERMAL POWER *.
| |
| ACTION: I C a.' With the indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes, either:
| |
| : 1. Restore the indicated AFD to within the above required target band limits, or
| |
| : 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
| |
| : b. With the indicated AFD oatside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce: r l 1. THERMAL POWER to less than 50% of RATED THERMAL POWER i within 30 minutes, and
| |
| : 2. The Power Range Nedtron Flux - High,'Setpoint.1 to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
| |
| 1
| |
| ^5ee Special Test Exception Specification 3.10.2.
| |
| # Surveillance testing of the Power Range Neutron Flux Channel may be performed i pursuant to~ Specification 4.3.1.1 provided the indicated AFD is maintaindd within the Acceptable, Operation Limits of-Figure 3.2-1. A total of 16 hours o'p'erition may be accumulated with the AFD outside of the above pequired target band during
| |
| ( testing without penalty deviation.
| |
| WOLF CREEK - UNIT 1 3/4'2-1
| |
| )
| |
| 9
| |
| | |
| POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| : c. With the indicated AFD outside of the above required target band for more than 1 hour of cumulated penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.
| |
| SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
| |
| : a. Monitoring the indicated AFD for each OPERABLE excore channel:
| |
| : 1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
| |
| : 2) At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status.
| |
| : b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to ex 4 t during the interval preceding each logging.
| |
| 4.2.1.2 The indicated AFD shall be considered outside of its target band when two er more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:
| |
| : a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and
| |
| : b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
| |
| 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.
| |
| The provisions of Specification 4.0.4 are not applicable.
| |
| 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.
| |
| WOLF CREEK - UNIT 1 3/4 2-2 l
| |
| | |
| E o a g a e w 4 O E
| |
| $ 5 E
| |
| 100 UNACCEPTABLE ( 11, 90) (11, 90) UNACCEPTABLE ac OPERATION OPERATION w
| |
| N O
| |
| E 5
| |
| E o ACCEPTABLE OPERATION 4"
| |
| m O
| |
| t i.ai. 50) (ai 503 6
| |
| 5
| |
| * o.
| |
| 20 0
| |
| -50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE ( Al)%
| |
| FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER WOLF CREEK - UNIT 1 3/4 2-3
| |
| | |
| POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 Fq (Z) shall be limited by the following relationships:
| |
| F0 (Z) $ [2.32] [K(Z)] for P > 0.5, and P
| |
| F9 (Z) 5 [4.64] [K(Z)] for P $ 0.5.
| |
| Where:
| |
| P = THERMAL POWERRATED THERMAL POWER , and K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
| |
| APPLICABILITY: MODE 1.
| |
| ACTION:
| |
| With F (Z) exceeding its limit:
| |
| 9
| |
| : a. Reduce THERMAL POWER at least 1% for each 1% q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit; and 0
| |
| : b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a,, above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. 9 O
| |
| WOLF CREEK - UNIT 1 3/4 2-4
| |
| | |
| t
| |
| )
| |
| :c o
| |
| t-
| |
| -ri n
| |
| :o m 1.2 m
| |
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| |
| _,i..... .._ ]
| |
| l
| |
| ! e I l ' I c- (6.0,1.0) ! !
| |
| l 2 .
| |
| ~ t
| |
| -4 1.0
| |
| } :
| |
| 1-
| |
| ' l M
| |
| .f' i. . .l ; ,
| |
| M(10.8, 0.94),
| |
| _. _. _ _. l 1 i
| |
| ! i l I
| |
| ; i ; '
| |
| i i, j ,
| |
| ! 6 . I j f . . . . - - . . . . . ..
| |
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| |
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| |
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| |
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| |
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| |
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| |
| 'i o . . . .! : ! -
| |
| w '
| |
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| |
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| |
| w N
| |
| ; -I-t 4---
| |
| (12.0,0.6,5)].- ,
| |
| 0.6 -j- - - F -
| |
| _J . '
| |
| 'L I
| |
| q j f, E
| |
| t '
| |
| - *b~ - - - * '
| |
| - N'*
| |
| N '
| |
| g . . . . . . ...g..
| |
| O ! '
| |
| N g . ! . . . . .. . ._. ._ ; .
| |
| , i ;-
| |
| U' l
| |
| _ 0.4 - -- - t--- - -t
| |
| * N .
| |
| ' 2 i . .
| |
| ! l .
| |
| i : I
| |
| - . . - i .. . '. - .. . . . . . .h .-.. S
| |
| . y_ _h.. 1 0.2 - i- '
| |
| . 4 l .-.. '.. '
| |
| . . - . . . -.- y . . .y .. . ;._. . -
| |
| ._. r-:
| |
| ---
| |
| * r i .n
| |
| ! t
| |
| .! -i_ ..g
| |
| ' ' I 0
| |
| O 2 4 6 8 10 12 14 16 CORE HEIGHT (FT)
| |
| FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT n
| |
| | |
| POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
| |
| 4.2.2.2 F xy shall be evaluated to determine if F (Z) is within its limit by:
| |
| q
| |
| : a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER;
| |
| : b. Increasing the measured F component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties;
| |
| : c. Comparing the F computed (F ) obtained in Specification 4.2.2.2b.,
| |
| above to:
| |
| : 1) The F limits for RATED THERMAL POWER (F P) for the appropriate xy measured core planes given in Specifications 4.2.2.2e. and f.,
| |
| below, and
| |
| : 2) The relationship:
| |
| F =F P [1+0.2(1-P)]
| |
| Where F ' is the limit for fractional THERMAL POWER operation RTP express as a function of F x and P is the fraction of RATED THERMAL POWER at which F xy w s measured.
| |
| : d. Remeasuring F xy according to the following schedule:
| |
| : 1) When F is greater than the F P limit for the appropriate measured core plane but less than the F relationship, additional P
| |
| power distribution maps shall be taken a d F y compared to F and F xy either:
| |
| a) Within 24 hours after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F y was last determined, or b) At least once per 31 EFPD, whichever occurs first.
| |
| O WOLF CREEK - UNIT 1 3/4 2-6
| |
| | |
| _ - . . _- .- . . .. - - __ ._= .
| |
| POWER DISTRIBUTION LIMITS ,
| |
| ' SURVEILLANCE-REQUIREMENTS (Continued)
| |
| : 2) When the F is less than or equal to the F limit for the y x appropriate measured core plane, additional power distribution maps'shall'be taken and F compared to F P and F xyl at least once per 31 EFPD.
| |
| RTP
| |
| : e. The F xy limits for RATED THERMAL POWER (Fxy ) shall be provided for all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.9;
| |
| : f. The F xy-limits of-Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
| |
| : 1) Lower core region from 0 to 15%, inclusive,
| |
| : 2) Upper. core region from 85 to 100%, inclusive,
| |
| \
| |
| 3)- Grid plane regions at 17.8 i 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%
| |
| and 74.9 1 2%,-inclusive, and
| |
| : 4) Core plane regions within i 2% of core height (i'2.88 inches) about. the bank demand position of the Bank "D" control rods.
| |
| L g .' With F exceeding Fxy, the effects of Fxy n Fq (Z) shall be evaluated to determine if F q (Z) is within its limits.
| |
| 4.2.2.3 When qF (Z) is measured for other than F xy determinations, an overall i measured F (Z) shall be obtained from a power. distribution map and increased 9
| |
| l~ by 3% to account for; manufacturing tolerances and further increased by 5% to account for measurement uncertainty.
| |
| l-I WOLF CREEK - UNIT 1 3/4 2-7
| |
| | |
| POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation.
| |
| Where:
| |
| N H
| |
| a* ,
| |
| R = 1.49 [1.0 + 0.2 (1.0 - P)]
| |
| b* P = THERMAL POWER , and RATED THERMAL POWER c.
| |
| F H = Measured values of F H obtained by using the movable incore detectors to obtain a power distribution map. The measured values of F H shall be used to calculate R since Figure 3.2-3 includes measurement uncertainties of 2.8% for flow and 4%
| |
| for incore measurement of F g.
| |
| APPLICABILITY: MODE 1.
| |
| ACTION:
| |
| With the combination of RCS total flow rate and R outside the region of acceptable operation shown on Figure 3.2-3:
| |
| : a. Within 2 hours either:
| |
| : 1. Restore the combination of RCS total flow rate and R to within the above limits, or
| |
| : 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours,
| |
| : b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours; and O
| |
| WOLF CREEK - UNIT 1 3/4 2-8
| |
| | |
| b MEASUREMENT UNCERTAINTIES OF2.8% FOR FLOW AND 4.0%
| |
| IN THIS FOR FIGURE INCORE MEASUREMENT OF F NA"AREINCLUDED 48 .
| |
| .: :. e. :.L,. ...
| |
| )J: .
| |
| . . . .:. ... L.
| |
| : }' .. .. . . ...
| |
| =
| |
| . . e. ..
| |
| : n.,
| |
| Ihi .: . !!!; t
| |
| ! t ii@
| |
| i,
| |
| .a au '
| |
| 46 ((.[ .
| |
| = : . -..
| |
| = ....a: . = - :-. .=. .. .. . n - -
| |
| =$ .. . t. : . .
| |
| ~-.
| |
| g:: := : , . . . . . . .
| |
| ..!:gi+ . . . . . . i'IE in ; ;j ACCEPTABLE ; ;
| |
| . ii:1 UNACCEPTABLE 2 :j OPERATION i mis m'
| |
| : ali- OPERATION 0- ...: REGION
| |
| ~
| |
| c ... .
| |
| REGION .
| |
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| 0.90 0.95 1.00 1.05 1.10 R=F AH N /1.49 (1.0 + 0.2(1.0-P)l J l
| |
| FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R FOUR LOOPS IN OPERATION WOLF CREEK - UNIT 1 3/42-9
| |
| | |
| POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| : c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:
| |
| : 1. A nominal 50% of RATED THERMAL POWER,
| |
| : 2. A nominal 75% of RATED THERMAL POWER, and
| |
| : 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.
| |
| SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
| |
| 4.2.3.2 The combination of indicated RCS total flow rate and R shall be determined to be within the region of acceptable operation of Figure 3.2-3:
| |
| : a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
| |
| : b. At least once per 31 Effective Full Power Days.
| |
| 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours when the most recently obtained value of R obtained per Specification 4.2.3.2, is assumed to exist.
| |
| 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.
| |
| 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.
| |
| 4.2.3.6 The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months.
| |
| WOLF CREEK - UNIT 1 3/4 2-10
| |
| | |
| POWER DISTRIBUTION LIMITS I
| |
| O J 3/4.2.4 QUADRANT POWER TILT RATIO V
| |
| LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANf POWER TILT RATIO shall not exceed 1.02.
| |
| APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *.
| |
| ACTION:
| |
| : a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
| |
| o
| |
| : 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
| |
| a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or 1
| |
| b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
| |
| 3 2. Within 2 hours either:
| |
| )
| |
| a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
| |
| : 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours, and
| |
| : 4. Identify and correct the cause of the out of-limit condition -
| |
| . subsequent POWER OPERATION prior to increasing above 50% THERMAL of RATED THERMAL POWER mPOWER;dy proceed provided that th i
| |
| QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%
| |
| or greater RATED THERMAL POWER.
| |
| "See Special Test Exception Specification 3.10.2.
| |
| , v WOLF CREEK - UNIT 1 3/4 2-11
| |
| | |
| POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| : b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
| |
| : 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
| |
| a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
| |
| : 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes;
| |
| : 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% oi RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
| |
| : 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or 'intil verified acceptable at 95%
| |
| or greater RATED THERMAL POWER.
| |
| : c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
| |
| l
| |
| : 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
| |
| a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
| |
| O WOLF CREEK - UNIT 1 3/4 2-12
| |
| | |
| POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| : 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
| |
| : 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER.
| |
| : d. The provisions of Specification 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ( ,/ 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
| |
| : a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and
| |
| : b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.
| |
| 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours.
| |
| ( -
| |
| WOLF CREEK - UNIT 1 3/4 2-13
| |
| | |
| POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:
| |
| a.
| |
| Reactor Coolant System T,yg, and
| |
| : b. Pressurizer Pressure.
| |
| APPLICABILITY: MODE 1.
| |
| ACTION:
| |
| With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours.
| |
| O WOLF CREEK - UNIT 1 3/4 2-14
| |
| | |
| i t
| |
| TABLE 3.2-1 DNB PARAMETERS l
| |
| i t LIMITS Four Loops in.
| |
| j PARAMETER Operation Indicated Reactor Coolant System T,yg 5 592.5*F Indicated Pressurizer Pressure > 2220 psig*
| |
| l l
| |
| i
| |
| (
| |
| i i
| |
| * Limit not applicable during either a THERMAL POWER ramp in excess of 5% of l RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of l RATED THERMAL POWER.
| |
| l I
| |
| ; WOLF CREEK - UNIT 1 3/4 2-15 l- i
| |
| | |
| (O)
| |
| ,,, 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
| |
| APPLICABILITY: As shown in Table 3.3-1.
| |
| ACTION:
| |
| As shown in Table 3.3-1.
| |
| i SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and
| |
| ! f- s) the automatic trip logic shall be demonstrated OPERABLE by the performance of
| |
| \s_s/ the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
| |
| 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
| |
| Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the
| |
| " Total No. of Channels" column of Table 3.3-1.
| |
| e 4
| |
| WOLF CREEK - UNIT 1 3/4 3-1 i-
| |
| | |
| TABLE 3.3-1 g REACTOR TRIP SYSTEM INSTRUMENTATION r-MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
| |
| : 1. Manual Reactor Trip 2 1 2 1, 2 1 E 2 1 2 3*, 4*, 5* 10 Z
| |
| ~ 2. Power Range, Neutron Flux
| |
| : a. High Setpoint 4 2 3 1, 2 2#
| |
| : b. Low Setpoint 4 2 3 1###, 2 2#
| |
| : 3. Power Range, Neutron Flux 4 2 3 1, 2 2#
| |
| High Positive Rate
| |
| : 4. Power Range, Neutron Flux, 4 2 3 1, 2 2#
| |
| High Negative Rate s
| |
| : 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3
| |
| : 6. Source Range, Neutron Flux
| |
| : a. Startup 2 1 2 2##** 4
| |
| : b. Shutdown 2 1 2 3**, 4, 5 5
| |
| : 7. Overtemperature AT Four Loop Operation 4 2 3 1, 2 6#
| |
| : 8. Overpower AT Four Loop Operation 4 2 3 1, 2 6#
| |
| : 9. Pressurizer Pressure-Low 4 2 3 1 6#
| |
| : 10. Pressurizer Pressure-High 4 2 3 1, 2 6#
| |
| 9 9 9
| |
| | |
| < O O TABLE 3.3-1 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION r-MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
| |
| [ 11. Pressurizer Water Level-High 3 2 2 1 7#
| |
| z Z 12. Reactor Coolant Flow - Low
| |
| , w j a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 7#
| |
| j any oper- each oper-ating loop ating loop j b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop in 1 7#
| |
| below P-8) two oper- each oper-ating loops ating loop f { 13. Steam Generator Water 4/sta. gen. 2/sta. gen. 3/sta. gen. in 1, 2 6#
| |
| Level-Low-Low in any oper- each oper-
| |
| { ating stm.
| |
| gen.
| |
| ating stm.
| |
| gen.
| |
| l 14. Undervoltage-Reactor Coolant l Pumps 4-2/ bus 2-1/ bus 3 1 6#
| |
| I
| |
| : 15. Underfrequency-Reactor Coolant Pumps 4-2/ bus 2-1/ bus 3 1 6#
| |
| l 16. Turbine Trip
| |
| : a. Low Fluid Oil Pressure 3 2 2 1 11# '
| |
| : b. Turbine Stop Valve Closure 4 4 1 1 11#
| |
| : 17. Safety Injection Input from ESF 2 1 2 1, 2 9
| |
| | |
| TABLE 3.3-1 (Centinued) g REACTOR TRIP SYSTEM INSTRUMENTATION 9
| |
| n MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
| |
| [ 18. Reactor Trip System Interlocks z
| |
| U a. Intermediate Range w Neutron Flux, P-6 2 1 2 2## 8
| |
| : b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 or P-13 Input 2 1 2 1 8
| |
| , c. Power Range Neutron g Flux, P-8 4 2 3 1 8 Y d. Power Range Neutron Flux, P-9 4 2 3 1 8
| |
| : e. Power Range Neutron Flux, P-10 4 2 3 1, 2 8
| |
| : f. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8
| |
| : 19. Reactor Trip Breakers 2 1 2 1, 2 9 2 1 2 3*, 4*, 5* 10
| |
| : 20. Automatic Trip and Interlock Logic 2 1 2 1, 2 9 2 1 2 3*, 4*, 5* 10 0 0 0
| |
| | |
| 7' 'S TABLE 3.3-1 (Continued) i /
| |
| \ '' TABLE NOTATIONS
| |
| *0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
| |
| **The boron dilution flux doubling signal may be blocked during reactor startup in accordance with normal operating procedures.
| |
| #The provisions of Specification 3.0.4 are not applicable.
| |
| ##Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
| |
| ###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
| |
| ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours.
| |
| ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
| |
| : a. The inoperable channel is placed in the tripped condition gNg within 1 hour;
| |
| \ )
| |
| 's / b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1; and
| |
| : c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL F0WER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.
| |
| ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
| |
| : a. Below the P-6 (Intermediate Range Neutron Flux Interlock)
| |
| Setpoint, restore the inoperable channel to OPERABLE status prior to increacing THERMAL POWER above the P-6 Setpoint; or
| |
| : b. Above the P-6 (Intermediate Range Neutron Flux Interlock)
| |
| Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing
| |
| ,_s THERMAL POWER above 10% of RATED THERMAL POWER.
| |
| ( ,) ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.
| |
| WOLF CREEK - UNIT 1 3/4 3-5
| |
| | |
| TABLE 3.3-1 (Continued)
| |
| ACTION STATEMENTS (Continued)
| |
| ACTION 5 - a. With the number of OPERABLE channels one less than the Mini-mum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reac-tor Trip Breakers, suspend all operations involving positive reactivity changes and verify valves BG-V178 and BG-V601 are closed and secured in position within the next hour.
| |
| : b. With no channels OPERABLE, open the Reactor Trip Breakers, suspend all operations involving positive reactivity' changes and verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within I hour and every 12 hours thereafter, and verify valves BG-V178 and BG-V601 are closed and secured in position within 4 hours and verified to be closed and secured in position every 14 days.
| |
| ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
| |
| : a. The inoperable channel is placed in the tripped condition within 1 hour; and
| |
| : b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1.
| |
| ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour.
| |
| ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
| |
| ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.
| |
| ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor trip breakers within the next hour.
| |
| ACTION 11 - With the number of OPERABLE channels less than the Total Number of Cleannels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour.
| |
| WOLF CREEK - UNIT 1 3/4 3-6
| |
| | |
| TABLE 3.3-2 g REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E
| |
| E FUNCTIONAL UNIT RESPONSE TIME lR 7 1. Manual Reactor Trip N.A.
| |
| E 2. Power Range, Neutron Flux 5 0.5 second*
| |
| Z
| |
| ~ 3. Power Range, Neutron Flux, High Positive Rate N.A.
| |
| : 4. Power Range, Neutron Flux, High Negative Rate 5 0.5 second*
| |
| : 5. Intermediate Range, Neutron Flux N.A.
| |
| : 6. Source Range, Neutron Flux N.A.
| |
| : 7. Overtemperature AT $ 6.0 seconds
| |
| * w E u
| |
| " 8. Overpower AT 5 6.0 seconds *
| |
| : 9. Pressurizer Pressure-Low 5 2.0 seconds
| |
| : 10. Pressurizer Pressure-High 5 2.0 seconds
| |
| : 11. Pressurizer Water Level-High N.A.
| |
| * Neutron detectors are exempt from response' time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
| |
| | |
| TABLE 3.3-2 (Continuad)
| |
| REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES g
| |
| 9 g FUNCTIONAL UNIT RESPONSE TIME A
| |
| ^ 12. Reactor Coolant Flow-Low e
| |
| E a. Single Loop (Above P-8) 5 1.0 second p b. Two Loops (Above P-7 and below P-8) 5 1.0 second w
| |
| : 13. Steam Generator Water Level-Low-Lew 5 2.0 seconds
| |
| : 14. Undervoltage-Reactor Coolant Pumps 5 1.5 seconds
| |
| : 15. Underfrequency-Reactor Coolant Pumps 5 0.6 second
| |
| : 16. Turbine Trip R a. Low Fluid Oil Pressure N.A.
| |
| A b. Turbine Stop Valve Closure N.A.
| |
| Y m 17. Safety Injecti^n Input from ESF N.A.
| |
| : 18. Reactor Trip System Interlocks N.A.
| |
| : 19. Reactor Trip Breakers N.A.
| |
| : 20. Automatic Trip and Interlock Logic N.A.
| |
| O O O
| |
| | |
| m O O TABLE 4.3-1 g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS r-TRIP 9 ANALOG ACTUATING MODES FOR E
| |
| * CHANNEL DEVICE WHICH
| |
| ' CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED E
| |
| Q 1. Manual Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3*, 4*, 5*
| |
| w
| |
| : 2. Power Range, Neutron Flux
| |
| : a. High Setpoint S D(2, 4) M N.A. N.A. 1, 2 M(3, 4)
| |
| Q(4, 6)
| |
| R(4, 5)
| |
| : b. Low Setpoint S R(4) M N.A. N.A. 1###, 2
| |
| : 3. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2
| |
| { High Positive Rate T
| |
| : 4. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 High Negative Rate
| |
| : 5. Intermediate Range, S R(4, 5) S/U(1),M N.A. N.A. 1###, 2 Neutron Flux
| |
| : 6. Source Range, Neutron Flux 5 R(4, 5, 12) S/U(1),M(9) N.A. N.A. 2##, 3, 4, 5
| |
| : 7. Overtemperature AT S R(13) M N.A. N.A. 1, 2
| |
| : 8. Overpower AT S R H N.A. N.A. 1, 2
| |
| : 9. Pressurizer Pressure-Low S R M N.A. N.A. 1
| |
| : 10. Pressurizer Pressure-High S R M N.A. N.A. 1, 2
| |
| : 11. Pressurizer Water Level-High S R M N.A. N.A. 1
| |
| : 12. Reactor Coolant Flow-Low S R M N.A. N.A. 1
| |
| | |
| TABLE 4.3-1 (C:ntinued) g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS r-TRIP 2 ANALOG ACTUATING MODES FOR S
| |
| CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED E
| |
| Z 13. Steam Generator Water Level- S R M N.A. N.A. 1, 2
| |
| - Low-Low
| |
| : 14. Undervoltage - Reactor N.A. R N.A. M N.A. 1 Coolant Pumps
| |
| : 15. Underfrequency - Reactor N.A. R N.A. M N.A. 1 Coolant Pumps
| |
| : 16. Turbine Trip R
| |
| : a. Low Fluid Oil Pressure N.A. R N.A. S/U(1, 10) N.A. 1
| |
| : b. Turbine Stop Valve N.A. R N.A. S/U(1, 10) N.A. 1 Y Closure E$
| |
| : 17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESF
| |
| : 18. Reactor Trip System Interlocks
| |
| : a. Intermediate Range Neutron Flux, P-6 N.A. R(4) M N.A. N.A. 2##
| |
| : b. Low Power Reactor Trips Block, P-7 N.A. R(4) M(8) N.A. N.A. 1
| |
| : c. Power Range Neutron Flux, P-8 N.A. R(4) M(8) N.A. N.A. 1
| |
| : d. Power Range Neutron Flux, P-9 N.A. R(4) M(8) N.A. N.A. 1 O O O
| |
| | |
| 4 s
| |
| TABLE 4.3-1 g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
| |
| ; r TRIP l E ANALOG ACTUATING MODES FOR
| |
| ! E
| |
| * CHANNEL DEVICE WHICH i CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE 1
| |
| [ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED z
| |
| G 18. Reactor Trip System Interlocks (Continued)
| |
| ~
| |
| l . e. Power Range ,
| |
| j Neutron Flux, P-10 N.A. R(4) M(8) N.A. N.A. 1, 2 i
| |
| ! f. Turbine Impulse Chamber
| |
| ! Pressure, P-13 N.A. R M(8) N.A. N.A. 1
| |
| : 19. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, 2, 3*, 4*, 5*
| |
| R 20. Automatic Trip and i
| |
| [ Interlock Logic N.A. N.A. N.A. N.A. M (7) 1, 2, 3*, 4*, 5*
| |
| s i
| |
| I i
| |
| i I
| |
| i 4
| |
| | |
| TABLE 4.3-1 (Continued)
| |
| TABLE NOTATIONS
| |
| *0nly if the Reactor Trip System breakers happen to be closed and the control rod drive system 13 capable of rod withdrawal.
| |
| ##Below P-6 (Internediate Range Neutron Flux Interlock) Setpoint.
| |
| ###Below P-10 (Lod Setpoint Power Range Neutron Flux Interlock) Setpoint.
| |
| (1) If not performed in previous 7 days.
| |
| (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 cr 1.
| |
| (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15%
| |
| of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applic-able for entry into MODE 2 or 1.
| |
| (4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
| |
| (5) Detector plateau curves shall be obtained, evaluated and compared to manu-facturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
| |
| (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provi-sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
| |
| (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
| |
| (8) With power greater than or equal to the interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window.
| |
| (9) Monthly surveillance in MODES 3*, 4* and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Monthly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period.
| |
| (10) Setpoint verification is not required.
| |
| (11) At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips.
| |
| (12) At least once per 18 months during shutdown, verify that on a simulated Boron Dilution Doubling test signal the normal CVCS discharge valves will close and the centrifugal charging pumps suction valves from the RWST will open within 30 seconds.
| |
| (13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.
| |
| O WOLF CREEK - UNIT 1 3/4 3-12
| |
| | |
| INSTRUMENTATION (v}j 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION h.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentatio;.
| |
| channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
| |
| APPLICABILITY: As shown in Table 3.3-3.
| |
| ACTION:
| |
| : a. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Trip Setpoint column but more conser-vative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
| |
| : b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conser-vative than the value shown in the Allowable Values column of Table 3.3-4, either:
| |
| : 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
| |
| : 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent Q with the Trip Setpoint value.
| |
| Equation 2.2-1 Z + R + S < TA Where:
| |
| Z = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel,
| |
| : c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
| |
| SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall he demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
| |
| 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function
| |
| , shall be demonstrated to be within the limit at least once per 18 months. Each O
| |
| t /
| |
| test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested V at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shnwn in the " Total No. of Channels" Column of Table 3.3-3.
| |
| WOLF CREEK - UNIT 1 3/4 3-13
| |
| | |
| TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM Q TOTAL NO. CHANNELS - CHANNELS APPLICABLE _
| |
| y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
| |
| , 1. Safety Injecticn, (Reactor - -
| |
| c- Trip, Phase "A" Isolation, '
| |
| i'i Feedwater Isolation, Compo-
| |
| * nent Cooling Water, Turbine H Trip, Auxiliary Feedwater-Motor-Driven Pump, Emergency Diesel Generator Operation, Containment Cooling, and Essential Service Wat r l Operation)
| |
| : a. Manual Initiation 2 1 2 1,2,3,4 18 w b. Automatic Actuation 2 1 2 1,2,3,4 14
| |
| ) Logic and Actuation w Relays (SSPS) h c. Containment 3 2 2 1,2,3 15*
| |
| Pressure-High-1
| |
| , d. Pressur izer 4 2 3 1,2,3 # 19*
| |
| l Pressure - Low
| |
| : e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 1,2,3 # 15*
| |
| Low any steam line ,
| |
| : 2. Containment Spray
| |
| : a. Manual Initiation 2 pair 1 pair 2 pair 1,2,3,4 18 operated simul-taneously
| |
| : b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays (SSPS)
| |
| : c. Containment Pressure- 4 2 3 1,2,3 16 High-3 O O O
| |
| | |
| O O O TABLE 3.3-3 (Continued) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
| |
| : 3. Containment Isolation e '
| |
| 3 a. Phase "A" Isolation w , 1) Manual Initiation 2 1 2 1,2,3,4 18
| |
| : 2) Automatic Actuation 2 1 2 1,2,3,4 14
| |
| , Logic and Actuation Relays (SSPS) i 3) Safety Injection See Item 1. above for all Safety Injection initiating functions and
| |
| . requirements.
| |
| ; b. Phase "B" Isolation
| |
| { 1) Manual Initiation 2 pair 1 pair 2 pair 1,2,3,4 18
| |
| , operated 4 simul-m taneously
| |
| : 2) Automatic Actuation 2 1 2 1,2,3,4 14 :
| |
| Logic and Actuation !
| |
| Relays (SSPS)
| |
| : 3) Containment 4 2 3 1,2,3 16 Pressure-High-3
| |
| : c. Containment Purge Isolation i 1) Manual Initiation 2 1 2 1,2,3,4 17
| |
| : 2) Automatic Actuation 2 1 2 1,2,3,4 17 Logic and Actuation Relays (SSPS) i
| |
| ! 3) Automatic Actuation 2 1 2 1,2,3,4 17 !
| |
| l Logic and Actuation !
| |
| l Relays (BOP ESFAS) l
| |
| : 4) Phase "A" Isolation See Item 3.a. for all Phase "A" Isolation initiating functions and ,
| |
| requirements.
| |
| | |
| TABLE 3.3-3 (Crntinued) c ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM 9 TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
| |
| : 4. Steam Line Isolation C
| |
| : a. Manual Initiation
| |
| : 1) Individual 1/ steam line 1/ steam line 1/ operating 1,2,3 23 steam line
| |
| : 2) System 2 1 2 1,2,3 22
| |
| : b. Automatic Actuation 2 1 2 1,2,3 21 Logic and Actuation Relays (SSPS) w c. Containment Pressure- 3 2 2 1,2,3 15*
| |
| 'A High-2
| |
| : d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15*
| |
| Pressure-Low any steam line
| |
| : e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 3## 15*
| |
| Negative Rate-High any steam line
| |
| : 5. Turbine Trip &
| |
| Feedwater Isolation
| |
| : a. Automatic Actuation 2 1 2 1, 2 27 Logic and Actuation Relay (SSPS)
| |
| : b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 19*
| |
| Water Level- in any oper- in each oper-High-High ating stm. gen. ating stm.
| |
| gen.
| |
| : c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
| |
| O O O
| |
| | |
| .. . . . - . = . ._. . . . - - . - . - ._-- . . - . . - . - - - . . - .
| |
| i TABLE 3.3-3 (Continued) f ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION R
| |
| n MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE ,
| |
| p FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION-
| |
| ' 6. Auxiliary Feedwater
| |
| : a. Manual Initiation 3(1/ pump) 1/ pump 1/ pump 1,2,3 24
| |
| : b. Automatic Actuation Logic 2 1 2 1,2,3 21 and Actuation Relays (SSPS)
| |
| : c. Automatic Actuatien 2 1 2 1,2,3 21 Logic and Actuation Relays (BOP ESFAS)
| |
| : d. Stm. Gen. Water Level-
| |
| , Low-Low k
| |
| , 1) Start Motor- '
| |
| 4 Driven Pumps 4/sta. gen. 2/sta. gen. 3/stal. gen. 1,2,3 19*
| |
| w in any opera- in each ting stm. gen. operating stm. gen.
| |
| : 2) Start Turbine-Driven Pump 4/sta. gen. 2/sta. gen. 3/sta. gen. 1,2,3 19*
| |
| in any in each 2 operating operating stm. gen. stm. gen.
| |
| : e. Safety Injection - Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions and requirements.
| |
| : f. Loss-of-Offsite Power -
| |
| Start Turbine-Driven Pump 2 1 2 1,2,3 22
| |
| | |
| TABLE 3.3-3 (Cantinued) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 9
| |
| MINIMUM E
| |
| " TOTAL NO. CHANNELS CHANNELS APPLICABLE m FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i 6. Auxiliary Feedwater (Continued)
| |
| E g. Trip of All Main H Feedwater Pumps -
| |
| s Start Motor-Driven Pumps 4-(2/ pump)"* 2-(1/ pump 3 1, 2 19 in same separation)
| |
| : h. Auxiliary Feed-water Pump Suction Pressure-Low (Transfer to ESW) 3 2 2 1,2,3 15*
| |
| M 7. Automatic Switchover to -
| |
| [ Containment Sump E
| |
| * a. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays (SSPS) l b. RWST Level - Low-Low 4 2 3 1,2,3,4 16 l Coincide t With l Safety Ir.jection See Item 1. above for Safety Injection initiating functions and requirements.
| |
| : 8. Loss of Power
| |
| : a. 4 kV Bus Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 19*
| |
| -Loss of Voltage
| |
| : b. 4 kV Bus Undervoltage 4/8us 2/ Bus 3/ Bus 1,2,3,4 19*
| |
| -Grid Degraded Voltage O O O
| |
| | |
| \
| |
| TABLE 3.3-3 (Continued) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION r-MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE h FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i 9. Control Room Isolation
| |
| : a. Manual Initiation 2 1 2 All 26
| |
| : b. Automatic Actuation
| |
| : Logic and Actuation Relays (SSPS) '2 1 2 1,2,3,4 26 jl
| |
| : c. Automatic Actuation 2 1 2 All 26 Logic and Actuation Relays (80P ESFAS)
| |
| : d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation initiating
| |
| { functions and requirements.
| |
| : 10. Solid-State Load Sequencer 2-1/ train 1/ train 2-1/ train 1, 2, 3, 4 25
| |
| : 11. Engineered Safety Features Actuation System Interlocks
| |
| : a. Pressurizer Pressure, 3 2 2 1,2,3 20 P-11
| |
| : b. Reactor Trip, P-4 4-2/ Train 2/ Train 2/ Train 1,2,3 22 l
| |
| | |
| TABLE 3.3-3 (Continued)
| |
| TABLE NOTATIONS
| |
| # Trip function may be blocked in this H0DE below the P-ll (Pressurizer Pressure Interlock) Setpoint.
| |
| ## Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.
| |
| *The provisions of Specification 3.0.4 are not applicable.
| |
| **0ne in Separation Group 1 and one in Separation Group 4.
| |
| ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least H01 STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.
| |
| ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour.
| |
| ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.
| |
| ACTION 17 - With less than the Minimum Channels OPERABLE requirement,
| |
| , operation may continue provided the containment purge supply l and exhaust valves are maintained closed.
| |
| ACTION 18 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| ACTION 19 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
| |
| : a. The inoperable channel is f aced in the tripped condition within 1 hour, and WOLF CREEK - UNIT 1 3/4 3-20
| |
| | |
| O TABLE 3.3-3 (Continued)
| |
| (d ACTION STATEMENTS (Continued)
| |
| : b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.2.1.
| |
| ACTION 20 - With less than the Minimum Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
| |
| ACTION 21 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.
| |
| ACTION 22 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE-status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTOOWN within the following O/
| |
| i" 6 hours.
| |
| ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoperable and take the action required by Specification 3.7.1.5.
| |
| ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, declare the affected auxiliary feedwater pump inoperable and take the ACTION required by Specification 3.7.1.2.
| |
| ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, declare the affected diesel generator and off-site power source inoperable and take the ACTION required by Specification 3.8.1.1.
| |
| ACTION 26 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or initiate and maintain operation of the Control Room Emergency Ventilation System.
| |
| ACTION 27 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.
| |
| WOLF CREEK - UNIT 1 3/4 3-21
| |
| | |
| TABLE 3.3-4 c ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8n O TOTAL SENSOR TRIP ALLOWABLE y FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE
| |
| : 1. Safety Injection (Reactor E Trip, Phase "A" Isolation, Z Feedwater Isolation, s Component Cooling Water, Turbine Trip, Auxiliary Feedwater-Motor-Driven Pump, Emergency Diesel Generator Operation, Containment Cooling, and Essential Service Water Operation)
| |
| : a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| {
| |
| T b. Automatic Actuation M Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| : c. Containment Pressure High-1 3.6 0.71 1.98 5 3.5 psig $ 4.5 psig
| |
| : d. Pressurizer Pressure -
| |
| Low 16.2 10.71 2.49 2 1830 psig 2 1815 psig
| |
| : e. Steam Line Pressure -
| |
| Low 19.6 14.81 1.93 2 615 psig 2 571 psig*
| |
| O O O
| |
| | |
| : O O TABLE 3.3-4 (Continued)
| |
| O g ENGINEERED SAFETY FEATURES ACTUATION' SYSTEM INSTRUMENTATION TRIP SETPOINTS 9
| |
| O TOTAL SENSOR TRIP ALLOWABLE y FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE
| |
| , 2. Containment Spray i z i
| |
| Z a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| ~
| |
| : b. Automatic Actuation Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| : c. Containment Pressure-High-3 4.3 0.71 1.98 5 27.0 psig 5 28.3 psig 1
| |
| : 3. Containment Isolation R
| |
| : a. Phase "A" Isolation Y
| |
| O 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| l 2) Automatic Actuation Logic and Actuation
| |
| ! Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| l 3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
| |
| : b. Phase "B" Isolation
| |
| : 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| : 2) Automatic Actuation Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| : 3) Containment Pressure-High-3 4.3 0.71 1.98 5 27.0 psig 5 28.3 psig
| |
| | |
| TABLE 3.3-4 (Continu-d) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 9
| |
| O TOTAL SENSOR TRIP ALLOWABLE h FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE
| |
| [ 3. Containment Isolation (Continued) z Z c. Containment H Purge Isolation
| |
| : 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| : 2) Automatic Actuation Logic and Actuation .
| |
| Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| : 3) Automatic Actuation -
| |
| U Logic and Actuation
| |
| [, Relays (BOP ESFAS) N.A. N.A. N.A. N.A. N.A.
| |
| $ 4) Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Values.
| |
| 4.SteamLineIsolation
| |
| : a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| : b. Automatic Actuation Logic and Actuation l Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| l
| |
| ! c. Containment Pressure-High-2 4.3 0.71 1.98 5 17.0 psig 5 18.3 psig l d. Steam Line Pressure-Low 19.6 14.81 1.93 2 615 psig 2 571 psig
| |
| : e. Steam Line Pressure-Negative Rate - High 3.0 0.5 0 5 100 psi 5 125 psi **
| |
| O -
| |
| O O
| |
| | |
| O O J i
| |
| ; TABLE 3.3-4 (Continued) l l g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r-
| |
| [ TOTAL SENSOR TRIP ALLOWABLE g FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE
| |
| ! E 5. Turbine Trip and i
| |
| Feedwater Isolation E
| |
| q a. Automatic Actuation g Logic and Actuation Re1ays (SSPS) N.A. N.A N.A. N.A. N.A.
| |
| : b. Steam Generator Water Level-High-High 5.0 2.18 2.51 < 78% of < 79.7% of i
| |
| Harrow range narrow range instrument instrument span span
| |
| : c. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
| |
| y 6. Auxiliary Feedwater
| |
| : a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
| |
| : b. Automatic Actuation Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N, A. N.A.
| |
| : c. Automatic Actuation Logic and Actuation Re1ays (80P.ESFAS) N.A. N.A. N.A. N.A. N.A.
| |
| : d. Steam Generator Water Level-Low-Low
| |
| : 1) Start Motor-Driven Pumps 23.5 21.18- 2.51 > 23.5% of > 22.3% of
| |
| ; Harrow range narrow range
| |
| ; instrument instrument span span
| |
| | |
| TABLE 3.3-4 (Continu:d)
| |
| ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 6
| |
| 6 c3 TOTAL SENSOR TRIP ALLOWABLE A FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S) SETPOINT VALUE __
| |
| E
| |
| , 6. Auxiliary Feedwater (Continued)
| |
| C 25
| |
| : 2) Start Turbine-Driven Pumps 23.5 21.18 2.51 1 23.5% of 2 22.3% of
| |
| "" narrow range narrow range instrument instrument span span
| |
| : e. Safety Injection -
| |
| Start Motor-Driven Pumps See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
| |
| : f. Loss-of-Offsite Power-Start Turbine-R3, Driven Pump N.A. N.A. N.A. N.A. N.A.
| |
| o, g. Trip of All Main Feed-4 water Pumps - Start
| |
| * Motor-Driven Pumps N.A. N.A. N.A. N.A. N.A.
| |
| : h. Auxiliary Feedwater Pump Suction Pressure-Low (Transfer to ESW) N.A. N.A. N.A. 1 21.60 psia 2 20.53 psia
| |
| : 7. Automatic Switchover to Containment Sump
| |
| : a. Automatic Actuation Logic and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A.
| |
| : b. RWST Level-Low-Low 3.4 1.21 1.86 > 36% of > 35.1% of Instrument Instrument span span Coincident with Safety Injection See Item 1. above for Safety Injection Trip Setpoints and Allowable Values.
| |
| O O - -
| |
| O-
| |
| | |
| J v G TABLE 3.3-4 (Continued) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 9
| |
| 2 TOTAL SENSOR TRIP ALLOWABLE FUNCTIONAL UNIT h ALLOWANCE (TA) Z_ ERROR (S) SETPOINT VALUE
| |
| [ 8. Loss of Power z
| |
| Z a. 4 kV Undervoltage
| |
| - -Loss of Voltage N.A. N.A. N.A. 2 83V (120V 1 74.7V (120V Bus)
| |
| Bus) w/1s w/1 + 0.2, -0.5s delay delay
| |
| : b. 4 kV Undervoltage
| |
| -Grid Degraded Voltage N.A. N.A. N.A. 2 106.9V 2 104.3V (120V Bus)
| |
| (120V Bus) w/119 1 11.6s delay w/119s delay l ,
| |
| : 9. Control Room Isolation
| |
| : a. Manual Initiation N.A. N.A. N.A. N.A. N.A O b. Automatic Actuation Logic and Actuation Re1ays (SSPS) N.A. N.A. N.A. N.A. N.A
| |
| : c. Automatic Actuation Logic and Actuation Relays (80P ESFAS) N.A. N.A. N.A. N.A. N.A.
| |
| : d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Values.
| |
| : 10. Solid-State Load Sequencer N.A. N.A. N.A. N.A. N.A.
| |
| : 11. Engineered Safety Features Actuation System Interlocks
| |
| : a. Pressurizer Pressure, P-11 N.A. N.A. N.A. 5 1970 psig < 1979 psig
| |
| : b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.
| |
| I
| |
| | |
| a TABLE 3.3-4 (Continued)
| |
| TABLE NOTATIONS
| |
| * Time constants utilized in the lead-lag controller for Steam Pressure-Low are ti > 50 seconds and T2 < 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.
| |
| **The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 3-28
| |
| | |
| TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES 4
| |
| INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
| |
| : 1. Manual Initiation
| |
| : a. Safety Injection (ECCS) N.A.
| |
| : b. Containment Spray N.A.
| |
| : c. Phase "A" Isolation N.A.
| |
| Phase "B" Isolation N.A.
| |
| d.
| |
| : e. Containment Purge Isolation N.A.
| |
| : f. Steam Line Isolation N.A.
| |
| : g. Feedwater Isolation N.A.
| |
| : h. Auxiliary Feedwater N.A.
| |
| : i. Essential Service Water N.A.
| |
| j- Containment Cooling N.A.
| |
| : k. Control Room Isolation N.A.
| |
| : 1. Reactor Trip N.A.
| |
| : m. Emergency Diesel Generators N.A.
| |
| : n. Component Cooling Water N.A.
| |
| : o. Turbi..e Trip N.A.
| |
| : 2. Containment Pressure-High-1
| |
| : a. Safety Injection (ECCS) 5 29(1)/12(4)
| |
| : 1) Reactor Trip 52
| |
| : 2) Feedwater Isolation <7
| |
| ! 3) Phase "A" Isolation 1.5(5)
| |
| : 4) Auxiliary Feedwater 5 60
| |
| ; 5) Essential Service Water 5 60(1)
| |
| : 6) Containment Cooling 5 60(1)
| |
| : 7) Component Cooling Water N.A.
| |
| : 8) Emergency Diesel Generators 5 14(6)
| |
| : 9) Turbine Trip N.A.
| |
| O WOLF CREEK - UNIT 1 3/4 3-29 3
| |
| | |
| TABLE 3.3-5 (Continued)
| |
| ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
| |
| : 3. Pressurizer Pressure-Low
| |
| : a. Safety Injection (ECCS) 5 29I1)/12(4)
| |
| : 1) Reactor Trip 52
| |
| : 2) Feedwater Isolation <7
| |
| : 3) Phase "A" Isolation 2(5)
| |
| : 4) Auxiliary Feedwater 5 60
| |
| : 5) Essential Service Water 5 60(1)
| |
| : 6) Containment Cooling 5 60 II)
| |
| : 7) Component Cooling Water N.A.
| |
| : 8) Emergency Diesel Generators 5 14(6)
| |
| : 9) Turbine Trip N.A.
| |
| : 4. Steam Line Pressure-Low
| |
| : a. Safety Injection (ECCS) 5 24(3)/12(4)
| |
| : 1) Reactor Trip 52
| |
| : 2) Feedwater Isolation <7
| |
| : 3) Phase "A" Isolation 2(5)
| |
| : 4) Auxiliary Feedwater 5 60
| |
| : 5) Essential Service Water < 60 II)
| |
| : 6) Containment Cooling 60(1)
| |
| : 7) Component Cooling Water N.A.
| |
| ; 8) Emergency Diesel Generators 5 14(6)
| |
| : 9) Turbine Trip N.A.
| |
| : b. Steam Line Isolation 5 2(5)
| |
| O WOLF CREEK - UNIT 1 3/4 3-30
| |
| | |
| i TABLE 3.3-5 (Continued)
| |
| ENGINEERED SAFETY FEATURES RESPONSE TIMES 1
| |
| I t
| |
| INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS l
| |
| : 5. Containment Pressure-High-3
| |
| : a. Containment Spray 5 32(1)/20(2)
| |
| : b. Phase "B" Isolation 5 31.5
| |
| : 6. Containment Pressure-High-2 l Steam Line Isolation 5 2(5)
| |
| : 7. Steam Line Pressure-Negative Rate-High Steam Line Isolation 5 2(5)
| |
| : 8. Steam Generator Water Level-High-High n Turbine Trip
| |
| : a. 5 2.5
| |
| : b. Feedwater Isolation 57
| |
| : 9. Steam Generator Water Level - Low-Low
| |
| : a. Start Motor-Driven Auxiliary Feedwater Pumps 5 60 l b. Start Turbine-Driven Auxiliary Feedwater Pumps 5 60 l
| |
| : 10. Loss-of-Offsite Power Start Turbine-Driven Auxiliary Feedwater N.A.
| |
| Pumps
| |
| : 11. Trip of All Main Feedwater Pumps Start Motor-Driven N.A Auxiliary Feedwater Pumps lO WOLF CREEK - UNIT 1 3/4 3-31
| |
| | |
| TABLE 3.3-5 (Continued)
| |
| ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
| |
| : 12. Auxiliary Feedwater Pump Suction Pressure-Low Transfer to Essential Service Water N.A.
| |
| : 13. RWST Level-Low-Low Coincident with Safety Injection Automatic Switchover to Containment < 60
| |
| ~
| |
| Sump
| |
| : 14. Loss of Power
| |
| : a. 4 kV Bus Undervoltage- 5 14 Loss of Voltage
| |
| : b. 4 kV Bus Undervoltage- -< 144 Grid Degraded Voltage
| |
| : 15. Phase "A" Isolation
| |
| : a. Control Room Isolation N.A.
| |
| : b. Containment Purge Isolation 5 2(5) l l
| |
| i I
| |
| O WOLF CREEK - UNIT 1 3/4 3-32
| |
| | |
| TABLE 3.3-5 (Continued)
| |
| TABLE NOTATIONS (1) Diesel generator starting and sequence loading delays included.
| |
| (2) Diesel generator starting delay not included. Offsite power available.
| |
| (3) Diesel generator starting and sequence loading delay included. RHR pumps not included.
| |
| (4) Diesel generator starting and sequence loading delays not included.
| |
| Offsite power available. RHR pumps not included.
| |
| (5) Does not include valve closure time.
| |
| (6) Includes time for diesel to reach full speed.
| |
| i r
| |
| i l
| |
| l f
| |
| l l
| |
| t k,
| |
| l WOLF CREEK - UNIT 1 3/4 3-33 l
| |
| | |
| _ TABLE 4.3-2
| |
| : ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION S SURVEILLANCE REQUIREMENTS 2 TRIP lE ANALOG ACTUATING MODES
| |
| ^ CHANNEL DEVICE MASTER SLAVE FOR WHICH
| |
| ' RELAY SURVEILLANCE CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY E FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED Z
| |
| : 1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Turbine Trip, Component Cooling Water, Auxiliary Feedwater-Motor-Driven Pump, Emergency Diesel Generator Operation, Containment Cooling, and R Essential Service Water Operation)
| |
| ] N.A. N.A. N.A. 1,2,3,4
| |
| : a. Manual Initiation N.A. N.A. N.A. R f N.A. 1, 2, 3, 4
| |
| : b. Automatic Actuation N.A. N.A. N.A. M(1) M(1) Q(3)
| |
| Logic and Actuation Relays (SSPS)
| |
| : c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 High-1
| |
| : d. Pressurizer Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 Low
| |
| : e. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 Low
| |
| : 2. Containment Spray
| |
| : a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
| |
| : b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2, 3, 4 Logic and Actuation Relays (SSPS)
| |
| : c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 High-3 O O O
| |
| | |
| m n
| |
| O TABLE 4.3-2 (Continued)
| |
| U c ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
| |
| { SURVEILLANCE REQUIREMENTS 2 TRIP R
| |
| * ANALOG ACTUATING MODES
| |
| ' CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE l E FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
| |
| ] 3. Containment Isolation
| |
| : a. Phase "A" Isolation j 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4 l 2) Automatic Actuation N.A N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2, 3, 4
| |
| { Logic and Actuation Relays (SSPS)
| |
| : 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
| |
| R s
| |
| : b. Phase "B" Isolation-y 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
| |
| !N 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays (SSPS)
| |
| : 3) Containment S R M N.A. N.A. N.A. N.A. 1,2,3 i Pressure-High-3
| |
| : c. Containment Purge Isolation
| |
| : 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
| |
| : 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2, 3, 4 Logic and Actuation Relays (SSPS)
| |
| : 3) Automatic Actuation N.A. N.A. N.A. N.A. M(1)(2) N.A. N.A. 1,2,3,4 Logic and Actuation Relays (B0P ESFAS)
| |
| : 4) Phase "A" See Item 3.a. above for all Phase "A" Isolation Surveillance Requirements.
| |
| Isolation
| |
| | |
| TABLE 4.3-2 (Continu d) c ENGINEERED SAFETY FEATURES ACTUATION SYSTFM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g TRIP p ANALOG ACTUATING MODES X CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
| |
| [ 4. Steam Line Isolation
| |
| : a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3
| |
| : b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3 Logic and Actuation Relays (SSPS)
| |
| : c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 High-2 y d. Steam Line Pr. assure- S R M N.A. N.A. N.A. N.A. 1,2,3 A Low
| |
| $ e. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 3 m Negative Rate-High
| |
| : 5. Turbine Trip and Feedwater Isolation
| |
| : a. Automatic Actuation N.A. N.A. N.A. N. A. M(1) M(1) Q(3) 1, 2 Logic and Actuation Relay (SSPS)
| |
| : b. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1, 2 Level-High-High
| |
| : c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
| |
| : 6. Auxiliary Feedwater
| |
| : a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3
| |
| : b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3 Logic and Actuation Relays (SSPS)
| |
| O O O
| |
| | |
| rh n TABLE 4.3-2 (Continued)
| |
| I i g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION r- SURVEILLANCE REQUIREMENTS n
| |
| 9 TRIP MODES A ANALOG ACTUATING
| |
| * CHANNEL DEVICE MASTER SLAVE FOR WHICH
| |
| ' OPERATIONAL OPERATIONAL ACTUATION RELAY. RELAY SURVEILLANCE CHANNEL CHANNEL E FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED Z
| |
| ~ 6. Auxiliary Feedwater (Continued)
| |
| : c. Automatic Actuation N.A. N.A. N.A. N.A. M(1)(2) N.A. N.A. 1,2,3 l Logic and Actuation Relays (BOP ESFAS)
| |
| : d. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1,2,3 Level-Low-Low
| |
| : e. Safety Injection See Item 1 above for all Safety Injection Surveillance Requirements
| |
| : f. Loss-Offsite Power N.A. R N.A. M N.A. N.A. N.A. 1,2,3 w g. Trip of All Main N.A. N.A. N.A R N.A. N.A. N.A. 1, 2 O Feedwater Pumps u
| |
| : h. Auxiliary Feedwater S R M N.A. N.A. N.A. N.A. 1,2,3 Pump Suction Pressure-Low
| |
| : 7. Automatic'Switchover to Containment Sump I a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2, 3, 4 Logic and Actuation
| |
| : Relays (SSPS)
| |
| : b. RWST Level - Low-Low S R M- N.A. N.A. N.A. N.A. 1,2,3,4 Coincident With j Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
| |
| ! 8. Loss of Power 4
| |
| : a. 4 kV Undervoltage - N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4 Loss of Voltage l
| |
| : b. 4 kV Undervoltage - N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4 Grid Degraded Voltage
| |
| | |
| TABLE 4.3-2 (Centinurd)
| |
| :c ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
| |
| { SURVEILLANCE REQUIREMENTS 9 TRIP E
| |
| * ANALOG ACTUATING MODES
| |
| ' CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE E FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED Z
| |
| ~ 9. Control Room Isolation
| |
| : a. Manual Initiation N.A. N.A. N.A. R H.A. N.A. N.A. All
| |
| : b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2, 3, 4 Logic and Actuation Relays (SSPS)
| |
| : c. Automatic Actuation N.A. N.A. N.A. N.A. M(1)(2) N.A. N.A. All Logic and Actuation Relays (BOP ESFAS)
| |
| : d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Surveillance Requirements.
| |
| {
| |
| Y 10. Solid-State Load Sequencer N.A. N.A. N.A. N.A. M(1)(2) N.A. N.A. 1,2,3,4
| |
| : 11. Engineered Safety Features Actuation System Interlocks
| |
| : a. Pressurizer Pressure, N.A. R M N.A. N.A. N.A. N.A. 1,2,3 P-11
| |
| : b. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 TABLE NOTATIONS (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
| |
| (2) Continuity check may be excluded from the ACTUATION LOGIC TEST.
| |
| (3) Except Relays K602, K620, K622, K624, K630, K740, and K741, which shall be tested at least once per 18 months during refueling and during each COLD SHUTOOWN exceeding 24 hours unless they have been tested within the previous 90 days.
| |
| O O O
| |
| | |
| INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits.
| |
| APPLICABILITY: As shown in Table 3.3-6.
| |
| ACTION:
| |
| : a. With a radiation monitoring channel Alarm / Trip Setpoint for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours or declare the channel inoperable.
| |
| : b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.
| |
| : c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at the fre-quencies shown in Table 4.3-3.
| |
| WOLF CREEK - UNIT 1 3/4 3-39
| |
| | |
| TABLE 3.3-6 RADIATION __ MONITORING INSTRUMENTATION FOR PLANT OPERATIONS G; MINIMUM n CHANNELS CHANNELS APPLICABLE ALARM / TRIP A FUMCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION 9
| |
| i 1. Containment C"
| |
| 25 a. Containment Atmosphere-
| |
| -d Gaseous Radioactivity-High (GT-RE-31 & 32) 1 2 All ### 26
| |
| : b. Gaseous Radioactivity-RCS Leakage Detection (GT-RE-31 & 32) N.A. 1 1,2,3,4 N.A. 29
| |
| : c. Particulate Radioactivity-w RCS Leakage Detection 32 (GT-RE-31 & 32) N.A. 1 1,2,3,4 N.A. 29 w
| |
| A 2. Fuel Building o
| |
| : a. Fuel Building Exhaust-l Gaseous Radioactivity-High (GG-RE-27 & 28) 2 ** ##
| |
| i 1 30 l b. Criticality-High I Radiation Level
| |
| : 1) Spent Fuel Pool 1 1
| |
| * 5 15 mR/h 28 (SD-RE-37 or 38)
| |
| : 2) New Fuel Pooi 1 1 1 15 mR/h 28 (SD-RE-35 or 36)
| |
| : 3. Control Room Air Intake-Gaseous Radioactivity-High (GK-RE-04 & 05) 1 2 All # 27 1
| |
| O O O
| |
| | |
| p TABLE 3.3-6 (Continued)
| |
| TABLE NOTATIONS
| |
| *With fuel in the respective fuel storage pool.
| |
| **With irradiated fuel in the fuel storage areas or fuel building.
| |
| # Trip Setpoint concentration value (pCi/cm3 ) is to be established such that the actual submersion dose rate would not exceed 2 mR/h in the control room.
| |
| ## Trip Setpoint concentration value (pCi/cma ) is to be established such that the actual submersion dose rate would not exceed 4 mR/h in the fuel building.
| |
| ### Trip Setpoint concentration value (pCi/cm3 ) is to be established such that the actual submersion dose rate would not exceed 9 mR/h in the containment building. The Setpoint value may be increased up to the equivalent limits of Specification 3.11.2.1 in accordance with the methodology and parameters in the ODCM during containment purge or vent provided the Setpoint value does not exceed twice the maximum concentration activity in the containment determined by the sample analysis performed prior to each release in accordance with Table 4.11-2.
| |
| ACTION STATEMENTS ACTION 26 - With less than the Minimum Channels OPERABLE requirement, opera-h tion may continue provided the containment purge valves are main-tained closed.
| |
| ACTION 27 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour isolate the Control Room Emergency Ventilation System and initiate operation of the Control Room Emergency Ventilation System in the recirculation mode.
| |
| ACTION 28 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint is provided in the fuel area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel building.
| |
| ACTION 29 - Must satisfy the ACTION requirements for Specification 3.4.6.1.
| |
| ACTION 30 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour isolate the Fuel Building Ventilation System and initiate operation of the Emergency Exhaust System to maintain the fuel building at a negative pressure.
| |
| l
| |
| \ O
| |
| ; d WOLF CREEK - UNIT 1 3/4 3-41
| |
| | |
| TABLE 4.3-3 m: RADIATION MONITORING INSTRUMENTATION FOR PLANT k OPERATIONS SURVEILLANCE REQUIREMENTS 2 '
| |
| 5
| |
| * ANALOG MODES FOR
| |
| ' CHANNEL WHICH SURVIL-CHANNEL CHANNEL OPERATIONAL LANCE IS E FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED w
| |
| H 1. Containment
| |
| : a. Containment Atmosphere-Gaseous Radioactivity-High (GT-RE-31 & 32) S R M All
| |
| : b. Gaseous Radioactivity-RCS Leakage Detection (GT-RE-31 & 32) S R M 1,2,3,4 w c. Particulate A Radioactivity -
| |
| w RCS Leakage Detection g (GT-RE-31 & 32) S R M 1,2,3,4 -
| |
| : 2. Fuel Building <
| |
| : a. Fuel Building Exhaust-Gaseous Radioactivity-High (GG-RE-27 & 28) S R M **
| |
| : b. Criticality-High.
| |
| Radiation Level -
| |
| : 1) Spent Fuel Pool S R M *
| |
| (SD-RE-37 & 38)
| |
| : 2) New Fuel Pool (SD-RE-35 & 36) S R M
| |
| : 3. Control Room Air Intake-Gaseous Radioactivity-High (GK-RE-04 & 05) S R M All
| |
| *With fuel in the respective fuel storage pool.
| |
| **With irradiated fuel in the fuel storage areas or fuel building.
| |
| O O O
| |
| | |
| INSTRUMENTATION d MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 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 equipm;nt to map these thimbles.
| |
| APPLICABILITY: When the Movable Incore Detection System is used for:
| |
| : a. Recalibration of the Excore Neutron Flux Detection System,
| |
| : b. Monitoring the QUADRANT POWER TILT RATIO, or
| |
| : c. Measurement of F H, Fq(Z) and F xy O ACTION:
| |
| : a. With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours by normalizing each detector output when required for:
| |
| : a. Recalibration of the Excore Neutron Flux Detection System, or
| |
| : b. Monitoring the QUADRANT POWER TILT RATIO, or
| |
| : c. Measurement of F H' I Q(Z), and Fxy.
| |
| WOLF CREEK - UNIT 1 3/4 3-43
| |
| | |
| INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.
| |
| APPLICABILITY: At all. times.
| |
| ACTION:
| |
| : a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.3.1 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 4.3-4.
| |
| 4.3.3.3.2 Each of the above required seismic monitoring instruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 10 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, fre-quency spectrum, and resultant effect upon facility features important to .
| |
| safety. I O
| |
| WOLF CREEK - UNIT 1 3/4 3-44 l
| |
| | |
| TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
| |
| : 1. Triaxial Peak Recording Accelerographs
| |
| : a. Radwaste Base Slab i 1.0g 1
| |
| : b. Control Room i 1.0g I
| |
| : c. ESW Pump Facility i 1.0g 1
| |
| : d. Ctmt Structure i 2.0g 1
| |
| : e. Auxiliary Bldg. SI Pump Suctions i 1.0g 1
| |
| : f. SGB Piping i 5.0g 1
| |
| : g. SGC Support i 1. 0g 1
| |
| : 2. Triaxial Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers (Active)
| |
| : a. Ctmt. Base Slab i 1.0g 1
| |
| : b. Ctmt. Oper. Floor 1.0g 1
| |
| : c. Reactor Support i 1.0g 1
| |
| : d. Aux. Bldg. Base Slab i 1.0g 1
| |
| : e. Aux. Bldg. Control Room Air Filter i 1.0g 1 k f. Free Field 1 0.5g 1
| |
| : 3. Triaxial Response-Spectrum Recorder (Passive)
| |
| Ctmt. Base Slab i 1. 0g 1
| |
| : 4. Triaxial Seismic Switches ACCELERATION LEVEL North East Vertical
| |
| : a. OBE Ctat. Base Slab 0.06g 0.06g 0.06g 1
| |
| : b. SSE Ctmt. Base Slab 0.15g 0.15g 0.16g 1
| |
| : c. OBE Ctmt. Oper. F1. 0.07g 0.07g 0.07g 1
| |
| : d. SSE Ctmt. Oper. F1. 0.16g 0.17g 0.16g 1
| |
| : e. System Trigger 0.01g 0.01g 0.01g 1 l
| |
| : i. .
| |
| WOLF CREEK - UNIT 1 3/4 3-45
| |
| | |
| TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
| |
| : 1. Triaxial Peak Recording Accelerographs
| |
| : a. Radwaste Base Slab N.A. R N.A.
| |
| : b. Control Room N.A. R N.A.
| |
| : c. ESW Pump Facility N.A. R N. A.
| |
| : d. Ctmt Structure N.A. R N.A.
| |
| : e. Auxiliary Bldg. SI Pump Suction N.A. R N.A.
| |
| : f. SGB Piping N.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 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 N.A. R N.A.*
| |
| : 4. Triaxial Seismic Switches
| |
| : a. OBE Ctmt. Base Slab M R SA
| |
| : b. SSE Ctmt. Base Slab M R SA
| |
| : c. OBE Ctmt. Oper. F1. M R SA
| |
| : d. SSE Ctmt. Oper. F1. 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 I of an ANALOG CHANNEL OPERATIONAL TEST. 1 01 l WOLF CREEK - UNIT 1 3/4 3-46 '
| |
| l
| |
| | |
| . . _ . - __ . . .. - =- .-
| |
| INSTRUMENTATION
| |
| 'O('''N METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteo.ological monitoring instrumentation channels in Table 3.3-8 shall be OPERA 61.E, APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With one or more required meteorological monitoring channels inoperable
| |
| 'for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| I
| |
| (
| |
| \' SURVEILLANCE REQUIREMENTS 4.3.3.4 -Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and
| |
| - CHANNEL CALIBRATION at the frequencies shown in Table 4.3-5.
| |
| 9 i
| |
| : i. ,
| |
| WOLF CREEK - UNIT 1 3/4 3-47
| |
| | |
| TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUM M LOCATION OPERABLE
| |
| : 1. Wind Speed Nominal Elev. 10m 1 Nominal Elev. 60m 1
| |
| : 2. Wind Direction Nominal Elev. 10m 1 Nominal Elev. 60m 1
| |
| : 3. Air Temperature - AT Nominal Elev. 10m-60m 1 l
| |
| l WOLF CREEK - UNIT 1 3/4 3-48 1
| |
| | |
| k TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ~
| |
| CHANNEL CHANNEL 4
| |
| INSTRUMENT CHECK CALIBRATION
| |
| : 1. Wind Speed l .a. Nominal Elev. 10m D SA
| |
| : b. Nominal Elev. 60m D SA i -
| |
| : 2. Wind Direction'
| |
| : a. Nominal Elev. 10m D SA i
| |
| j b. Nominal Elev.' 60m D SA I
| |
| : 3. Air Temperature - AT l a. Nominal Elev. 10-60m D SA l.
| |
| i
| |
| (
| |
| s
| |
| ' WOLF CREEK - UNIT 1 3/4 3-49 :
| |
| *F-N'-rc''M-*pr'N1"P-2
| |
| | |
| INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels given in Table 3.3-9 and the auxiliary shutdown panel (ASP) controls shall be OPERABLE with readouts displayed external to the control room.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| : a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : b. With the ASP controls inoperable, restore the inoperable ASP controls to OPERABLE status within 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : c. The provisions of Specification 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies given in Table 4.3-6.
| |
| 4.3.3.5.2 The ASP controls shall be demonstrated OPERABLE at least once per 18 months by operating each actuated component from the ASP.
| |
| 4.3.3.5.3 The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 for the turbine-driven auxiliary feedwater pump or tl.e atmospheric dump valves.
| |
| O WOLF CREEK - UNIT 1 3/4 3-50
| |
| | |
| /''')
| |
| ~'
| |
| TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION TOTAL NO. MINIMUM READ 0UT OF CHANNELS INSTRUMENT LOCATION CHANNELS OPERABLE
| |
| : 1. RCS Pressure-Wide Range ASP
| |
| * 2 1
| |
| : 2. Reactor Coolant Temperature-Cold Leg ASP
| |
| * 4 1
| |
| : 3. Source Range, Neutron Flux # ASP
| |
| * 2 1
| |
| : 4. Reactor Trip Breaker Indication RTS** 1/ trip breaker 1/ trip breaker
| |
| : 5. Reactor Coolant Temperature - ASP
| |
| * 2 1 Hot Leg
| |
| : 6. Reactor Coolant Pump Breakers *** 1/ pump 1/ pump
| |
| : 7. Pressurizer Pressure ASP
| |
| * 1 1
| |
| : 8. Pressurizer Level ASP
| |
| * 2 1
| |
| : 9. Steam Generator Pressure ASP
| |
| * 2/stm. gen. 1/stm. gen.
| |
| s
| |
| : 10. Steam Generator Level ASP
| |
| * 2/stm. gen. 1/stm. gen.
| |
| : 11. Auxiliary Feedwater Flow Rate ASP
| |
| * 4 1
| |
| : 12. Auxiliary Feedwater Suction ASP
| |
| * 3 1 Pressure l
| |
| * Auxiliary Shutdown Panel.
| |
| ** Reactor Trip Switchgear.
| |
| ***13.8 kV Switchgear.
| |
| Not' required OPERABLE in MODE 1 or in MODE 2 above P-6 Setpoint.
| |
| (
| |
| WOLF CREEK - UNIT 1 3/4 3-51
| |
| | |
| TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
| |
| : 1. RCS Pressure - Wide Range M R
| |
| : 2. Reactor Coolant Temperature - Cold Leg M R
| |
| : 3. Source Range, Neutron Flux # M R
| |
| : 4. Reactor Trip Breaker Indication M N.A.
| |
| : 5. Reactor Coolant Temperature - Hot Leg M R
| |
| : 6. Reactor Coolant Pump Breakers N.A. N.A.
| |
| : 7. Pressurizer Pressure M R
| |
| : 8. Pressurizer Level M R
| |
| : 9. Steam Generator Pressure M R
| |
| : 10. Steam Generator Level M R
| |
| : 11. Auxiliary Feedwater Flow Rate M R
| |
| : 12. Auxiliary Feedwater Pump Suction Pressure M R Not required OPERABLE in MODE 1 or in MODE 2 above P-6 Setpoint.
| |
| l l
| |
| l l
| |
| l l
| |
| WOLF CREEK - UNIT 1 3/4 3-52
| |
| | |
| A INSTRUMENTATION i
| |
| ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| : a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : b. With the numb 2r of OPERABLE accident monitoring instrumentation channels, except the containment radiation level monitor and the unit vent - high range noble gas monitor, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoper-able channel (s) to OPERABLE status within 48 hours; otherwise, be in g
| |
| v) at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : c. With the number of OPERABLE channels for the containment radiation level monitor or the unit vent - high range noble gas monitor 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 hours and either restore the inoperable channel to OPERABLE status within 7 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days that provides actions taken, cause of the inoperability and plans and schedule for restoring the channels to OPERABLE status,
| |
| : d. 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 the frequencies shown in Table 4.3-7.
| |
| G i
| |
| WOLF CREEK - UNIT 1 3/4 3-53
| |
| | |
| TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION 6
| |
| 9 TOTAL MINIMUM n NO. OF CHANNELS y INSTRUMENT CHANNELS OPERABLE 7 1. Containment Pressure g a) Normal Range 2 1 Z b) Extended Range 2 1
| |
| : 2. Reactor Coolant Outlet Temperature - THOT (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
| |
| : 7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator
| |
| {
| |
| y 8. Steam Generator Water :.evel - Wide Range 1/ steam generator 1/ steam generator T 9. Refueling Water Storage Tank Water Level 2 1
| |
| : 10. Containment Hydrogen Concentration Level 2 1
| |
| : 11. Auxiliary Feedwater Flow Rate 1/ steam generator 1/ steam generator
| |
| : 12. PORV Position Indicator
| |
| * 1/ Valve 1/ Valve
| |
| : 13. PORV Block Valve Position Indicator ** 1/ Valve 1/ Valve
| |
| : 14. Safety Valve Position Indicator 1/ Valve 1/ Valve
| |
| : 15. Containment Water Level 2 1
| |
| : 16. Containment Radiation Level (High Range) N.A. I
| |
| : 17. Thermocouple / Core Cooling Detection System 4/ core quadrant 2/ core quadrant
| |
| : 18. Unit Vent - High Range Noble Gas Monitor N.A. 1
| |
| *Not applicable if the associated block valve is in the closed position.
| |
| **Not applicable if the block valve is verified in the closed position and power is removed.
| |
| O O O
| |
| | |
| TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
| |
| Q; CHANNEL CHANNEL n INSTRUMENT CHECK CALIBRATION x
| |
| A 1. Containment Pressure M R
| |
| $ 2. Reactor Coolant Outlet Temperature - TH0T (Wide Range) M R
| |
| : 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R H 4. Reactor Coolant Pressure - Wide Range M R
| |
| : 5. Pressurizer Water Level M R
| |
| : 6. Steam Line Pressure M R
| |
| : 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
| |
| $ 10. Containment Hydrogen Concentration Level M R y 11. Auxiliary Feedwater Flow Rate M R E 17. PORV Position Indicator
| |
| * M N.A.
| |
| : 13. PORV Block Valve Position Indicator ** M N.A.
| |
| : 14. Safety Val e Position Indicator M N.A.
| |
| : 15. Containment Water Level M R
| |
| : 16. Containment Radiation Level (High Range) M R***
| |
| : 17. Thermocouple / Core Cooling Detection System M R
| |
| : 18. Unit Vent - High Range Noble Gas Monitor M R
| |
| *Not applicable if the associated block valve is in the closed position.
| |
| **Not applicable if the block valve is verified in the closed position and power is removed.
| |
| *** CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.
| |
| | |
| INSTRUMENTATION CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.7 Two independent Chlorine Detection Systems, with their Alarm / Trip Setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE.
| |
| APPLICABILITY: All MODES.
| |
| ACTION:
| |
| : a. With one Chlorine Detection System inoperable, restore the inoperable system to OPERABLE status within 7 days or within the next 6 hours initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation.
| |
| : b. With both Chlorine Detection Systems inoperable, within 1 hour initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation.
| |
| : c. The provisions of Specification 3.0.4 are not applicable.
| |
| O SURVEILLANCE REQUIREMENTS 4.3.3.7 Each Chlorine Detection System shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
| |
| O WOLF CREEK - UNIT 1 3/4 3-56
| |
| | |
| C'; INSTRUMENTATION t /
| |
| '~#
| |
| FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERABLE.
| |
| APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE.
| |
| ACTION:
| |
| : a. With any, but not more than one-half the total in any fire zone, Function A fire detection instruments shown in Table 3.3-11 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.
| |
| : b. With more than one-half of the Function A fire detection instruments in any fire zone shown in Table 3.3-11 inoperable or with any
| |
| /'~'N Function B fire detection instruments shown in Table 3.3-11,
| |
| (' ') inoperable, or with any two or more adjacent fire detection instruments shown in Table 3.3-11 inoperable, within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5. -
| |
| : c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months.
| |
| 4.3.3.8.2 The NFPA Standard 720 supervised circuits supervision ariaciated with the detector alarms of each of the above required fire detect an instruments shall be demonstrated OPERABLE at least once per 6 months.
| |
| O)
| |
| \
| |
| v WOLF CREEK - UNIT 1 3/4 3-57
| |
| | |
| TABLE 3.3-11 FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS
| |
| * INSTRUMENT LOCATION ZONE HEAT FLAME SM0KE (x/y) (x/y) (x/y) 1101-Aux. Bldg. 1974' Gen. Fir. 11 100 0/11 1102-Chiller & Surge Tks. Area 100 0/4 1102-Chiller & Surge Tks. Area 101 2/0 1107-Cent. Charg. Pmp. Rm. B 101 2/0 1108-Safety Inj. Pmp. Rm. B 101 2/0 1109-Res. Ht. Remov. Pmp. Rm. B 101 1/0 1110-Ctmt. Spray Pmp. Rm. B 101 1/0 1111-Res. Ht. Remov. Pmp. Rm. A 101 1/0 1112-Ctmt. Spray Pmp. Rm. A 101 1/0 1113-Safety Inj. Pmp. Rm. A 101 2/0 1114-Cent. Charg. Pmp. Rm. A 101 2/0 1115-Pos. Disp. Charg. Pmp. Rm. 101 2/0 1116, 1117-Boric Acid Tk. Rms. 101 2/0 1116, 1117-Boric Acid Tk. Rms. 101 2/0 1120-Aux. Bldg.1974' Gen. F1r. #2 101 4/0 1122-Aux. Bldg.1974' Gen. Fir. #3 100 0/3 1122-Aux. Bldg. 1974' Gen. Fir. #3 101 5/0 1126-Boron Inj. Tk. & Pmp. Rm. 101 1/0 1127-Stair A-Z 109 1/0 1128-Aux. Feedwater Pump Rm. Basement 117 2/0 1130-Aux. Bldg. 1974' N. Corr. 100 0/2 1206-W. Pipe Chase Below AFWP Area 117 2/0 1203-Aux. Bldg. Elec. Chase S. 1988' 117 1/0 1301-Aux. Bldg. 2000' Corridor #1 103 0/10 1301-Aux. Bldg. 2000' Corridor #1 117 2/0 1311-Aux. Bldg. Sampling Rm. 117 2/0 1312-Boron Meter /RC Activity Mon. Rm. 103 0/1 1314-Aux. Bldg. 2000' Corridor #3 103 0/3 1314-Aux. Bldg. 2000' Corridor #3 117 2/0 1315-Ctmt. Spray Add. Tk. Area 103 0/2 1316-V1v. Rm. by Seal Wtr. Ht. Exch. 103 0/1 1320-Aux. Bldg. 2000' Corridor #4 103 0/3 1321-Aux. Bldg. 2000' S. Exit Vest. 103 0/1 1322-Pipe Pene. Rm. B 117 5/0 1323-Pipe Pene. Rm. A 117 6/0 1325-Aux. FW Pmp. Rm. B 117 2/0 1326-Aux. FW Pmp. Rm. A 117 2/0 1331-Aux. FW Pmp. Rm. C 111 2/0 1 1331-Aux. FW Pmp. Rm. C 117 1/0 l 1335-Aux. Bldg. Elec. Chase N. 2000' 117 1/0 1336-Aux. Bldg. Elec. Chase S. 2000' 117 1/0 1401-Comp. Cool. Pmp. & Ht. Exch. B 118 5/0 1402-Aux. Bldg. 2026' Corridor #1 104 1403-MG Set Rm. 105 0/6(1) ,
| |
| 1403-MG Set Rm. 112 0/9(1) 0/9 WOLF CREEK - UNIT 1 3/4 3-58
| |
| | |
| TABLE 3.3-11 (Continued)
| |
| (O)
| |
| U FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS
| |
| * INSTRUMENT LOCATION ZONE HEAT FLAME SM0KE (x/y) (x/y) (x/y) 1405-Chemical Stg. Area 118 6/0 1406-Comp. Cool. Pmp. & Ht. Exch. A 104 0/1 1406-Comp. Cool. Pmp. & Ht. Exch. A 118 2/0 1408-Aux. Bldg. 2026' Corridor #2 104 0/9 1408-Aux. Bldg. 2026' Corridor #2 118 5/0 1409-Elec. Pene. Rm. B 106 0/4 f1) 1409-Elec. Pene. Rm. B 113 0/4 f1) 1410-Elec. Pene. Rm. A 107 0/8 f1) 1410-Elec. Pene. Rm. A 114 0/8 f1) 1413-Aux. Shutdown Pnl. Rm. 118 4/0 1501-Ctrl. Rm. A/C & Filt. Units B 110 10/0 1504-Ctmt. Purge'Exh. & Mech. Equip. B 108 18/0 1506-Cmt. Purge Sup. AHU Rm. A 109 18/0 1507-Personnel Hatch Area 108 3/0 1508-Main Steam Iso. Valve Rm #1 115 1/0 1509-Main Steam Iso. Valve Rm #2 115 1/0 p- 1512-Ctr1. Rm. A/C & Filt. Units A 110 10/0
| |
| (- 1513-Ctrl. Bldg.' Vent Sup. A/C Unit Rm.
| |
| 1513-Aux. Bldg. Duct 2047'6" 109 119 3/0 1/0 N.A.-Containment ** 201 1/0(2)
| |
| N.A.- Containment ** 202 2/0(2)
| |
| N.A.-Containment ** 203 1/0(2)
| |
| N.A.-Containment ** 204 1/0(2)
| |
| N.A.-Containment ** 206 N.A.-Containment ** 215 3/0((2)
| |
| N.A.-Containment ** 216 1/0(2) 1/0 2)
| |
| N.A.-Containment ** 217 2)
| |
| N.A.-Containment ** 218 1/0(2) 1/0(
| |
| N.A.-Containment ** 219 4/0 N.A.-Containment ** 220 1/0(2) 3101-Ctrl. Bldg.1974' Pipe Space 300 11/0 3105-Ctr1. Bldg. Elec. Chase S. 1974' 300 1/0 3106-Ctrl. Bldg. Elec. Chase N. 1974' 300 1/0 N.A.-Area Above Access Control 301 12/0 3229-Ctr1. Bldg. Elec. Chase S. 1984' 300 1/0 3230-Ctrl. Bldg. Elec. Chase N. 1984' 300 1/0 3301-ESF Swgr. Rm. #1 314 1) 3301-ESF Swgr. Rm. #1 315 0/7(1) 3302-ESF Swgr. Rm. #2 316 0/7(1) 3302-ESF Swgr. Rm. #2 317 0/5(1) 0/5 I 3305-Ctr1. Bldg. Elec. Chase S. 2000' 301 1/0 3306-Ctr1. Bldg. Elec. Chase N. 2000' 301 1/0 3403-Non-Vit. Swgr. & Xfmr. Rm. #1 304 0/1(1)
| |
| V 3403-Non-Vit. Swgr. & Xfmr. Rm. #1 305 0/1 3404-Switchboard Rm. #4 321 0/2 WOLF CREEK - UNIT 1 3/4 3-59
| |
| | |
| TABLE 3.3-11 (Continued)
| |
| FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS
| |
| * INSTRUMENT LOCATION ZONE HEAT FLAME SM0KE (x/y) (x/y) (x/y) 3404-Switchboard Rm. #4 322 0/2(1) 3405-Battery Rm. #4 303 2/0 3407-Battery Rm. #1 303 2/0(y) 3408-Switchboard Rm. #1 325 0/2 3408-Switchboard Rm. #1 326 0/2 I1) 3409-Non-Vit. Swgr. & Xfmr. Rm. #2 323 I
| |
| 3409-Non-Vit. Swgr. & Xfmr. Rm. #2 327 0/1(I) 0/1 1) 3410-Switchboard Rm. #2 324 I
| |
| 3410-Switchboard Rm. #2 328 0/2(1) 0/2 1) 3411-Battery Rm. #2 303 2/0 3413-Battery Rm. #3 303 1/0 3414-Switchboard Rm. #3 318 0/2(1) 3414-Switchboard Rm. #3 320 0/2(1) 3415-Acc. Ctrl. & Elec. Equip. A/C 303 4/0 Units #1 3416-Acc. Ctrl. & Elec. Equip. A/C 303 4/0 Units #2 3418-Ctr1. Bldg. Elec. Chase S. 2016' 303 1/0 3419-Ctr1. Bldg. Elec. Chase N. 2016' 303 1/0 3414-Ctrl. Bldg. Elec. Chase N. 2016' 303 1/0 3410-Ctrl. Bldg. Elec. Chase S. 2016' 303 1/0 3501-Lower Cable Spreading Rm. 306 0/13 3504-Ctrl. Bldg. Elec. Chase N. 2032' 303 1/0 3505-Ctri. Bldg. Elec. Chase S. 2032' 303 1/0 3501-Ctrl. Bldg. Elec. Chase N. 2032' 303 1/0 3501-Ctri. Bldg. Elec. Chase S. 2032' 303 1/0 3601-Control Room 308 4/0 3601-Control Room 309 I
| |
| 3601-Control Room 319 0/7(I) 0/7 1) 3601-Control Room 329 20/0 3602-Pantry 308 1/0 3603-Shift Supv. Office 308 1/0 3605-Equipment Cabinet Area 308 15/0 3606-Emerg. Equip. Storage Rm. 308 1/0 3608-Janitor's Closet 308 1/0 3609-SAS Rm. 308 1/0 3617-Ctrl. Bldg. Elec. Chase S. 2047'6" 308 1/0 3618-Ctrl. Bldg. Elec. Chase N. 2047'6" 308 1/0 3605-Ctrl. Bldg. Elec. Chase S. 2047'6" 308 1/0 3801-Upper Cable Spreading Rm. 307 0/18 3804-Ctri. Bldg. Elec. Chase S. 2073'6" 308 1/0 3801-Ctr1. Bldg. Elec. Chase S. 2073'6" 308 1/0 5201-W. Diesel Gen. Rm. 501 4/0 5201-W. Diesel Gen. Rm. 502 0/8 5203-E. Diesel Gen. Rm. 500 4/0 WOLF CREEK - UNIT 1 3/4 3-60
| |
| | |
| i i
| |
| 1~
| |
| -TABLE 3.3-11 (Continued) g FIRE-DETECTION INSTRUMENTS l TOTAL NUMBER-OF INSTRUMENTS
| |
| * i -
| |
| INSTRUMENT LOCATION ZONE HEAT FLAME SM0KE (x/y) (x/y) ~(x/y) ,
| |
| . . .r
| |
| ; 5203-E. Diesel Gen. Rm. 503 0/8 j 6102-Fuel-B1dg. Railroad Bay 600 0/8 6104-Fuel. Pool Cool..HX Rm. B 601 6/0
| |
| ; 6105-Fuel. Pool Cool. HX Rm. A 601 6/0 ,
| |
| l 6202-Elec. Equipment Rm. -601- 3/0 +
| |
| ; 6203-Air Handling Equip. Rm. . 601 3/0
| |
| ! 6301-Fuel Bldg. 2047'6" Gen. Flr 602 2/0
| |
| , 6303-Fuel Bldg..Exh. Filt. Absorb. 601 2/0 j: Rm. A i 6304-Fuel Bldg. Exh. Filt. Absorb. 601 2/0 !
| |
| r -Rm. B '
| |
| l N.A.-ESW Pumphouse Train B 002 3/0 i N.A.-ESW Pumphouse Train A 001 3/0
| |
| : N.A.-ESF Transformer XNB01 016 0/6 ,
| |
| !. -N.A.-ESF Transformer XNB02 017 0/6 TABLE NOTATIONS
| |
| *(x/y): x is number of-Function A (early warning fire detection and
| |
| . notification only) instruments.
| |
| y is number of Function B (actuation of fire suppression.sys- j
| |
| !_ tems and early warning and notification) instruments.
| |
| :. **The fire detection instruments located within the containment are not i required to be OPERABLE during the performance of Type A containment j; leakage rate tests.
| |
| y l (1) Zone is associated with a Halon protected space. Each space has two i separate detection circuits (zones). One zone, in its entirety, needs I
| |
| to remain'0PERABLE.
| |
| ~
| |
| l (2) Line-type heat detector.
| |
| l- .
| |
| : j. -
| |
| l i
| |
| E WOLF CREEK - UNIT 1 3/4 3 i-I l' _ _ - - , _ , . . _ - . , . , _ _ _ , _ _ _ - - _ _ _ . _ _ _ _ _ _ _ . _ . . _ - - - . - . ~ - - - - - . -
| |
| | |
| 4 INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.9 The Loose-Part Detection System shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACIT0N:
| |
| : 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 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 Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.9 Each channel of the Loose-Part Detection System shall be demonstrated OPERABLE by performance of:
| |
| : a. A CHANNEL CHECK at least once per 24 hours,
| |
| : b. An ANALOG CHANNEL OPERATIONAL TEST except for verification of Setpoint at least once per 31 days, and
| |
| : c. A CHANNEL CfiLIBRATION at least once per 18 months.
| |
| O WOLF CREEK - UNIT 1 3/4 3-62
| |
| | |
| INSTRUMENTATION V RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to
| |
| ! ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm /
| |
| Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the 0FFSITE DOSE CALCULATION MANUAL (00CM).
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
| |
| p b. With less than the minimum number of radioactive liquid effluent s j monitoring instrumentation channels OPERABLE, take the ACTION shown V in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE
| |
| , status within the time specified in the ACTION, or explain in the l next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7, why this inoperability was not corrected within the time specified.
| |
| : c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS i
| |
| 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequen-cies shown in Table 4.3-8.
| |
| t m
| |
| WOLF CREEK - UNIT 1 3/4 3-63
| |
| | |
| TABLE 3.3-12 g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION r-MINIMUM E CHANNELS y INSTRUMENT OPERABLE ACTION
| |
| : 1. Radioactivity Monitors Providing Alarm and E Automatic Termination of Release
| |
| [ a. Liquid Radwaste Discharge Monitor (HB-RE-18) 1 31
| |
| : b. Steam Generator Blowdown Discharge Monitor (BM-RE-52) 1 32
| |
| : c. Turbine Building Drain Monitor (LE-RE-59) 1 32
| |
| : d. Secondary Liquid Waste System Monitor (HF-RE-45) 1 33
| |
| : 2. Flow Rate Measurement Devices
| |
| : a. Liquid Radwaste Discharge Line s
| |
| : 1) Waste Monitor Tank A Discharge Line 1 34
| |
| $ 2) Waste Monitor Tank B Discharge Line 1 34
| |
| : b. Steam Generator Blowdown Discharge Line 1 34
| |
| : c. Secondary Liquid Waste System Discharge Line 1 34 l
| |
| l O O O
| |
| | |
| TABLE 3.3-12 (Continued)
| |
| IA) ACTION STATEMENTS ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:
| |
| : a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
| |
| : b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
| |
| Otherwise, suspend release of radioactive effluents via this pathway.
| |
| ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for principle gamma emitters and I-131 at a lor r limit of detection as specified in Table 4.11-1:
| |
| : a. At least once per 12 hours when the specific activity of the n secondary coolant is greater than 0.01 microcurie / gram DOSE (b i EQUIVALENT I-131, or
| |
| : b. At least once per 24 hours when the specific activity of the riecondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
| |
| ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that prior to initiating a release:
| |
| : a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
| |
| : b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
| |
| Otherwise, suspend release of radioactive effluents via this pathway.
| |
| ACTION 34 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curve's generated in place may be used to estimate flow.
| |
| O V
| |
| WOLF CREEK - UNIT 1 3/4 3-65
| |
| | |
| TABLE 4.3-8 g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG 2 CHANNEL 5 CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT 7 CHECK CHECK CALIBRATION TEST E 1. Radioactivity Monitors Providing Q Alarm and Automatic Termination
| |
| ~ of Release
| |
| : a. Liquid Radwaste Discharge Monitor D P R(2) Q(1)
| |
| (HB-RE-18)
| |
| : b. Steam Generator Blowdown Discharge D M R(2) Q(1)
| |
| Monitor (BM-RE-52)
| |
| : c. Turbine Building Drain Monitor R
| |
| a (LE-RE-59) D M R(2) Q(1)
| |
| T d. Secondary Liquid Waste System D P R(2) Q(1) 8 Monitor (HF-RE-45)
| |
| : 2. Flow Rate Measurement Devices
| |
| : a. Liquid Radwaste Discharge Line D(3) N.A. R N.A.
| |
| : b. Steam Generator Blowdown Discharge Line D(3) N.A. R N.A.
| |
| : c. Secondary Liquid Waste System D(3) N.A. R N.A.
| |
| Discharge Line O O O
| |
| | |
| TABLE 4.3-8 (Continued)
| |
| /'''-)
| |
| ('
| |
| TABLE NOTATIONS (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur as i appropriate if any of the following conditions exists:
| |
| : a. Instrument indicates measured levels above the Alarm / Trip Setpoint (isolation and alarm), or
| |
| ! b. Circuit failure (alarm only), or
| |
| , c. Instrument indicates a downscale failure (al 1 only), or
| |
| : d. Instrument controls not set in operate mode (alarm only).
| |
| (2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference (gas or liquid and solid) standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.
| |
| These standards shall permit calibrating the system over its intended range of energy, measurement range, and establish monitor response to a solid calibration source. For subsequent CHANNEL CALIBRATION, NBS trace-able standard (gas, liquid, or solid) may be used; or a gas, liquid, or
| |
| -- solid source that has been calibrated by relating it to equipment that was previously (within 30 days) calibrated by the same geometry and type of
| |
| ,7 s\s_-} source standard traceable to NBS.
| |
| (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on
| |
| ; days on which continuous, periodic, or batch releases are made.
| |
| i b
| |
| O WOLF CREEK - UNIT 1 3/4 3-67
| |
| . . , , - - - - . , . - ,..,.n -.- -. --., . - - - . - . _ - , . . . - _ . . ---
| |
| | |
| INSTRUMENTATION RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION -
| |
| LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specifications 3.11.2.1 and 3.11.2.5 are not exceeded.
| |
| The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
| |
| APPLICABILITY: As shown in Table 3.3-13.
| |
| ACTION:
| |
| : a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.
| |
| : b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Specifica-tion 6.9.1.7, why this inoperability was not corrected within the time specified.
| |
| : c. The provisions of Specifications 3.0.3 ana 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequen-cies shown in Table 4.3-9.
| |
| O WOLF CREEK - UNIT 1 3/4 3-68
| |
| | |
| O O TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 6
| |
| G MINIMUM CHANNELS APPLICABILITY ACTION g INSTRUMENT OPERABLE
| |
| : 1. WASTE GAS HOLDUP SYSTEM Explosive Gas i Monitoring System
| |
| ** 44
| |
| : a. Hydrogen Monitor 1/recombiner
| |
| * ** 42
| |
| : b. Oxygen Monitor 2/recombiner
| |
| : 2. Unit Vent System
| |
| * 40
| |
| : a. Noble Gas Activity Monitor - Providing 1 Alarm (GT-RE-21)
| |
| * 43
| |
| : b. Iodine Sampler 1 w
| |
| * 43 1 c. Particulate Sampler 1 w
| |
| * 45
| |
| : d. Flow Rate N.A.
| |
| h
| |
| * 39
| |
| : e. Sampler Flow 9 ate Monitor 1
| |
| : 3. Containment Purge System
| |
| : a. Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release *
| |
| (GT-RE-22, GT-RE-33) 1 41
| |
| * 43
| |
| : b. Iodine Sampler 1
| |
| * 43
| |
| : c. Particulate Sampler 1
| |
| * 45
| |
| : d. Flow Rate N.A.
| |
| * 39
| |
| : e. Sampler Flow Rate Monitor 1
| |
| | |
| TABLE 3.3-13 (Continued) c RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 8
| |
| MINIMUM CHANNELS Q INSTRUMENT OPERABLE APPLICABILITY ACTION A
| |
| : 4. Radwaste Building Vent System 7
| |
| E a. Noble Gas Activity Monitor - Providing Z Alarm and Automatic Termination of g Release (GH-RE-10) 1
| |
| * 38, 40
| |
| : b. Iodine Sampler 1
| |
| * 43
| |
| : c. Particulate Sampler 1
| |
| * 43
| |
| : d. Flow Rate N.A.
| |
| * 45
| |
| : e. Sampler Flow Rate Monitor 1
| |
| * 39 R.
| |
| Y 8
| |
| O O O
| |
| | |
| TABLE 3.3-13 (Continued)
| |
| (g3 TABLE NOTATIONS
| |
| * At all times.
| |
| ** During WASTE GAS HOLDUP SYSTEM operation.
| |
| ACTION STATEMENTS ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the' release:
| |
| : a. At least two independent samples of the tank's contents are analyzed, and
| |
| : b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.
| |
| Otherwise, suspend release of radioactive effluents via this pathway.
| |
| / ACTION 39 - With the number of channels OPERABLE less than required by the f Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours.
| |
| 'Q ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours.
| |
| ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.
| |
| ACTION 42 - With the Outlet Oxygen Monitor channel inoperable, operation of the system may continue provided grab samples are t u n and analyzed at least once per 24 hours. 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 hours during degassing operations and at least once per 24 hours during other operations.
| |
| ACTION 43 - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sample equipment as required in Table 4.11-2.
| |
| ACTION 44 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirements, suspend oxygen supply to l
| |
| the recombiner.
| |
| ACTION 45 - Flow rate for this system shall be based on fan status and l operating curves or actual measurements.
| |
| t l
| |
| WOLF CREEK - UNIT 1 3/4 3-71 l
| |
| l
| |
| | |
| TABLE 4.3-9 r RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS S
| |
| ANALOG 9 CHANNEL MODES FOR kHICH E CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION 7 TEST IS REQUIRED E 1. WASTE GAS HOLDUP SYSTEM Explosive Z Gas Monitoring System
| |
| : a. Inlet Hydrogen Monitor D N.A. Q(4) M **
| |
| : b. Outlet Hydrogen Monitor D N.A. Q(4) M **
| |
| : c. Inlet Oxygen Monitor D N.A. Q(5) M **
| |
| : d. Outlet Oxygen Monitor D N.A. Q(6) M **
| |
| : 2. Unit Vent System R
| |
| : a. Noble Gas Activity Monitor- D M R(3) Q(2)
| |
| * Providing Alarm (GT-RE-21)
| |
| (a g b. Iodine Sampler W N.A. N.A. N.A. *
| |
| : c. Particulate Sampler W N.A. N.A. N.A. *
| |
| : d. Flow Rate N.A. N.A. R(7) N.A. *
| |
| : e. Sampler Flow Rate Monitor D N.A. R Q
| |
| : 3. Containment Purge System
| |
| : a. Noble Gas Activity Monitor -
| |
| Providing Alarm and Automatic Termination of Release (GT-RE-22, GT-RE-33) D P R(3) Q(1) *
| |
| : b. Iodine Sampler W N.A. N.A. N.A. *
| |
| : c. Particulate Sampler W N.A. N.A. N.A. *
| |
| : d. Flow Rate N.A. N.A. R(7) N.A. *
| |
| : e. Sampler Flow Rate Monitor D N.A. R N.A.
| |
| * O O O
| |
| | |
| i TABLE 4.3-9 (Continued) g RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS G
| |
| ANALOG i E CHANNEL MODES FOR WHICH i E CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE ?
| |
| ; 7 INSTRUMENT CHECK ,_ CHECK CALIBRATION TEST- IS REQUIRED E 4. Radwaste Building Vent System H
| |
| s a. Noble Gas Activity Monitor - D, P M, P R(3) Q(1)
| |
| * Providing Alarm and Automatic Termination of Release (GH-RE-10) '
| |
| : b. Iodine Sampler W N.A. N.A. N.A. *
| |
| : c. Particulate Sampler W N.A. N.A. N.A.
| |
| * l d. Flow Rate N.A. N.A. R(7) N.A.
| |
| * i R '
| |
| : e. Sampler Flow Rate Monitor D N.A. R N.A. * ,
| |
| S 4 ;
| |
| I i
| |
| I
| |
| | |
| TABLE 4.3-9 (Continued)
| |
| TABLE NOTATIONS n
| |
| * At all times.
| |
| ** During WASTE GAS HOLDUP SYSTEM operation.
| |
| (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation as appropriate occur if any of the following conditions exists:
| |
| : a. Instrument indicates measured levels above the Alarm / Trip Setpoint (isolation and alarm), or
| |
| : b. Circuit failure (alarm only), or
| |
| : c. Instrument indicates a downscale failure (alarm only) or
| |
| : d. Instrument controls not set in operate mode (alarm only).
| |
| (2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any one or combination of the following conditions exists:
| |
| : a. Instrument indicates measured levels above the Alarm Setpoint
| |
| : b. Circuit failure
| |
| : c. Instrument indicates a downscale failure
| |
| : d. Instrument controls not set in operate mode.
| |
| (3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference (gas or liquid and solid) standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.
| |
| These standards shall permit calibrating the system over its intended range of energy, measurement range, and establish monitor response to a solid calibration source. For subsequent CHANNEL CALIBRATION, NBS traceable standard (gas, liquid, or solid) may be used; or a gas, liquid, or solid source that has been calibrated by relating it to equipment that was previously (within 30 days) calibrated by the same geometry and type of source traceable to NBS.
| |
| (4) The CHANNEL CALIBRt. TION shall include the use of standard gas samples containing a nominal:
| |
| : a. One volume percent hydrogen, balance nitrogen, and
| |
| : b. Four volume percent hydrogen, balance nitrogen.
| |
| WOLF CREEK - UNIT 1 3/4 3-74
| |
| | |
| TABLE 4.3-9 (Continued)
| |
| TABLE NOTATIONS (Continued)
| |
| (5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
| |
| : a. One volume percent oxygen, balance nitrogen, and
| |
| : b. Four volume percent oxygen, balance nitrogen.
| |
| (6) The CHANNEL CALIBRATION shall include 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.
| |
| (7) If flow rate is determined by exhaust fan status and fan performance curves, the following surveillance operations shall be performed at least once per 18 months:
| |
| : a. The specific vent flows by direct measurement, or
| |
| ,[ b. _The differential pressure across the exhaust fan and vent flow
| |
| \ established by the fan's " flow-AP" curve, or
| |
| : c. The fan motor horsepower measured and vent flow established by th?
| |
| fan's " flow-horsepower" curve.
| |
| I WOLF CREEK - UNIT 1 3/4 3-75
| |
| | |
| INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2,* and 3.*
| |
| ACTION:
| |
| : a. With one stop valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoper-able, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam lines or isolate the turbine from the steam supply within the next 6 hours.
| |
| : b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours isolate the turbine from the steam supply.
| |
| SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.
| |
| 4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:
| |
| : a. At least once per 7 days by cycling 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 l 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 disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or Corrosion.
| |
| *Not applicable in MODE 2 or 3 with all main steam line isolation valves and associated bypass valves in the the closed position and all other steam flow paths to the turbine isolated.
| |
| WOLF CREEK - UNIT 1 3/4 3-76
| |
| | |
| 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops'shall be in operation.
| |
| APPLICABILITY: MODES 1 and 2.*
| |
| ACTION:
| |
| With less than the above required reactor coolant loops in operation, be in at least HOT STANOBY within 6 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours.
| |
| 4
| |
| *See.Special Test Exception Specification 3.10.4.
| |
| O
| |
| ! WOLF CREEK - UNIT 1 3/4 4-1
| |
| | |
| REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least three of the reactor coolant loops listed below shall be OPERABLE and at least two of these reactor coolant loops shall be in operation:*
| |
| : a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,
| |
| : b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,
| |
| : c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and
| |
| : d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
| |
| APPLICABILITY: MODE 3.**
| |
| ACTION:
| |
| : a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
| |
| : b. With only one reactor coolant loop in operation, restore at least two loops to operation within 72 hours or within 1 hour open the Reactor Trip System breakers.
| |
| : c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.
| |
| SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
| |
| 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side wide range water level to be greater than or equal to 10% at least once per 12 hours.
| |
| 4.4.1.2.3 At least two reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours.
| |
| *All reactor coolant pumps may be deenergized for up to 1 hour provided:
| |
| (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
| |
| **See Special Test Exception Specification 3.10.4.
| |
| WOLF CREEK - UNIT 1 3/4 4-2
| |
| | |
| REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops.shall be in operation:*
| |
| : a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,**
| |
| : b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,**
| |
| : c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,**
| |
| : d. . Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,**
| |
| : e. RHR Loop A, and
| |
| : f. RHR Loop B.
| |
| APPLICABILITY: MODE 4.
| |
| s ACTION:
| |
| : a. With less than the above required reactor coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours.
| |
| : b. With no reactor coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
| |
| *All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
| |
| **A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures. .
| |
| l O
| |
| WOLF CREEK - UNIT 1 3/4 4-3
| |
| - . . . c
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pump, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
| |
| 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side wide range water level to be greater than or equal to 10% at least once per 12 hours.
| |
| 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 4-4
| |
| | |
| REACTOR COOLANT SYSTEM A
| |
| COLD SHUTOOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:
| |
| : a. One additional RHR loop shall be OPERABLE #, or
| |
| : b. The secondary side water level of at least two steam generators shall be greater than 10% of the wide range.
| |
| APPLICABILITY: MODE 5 with reactor coolant loops filled ##.
| |
| ACTION:
| |
| : a. With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator level as soon as possible.
| |
| : b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR p loop to operation.
| |
| SURVEILLANCE REQUIREMENTS 4.4.1.4.1.3 The secondary side water level of at least two steam generators when requirt shall be determined to be within limits at least once per 12 hours.
| |
| 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
| |
| #0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
| |
| ##A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures.
| |
| *The RHR pump may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
| |
| WOLF CREEK - UNIT 1 3/4 4-5
| |
| | |
| REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE # and at least one RHR loop shall be in operation.*
| |
| APPLICABILITY: MODE 5 with Reactor Coolant loops not filled.
| |
| ACTION:
| |
| : a. With less than the above requirad RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
| |
| : b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
| |
| SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and O
| |
| circulating reactor coolant at least once per 12 hours.
| |
| #0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
| |
| *The RHR pump may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
| |
| O WOLF CREEK - UNIT 1 3/4 4-6
| |
| | |
| , REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig i EE.*
| |
| APPLICABILITY: MODES 4 and 5.
| |
| ACTION:
| |
| With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving pocitive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.
| |
| 1 l-SURVEILLANCE REQUIREMENTS
| |
| , 4.4.2.1 No additional requirements other than those required by Specification 4.0.5.
| |
| l l
| |
| *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
| |
| b O
| |
| WOLF CREEK - UNIT 1 3/4 4-7 l.- , - . .-,
| |
| | |
| REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig i 1%.*
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by O
| |
| Specification 4.0.5.
| |
| *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
| |
| O WOLF CREEK - UNIT 1 3/4 4-8
| |
| | |
| s p s v) g REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with at least two groups of backup pressurizer heaters each having a capacity of at least 150 kW and a water level of less than or equal to 92% (1657 cubic feet).
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one group of backup pressurizer heaters inoperable, restore at least two groups of backup heaters to OPERABLE status within 72 hours or be in at least HOT STANDBY _ within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
| |
| : b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| O V
| |
| SURVEILLANCE REQUIREMENTS i
| |
| 4.4.3.1 The pressurizer water level shall be determined to be within its limit at least once per 12 hours.
| |
| 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at least once per 92 days.
| |
| l l
| |
| l l
| |
| I 1G WOLF CREEK - UNIT 1 3/4 4-9 1
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.*
| |
| ACTION:
| |
| : a. With one or more PORV(s) inoperable because of excessive seat leakage,
| |
| , within 1 hour either restore the PORV(s) to OPERABLE status or close l the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour either restore the PORV to OPERABLE status, or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following t
| |
| 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
| |
| : d. With one or more block valve (s) inoperable, within 1 hour: (1) restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV and remove power from its associated solenoid valve; and (2) apply ACTION b. or
| |
| : c. above, as appropriate, for the isolated PORV(s).
| |
| : e. The provisions of Specification 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION.
| |
| 4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per l 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.
| |
| *With all RCS cold leg temperatures abovc 368 F.
| |
| WOLF CREEK - UNIT 1 3/4 4-10
| |
| | |
| l n
| |
| REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3 and 4.
| |
| ACTION:
| |
| With one or more steam generators inoperable, restore the inoperable steam generator (s) to OPERABLE status prior to increasing T,yg above 200 F.
| |
| SURVEILLANCE REQUIREMENTS _
| |
| 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
| |
| b Q 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
| |
| 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam '
| |
| generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include 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:
| |
| i j
| |
| : a. Where experience in similar plants with similar water chemistry indicates critical 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 l shall include:
| |
| l [N l
| |
| WOLF CREEK - UNIT 1 3/4 4-11
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
| |
| : 2) Tubes in those areas where experience has indicated potential problems, and
| |
| : 3) A tube inspection (pursuant to Specification 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 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 4.4-2) 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.
| |
| The results of each sample inspection shall be classified into one of the following three categories:
| |
| Category Inspection Results C-1 Less than 5% of the total tubes 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 total tubes inspected are degraded tubes.
| |
| C-3 More than 10% of the total tubes inspected are 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.
| |
| O WOLF CREEK - UNIT 1 3/4 4-12
| |
| | |
| .i
| |
| ("N REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
| |
| : a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.
| |
| Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the pre-service inspection, result in all inspection 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 4.4-2 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 shall apply until the subsequent inspections satisfy the criteria of
| |
| (~N Specification 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 4.4-2 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 limits of l Specification 3.4.6.2, or l
| |
| l 2) A seismic occurrence greater than the Operating Basis Earthquake, or
| |
| : 3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or
| |
| : 4) A main steam line or feedwater line break.
| |
| 1 V
| |
| WOLF CREEK - UNIT 1 3/4 4-13
| |
| | |
| REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
| |
| : a. As used in this specification:
| |
| : 1) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;
| |
| : 2) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube;
| |
| : 3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
| |
| : 4) % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
| |
| : 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective;
| |
| : 6) Plugging Limit 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 structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specificatibn 4.4.5.3c., above;
| |
| : 8) Tube Inspection means an. inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and O
| |
| WOLF CREEK - UNIT 1 3/4 4-14
| |
| | |
| m 1
| |
| d REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques ex-pected to be used during 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 4.4-2.
| |
| 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 Specification 6.9.2;
| |
| /%
| |
| ) b. The complete results of the steam generator tube inservice inspection
| |
| -(d shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
| |
| : 1) Number and extent of tubes inspected,
| |
| : 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and i
| |
| : 3) Identification of tubes plugged.
| |
| i
| |
| : c. Results of steam generator tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of l 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.
| |
| l O
| |
| WOLF CREEK - UNIT 1 3/4 4-15
| |
| | |
| E 9
| |
| n A TABLE 4.4-1 9
| |
| MINIMUM NUMBER OF STEAM GENERATORS TO BE E
| |
| q INSPECTED DURING INSERVICE INSPECTION Preservice inspection No Yes No. of Steam Generators per Unit Two Three Four Two Three Four First inservice Inspection All One Two Two w Second & Subsequent inservice inspections One l One l One 2 One 3 1
| |
| p TABLE NOTATIONS 5 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 circum-stances the sample sequence shall be modified to inspect the most severe conditions.
| |
| : 2. The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections 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 described in 1 above.
| |
| O O O
| |
| | |
| ^x
| |
| \v
| |
| :c
| |
| * S, TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION l
| |
| 5 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION C
| |
| 3
| |
| -e Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N. A. N. A. N. A. N. A.
| |
| ! S Tubes per
| |
| ] S. G.
| |
| C-2 Plug defective tubes C-1 None N. A. N. A.
| |
| and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G.
| |
| Perform action for w C-3 C-3 result of first s sample
| |
| ! 4
| |
| , Perform action for i
| |
| C-3 C-3 result of first N. A. N. A.
| |
| Z sample C-3 Inspect all tubes in All other this S. G., plug de. S. G.s are None . - g fective tubes and C-1 inspect 2S tubes in Some S. G.s Perform action for each other S. G- N. A. N. A.
| |
| C-2 but no C-2 result of second additional sample Notification to NRC S. G. are pursuant to @50.72 C-3 (b)(2) of 10 CF R Additional Inspect all tubes in Part 50 S. G. is C-3 each S. G. and plug defective tubes.
| |
| Notification to N RC N. A. N. A.
| |
| pursuant to @50.72 (b)(2) of 10 CF R Part 50 S=3 % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected
| |
| , n during an inspection
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
| |
| : a. The Containment Atmosphere Particulate Radioactivity Monitoring System,
| |
| : b. The Containment Normal Sump Level Measurement System, and
| |
| : c. Either the Containment Air Cooler Condensate Flow Rate or the Containment Atmosphere Gaseous Radioactivity Monitoring System.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4..
| |
| ACTION:
| |
| With only two of the above required Leakage Detection Systems OPERABLE, opera-tion may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for gaseous and particulate radioactivity or a gamma isotopic analysis of the conta',nment atmosphere is performed using the Post Accident Sampling System at least once per 24 hours when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
| |
| : a. Containment Atmosphere Gaseous and Particulate Monitoring System performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,
| |
| : b. Containment Normal Sump Level Measurement System performance of CHANNEL CALIBRATION at least once per 18 months, and
| |
| : c. Containment Air Cooler Condensate Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.
| |
| WOLF CREEK - UNIT 1 3/4 4-18
| |
| | |
| (~N REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
| |
| : a. No PRESSURE BOUNDARY LEAKAGE,
| |
| : b. 1 gpm UNIDENTIFIED LEAKAGE,
| |
| : c. 1 gpm total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator,
| |
| : d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
| |
| : e. 8 gpm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig, and
| |
| : f. 1 gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*
| |
| \ APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| : a. With any PRESSURE B0UNDARY LEAKAGE, be in at least H0T STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, reduce the leakage rate to within limits within 4 hours, or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 12 hours with an RCS pressure of less than 600 psig.
| |
| * Test pressures less than 2235 psig but greater than 150 psig are allowed.
| |
| Observed leakage shall be adjusted for the natural test pressure up to O 2235 osig assuming the leakage to be directly proportional to pressure is differential to the one-half power.
| |
| A WOLF CREEK - UNIT 1 3/4 4-19 i
| |
| | |
| l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
| |
| : a. Monitoring the containment atmosphere gaseous or particulate radioactivity monitor at least once per 12 hours;
| |
| : b. Monitoring the containment normal sump inventory and discharge at least once per 12 hours;
| |
| : c. Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;
| |
| : d. Performance of a Reactor Coolant System water inventory balance at
| |
| .least once per 72 hours; and
| |
| : e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours.
| |
| 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:
| |
| : a. At least once per 18 months,
| |
| : b. Prior to entering MODE 2 whenever the unit has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months,
| |
| : c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and
| |
| : d. Within 24 hours following valve actuation due to automatic or manual action or flow through the valve except for valves BBPV8702 A/8 and EJHV8701 A/B.
| |
| The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
| |
| O WOLF CREEK - UNIT 1 3/4 4-20
| |
| | |
| TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION BBV8948 A, B, C, D SI/RHR/Accum. Cold Leg Injection BBV8949 A, B, C, O SI/RHR Hot Leg Injection BBV001, 022, 040, 059 Bit Cold Leg Injection BBPV8702 A, B RHR Normal Suction EJV8841 A, B RHR Hot Leg Recirc Ctat ISO EJHV8701 A, B RHR Normal Suctica EMV001, 002, 003, 004 SI Hot : 9g Inj Ctat ISO EM8815 BIT Inj. Ctmt Isolation EPV010, 020, 030, 040 SI Cold Leg Inj Ctmt ISO EPV8813, A, B, C, D RHR Cold Leg Inj Ctmt 150 EPV8956 A, B, C, D Accum Inj Isolation O
| |
| O WOLF CREEK - UNIT 1 3/4 4-21
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained witnin 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 parameter in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| At All Other Times:
| |
| With the concentration of either chloride or fluoride in the Reactor
| |
| ; Coolant System in excess of its Steady-State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the I structural integrity 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 i
| |
| 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 4.4-3.
| |
| l l
| |
| 9 WOLF CREEK - UNIT 1 3/4 4-22
| |
| | |
| - l I
| |
| l' i i TABLE 3.4-2
| |
| (
| |
| i- REACTOR COOLANT SYSTEM ,
| |
| i
| |
| ; CHEMISTRY LIMITS i-l-
| |
| ) -STEADY-STATE TRANSIENT :
| |
| ! PARAMETER LIMIT LIMIT ;
| |
| i- Dissolved Oxygen
| |
| * 5 0.10 ppm 5 1.00 ppm j' - Chloride 1 0.15 ppm i 1.50 ppm i
| |
| Fluoride 1 0.15 ppm i 1.50 ppm i
| |
| s P
| |
| l-i- i f
| |
| * Limit not applicable with T,y less than or equal to 250 F. ,
| |
| i i
| |
| i I ,
| |
| b g.
| |
| WOLF CREEK - UNIT 1 3/4 4-23
| |
| | |
| TABLE 4 4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS SAMPLE AND PARAMETER ANALYSIS FREQUENCY Dissolved Oxygen
| |
| * At least once per 72 hours Chloride At least once per 72 hours Fluoride At least once per 72 hours
| |
| *Not required with T avg less than or equal to 250*F O
| |
| i i
| |
| l l
| |
| l l
| |
| 9 WOLF CREEK - UNIT 1 3/4 4-24
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
| |
| : a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and
| |
| : b. Less than or equal to 100/E microCuries per gram of gross radioactivity.
| |
| APPLICABILITY: MODES 1, 2, 3, 4, and 5.
| |
| ACTION:
| |
| MODES 1, 2, and 3*:
| |
| : a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours provided that the cumula-tive operating time under these circumstances .does not exceed 800 hours in any consecutive 12-month period. The provisions of h Specification 3.0.4 are not applicable;
| |
| : b. With the total cumulative operating time at a reactor coolant specific activity greater than 1 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive 6-month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours above this limit.
| |
| The provisions of Specification 3.0.4 are not applicable;
| |
| : c. With the specific activity of the reactor coalant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500 F within 6 hours; and
| |
| : d. With the specific activity of the reactor coolant greater than 100/E microCuries per gram of gross radioactivity, be in at least HOT STANDBY with T avg less than 500 F within 6 hours.
| |
| *With T avg greater than or equal to 500 F.
| |
| L)
| |
| WOLF CREEK - UNIT 1 3/4 4-25
| |
| | |
| REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| MODES 1, 2, 3, 4, and 5:
| |
| With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E_
| |
| microCuries per gram of gross radioac.tivity, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days with a copy to the Director, Nuclear Reactor Regula-tion, Attention: Chief, Core Performance Branch, and Chief, Accident Evaluation Branch, U.S. Nuclear Regulatory Commission, Washington, D.C.
| |
| 20555. This report shall contain the results of the specific activity analyses together with the following information:
| |
| : a. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded,
| |
| : b. Results of: (1) the last isotopic analysis for radioiodine performed prior to exceeding the limit, (2) analysis while limit was exceeded and (3) one analysis after the radiciodine activity was reduced to less than the limit, including for each isotopic analysis, the date and time of sampling and the radioiodine concentrations,
| |
| : c. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded,
| |
| : d. History of degassing operations, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and
| |
| : e. The time duration when the specific activity of the primary coolant exceeded 1 microcurie per gram DOSE EQUIVALENT I-131.
| |
| SURVEILLANCE REQUIREMENTS 1
| |
| 4.4.8 The specific activity of the primary coolant shall be determined to be f within the limits by performance of the sampling and analysis program of l Table 4.4-4. l 9
| |
| WOLF CREEK - UNIT 1 3/4 4-26
| |
| | |
| . i .. , . , , .
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| s 5 250 ,
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| ~
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| MO D . . iUNACCEPTABLE .
| |
| < , ,' :, . i '
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| O. .P. E, RATION.
| |
| g .
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| , . 4 . ,
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| ,9 .
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| i4 . .
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| 9 Q . , L -1 s .
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| W * . ..
| |
| g ^
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| . . s. ...,
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| $150 s :h .i O :
| |
| O
| |
| * U i g
| |
| O g . .
| |
| o . . .
| |
| m 100 g CC i E,- -
| |
| l
| |
| $ ACCEPTABLE .
| |
| "j OPERATION 4 50 5
| |
| 0 m
| |
| w l
| |
| 8 I
| |
| 0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER 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 >l pCi/ gram DOSE EQUIVALENT I-131 x
| |
| WOLF CREEK - UNIT 1 3/4 4-27
| |
| | |
| TABLE 4.4-4 -
| |
| =c REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE E2 AND ANALYSIS PROGRAM n
| |
| 53 TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE A AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED n
| |
| i 1. Gross Radioactivity At least once per 72 hours 1,2,3,4 g Determination El 2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days 1 ra LENT I-131 Concentration
| |
| : 3. Radiochemical for E Determination 1 per 6 months
| |
| * 1
| |
| : 4. Isotopic Analysis for Iodine a) Once per 4 hours, 1#, 2#, 3#, 4#, 5#
| |
| Including I-131, I-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE EQUIVALENT I-131 02 or 100/E pCi/ gram of gross 30 radioactivity, and i b) One sample between 2 1,2,3
| |
| $ and 6 hours following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period.
| |
| #Until the specific activity of the Reactor Coolant System is restored within its limits
| |
| * Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
| |
| O O O
| |
| | |
| 3/4.4.9 PRESSURE / TEMPERATURE LIMITS v
| |
| REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION !
| |
| 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and nydrostatic testing with:
| |
| : a. A maximum heatup of 100 F in any 1-hour period,
| |
| : b. A maximum cooldown of 100*F in any 1-hour period, and
| |
| : c. A maximum temperature change of less than or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| ) With any of the above limits exceeded, restore the temperature and/or pressure
| |
| (_,/ to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or bo in at least HOT STANDBY within the next 6 hours and reduce the RCS Tavg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 mir.utes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
| |
| 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4.
| |
| O WOLF CREEK - UNIT 1 3/4 4-29
| |
| | |
| MATERIAL PROPERTY BASIS COPPER CONTENT : ASSUMED TO BE 0.10 WT%
| |
| RTuny INITIAL :40*F RTggy AFTER 16 EFPY :1/4T,110*F 3/4T,87'F CURVE APPLICABLE FOR HEATUP RATES UP TO 60'F/HR AND 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.
| |
| I LEAK TEST l f f I I
| |
| ''"'h / ; f j
| |
| \' / / J I ;
| |
| CRITICALITY LIMIT f f [ '
| |
| f f - BASED ON 60*/HR ---
| |
| ) ) ;
| |
| )") -
| |
| HEATUP CURVE
| |
| - ( ) I i i [ f) {
| |
| $ I I f f 5 / /J /
| |
| W [ (( [
| |
| CRITICALITY LIMIT BASED ON 10&F/HR--
| |
| 5 ) [ ) k' HEATUP CURVE e / / / /
| |
| 8 i / / /
| |
| g / // /
| |
| 1000
| |
| / '
| |
| < < ' (
| |
| HEATUP_
| |
| CURVES
| |
| / [ ; [
| |
| } { f 1 ! CRITICALITY LIMIT BASED ON __
| |
| [ / INSERVICE HYDROSTATIC TEST l l / / -
| |
| d TEMPERATURE (255'F) FOR THE --
| |
| __ 60'F/HR c- , / SERVICE PERIOD UP TO 16 EFPY 100'F/HR I 0
| |
| 0 100 200 300 400 500 INDICATED AVERAGE TEMPERATURE (DEG. F)
| |
| FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY WOLF CREEK - UNIT 1 3/4 4-30
| |
| | |
| \4
| |
| .4......u .. :
| |
| (v/
| |
| ..i~. MATERIAL PROPERTY BASIS .-
| |
| t.
| |
| 6 COPPER CONTENT : ASSUMED TO BE 0.10 WT% .
| |
| RT NOT INITIAL : 40T T t.- -
| |
| AFTER 14 EFPY: 1/4T,110T T C RT NOT
| |
| +-
| |
| E 3/4T. SPF ?
| |
| M 5..
| |
| +- CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100T/HR FOR THE SERVICE 1 PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 10T and 80 PSIG FOR POSSIBLE :
| |
| .T. INSTRUMENT ERRORS.
| |
| r.
| |
| . ... . .. ...a, u..... .
| |
| _m.%. ..- ...
| |
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| |
| .9 g.. ...+
| |
| 0 100 200 300 400 500 INDICATED AVERAGE TEMPERATURE (DEG. F)
| |
| FIGURE 3.4-3 REACTOR C0OLANT SYSTEM C00LD0WN LIMITATIONS APPLICABLE UP TO 16 EFPY
| |
| \v)
| |
| WOLF CREEK - UNIT 1 3/4 4-31
| |
| | |
| TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY)
| |
| U 58.5 4.00 1st Refueling Y 241 3.69 5 V 61 3.69 9 X 238.5 4.00 15 W 121.5 4.00 Standby Z 301.5 4.00 Standby 9
| |
| O WOLF CREEK - UNIT 1 3/4 4-32
| |
| | |
| l
| |
| ' 's REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
| |
| 1
| |
| : a. A maximum heatup of 100*F in any 1-hour period,
| |
| : b. A maximum cooldown of 200 F in any 1-hour period, and
| |
| : c. A maximum spray water temperature differential of 583*F.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig O,, within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation.
| |
| i i
| |
| I WOLF CREEK - UNIT 1 3/4 4-33
| |
| | |
| REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:
| |
| : a. Two residual heat removal (RHR) suction relief valves each with a Setpoint of 450 psig i 3%, or
| |
| : b. Two power-operated relief valves (PORVs) with Setpointc which do not exceed the limit established in Figure 3.4-4, or
| |
| : c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square inches.
| |
| APPLICABILITY: MODE 3 when the temperature of any RCS cold leg is less than or equal to 368 F, MODES 4 and 5, and MODE 6 with the reactor vessel head on.
| |
| ACTION:
| |
| : a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours,
| |
| : b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 2 square inch vent within 8 hours.
| |
| : c. In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence. ,
| |
| 1
| |
| : d. The provisions of Specification 3.0.4 are not applicable.
| |
| 1 1
| |
| 01 WOLF CREEK - UNIT 1 3/4 4-34
| |
| | |
| REACTOR COOLANT SYSTEM
| |
| (
| |
| D SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
| |
| : a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;
| |
| : b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and
| |
| : c. Verifying the PORV. isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.
| |
| 4 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when i the RHR suction relief valves are being used for cold overpressure protection as follows:
| |
| : a. For RHR suction relief valve 87088:
| |
| : 1) By verifying at least once per 31 days that RHR RCS Suction
| |
| , ['q Isolation Valve (RRSIV) 8701B is open with power to the 4
| |
| valve operator removed, and
| |
| : 2) By verifying at least once per 12 hours that RRSIV 87028 is open.
| |
| : b. For RHR suction relief valve 8708A:
| |
| : 1) By verifying at least once per 31 days that RRSIV 8702A is open with power to the valve operator removed, and
| |
| : 2) By verifying at least once per 12 hours that RRSIV 8701A is open.
| |
| : c. Testing pursuant to Specification 4.0.5.
| |
| 4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours
| |
| * when the vent (s) is being used for overpressure protection.
| |
| *Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at
| |
| {v least once per 31 days.
| |
| WOLF CREEK - UNIT 1 3/4 4-35
| |
| .- - .. - - .. . _ . . ._ ~
| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| 2000
| |
| _._.m,_.. .._... .._.. _....
| |
| . . . ~ _. .. . .m.... ._ .. .. _ _ . .m. _ . . . ..... ...._. .._ . ._.._. E._.. ._ . . _..-.
| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| _ +_~~
| |
| RTD SMAX ..._....
| |
| . .. _....t_.....
| |
| m.. . - . . .
| |
| z_ . . . . _
| |
| _ . _1 . . . . . .
| |
| ._.... (*F) (psig) _.... _.._..
| |
| O
| |
| : o. 1500 70 530
| |
| . ._..t....
| |
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| |
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| |
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| |
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| |
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| |
| 213 575 - - .
| |
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| |
| _m. . . . _ .
| |
| _... . 263
| |
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| |
| 785 ~..
| |
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| |
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| |
| ...~..
| |
| . 313 . . . . .
| |
| _._ _.++._._ ._ . ..._.t._ __, 1335 _ . _ . ._. . . _ .._ _ . . _. . . . . _. . . _ . . . ... _.._
| |
| . . ~ . . 363 2440 ._~ . . _ ~ ~__ _-
| |
| : : : +
| |
| . . = . _ __. _ . ._. . . _ _ . _ . . . .._
| |
| m _
| |
| ... _...... . _. _.... ._... L. .
| |
| -_. .... .+
| |
| _...~._ - m.. _._
| |
| .m_. . _...
| |
| _ . . __ .. . .. . ...t. ... ..
| |
| 1000 .. _._ ....._ . . . . .. ..... .. _ . . . . _ . _. . . . . .
| |
| : m. _. _._ ....._.. _.. _ ... _.__
| |
| . . . . -... . . _. .. ._ __ .. .._J_...
| |
| . . . .~ . . ._.
| |
| _ .. m ._. . _ . . . _. . - . .. . .. . . .. . .... .. .
| |
| ~ . - . . . . . . ._ . .. ._..
| |
| .: _.._ : ._ . _.__. _.. .. . _ . . ._... E. L..
| |
| ..~_. :
| |
| . . ~
| |
| _._t__.
| |
| ...~.
| |
| __. _...~.._ ._...
| |
| . . . -~._ .. .. . . . . . . _ .
| |
| . _.+ . . _-..w
| |
| ._ t
| |
| . t..
| |
| .... . .. m.t.. _...t._..
| |
| m g,_
| |
| 500 __.. .. ..... _. . . . . _ . . ..... . ;
| |
| ...L.__.__ .tm. .
| |
| .... . . - . ...~ . ....
| |
| .. _ ~-_ - ._._.
| |
| _ . ..- . . . . _ ._m..
| |
| .~ ._.. .... ....
| |
| . ...=.m_.= . . . ... , _ .. _ . . . . .-.
| |
| ._... ... . ~.... -. _ .....
| |
| _ ._ _ _ . .. _... _ . ....t_....
| |
| .....t...
| |
| ._.L.... . . . .... , . . ..
| |
| .m..
| |
| 100 200 300 400 MEASURED RTD TEMPERATURE, 'F FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM WOLF CREEK - UNIT 1 3/4 4-36 s
| |
| | |
| REACTOR COOLANT SYSTEM lA) v 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.
| |
| APPLICABILITY: All MODES.
| |
| ACTION:
| |
| : a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.
| |
| [ b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate A the affected component (s) prior to increasing the Reactor Coolant
| |
| (
| |
| ) System temperature above 200 F.
| |
| : c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
| |
| : d. The provisions of Specification 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of 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.
| |
| N U
| |
| WOLF CREEK - UNIT 1 3/4 4-37
| |
| | |
| REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one reactor vessel head vent path consisting of at least two valves in series powered from emergency busses shall be OPERABLE and closed.
| |
| APPLICABILITY: 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 hours and in COLD SriUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:
| |
| : a. Verifying all manual isolation 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 SHUTOOWN or REFUELING, and
| |
| : c. Verifying flow through the reactor vessel head vent paths during venting during COLD SHUTDOWN or REFUELING.
| |
| O WOLF CREEK - UNIT 1 3/4 4-38
| |
| | |
| ('
| |
| 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:
| |
| : a. The isolation valve open and power removed,
| |
| : b. A contained borated water volume of between 6122 and 6594 gallons,
| |
| : c. A boron concentration of between 1900 and 2100 ppm, and
| |
| : d. A nitrogen cover pressure of between 585 and 665 psig.
| |
| APPLICABILITY: MODES 1, 2, and 3*.
| |
| ACTION:
| |
| : a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next
| |
| ( 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
| |
| : a. At least once per 12 hours by:
| |
| : 1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and
| |
| : 2) Verifying that each accumulator isolation valve is open.
| |
| * Pressurizer pressure above 1000 psig.
| |
| f b
| |
| WOLF CREEK - UNIT 1 3/4 5-1 ;
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 70 gallons by verifying the boron concentration of the accumulator solution, and
| |
| : c. At least once per 31 days when the RCS pressure is above 1000 psig by verifying that the circuit breaker supplying power to the isolation valve operator is open.
| |
| 4.5.1.2 Each accumulator water level and pressure channel shall be demonstrated OPERABLE at least once per 18 months by the performance of a CHANNEL CALIBRATION.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 5-2
| |
| | |
| (
| |
| \s_,/ EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
| |
| J
| |
| : a. One OPERABLE centrifugal charging pump,
| |
| : b. .One OPERABLE Safety Injection pump,
| |
| : c. One OPERABLE RHR heat exchanger,
| |
| : d. One OPERABLE RHR pump, and
| |
| : e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation f'~'s phase of operation.
| |
| ?
| |
| \ ') APPLICABILITY: MODES 1, 2, and 3.*
| |
| ACTION:
| |
| : a. With one ECCS subsystem inoperable, restore the inoperable subsystem to CPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall De prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
| |
| *The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pumps and the Safety Injection pumps declared inoperable pursuant to Specification 4.5.3.2 provided the centrifugal charging pumps and the Safety Injection pumps are restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs s exceeding 375 F, whichever comes first.
| |
| h O\
| |
| WOLF CREEK - UNil 1 3/4 5-3
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
| |
| : a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
| |
| Valve Number Valve Function Valve Position BN-HV-8813 Safety Injection to Open RWST Isolation Vlv EM-HV-8802A(B) SI Pump Discharge Closed Hot Leg Iso Vlvs EM-HV-8835 Safety Injection Open Cold Leg Ico Valve EJ-HV-8840 RHR/SI Hot Leg Closed Recirc Iso Valve EJ-HV-8809A RHR to Accum Inj Open Loops 1 & 2 Iso Vlv EJ-HV-8809B RHR to Accum Inj Open Loops 3 & 4 Iso Viv
| |
| : b. At least once per 31 days by:
| |
| : 1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and
| |
| : 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
| |
| : c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
| |
| : 1) For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and
| |
| : 2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
| |
| : d. At least once per 18 months by:
| |
| : 1) Verifying automatic isolation and interlock action of the RHR System from the Reactor Coolant System by ensuring that:
| |
| a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 750 psig the interlocks will cause the valves to automatically close.
| |
| WOLF CREEK - UNIT 1 3/4 5-4
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
| |
| : e. .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/or on Automatic Switchover to Containment Sump from RWST Level-Low-Low coincident with Safety Injection test signal; and
| |
| : 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
| |
| a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.
| |
| : f. By verifying that each of the following pumps develops the required differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:
| |
| : 1) Centrifugal charging pump > 2400 psid,
| |
| : 2) Safety Injection pump > 1445 psid, and
| |
| : 3) RHR pump > 165 psid.
| |
| : g. By verifying the correct position of each mechanical position stop for the following ECCS throttle valves:
| |
| : 1) Within 4 hours following completion of each valve stroking i operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and
| |
| : 2) At least once per 18 months.
| |
| I HPSI System CVCS System i
| |
| Valve Numbers Valve Numbers EMV095 EMV109 BGV-198 O EMV096 EMV110 BGV-199 EMV097 EMV089 BGV-200 V EMV098 EMV090 BGV-201 EMV107 EMV091 BGV-202 EMV108 EMV092 WOLF CREEK - UNIT 1 3/4 5-5
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : h. By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
| |
| : 1) For centrifugal charging pump lines, with a single pump running:
| |
| a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 346 gpm, and b) The total pump flow rate is less than or equal to 556 gpm.
| |
| : 2) For Safety Injection pump lines, with a single pump running:
| |
| a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 459 gpm, and b) The total pump flow rate is less than or equal to 665 gpm.
| |
| : 1. By performing a flow test, during shutdown, following completion of modifications to the RHR subsystems that alter the subsystem flow characteristics and verifying that the RHR pump lines, with a single pump running:
| |
| : 1) The sum of the injection line flow rates is greater than or equal to 3800 gpm, and
| |
| : 2) The total pump flow rate is less than or equal to 5500 gpm.
| |
| O WOLF CREEK - UNIT 1 3/4 5-6
| |
| | |
| (3 EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350 F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
| |
| : a. One OPERABLE centrifugal charging pump,
| |
| : b. One OPERABLE RHR heat exchanger,
| |
| : c. One OPERABLE RHR pump, and
| |
| : d. An OPERABLE flow path capable of taking suction from the RWST upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
| |
| APPLICABILITY: MODE 4.
| |
| ACTION:
| |
| V a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the RWST, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.
| |
| : b. With no ECCS subsystem OPERABLE because of the inoperability of either the RHR heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T,yg less than 350 F by use of alternate heat removal methods.
| |
| : c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for
| |
| ; each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
| |
| 4 O
| |
| ! WOLF CREEK - UNIT 1 3/4 5-7 4 ,- -
| |
| , e, - - - -
| |
| .-r- -na- - - - . . - -.- , - - , , - , - - - - - , , - - - - , - - - - - - - - - - - - - - - - - -
| |
| | |
| EMERGENCY CORE COGLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.
| |
| 4.5.3.2 All centrifugal charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps, shall be demonstrated inoperable
| |
| * by verifying that the motor circuit breakers are secured in the open position within 4 hours after entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325 F whichever comes first, and at least once per 31 days thereafter.
| |
| O
| |
| *An inoperable pump may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
| |
| O WOLF CREEK - UNIT 1 3/4 5-8
| |
| | |
| ' EMERGENCY CORE' COOLING SYSTEMS 3/4.5.4 ECCS SUBSYSTEMS - T,yg < 200*F
| |
| . LIMITING CONDITION FOR 0PERATION 3.5.4 'All Safety Injection pumps shall be inoperable.
| |
| APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on.
| |
| -ACTION:
| |
| _ With a Safety Injection pump OPERABLE,~ restore all Safety Injection pumps to
| |
| .an_ inoperable status within 4 hours.
| |
| : SURVEILLANCE REQUIREMENTS
| |
| ~
| |
| L.4.5.4' 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.
| |
| *An inoperable pump may be 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 3/4 5-9
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:
| |
| : a. A minimum contained borated water volume of 394,000 gallons,
| |
| : b. A boron concentration of between 2000 and 2100 ppm of boron,
| |
| : c. A minimum solution temperature of 37 F, and
| |
| : d. A maximum solution temperature of 100 F.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstraiad OPERABLE:
| |
| : a. At least once per 7 aays by:
| |
| : 1) Verifying the contained borated water volume in the tank, and
| |
| : 2) Verifying the boron concentration of the wate'r.
| |
| : b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 37 F or greater than 100 F.
| |
| O WOLF CREEK - UNIT 1 3/4 5-10
| |
| | |
| .n 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION: ,
| |
| Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour or be in at le SHUTDOWN within the f'gst HOT ollowing STANDBY within the next 6 hours and in COLD 30 hours.
| |
| SURVEILLANCE REQUIREMENTS tO V
| |
| -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 0PERABLE 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 Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L,.
| |
| *Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more O often than once per 92 days.
| |
| WOLF CREEK - UNIT 1 3/4 6-1
| |
| | |
| l l
| |
| l CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE l
| |
| LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:
| |
| : a. An overall integrated leakage rate of: i
| |
| : 1) Less than or equal to L a, 0.20% by weight of the containment I air per 24 hours at P,, 48 psig, or
| |
| : 2) Less than or equal to L t, 0.020% by weight of the containment air per 24 hours at Pt , 24 psig.
| |
| b.
| |
| A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P3 , 48 psig.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With either the measured overall integrated containment leakage rate exceeding 0.75 La or 0.75 L ,tas applicable, 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 integrated leakage rate to less than 0.75 L r less than a
| |
| Lt , as applicable, and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200 F.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fled in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:
| |
| : a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 1 10 month intervals during shutdown at a pressure not less than either P,, 48 psig, or P , 24 psig, during each 10 year t
| |
| service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection; WOLF CREEK - UNIT 1 3/4 6-2
| |
| | |
| R.
| |
| J E s
| |
| \ : CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1 .
| |
| : b. If any periodic Type A test fails to meet either.0.75 L, 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.
| |
| ? 5 . meet either 0.75 L, or 0.75 Lt, a Type A test shall be performed at
| |
| -least every 18 months until two consecutive Type A tests meet either
| |
| -0.75 L, or 0.75 Lt at which time the above test schedule may be resumed;
| |
| ~
| |
| : c. The accuracy of each Type A test shall be verified by a supplemental !
| |
| test which:
| |
| J l-
| |
| : 1) Confirms the accuracy of the test by verifying that the supple- ;
| |
| j.
| |
| mental test result, L , cminus the sum of the Type A and the super-j' imposed leak, L,, is equal to or less than 0.25 L, or 0.25 L t3 I 2) Has a duration sufficient to establish accurately the change in leakage rate between the' Type A test and the supplemental test; and i
| |
| l 3) Requires that the rate at which gas is injected into the contain- i l j ment or bled from the containment during the supplemental test is
| |
| {
| |
| 's between 0.75 L, and 1.25 L, or 0.75 Lt and 1.25 Lt*
| |
| 'i '
| |
| : d.. Type B and C tests shall be conducted with. gas at a pressure not less 1
| |
| : j. than P,, 48 psig, at intervals no greater than 24 months except for ;
| |
| I tests involving:
| |
| !. 1) Air locks, F :
| |
| I
| |
| : 2) Purge supply and exhaust isolation valves with resilient material seals, and l
| |
| i 3)- Valves pressurized with fluid from a seal system.
| |
| : e. Air locks shall be tested and demonstrated OPERABLE by the requirements i of Specification 4.6.'1.3; i f. Purge' supply and exhaust isolation valves with resilient material i i seals shall be tested and demonstrated OPERABLE by the requirements I
| |
| of Specification 4.6.1.7.2'and 4.6.1.7.4, as applicable;
| |
| : g. Leakage from isolation valves that are sealed with fluid from a seal (
| |
| i system may be excluded, subject to the provisions of Appendix J, l
| |
| Section III.C.3, when determining the combined leakage rate provided
| |
| ' the seal system and valves are pressurized to at least 1.10 P
| |
| : (53psig),andthesealsystemcapacityisadequatetomaintain
| |
| \ system pressure for at least 30 days; and
| |
| : h. The provisions of Specification 4.0.2 are not applicable.
| |
| ; WOLF CREEK - UNIT 1 3/4 6-3 i
| |
| f.
| |
| | |
| CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:
| |
| : a. Both doors closed except when the air lock is being used for normal transit entry and exits through the containment, then at least one air lock door shall be closed, and
| |
| : b. An overall air lock leakage rate of less than or equal to 0.05 L a at P , 48 psig.
| |
| a APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| : a. With one containment air lock door inoperable:
| |
| : 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed,
| |
| : 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days,
| |
| : 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and
| |
| : 4. The provisions of Specification 3.0.4 are not applicable.
| |
| : b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperr'le air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| O WOLF CREEK - UNIT 1 3/4 6-4
| |
| | |
| -p 4 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
| |
| : a. Within 72 hours following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying that the seal leakage is less than 0.005 L, as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at e constant pressure of greater than or equal to 10 psig;
| |
| : b. By conducting overall air lock leakage tests at not less than P,,
| |
| 48 psig, and verifying the overall air lock leakage rate is within its limit:
| |
| : 1) At least once per 6 months,# and
| |
| : 2) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.*
| |
| : c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
| |
| The provisions of Specification 4.0.2 are not applicable.
| |
| *This represents an exemption to Appendix J, Paragraph III.D.2.(b)(ii), of 10 CFR Part 50.
| |
| WOLF CREEK - UNIT 1 3/4 6-5
| |
| | |
| CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between
| |
| +1.5 and -0.3 psig.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| O SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours.
| |
| O WOLF CREEK - UNIT 1 3/4 6-6
| |
| | |
| /~) CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120*F.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the containment average air temperature greater than 120*F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at the following locations and shall be determined at least once per 24 hours:
| |
| Location
| |
| : a. Containment Cooler Inlet located near NNE wall (El 2068'-8"),
| |
| : b. Containment Cooler Inlet located near West wall (El 2068'-8"),
| |
| : c. Containment Cocler Inlet located near NNW wall (El 2068'-8"), and
| |
| : d. Containment Cooler Inlet located near East wall (El 2068'-8").
| |
| WOLF CREEK - UNIT 1 3/4 6-7
| |
| | |
| CONTAINMENT SYSTEMS CONTAINMENI VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| : a. With more than one tendon with an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit or with one tendon below 90% of the predicted lower limit, restore the tendon (s) to the required level of integrity 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 Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : b. With any abnormal degradation of the structural integrity other than ACTION a. at a level below the acceptance criteria of Specifica-tion 4.6.1.6, restore the containment vessel to the required level of integrity within 72 hours and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Tendons. The containment vessel tendons' struc-tural integrity shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at 5 year intervals thereafter. (he tendons' structural integrity 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 predicted limits for each. For each subsequent inspec-tion one tendon from each group may be kept unchanged to develop a history and to correlate the observed data. If the observed lif t-off force of any one tendon in the original sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon should be checked for their lift off forces. If both of these adjacent tendons are found to be within their predicted limits, all three tendons should be restored to the required level of integrity. This single deficiency may be conside, red unique and acceptable. Unless there is abnormal degradation of the containment vessel during the first three inspec-tions, the sample population for subsequent inspections shall include at least 6 tendons (3 inverted U and 3 hoop);
| |
| WOLF CREEK - UNIT 1 3/4 6-8
| |
| | |
| ) CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (inverted U and hoop). A randomly selected tendon from each group shall be completely deten-sioned in order to identify broken or damaged wires and determining that over the entire length of the removed wire that:
| |
| : 1) The tendon wires are free of corrosion, cracks, and damage,
| |
| : 2) There are no changes in the presence or physical appearance of the sheathing filler grease, and
| |
| : 3) A minimum tensile strength of 240,000 psi (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 of any one of the wire samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment vessel structure.
| |
| : c. Performing tendon retensioning of those tendons detensioned for i i inspection to their observed lift-off force with a tolerance limit of +6%. During retensioning of these tendons, the changes in load and elongation should 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 5% from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages;
| |
| : d. Assuring the observed lif t-off stresses adjusted to account for elastic losses exceed the average minimum design value given below:
| |
| Inverted U 139 ksi Hoop: Cylinder 147 ksi Dome 134 ksi
| |
| : e. Verifying the OPERABILITY of the sheathing filler grease by assuring:
| |
| : 1) No voids in excess of 5% of the net duct volume,
| |
| : 2) Minimum grease coverage exists for the different parts of the anchorage system, and
| |
| : 3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer, g
| |
| v
| |
| )
| |
| WOLF CREEK - UNIT 1 3/4 6-9
| |
| | |
| CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces. The structural integrity of the end anchorages of all tendons inspected pursuant to Specifi-cation 4.6.1.6.1 and the adjacent concrete surfaces shall be demonstrated by determining through inspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjacent to the end anchorages. Inspections of the concrete shall be performed during the Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment vessel is at its maximum test pressure.
| |
| 4.6.1.6.3 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, sr.all be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 6-10
| |
| | |
| A I CONTAINMENT SYSTEMS b
| |
| CONTAINMENT VENTILATION SYSTEM LIMITING' CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valves shall be OPERABLE and:
| |
| : a. Each 36-inch containment shutdown purge supply and exhaust isolation valve shall be closed and blank flanged, and
| |
| : b. The 18-inch containment mini purge supply and exhaust isolation valve (s) may be open for up to 2000 hours during a calendar year.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| : a. With a 36-inch containment purge supply and/or exhaust isolation valve open or not blank flanged, close and/or blank flange that valve or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
| |
| : b. With the 18-inch containment mini purge supply and/or exhaust isolation valve (s) open for more than 2000 hours during a calendar year, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
| |
| : c. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.2 and/or 4.6.1.7.4, restore the inoperable valve (s) to OPERABLE status within 24 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
| |
| \
| |
| WOLF CREEK - UNIT 1 3/4 6-11
| |
| | |
| 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 pressurized 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-mined pursuant to Specification 4.6.1.2d., for all other Type B and C penetra-tions, the combined leakage rate is less than 0.60 L " a 4.6.1.7.3 The cumulative time that all 18-inch containment mini purge supply and/or exhaust isolation valves have been open during a calendar year shall be 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 La when pressurized to Pa '
| |
| *Except valves and flanges which are located inside containment. These valves shall be verified to be closed with their blank flanges installed prior to entry into MODE 4 following each COLD SHUTOOWN.
| |
| O WOLF CREEK - UNIT 1 3/4 6-12
| |
| | |
| l l
| |
| j N
| |
| '(
| |
| V) CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Containment Spray System capable of taking suction from the RWST and transfer-ring suction to the containment sump.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With one Containment Spray System inoperable, restore the inoperable Contain-ment Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Containment Spray Sys-tem to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.-
| |
| \ ,
| |
| )
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:
| |
| : a. At least once per 3: days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
| |
| : b. By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 250 psig when tested pursuant to Specification 4.0.5;
| |
| : c. At least once per 18 months during shutdown, by:
| |
| : 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-High-3 (CSAS) test signal, and i
| |
| : 2) Verifying that each spray pump starts automatically on a Containment Pressure-High-3 (CSAS) test signal,
| |
| : d. At least once per 5 years by performing an air or smoke flow test Q through each spray header and verifying each spray nozzle is unobstructed.
| |
| WOLF CREEK - UNIT 1 3/4 6-13
| |
| | |
| CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:
| |
| : a. A spray additive tank containing a volume of between 4340 and 4540 gallons of between 28 and 31% by weight Na0H solution, and
| |
| : b. Two spray additive eductors each capable of adding Na0H solution from the chemical additive tank to a Containment Spray System pump flow. .
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTD0WN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
| |
| : b. At least once per 6 months by:
| |
| : 1) Verifying the contained solution volume in the tank, and
| |
| : 2) Verifying the concentration of the NaOH solution by chemical analysis,
| |
| : c. 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 Containment Pressure-High-3 (CSAS) test signal; and
| |
| : d. At least once por 5 years by verifying:
| |
| : 1) Each eductor flow rate is greater than or equal to 52 gpm using the RWST as the test source throttled to 17 psig at eductor inlet, and
| |
| : 2) The lines between the spray additive tank and the eductors are not blocked by verifying flow.
| |
| WOLF CREEK - UNIT 1 3/4 6-14
| |
| | |
| Y V CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 Two independent groups of containment cooling fans shall be OPERABLE with two fan systems to each group.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| : a. With one group of the above required containment cooling fans inoperable and both Containment Spray Systems OPERABLE, restore the inoperable group of cooling fans to OPER~ABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
| |
| : b. With two groups of the above required containment cooling fans inoperable and both Containment Spray Systems OPERABLE, restore at least one group of cooling fans to OPERABLE status within 72 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore both above required
| |
| . groups of cooling fans to OPERABLE status within 7 days of initial N loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : c. With one group of the above required containment cooling fans inoperable and one Containment Spray System inoperable, restore the inoperable Containment Spray System to OPERABLE status within 72 hours or be in at least HOT STAN06f within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore the inoperable group of containment cooling fans to OPERABLE status within 7 days of ini-tial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.3 Each group of containment cooling fans shall be demonstrated OfERABLE:
| |
| : a. At least once per 31 days by:
| |
| : 1) Starting each non-operating fan group from the control room, and verifying that each fan group operates for at least 15 minutes, and
| |
| : 2) Verifying a cooling water flow rate of greater than or equal to 2200 gpm to each cooler group,
| |
| : b. At least once per 18 months by verifying that on a Safety Injection y test signal, the fans start in slow speed or, if operating, shift to slow speed and the cooling water flow rate increases to at least 4000 gpm to each cooler group.
| |
| WOLF CREEK - UNIT 1 3/4 6-15
| |
| | |
| CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With one or more of the 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 hours, or
| |
| : b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
| |
| : c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange, or
| |
| : d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS ,
| |
| 4.6.3.1 The containment isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to sersico af ter 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.
| |
| O WOLF CREEK - UNIT 1 3/4 6-16
| |
| | |
| CONTAINMENT SYSTEMS v
| |
| SURVEILLANCE REQUIREMENTS (Continued) 1 4.6.3.2 Each containment iso',ation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUT 00WN or REFUELING MODE at least once !
| |
| per 18 months by:
| |
| ; a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position; '
| |
| 4
| |
| ; b. Verifying that on a Phase "B" Isolation test signal, each i
| |
| { Phase "B" isolation valve actuates to its isolation position; and ;
| |
| : c. Verifying that on a Containment Purge Isolation test signal, each purge supply and exhaust isolation valve actuates to its
| |
| ! isolation position.
| |
| 4.6.3.3 The isolation time of each power operated or automatic valve of ,
| |
| Table 3.6-1 shall be determined to be within its limit when tested pursuant to i Specification 4.0.5.
| |
| O l
| |
| l
| |
| }
| |
| I i
| |
| l l
| |
| l j !
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| i WOLF CREEK - UNIT 1 3/4 6-17
| |
| | |
| TABLE 3.6-1 CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 1. Phase "A" Isolation (active)
| |
| P-62 BB HV-8026** PRT Nitrogen C 10 Iso Valve P-62 B3 HV-8027** PRT Nitrogen C 10 Iso Valve BG HV-8100 Seal Water Return C 10 P-24 CTMT Iso Valve P-24 BG HV-8112 Seal Water Return C 10 CTMT Iso Valve BG HV-8152 Letdown System CTMT C 10 P-23 Iso Valve BG HV-8160 Letdown System CTMT C 10 P-23 Iso Valve BL HV-8047 Reactor Makeup Water C 10 P-25 CTMT Iso Valve EJ HCV-8825** RHR to SI Test Line A 10 P-21 Iso Valve EJ HCV-8890A** RHR A to SI Pumps Test A 13 P-82 Line Iso Valve EJ HCV-8890B** RHR B to SI Pumps Test A 13 P-27 Line Iso Valve P-49 EH HV-8823** SI/AccumulatorInjection A 10 Test Line Iso Valve Safety Injection Pump B A 10 P-48 EM HV-8824**
| |
| Test Line Iso Valve ,
| |
| CCThe provisions of Specification 3.0.4 are not applicable.
| |
| O WOLF CREEK - UNIT 1 3/4 6-18
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 1. Phase "A" Isolation (active) - (Continued)
| |
| P-88 EM HV-8843** Boron Injection Up- A 10 stream Test Line Iso P-92 EM HV-8871** SI Test Line to RWST C 10 Iso Valve P-87 EM HV-8881** Safety Injection Pump A 10 Test Line Iso Valve P-92 EM HV-8964** SI Test Line System C 10 Outside CTMT Iso P-99 GS HV-3 Hydrogen Analyzer B A,C 5 Inlet Iso P -99 GS HV-4 Hydrogen Analyzer B A,C 5 O P-99 GS HV-5 Inlet Iso Hydrogen Analyzer B A,C 5 Inlet Iso P-56 GS HV-8 Hydrogen Analyzer B A,C 5 Disch Iso P-56 GS HV-9 Hydregen Analyzer B A,C $
| |
| Diset. Iso P-101 GS HV-12 Hydrogen Analyzer A A,C 5 Inlet Iso P-101 GS HV-13 Hydrogen Analyzer A A,C 5 Inlet Iso P-101 GS HV-14 Hydrogen Analyzer A A,C 5 Inlet Iso P-97 GS HV-17 Hydrogen Analyzer A AC 5 Disch Iso P 97 GS HV 18 Hydrogen Analyzer A A,C 5 Disch Iso
| |
| **The provisions of Spectfication 3.0.4 are not applicable.
| |
| WOLF CREEK - UNIT 1 3/4 6-19
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 1. Phase "A" Isolation (active) - (Continued)
| |
| P-101 GS HV-31 Sample Line to CTMT A,C 5 Atmos Monitor P-101 GS HV-32 Sample Line to CTMT A,C 5 Atmos Monitor P-97 GS HV-33 Hydrogen Sample Return A,C 5 From PASS P-97 GS HV-34 Hydrogen Sample Return A,C 5 From PASS P-99 GS HV-36 Sample Line to CTHT A,C 5 Atmos Monitor P-99 GS HV-37 Sample Line to CTMT A,C 5 Atmos Monitor P-56 GS HV-38 Sample Return CTHT A,C 5 Atmos Monitor P-56 GS HV-39 Sample Return CTMT A,C 5 Atmos Monitor P-44 HB HV-7126 RCOT Vent Inside CTMT C 10 P-26 HB HV-7136 RCOT Pumps Olsch Hdr C 10 Outside CTHT Iso P-44 HB HV-7150 RCOT Vent Outside C 10 CTMT P-26 HB HV 7176 RCOT Pumps Olsch Hrd C 10 Inside CTMT 150 P-30 KA FV-29 Reactor B1dg Instr Air C 5 Supply Outside CTHT Iso P-32 LF FV-95 CTMT Normal Sumps to C 30 Floor Drain Tank Inside CTHT Iso WOLF CREEK - UNIT 1 3/4 6-20
| |
| | |
| im
| |
| ( ) TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 1. Phase "A" Isolation (active) - (Continued)
| |
| P-32 LF FV-96 CTMT Normal Sumps to C 4 Floor Drain Tank Out-side CTMT Iso P-93 SJ HV-5** PZR/RCS Liquid Sample C 5 Inner CTMT Iso P-93 SJ HV-6** PZR/RCS Liquid Sample C 5 Outer CTMT Iso P-69 SJ HV-12** PZR Vapor Sample Inner C 5 CTMT Iso P-69 SJ HV-13** PZR Vapor Sample Outer C 5 CTMT Iso P-95 SJ HV-18** Accumulator Sar.ple C 5 C 's Inner CTMT Iso LId P-95 SJ HV-19** Accumulator Sample C 5 Outer CTFT Iso p-93 SJ HV-127** PZR/RCS Liquid Sample C 5 Outer CTMT Iso P-64 SJ HV-128** PZR/RCS Liquid Sample A,C 5 Inner CTMT Iso P-64 SJ HV-129** PZR/RCS Liquid Sample A,C 5 Outer CTMT Iso P-64 SJ HV-130** PZR/RCS Liquid Sample A,C 5 Outer CTMT Iso Valve P-57 SJ HV-131** PASS Olscharge to RCDT A,C 5 P-57 SJ HV-132** PASS Olscharge to RCOT A,C 5
| |
| : 2. Phase "A" Isolation (passive)*
| |
| t 58 EM HV-8888** Accumulator Tank F111 C 5 Line Iso Valve p S May be opened on an intermittent basis under administrative control.
| |
| **The provisions of Specification 3.0.4 are not appIlcable.
| |
| WOLF CREEK - UNIT 1 3/4 6-21
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 2. Phase "A" Isolation (passive)* - (Continued)
| |
| P-16 EN HV-Ol** CTMT Recirc Sump to CTMT A 30 Spray Pump A Iso P-13 EN HV-07** CTMT Recirc Sump to CTMT A 30 Spray Pump B Iso P-45 EP HV-8880** CTMT Nitrogen Supply C 10 Iso Valve P-65 GS HV-20** Hydrogen Purge Inner C 5 CTMT Iso P-65 GS HV-21** Hydrogen Purge Outer C 5 CTMT Iso P-67 KC HV-253** Fire Protection System C 30 Hdr Outer CTMT Iso
| |
| : 3. Phase "B" Isolation (active)
| |
| P-74 EG HV-58 CCW to RCS Iso C 30 P-75 EG HV-59 CCW Return From C 30 RCS Iso P-75 EG HV-60 CCW Return From C 30 RCS Iso P-76 EG HV-61 CCW Return From C 30 RCS Iso P-76 EG HV-62 CCW Return From C 30 RCS Iso
| |
| : 4. Containment Purge Isolation (active)
| |
| V-161 GT HZ-4*** CTMT Mini-Purge C 3 Supply Outside CTMT Iso V-161 GT HZ-5*** CTHT Mini-Purge C 3 Supply inside CTMT 150
| |
| *May be opened on an intermittent basis under administrative control.
| |
| C2The pruvisions of Specification 3.0.4 are not applicabic.
| |
| * 2The provisions of Specification 3.0.4 are not applicable provided the penetration is isolated by two passive devices.
| |
| WOLF CREEK - UN!i 1 3/4 6-22
| |
| | |
| n TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM ,
| |
| TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 4. Containment Purge Isolation (active) - (Continued)
| |
| V-160 GT HZ-11*** CTMT Mini-Purge C 3 Exh Inside CTMT Iso V-160 GT HZ-12*** CTMT Mint-Purge C 3 Exh Outside CTMT Iso
| |
| : 5. Containment Purge Isolation (passive)
| |
| V-161 GT HZ-6*** CTMT S/D Purge C 10 Supply Outside CTMT Iso V-161 GT HZ-7*** CTMT S/0 Purge C 10 Supply Inside CTMT Iso V-160 GT HZ-8*** CTMT S/D Purge C 10 Exh Inside CTMT Iso V-160 GT HZ-9*** CTMT S/D Purge C 10 Exh Outside CTMT Iso
| |
| : 6. Remote Manual P-41 BB HV-8351A RCP A Seal Water C N.A.
| |
| Supply P-22 BB HV-83518 RCP B Seal Water C N.A.
| |
| Supply P-39 BB HV 8351C RCP C Seal Water C N.A.
| |
| Supply P-40 BB HV 83510 RCP O Seal Water C N.A.
| |
| Supply P-79 BB PV-8702A RCS Hot Leg 1 to RHR A N.A.
| |
| Pump A Suction
| |
| ***The provisions of Specification 3.0.4 are not applicable provided the penetration is isolated by two passive devices.
| |
| WOLF CREEK - UNIT 1 3/4 6-23
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 6. Remote Manual - (Continued)
| |
| P-52 BB PV-87028 RCS Hot Leg 4 to RHR A N.A.
| |
| Pump B Suction P-15 EJ HV-23** PASS Sump Sample CTMT C 5 Iso P-15 EJ HV-25** PASS Sump Sample CTMT C 5 Iso P-14 EJ HV-24** PASS Sump Sample CTMT C 5 Iso P-14 EJ HV-2G** PASS Sump Sample CTMT C 5 Iso P-71 EF HV-31 ESW Supply To C N.A.
| |
| Containment Coolers P-28 EF HV-32 ESW Supply To C N.A.
| |
| Containment Coolers P-71 EF HV-33 ESW Supply To C N.A.
| |
| Containment Coolers
| |
| 'P-28 Er HV-34 ESW Supply To C N.A.
| |
| Containment Coolers P-73 EF HV-45 ESW Return From C N.A.
| |
| Containment Coolers P-29 EF HV-46 ESW Return From C N.A.
| |
| Containment Coolers P-73 EF HV-41 ESW Return From C N.A.
| |
| Containment Coolers C
| |
| The provisions of Specification 3.0.4 are not applicable.
| |
| WOLF CREEK - UNIT 1 3/4 6-24
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES ,
| |
| MAXIMUM TYPE LEAK- ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED. (Seconds)
| |
| : 6. Remote Manual - (Continued)
| |
| P-29 EF HV-48 ESW Return From C N.A.
| |
| : Containment
| |
| ] Coolers i P-73 EF HV-49 ESW Return From C N.A.
| |
| i Containment Coolers P-29 EF HV-50 ESW Return From C N.A.
| |
| l Containment l t
| |
| Coolers l.
| |
| P-74 EG HV-127* CCW Supply to RCP C N.A.
| |
| 1
| |
| -P-75 EG HV-130* CCW Return From RCP C N.A. !
| |
| 'P-75 EG HV-131* CCW Return From RCP C N.A.
| |
| P-76 EG HV-132* CCW Return From RCP C N.A.
| |
| Thermal Barriers t
| |
| ~
| |
| P-76 EG HV-133* CCW Return from RCP C N.A.
| |
| : Thermal Barrier
| |
| *P-79 EJ HV-8701A RCS Hot Leg 1 to RHR A N.A.
| |
| Pump A Suction .
| |
| 1 a- P-52 EJ HV-8701B RCS Hot Leg 4 to RHR A N.A.
| |
| ! Pump B Suction P-82 EJ HV-8809A RHR Pump A Cold Leg A N.A.
| |
| Injection Iso Valve P-27 EJ HV-88098 RHR Pump B Cold Leg A N.A. i
| |
| ! Injection Iso Valve
| |
| } P-15 EJ HV-8811A CTMT Recirc Sump to A N.A. !
| |
| j RHR Pump A Suction i,
| |
| *These valves were assumed to be closed during the accident analysis, and are I
| |
| ; normally closed but may'be opened on an intermittent basis under administrative
| |
| , control.
| |
| l WOLF CREEK - UNIT 1 3/4 6-25 1
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 6. Remote Manual - (Continued)
| |
| P-14 EJ HV-8811B CTMT Recirc Sump to A N. A.
| |
| RHR Pump B Suction P-21 EJ HV-8840 RHR Hot Leg Recirc A N.A.
| |
| Iso Valve P-87 EM HV-8802A* SI Pump A Disch Hot A N.A.
| |
| Leg Iso Valve P-48 EM HV-88028* SI Pump B Disch Hot A N. A.
| |
| Leg Iso Valve P-49 EM HV-8835 SI Pumps Disch to A N.A.
| |
| Cold Leg Iso Valve P-89 EN HV-6 CTMT Spray Pump A A N.A.
| |
| Discharge Iso Valve P-66 EN HV-12 CTMT Spray Pump B A N.A.
| |
| Discharge Iso Valve
| |
| : 7. Active for SIS P-80 BG HV-8105 CVCS Charging Line C 10 P-88 EM HV-8801A Boron Injection to A N.A.
| |
| RCS Cold Legs P-88 EM HV-8801B Baron Injection to A N.A.
| |
| RCS Cold Legs
| |
| : 8. Hand-Operated and Check Valves P-41 BB V-118 RCP A Seal C N.A.
| |
| Water Supply P-22 BB V-148 RCP B Seal C N.A.
| |
| Water Supply P-39 BB V-178 RCP C Seal C N.A.
| |
| Water Supply P-40 BB V-208 RCP D Seal C N.A.
| |
| Water Supply Cihese valves were assun,ed to be closed during the accident analysis and are normally closed but may be opened on an intermittent basis under administrative control.
| |
| WOLF CREEK - UNIT 1 3/4 6-26
| |
| | |
| i 5
| |
| 77- . .
| |
| TABLE 3.6-1 (Continued) 4
| |
| . CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ~ ISOLATION TIME PENETRATIONS'. VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 8. . Hand-0perated and Check Valves - (Continued) i P-24 BG V-135 RCP Seal Water Return C N.A.
| |
| ! P-80 BG 8381 CVCS Charging Line C N.A. -
| |
| P-25 BL 8046- Reactor Makeup Water C - N.A.
| |
| Supply P-78 - BM V-045 Steam Generator Drain C N.A.
| |
| Line Iso Valve t
| |
| P-78 BM V-046 Steam Generator Drain C N.A.
| |
| Line Iso Valve i
| |
| ^
| |
| = P . EC V-083 Refueling Pool Supply C N.A.
| |
| j
| |
| ~
| |
| From Fuel Pool C1eanup ,
| |
| (
| |
| P-53 EC V-084- Refueling Pool Supply C N.A.
| |
| From Fuel Pool Cleanup P-54 EC V-087 ' Refueling Pool C N.A.
| |
| ;- Return to Fuel l- ,
| |
| Pool Cooling
| |
| - P-54 EC V-088 Refueling Pool C N.A.
| |
| Return to Fuel Pool Cooling P EC V-095 Refueling Pool C N.A.
| |
| Skimmers To Fuel Pool Cooling Loop P-55 EC V-096 Refueling Pool C N.A.
| |
| Skimmers To Fuel Pool Cooling Loop l .
| |
| P-74 EG V-204 CCW Supp!y to RCP C N.A.
| |
| t-P-82 EP 8818A RHR Pump to Cold A N.A.
| |
| i Leg 1 Injection EP 88188 RHR Pump to Cold A N.A.
| |
| P-82
| |
| ( Leg 2 Injecticn
| |
| - - y
| |
| . WOLF CREEK - UNIT 1- 3/4 6-27
| |
| ---.-,_..,-...-....-w--,-.-.---r-,n,,-e- ., ,.,_---3 ,,%.,-,e . . . -
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 8. Hand-0perated and Check Valves - (Continued)
| |
| P-27 EP 8818C RHR Pump to Cold A N.A.
| |
| Leg 3 Injection P-27 EP 8818D RHR Pump to Cold A N.A.
| |
| Leg 4 Injection P-21 EJ 8841A RHR Pump Disch to A N.A.
| |
| RCS Hot Leg 2 P-21 EJ 88418 RHR Pump Disch to A N.A.
| |
| RCS Hot Leg 3 P-87 EM V-001 SI Pump Hot Leg 1 A N.A.
| |
| Injection P-87 EM V-002 SI Pump Hot Leg 2 A N.A.
| |
| Injection P-48 EM V-003 SI Pump Hot Leg 3 A N.A.
| |
| Injection P-48 EM V-004 SI Pump Hot Leg 4 A N.A.
| |
| Injection P-58 EM V-006 Accumulator Fill Line C N.A.
| |
| From SI Pumps P-49 EP V-010 SI Pump Disch to Cold A N.A.
| |
| Leg 1 P-49 EP V-020 SI Pump Disch to Cold A N.A.
| |
| Leg 2 P-49 EP V-030 SI Pump Disch to Cold A N.A.
| |
| Leg 3 P-49 EP V-040 SI Pump Disch to Cold A N.A.
| |
| Leg 4 P-88 EM V-8815 BIT to RCS Cold Leg A N.A.
| |
| Injection P-89 EN V-013 CTMT Spray Pump A A N.A.
| |
| to CTMT Spray Nozzles WOLF CREEK - UNIT 1 3/4 6-28
| |
| | |
| TABLE 3.6-1 (Continued)
| |
| C0N'TAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIMI PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)
| |
| : 8. _ Hand-Operated and Check Valves - (Continued)
| |
| P-66 EN V-017 CTMT Spray Pump B A N.A.
| |
| to CTMT Spray Nozzles P-45 EP V-046 Accumulator Nitrogen C N.A.
| |
| Supply Line P-43 HD V-016 Auxiliary Steam to C N.A.
| |
| Decon System P-43 HD V-017 Auxiliary Steam to C N.A.
| |
| Decon System P-63 KA V-039 Rx Bldg Service Air C N.A.
| |
| Supply
| |
| \s -
| |
| P-63 KA V-118 Rx Bldg Service Air C N.A.
| |
| Supply P-98 KB V-001 Breathing Air Supply. C N.A.
| |
| to RX Bldg P-98 KB V-002 Breathing Air Supply C N.A.
| |
| to RX Bldg P-30 KA V-204 Rx Bldg Instrument C N.A.
| |
| Air Supply P-67 KC V-478 Fire Protection C N.A.
| |
| Supply to RX Bldg P-57 SJ V-111 Liquid Sample from A,C N.A.
| |
| PASS to RCDT d
| |
| a J
| |
| WOLF CREEK - UNIT 1 3/4 6-29
| |
| | |
| 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 H.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 Q*The provisions of Specification 3.0.4 are not applicable.
| |
| QQQThese valves are included for table completeness. The requirements of Specification 3.6.3 do not apply; instead, the requirements of Specification 3.7.1.5 and Specification 3.3.2 apply to the Main Steam Isolation Valves and Main Feedwater Isolation Valves, respectively.
| |
| WOLF CREE % - UNIT 1 3/4 6-30
| |
| | |
| J (
| |
| CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL ,
| |
| HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION l
| |
| I 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE.
| |
| l l APPLICABILITY: MODES 1 and 2.
| |
| ACTION:
| |
| i
| |
| : a. With one containment hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.
| |
| : b. With both hydrogen analyzers inoperable, restore at least one analyzer to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours.
| |
| l SURVEILLANCE REQUIREMENTS 4.6.4.1 Each containment hydrogen analyzer shall be demonstrated OPERABLE by the performance 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 CALIBRATION using sample gas containing ten volume percent hydrogen, balance nitrogen.
| |
| l I
| |
| l l
| |
| l
| |
| ; WOLF CREEK - UNIT 1 3/4 6-31
| |
| | |
| CONTAINMENT SYSTEMS HYDROGEN CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.4.2 A Hydrogen Control System shall be OPERABLE with two independent Hydrogen Recombiner Systems.
| |
| APPLICABILITY: MODES 1 and 2 ACTION:
| |
| With one of the two independent Hydrogen Recombiner Systems inoperable, restore the inoperable Hydrogen Recombiner System to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:
| |
| : a. At least once per 6 months by verifying, during a Hydrogen Recombiner System functional test, that the heater air temperature increases to greater than or equal to 1150 F within 5 hours; and
| |
| : b. At least once per 18 months by:
| |
| : 1) Performing a CHANNEL CALIBRATION of all hydrogen recombiner system instrumentation and control circuits,
| |
| : 2) Verifying through a visual examination that there is no evidence of abnormal conditions within the hydrogen recombiner system enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and
| |
| : 3) Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
| |
| O WOLF CREEK - UNIT 1 3/4 6-32
| |
| | |
| 3/4.7 PLANT SYSTEMS
| |
| -s 3/4.7.1 TURBINE CYCLE
| |
| (
| |
| SAFETY VALVES LIMITING CONDITION FOR OPERATION 1
| |
| 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| : a. With four reactor coolant loops and associated steam generators in
| |
| , operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hnurs, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is. reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : b. The provisions of Specification 3.0.4 are not applicable.
| |
| i SURVEILLANCE REQUIREMENTS 4
| |
| 4.7.1.1 No additional requirements other than those required by Specification 4.0.5.
| |
| , WOLF CREEK - UNIT 1 3/4 7-1
| |
| | |
| TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RAN3E SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (Percent of RATED THERMAL POWER) 1 87 2 65 3 44 O
| |
| O WOLF CREEK - UNIT 1 3/4 7-2
| |
| | |
| 1 4
| |
| TABLE 3.7-2
| |
| ,'' O STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (11%)* ORIFICE SIZE j
| |
| Loop 1 Loop 2 Loop 3 Loop 4
| |
| ; V055 V065 V075 V045 1185 psig 16.0 sq. in.
| |
| l V056 V066 V076 V046 1197 psig 16.0 sq. in.
| |
| ! V057 V067 V077 V047 1210 psig 16.0 sq. in. ,
| |
| V058 V068 V078 V048 1222 psig 16.0 sq. in, l
| |
| V059 V069 V079 V049 1234 psig 16.0 sq. in, i
| |
| i l
| |
| l 1
| |
| i i
| |
| The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressute.
| |
| I i
| |
| e t
| |
| ( i I
| |
| i WOLF CREEK - UNIT 1 3/4 7-3
| |
| | |
| PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
| |
| : a. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and
| |
| : b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours.
| |
| : b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| : c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
| |
| SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days on a STAGGERED TEST BASIS by:
| |
| : 1) Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1535 psig on recirculation flow when tested pursuant to Specification 4.0.5;
| |
| : 2) Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1625 psig at a flow of greater than or equal to 120 gpm when the secondary steam supply pressure is greater than 900 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; WOLF CREEK - UNIT 1 3/4 7-4
| |
| | |
| PLANT SYSTEMS 73
| |
| ! SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and
| |
| : 4) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER.
| |
| : b. At least once per 18 months during shutdown by:
| |
| : 1) Verifying that each automatic valve in the ESW supply to the auxiliary feedwater pumps actuates to its full open position upon receipt of an Auxiliary Feedwater Pump Suction Pressure-Low '
| |
| test signal,
| |
| : 2) Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal, and
| |
| : 3) Verifying that each auxiliary feedwater motor-operated discharge x valve limits the flow to each steam generator from the motor-driven pump to less than or equal to 320 gpm.
| |
| 4.7.1.2.2 An auxiliary feedwater flow path shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to at least two steam generators from one auxiliary feedwater pump.
| |
| i t
| |
| WOLF CREEK - UNIT 1 3/4 7-5
| |
| | |
| PLANT SYSTEMS CONDENSATE ORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a contained water volume of at least 281,000 gallons.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| With the CST inoperable, within 4 hours either:
| |
| : a. Restore the CST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHl'TDOWN within the following 6 hours, or
| |
| : b. Demonstrate the OPERABILITY of the Essential Service Water (ESW) Sys-tem as a backup supply to the auxiliary feedwater pumps and restore the CST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| SURVEILLANCE REQUIREMENTS O
| |
| 4.7.1.3.1 The CST shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.
| |
| 4.7.1.3.2 The ESW System shall be demonstrated OPERABLE at least once per 12 hours by verifying that the ESW System is in operation whenever the ESW System is the supply source for the auxiliary feedwater pumps.
| |
| O WOLF CREEK - UNIT 1 3/4 7-6
| |
| | |
| PLANT SYSTEMS
| |
| 'g SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie / gram DOSE EQUIVALENT I-131.
| |
| APPLICABILITY:. MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the specific activity of the Secondary Coolant System greater than 0.1 microcurie / pram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| O SURVEILLANCE REQUIREMENTS
| |
| ,O 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis
| |
| . program of Table 4.7-1.
| |
| (
| |
| WOLF CREEK - UNIT 1 3/4 7-7 l
| |
| | |
| TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY
| |
| : 1. Gross Radioactivity Determination At least once per 7?. hours.
| |
| : 2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radioactivity determination indicates concentrations greater than 10% of the allowable limit for radioiodines.
| |
| b) Once per 6 months, when-ever the gross radioactivity determination indicates concentrations less than or equal to 10% of the allowable limit for radioiodines.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 7-8
| |
| | |
| n PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSLIV) shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| MODE 1:
| |
| With one MSLIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| MODES 2 and 3:
| |
| With one MSLIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSLIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
| |
| WOLF CREEK - UNIT 1 3/4 7-9
| |
| | |
| PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 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 either coolant in the steam generator is greater than 200 psig.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied:
| |
| : a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and
| |
| : b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.
| |
| Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F.
| |
| SURVEILLANCE REQUIREMENTS 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 reactor or secondary coolant is less than 70*F.
| |
| O WOLF CREEK - UNIT 1 3/4 7-10 l l
| |
| | |
| PLANT SYSTEMS
| |
| / \ 3/4.7.3 COMPONENT COOLING WATER SYSTEM wj LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment
| |
| (~] that is not locked, sealed, or otherwise secured in position is in
| |
| ! its correct position. In addition, an ANALOG CHANNEL OPERATIONAL (d
| |
| TEST of the surge tank level and flow instrumentation which provide automatic isolation of the non-nuclear safety-related portion of the system shall be performed at least once per 31 days;
| |
| : b. At least once per 18 months during shutdown, by verifying that:
| |
| : 1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety-related portion of the system actuates to its correct position on a Safety Injection and on a simulated High Flow and Low Surge Tank Level test signal, and
| |
| : 2) Each OPERABLE Component Cooling Water System pump starts auto-matically on a Safety Injection and Loss-of-Power test signal,
| |
| : c. At least once per 18 months during shutdown, by performing a CHANNEL CALIBRATION of the surge tank level and flow instrumentation which provide automatic isolation of the non-nuclear safety-related portion of the system.
| |
| /
| |
| \ ]
| |
| 'O WOLF CREEK - UNIT 1 3/4 7-11
| |
| | |
| PLANT SYSTEMS 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent essential service water (ESW) loops shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With only one ESW loop OPERABLE, restore at least two ESW loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.7.4 At least two ESW loops shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is r.ot locked, sealed, or otherwise secured in position is in its correct position. In addition, at least once per 31 days, an ANALOG CHANNEL OPERATIONAL TEST of the differential pressure instrumentation for automatic isolation of the ESW to the air compressors shall be performed;
| |
| : b. At least once per 18 months during shutdown, by verifying that:
| |
| : 1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety-related portion of the system actuates to its correct position on a Loss-of-Power or Safety Injection test signal and on a simulated High Differential Pressure test signal, and
| |
| : 2) Each Essential Service Water System pump starts automatically on a Safety Injection, Low Suction Pressure (AFW pumps), and Loss-of-Power test signal,
| |
| : c. At least once per 18 months during shutdown, by performing a CHANNEL CALIBRATION of the differential pressure instrumentation for automatic isolation of ESW to the air compressors.
| |
| O WOLF CREEK - UNIT 1 3/4 7-12
| |
| | |
| PLANT SYSTEMS
| |
| /^hf i 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with:
| |
| : a. The crest of the UHS dam below the rip rap cover and corresponding water level at t,' above elevation 1070 Mean Sea Level, USGS datum, ,
| |
| and '
| |
| : b. The plant inlet tater temperature of less than or equal to 90*F.
| |
| L APPLICA8ILITY: MODES 1, 2 3, and 4.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| C'N SURVEILLANCE REQUIREMENTS 4.7.5 The UHS shall be determined OPERABLE:
| |
| : a. At least once per 24 hours by verifying the above required water temperature and water level to be within their limits, and
| |
| : b. At least once per 12 months by verifying that the crest of the UHS dam below the rip rap cover is at or above elevation 1070 Mean Sea Level, USGS datum.
| |
| WOLF CREEK - UNIT 1 3/4 7-13
| |
| | |
| PLANT SYSTEMS 3/4.7.6 CONIROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two independent Control Room Emergency Ventilation Systems shall be OPERABLE.
| |
| APPLICABILITY: All MODES.
| |
| ACTION:
| |
| MODES 1, 2, 3 and 4:
| |
| With one Control Room Emergency Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| MODES 5 and 6:
| |
| : a. With one Control Room Emergency Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Rooia Emergency Ventilation System in the recirculation mode,
| |
| : b. With both Control Room Emergency Ventilation Systems inoperable, or with the OPERABLE Control Room Emergency Ventilation System, required to be in the recirculation mode by ACTION a., not ccpable of being powered by an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
| |
| SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Emergency Ventilation System shall be demonstrated OPERABLE:
| |
| : a. At least once per 12 hours by verifying that the control room air temperature is less than or equal to 84 F;
| |
| : b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers of both the Filtration and Pressurization Systems and verifying that the Pressurization System operates for at least 10 continuous hours with the heaters operating; O
| |
| WOLF CREEK - UNIT 1 3/4 7-14
| |
| | |
| PLANT SYSTEMS
| |
| () SURVEILLANCE REQUIREMENTS (Continued)
| |
| : c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
| |
| : 1) Verifying that the Control Room Emergency Ventilation System satisfies the in place penetration and bypass leakage testing acceptance criteria; of less than 1% for HEPA filters and 0.05%
| |
| for charcoal adsorbers and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2000 cfm 110% at greater than or equal to 6.6 inches Water Gauge (W.G.) (dirty filter) for the Filtration System and 2200 cfm 110% at greater than or equal to 3.8 inches W.G.
| |
| (dirty filter) for the Pressurization System with 500 cfm 110%
| |
| going through the Pressurization System filter adsorber unit;
| |
| : 2) Verifying, within 31 days af ter removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision
| |
| -, 2, March 1978, for a methyl iodide penetration of less than 1%; and (d'
| |
| )
| |
| : 3) Verifying system flow rate of 2000 cfm 110% at greater than or equal to 6.6 inches W.G. (dirty filter) for the Filtration System and 2200 cfm 110% at greater than or equal to 3.8 inches W.G. (dirty filter) for the Pressurization System with 500 cfm 110% going through the Pressurization System filter adsorber unit during system operation when tested in accordance with ANSI N510-1980.
| |
| : d. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal, that a laboratory analysis of a represen-tative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;
| |
| : e. At least once per 18 months by:
| |
| : 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.6 inches Water Gauge while operating the system at a finw rate of 2000 cfm 110% for the Filtration System and 500 cfm 110% for the Pressurization System filter adsorber unit, q 2) Verifying that on a Control Room Ventilation Isolation or High
| |
| [C t Gaseous Radioactivity test signal, the system automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks, WOLF CREEK - UNIT 1 3/4 7-15
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 3) Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere during system operation,
| |
| : 4) Verifying that the Pressurization System filter adsorber unit heaters dissipate 15 1 2 kW in the Pressurization System when tested in accordance with ANSI H510-1975, and
| |
| : 5) Verifying that on a High Chlorine test signal, the system auto-matically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks within 15 seconds,
| |
| : f. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers in accordance with ANSI N510-1975 (however Prerequisite Testing, Sections 8 and 9 shall be in accord-ance with ANSI N510-1980) for a 00P test aerosol while operating the system at a flow rate of 2000 cfm 110% for the Filtration System and 500 cfm 110% for the Pressurization System filter adsorber unit; and
| |
| : g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers in accordance with ANSI N510-1975 (however Prerequisite Testing, Sections 8 and 9 shall be in accordance with ANSI N510-1980) for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 2000 cfm 110% for the Filtration System and 500 cfm 110% for the Pressurization System filter adsorber unit.
| |
| O WOLF CREEK - UNIT 1 3/4 7-16
| |
| | |
| m -
| |
| PLANT SYSTEMS i
| |
| ) 3/4.7.7 EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent Emergency Exhaust Systems shall be OPERABLE.
| |
| APPLICABILITY: "' DES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With one Emergency Exhaust System inoperable, restore the inoperable Emergency Exhaust System to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS _
| |
| 4.7.7 Each Emergency Exhaust System shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal (oV) adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;
| |
| : b. At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system, by:
| |
| : 1) Verifying that the Emergency Exhaust System satisfies the in-place penetration and bypass leakage testing acceptance cri-teria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers and uses the test procedure guidance in Regulatory Positions C.S.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 9000 cfm 110% at > 7.2 inches W.G. (dirty filter);
| |
| : 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;
| |
| V(D WOLF CREEK - UNIT 1 3/4 7-17
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 3) Verifying a system flow rate of 9000 cfm 110% at > 7.2 inches W.G. (dirty filter) during system operation when tested in accordance with ANSI N510-1980.
| |
| : c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;
| |
| : d. At least once per 18 months by:
| |
| : 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorbor banks of less than or equal to 7.2 inches Water Gauge while operating the system at a flow rate of 9000 cfm 110%,
| |
| : 2) Verifying that the system maintains the Fuel Buildir.g at a negative pressure of greater than or equal to inch Water Gauge relative to the outside atmosphere during system operation,
| |
| : 3) Verifying that the system starts on a Safety Injection test signal, and
| |
| : 4) Verifying that the heaters dissipate 37 1 3 kW when tested in accordance with ANSI N510-1975.
| |
| : e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers in accordance with ANSI N510-1975 (however Prerequisite Testing, Sections 8 and 9 shall be in accordance with ANSI N510-1980) for a DOP test aerosol while operating the system at a flow rate of 9000 cfm 110%; and
| |
| : f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers in accordance with ANSI N510-1975 (however Prerequisite Testing, Sections 8 and 9 shall bo in accordance with ANSI N510-1980) for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 9000 cfm 110%.
| |
| O WOLF CREEK - UNIT 1 3/4 7-18
| |
| | |
| PLANT SYSTEMS
| |
| ~
| |
| ! 3/4.7.8 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.8 All snubbers shall be OPERABLE. The only snubbers excluded from 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.
| |
| APPLTCABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.
| |
| ACTION:
| |
| With one or more snubbers inoperable on any system, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.8g on the attached comqonent or declare the attached system inopera-ble and follow the appropriate ACTION statement for that system.
| |
| SURVEILLANCE REQUIREMENTS (d 4.7.8 Each snubber shall be demonstrated OPERABLE by performance of the follow-ing augmented inservice inspection program and the requirements of Specifica-tion 4.0.5.
| |
| : a. Inspection Types As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.
| |
| : b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed after 4 months but within 10 months of commencing POWER OPERATION and shall include all hydraulic and mechanical snubbers. If all snubbers of each type are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection of that type shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given type shall be performed in accordance with the following schedule:
| |
| Ch WOLF CREEK - UNIT 1 3/4 7-19
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| No. of Inoperable Snubbers of Each Subsequent Visual Type per Inspection Period Inspection Period *#
| |
| 0 18 months i 25%
| |
| 1 12 months i 25%
| |
| 2 6 months i 25%
| |
| 3,4 124 days 1 25%
| |
| 5,6,7 62 days 25%
| |
| 8 or more 31 days i 25%
| |
| : c. Visual Inspection Acceptance Criteria Visual inspections shall verify 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 visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespec-tive of type that may be generically susceptible; or (2) the affected snubbar is functionally tested in the as-found condition and determined OPERABLE per Specification 4.7.8f. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers.
| |
| : d. Transient Event inspections An inspection shall be performed of all hydraulic and mechanical snub-bers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of opera-tional data and a visual inspection of the systems within 6 months following such an event. In additiun to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following:
| |
| (1) manually induced snubber movement; (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.
| |
| *The inspection interval for each type of snubber shall not be lengthened more than one step at a time unless a generic problem has been identified and corrected; in that event the inspection interval may be lengthened one step the first time and two steps thereafter if no inoperable snubbers of that type are found.
| |
| #The provisions of Specification 4.0.2 are not applicable.
| |
| WOLF CREEK - UN!f 1 3/4 7-20
| |
| | |
| PLANT SYSTEMS i
| |
| l SURVEILLANCE REQUIREMENTS (Continued)
| |
| : e. Functional 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 t.5anged during the test period. The NRC Regional Administrator shall be notified in writing 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 criteria of Specification 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 4.7-1. " (. " is the s total number of snubbers of a type found not meeting the accep-( ) tance requirements of Specification 4.7.8f. The cumulative s' 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 4.7-1. If at any time the point plotted falls in the " Reject" region, all snubbers of that type shall be function-ally tested. If at any time the point plotted falls in the
| |
| " Accept" region, testing of snubbers of that type may be terminated. When the peint plotted lies in the " Continue ler, ting" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the
| |
| " Reject" region, or all the snubbers of that type have been tested; or
| |
| : 3) An initial representative sample of 55 snubbers shall be func-tionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the factor, 1 + C/2 where "C" is the number of snubbers found which do not meet the functional test accept-ance criteria. The results from this sample plan shall be plotted using an " Accept" line which follows the equation N =
| |
| 55(1 + C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept" line, testing of that type of snubber may be J'")
| |
| (
| |
| terminated. If the point plotted falls above the " Accept" line, testing must continuo until the point falls in the " Accept" region or all the snubbers of that type have been tested.
| |
| WOLF CREEK - UNIT 1 3/4 7-21
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : e. Functional Tests (Continued)
| |
| Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment 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 functio- testing, additional sampling is required due to failure of only one 6ype 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 thruugh established methods.
| |
| : g. Service Life Monitoring Program 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 offort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
| |
| O WOLF CREEK - UNIT 1 3/4 7-22
| |
| | |
| <, g gy}
| |
| - ~ ~ ~ ^ ^ -- - - ' ~ ~
| |
| ~ .
| |
| --m3 7^
| |
| . qty - 3'3 " %. ' g.
| |
| y
| |
| , e,
| |
| .; *% n 1-y s" N LPLANT SYSTEMS' '9 A t
| |
| r r ;6
| |
| % ~. A 1 .
| |
| ; SURVEILLANCE REQUIREMENTS (Continued)
| |
| W w
| |
| c ,
| |
| .g'W
| |
| : g. 9 dr$ ice'L'if h nitoring Program (Continued) g yn F0r the Uiubbers found inoperable, an engineering evaluation shall ,
| |
| 4, ; A, be performed on the components to which the inoperable snubbers are h The purpose;of this engineering evaluation shall be to gycdetermine attached.;if the components to which the inoperable snubbers are
| |
| ~
| |
| y .
| |
| . ~A ' W * < attached were adverselycaffecte'd by theJinoperability of the snubbers i;;
| |
| -in order to ensure that the component remains capable of meeting the fdesignedsery, ice., - ,
| |
| e , y n >
| |
| 4 9 If any, snubber selected for f,uritionals testing either fails to y.h..
| |
| e lock,up or fails to move, i.e:, frozen-in place, the cause will be Q4*
| |
| rM i
| |
| .. evaluated"and, if caused by r.anufacturer or design deficiency,~all 3 snub 5ers of the same' type subject to the same defect shall be func-U,. '
| |
| a *. tionally tested. This: testing. requirement shall be independent of
| |
| : .sg the reqd diments stated in' Specification 4.7.8e. for snubbers'not
| |
| [ p,, meetinf,the functional test acceptance criteria.
| |
| 3 L h.' Functional Testing of Repaired and Replaced Snubbers A Snubbers which fail the visual inspection or the functional test Jacceptance criteria.shuil be repaired or replaced. . Replacement i
| |
| \
| |
| 4 snubbss and snubbers which have repairs which might affect the
| |
| 'funci.ioUal Test results shall be tested.to meet the functional test
| |
| ~
| |
| s 4 4
| |
| : h. ii ~
| |
| 'ceiteria,before installation in the unit. Mechanical snubbers shall haCe metJthe acceptance criteria subsequent to their most recent
| |
| ~~
| |
| .A'
| |
| ^> g service,'shsthe freedom-of motion test must have been performed
| |
| ~
| |
| ,4 *r3 'within 12 months before being installed in the unit.
| |
| n% , g .
| |
| i
| |
| :,-* N _i. ._ Snubber Service Life Program
| |
| .The servicelife of hydraulic and mechanical-snubbers shall be moni- '
| |
| : s. dtired to ensure that the service life is not exceeded between sur-
| |
| 'q.- Y ''m '
| |
| \pillance.i.nspections. The maximum expected service life for varous l' i _s . seals, springs, and other critical parts shall be determined'and
| |
| -M h" I established.bised o'n engineering information and shall be extended 1ED or shortened based on monitored test results and failure history.
| |
| .. Critic'ah paqts shall be replaced so that the maximum service life will
| |
| " M,('il - ~
| |
| -nop be 4kceuded during a period when the sn.ubber is required to be 3' -CPERABLE. S he parts replacements shall be documented and the docu-1 mentat Nd sh311-be retained in accordance with Specification 6.10.2.
| |
| , +
| |
| W s v q g &
| |
| [-lY p > '
| |
| s
| |
| \}'
| |
| ' ts e..
| |
| c q _ t
| |
| -t i. ,
| |
| l WOLF CREEK - UNIT 1 3/4 7-23 i
| |
| j -
| |
| ;a .- s
| |
| ,J& ,.~
| |
| ~ ~
| |
| . ,w.__m, *----- ..m* -, 4m
| |
| | |
| 10 h
| |
| 9 8
| |
| REJECT j /
| |
| C s g,yp V
| |
| *0 r
| |
| / ? s"rIL"o"" / G 2 -
| |
| ACCEPT 1 r 0 10 20 30 40 50 60 70 80 90 100 N
| |
| FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST O
| |
| WOLF CREEK - UNIT 1 3/4 7-24
| |
| | |
| 7.
| |
| PLANT SYSTEMS
| |
| '('D!-
| |
| t 3/4.7.9 SEALED SOURCE CONTAMINATION
| |
| %/
| |
| LIMITING CONDITION FOR OPERATION 3.7.9 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma-emitting material or 5 microCuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:
| |
| : 1. Decontaminate and repair the sealed source, or
| |
| : 2. Dispose of the sealed source in accordance with Commission Regulations.
| |
| O b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 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.
| |
| . 4.7.9.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below,
| |
| : a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:
| |
| : 1) With a half-life greater than 30 days (excluding Hydrogen 3),
| |
| and
| |
| / 2) In any form other than gas.
| |
| l l
| |
| WOLF CREEK - UNIT 1 3/4 7-25
| |
| | |
| 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.
| |
| 4.7.9.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 7-26 l
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| | |
| PLANT SYSTEMS (O
| |
| v
| |
| ) 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM 4
| |
| LIMITING CONDITION FOR OPERATION 3.7.10.1 The Fire Suppression Water System shall be OPERABLE with:
| |
| : a. At least two fire suppression pumps, each with a capacity of >3300 gpm, with their discharge aligned to the fire suppression header, and
| |
| : b. An OPERABLE flow path capable of taking suction from the Wolf Creek Generating Station cooling lake and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.10.2 and 3.7.10.4.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| (' a. With one pump inoperable, restore the inoperable equipment to OPERABLE status within 7 days. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| : b. With the Fire Suppression Water System otherwise inoperable establish a backup Fire Suppression Water System within 24 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.7.10.1.1 The Fire Suppression Water System shall be demonstrated OPERABLE:
| |
| : a. At least once per 7 days by verifying that lake level exceeds 1075 feet,
| |
| : b. At least once per 31 days on a STAGGERED TEST BASIS by starting the electric motor-driven pump and operating it for at least 15 minutes on recirculation flow,
| |
| : c. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position, 4
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| WOLF CREEK - UNIT 1 3/4 7-27
| |
| | |
| PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : d. At least once per 6 months by performance of a yard loop and fire hydrant flush,
| |
| : e. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
| |
| : f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
| |
| : 1) Verifying that each pump develops at least 3300 gpm at a system i pressure of 80 psig,
| |
| : 2) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and
| |
| : 3) Verifying that the electric driven fire pump starts on a start, signal initiated on decreasing header pressure of 75 psig and the diesel driven fire pump starts on a start signal on decreasing header pressure of 70 psig after 10 second time delay to avoid simultant:ous start of both pumps.
| |
| : g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.
| |
| 4.7.10.1.2 Each fire pump diesel engine shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days by verifying:
| |
| : 1) The fuel storage tank contains at least 200 gallons of fuel, and
| |
| : 2) The diesel starts from ambient conditions and operates for at least 30 minutes on recirculation flow.
| |
| : b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-1975, is within the acceptable limits specified in Table 1 of ASTM 0975-1977 when checked for viscosity, water, and sediment; and
| |
| : c. At least once per 18 months, during shutdown, by subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.
| |
| O WOLF CREEK - UNIT 1 3/4 7-28
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| | |
| PLANT SYSTEMS' SURVEILLANCE REQUIREMENTS (Continued) 4.7.10.1.3 Each fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:
| |
| : a. At least once per 7 days by verifying that:
| |
| : 1) The electrolyte level of each battery is above the plates, and
| |
| : 2) The overall battery voltage is greater than or equal to 24 volts.
| |
| : b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and
| |
| : c. At least once per 18 months, by verifying that:
| |
| : 1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and
| |
| : 2) The battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.
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| O O
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| e WOLF CREEK - UNIT 1 3/4 7-29
| |
| | |
| PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.10.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:
| |
| : a. Wet Pipe Sprinkler Systems Building Elevation Area Protected Auxiliary 2000/2026 North Electric Cable Chase Auxiliary 1988/2000/2026 South Electric Cable Chase Control 1974 - 2073 Vertical Electrical Chases Control 1974 Pipe Space and Tank Room Control 1992 Cable Area Above Access Control
| |
| : b. Pre-Action Sprinkler Systems Building Elevation Area Protected Auxiliary 1974 Cable Trays
| |
| * Auxiliary 2000 Cable Trays
| |
| * Auxiliary 2026 Cable Trays
| |
| * Control 2032 Lower Cable Spreading Room Control 2073 Upper Cable Spreading Room Reactor 2026 North Cable Penetration Area Reactor 2026 South Cable Penetration Area Diesel Gen. (E) 2000 East Diesel Generator Room Diesel Gen. (W) 2000 West Diesel Generator Room
| |
| : c. Water Sprays Systems Building Elevation Area Protected Auxiliary 2000 Auxiliary Feedwater Pump Turbine ESF Transformer Grade Transformer XNB01*
| |
| ESF Transformer Grade Transformer XNB02*
| |
| APPLICABILITY: Whenever equipment protected by the Spray / Sprinkler System is required to be OPERABLE.
| |
| ACTION:
| |
| : a. With one or more of the above required Spray and/or Sprinkler Systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol,
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.7.10.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE:
| |
| * Areas contain redundant systems or components which could be damaged.
| |
| l WOLF CREEK - UNIT 1 3/4 7-30
| |
| | |
| l n PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : a. At least once per 31 days, by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position,
| |
| : b. At least once per 12 months, by cycling each testable valve in the flow path through at least one complete cycle of full travel,
| |
| : c. At least once per 18 months:
| |
| : 1) By performing a system functional test which includes simulated automatic actuation of the system, and:
| |
| a) Verifying that the automatic valves in the flow path actuate to their correct positions on a Simulated Fire test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.
| |
| : 2) By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity, and
| |
| : 3) By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.
| |
| : d. At least once per 3 years by performing an air or water flow test through each open head spray / sprinkler header and verifying each open head spray / sprinkler nozzle is unobstructed.
| |
| l l
| |
| l 1
| |
| i i
| |
| i v WOLF CREEK - UNIT 1 3/4 7-31
| |
| | |
| PLANT SYSTEMS HALON SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.10.3 The following Halon Systems shall be OPERABLE:
| |
| Building Elevation Area Protected Auxiliary 2026 North Electrical Penetration Room Auxiliary 2026 South Electrical Penetration Room Auxiliary 2026 Load Center and M. G. Sets Room
| |
| * Control 2000 ESF Switchgear Rooms
| |
| * Control 2016 Switchgear Rooms Control 2047 Control Room Cable Trenches and Chases APPLICABILITY: Whenever equipment protected by the Halon System is required to be OPERABLE.
| |
| ACTION:
| |
| : a. With one or more of the above required Halon systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.7.10.3 Each of the above required Halon Systems shall be demonstrated OPERABLE:
| |
| : a. At least once per 6 months by verifying Halon storage tank weight (or level) to be at least 95% of full charge weight and pressure to be at least 90% of full charge pressure, and
| |
| : b. At least once per 18 months by verifying the system, including associated Ventilation System fire dampers and fire door release mechanisms, actuates manually and automatically, upon receipt of a simulated actuation signal.
| |
| * Areas contain redundant systems or components which could be damaged.
| |
| O WOLF CREEK - UNIT 1 3/4 7-32
| |
| | |
| PLANT SYSTEMS
| |
| , ) FIRE HOSE STATIONS v
| |
| LIMITING CONDITION FOR OPERATION 3.7.10.4 The fire hose stations given in Table 3.7-3 shall be OPERABLE.
| |
| APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.
| |
| ACTION:
| |
| : a. With one or more of the fire hose stations shown in Table 3.7-3 inoperable, provide equivalent capacity backup hose protection to the unprotected area from the spare hose connection on the adjacent OPERABLE standpipe. If two standpipe hose connections are not available at the adjacent OPERABLE hose station (s), provide gated wye (s) to ensure continued OPERABILITY of the affected hose station.
| |
| Where it can be demonstrated that the physical routing of the backup hose would result in a recognizable hazard to operating technicians, plant equipment, or the hose itself, or would require the blocking open of a fire door, the hose shall be stored at the point of origin and properly identified as to its intended use. The above action shall be accomplished within 1 hour if the inoperable fire hose is
| |
| 'n the primary means of fire suppression; otherwise route the additional
| |
| } hose within 24 hours.
| |
| ''d b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.7.10.4 Each of the fire hose stations given in Table 3.7-3 shall be demonstrated OPERABLE:
| |
| : a. At least once per 31 days, by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station,
| |
| : b. At least once per 18 months, by:
| |
| : 1) Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station,
| |
| : 2) Removing the hose for inspection and reracking, and
| |
| : 3) Inspecting all gaskets and replacing any degraded gaskets in the couplings.
| |
| : c. At least once per 3 years, by:
| |
| : 1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and
| |
| : 2) Conducting a hose hydrostatic test at a pressure of 150 psig or
| |
| ( 'N at least 50 psig above maximum fire main operating pressure,
| |
| \ ) whichever is greater.
| |
| WOLF CREEK - UNIT 1 3/4 7-33
| |
| | |
| ABLE 3.7-3 FIRE HOSE STATIONS BUILDING ELEVATION AREA HOSE RACK Auxiliary 1974 1122 KC-HR-051 Auxiliary 1974 1122 KC-HR-047 Auxiliary 1974 1120 KC-HR-031 Auxiliary 1974 1120 KC-HR-025#
| |
| Auxiliary 1974 1101 KC-HR-023#
| |
| Auxiliary 1974 1101 KC-HR-040 Auxiliary 1974 1101 KC-HR-042 Auxiliary 1988 1201 KC-HR-024 Auxiliary 2000 1329 KC-HR-111 Auxiliary 2000 1320 KC-HR-048 Auxiliary 2000 1320 KC-HR-046#
| |
| Auxiliary 2000 1314 KC-HR-030 Auxiliary 2000 1321 KC-HR-029#
| |
| Auxiliary 2000 1301 KC-HR-035#
| |
| Auxiliary 2000 1301 KC-HR-039 Auxiliary 2000 1301 KC-HR-041#
| |
| Auxiliary 2026 1408 KC-HR-049 Auxiliary 2026 1408 KC-HR-044 Auxiliary 2026 1408 KC-HR-032#
| |
| Auxiliary 2026 1408 KC-HR-026#
| |
| Auxiliary 2026 1401 KC-HR-034 Auxiliary 2026 1403 KC-HR-037#
| |
| Auxiliary 2047 1506 KC-HR-050 Auxiliary 2047 1513 KC-HR-043 Auxiliary 2047 1506 KC-HR-045 Auxiliary 2047 1501 KC-HR-038 Auxiliary 2047 1504 KC-HR-033 Auxiliary 2047 1502 KC-HR-027 Auxiliary 2064 1119 KC-HR-028#
| |
| Control 1974 3101 KC-HR-002#
| |
| Control 1974 3101 KC-HR-014#
| |
| Control 1984 3204 KC-HR-015#
| |
| Control 1984 3221 KC-HR-001#
| |
| Control 2000 3301 KC-HR-004#
| |
| Control 2000 3301 KC-HR-017#
| |
| Control 2000 3302 KC-HR-016#
| |
| Control 2016 3401 KC-HR-005 Control 2016 3401 KC-HR-019 Control 2016 3401 KC-HR-018 O
| |
| WOLF CREEK - UNIT 1 3/4 7-34
| |
| | |
| l TABLE 3.7-3 (Continued)
| |
| FIRE HOSE STATIONS SUILDING ELEVATION AREA HOSE RACK Control 2032 3501 KC-HR-006#
| |
| Control 2032 3501 KC-HR-020#
| |
| Control 2047 3604 KC-HR-007 Control 2047 3616 KC-HR-021 Control 2073 3801 KC-HR-008#
| |
| Control 2073 3801 KC-HR-022#
| |
| Reactor 2000 2201 KC-HR-120*
| |
| Reactor 2000 2201 KC-HR-131* ,
| |
| Reactor 2000 2201 KC-HR-124*
| |
| Reactor 2000 2201 KC-HR-129*
| |
| Reactor 2026 N.A. KC-HR-121*
| |
| Reactor 2026 N.A. KC-HR-132*#
| |
| Reactor 2026 N.A. KC-HR-125*
| |
| Reactor 2026 N.A. KC-HR-130*
| |
| Reactor 2047 N.A. KC-HR-128*
| |
| Reactor 2047 N.A. KC-HR-122*
| |
| Reactor 2047 N.A. KC-HR-126" Reactor 2068 N.A. KC-HR-123*
| |
| Reactor 2068 N.A. KC-HR-127*
| |
| (q
| |
| -L/
| |
| Fuel Fuel Fuel 2000 2000 2000 6102 6102 6102 KC-HR-142#
| |
| KC-HR-054#
| |
| KC-HR-143 Fuel 2000 6104 KC-HR-057 Fuel 2026 G201 KC-HR-133 Fuel 2026 6203 KC-HR-052 Fuel 2047 6301 KC-HR-055# '
| |
| Fuel 2047 6302 KC-HR-056#
| |
| Fuel 2047 6301 KC-HR-053#
| |
| ESW 2000 N.A. K'. HR-140 ESW 2000 N.A. KC-HR-141 TABLE NOTATIONS i # Secondary means of fire suppression to Water Sprays / Deluge or Halon Systems.
| |
| * Fire hose for station to be stored external to Reactor Building.
| |
| WOLF CREEK - UNIT 1 3/4 7-35
| |
| | |
| PLANT SYSTEMS 3/4.7.11 FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.11 All fire barrier penetrations (walls, floor / ceilings, cable tray enclosures, and other fire barriers) separating safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire-rated assembly penetrations (fire doors, fire windows, fire dampers, cable, piping, and ventilation duct penetra-tion seals) shall be OPERABLE.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With one or more of the above required fire barrier penetrations inoperable, within 1 hour establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the inoperable fire barrier and establish an hourly fire watch patrol.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.7.11.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of:
| |
| : a. The exposed surfaces of each fire rated assembly,
| |
| : b. Each fire window / fire damper and associated hardware, and
| |
| : c. At least 10% of each type (electrical and mechanical) of sealed penetration. If apparent changes in apoearance or abnormal degrada-tions are found, a visual inspection of an additional 10% of each type of sealed penetration shall be made. This inspection process shall continue until a 10% sample with no apparent changes in appear-ance or abnormal degradation is found. Samples shall be selected such that each penetration seal will be inspected every 15 years.
| |
| 4.7.11.2 Each of the above required fire doors shall be verified OPERABLE by inspecting the automatic hold-open, release and closing mechanism and latches at least once per 6 months, and by verifying:
| |
| : a. The OPERABILITY of the fire Door Supervision System for each electrically supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days,
| |
| : b. That each locked closed fire door without electrical supervision is closed at least once per 7 days,
| |
| : c. That doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours and performing a functional test at least once per 18 months, and
| |
| : d. That each unlocked fire door without electrical supervision is closed at least once per 24 hours.
| |
| WOLF CREEK - UNIT 1 3/4 7-36
| |
| | |
| , PLANT SYSTEMS A
| |
| j( 3/4.7.12 AREA TEMPERATURE MONITORING l LIMITING CONDITION FOR OPERATION 3.7.12 The temperature limit of each area given in Table 3.7-4 shall not be !
| |
| exceeded for more'than 8 hours or by more than 30'F. '
| |
| APPLICABILITY: 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 3.7-4 for more than 8 hours, prepare and submit to the Commission within 30 days, pursuant to 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 Specifications 3.0.3 and 3.0.4 are not applicable. ,
| |
| ; b. With one or more areas exceeding the temperature limit (s) shown in (q
| |
| 'd
| |
| )
| |
| Table 3.7-4 by more than 30*F, prepare and submit a Special Report as required by ACTION a. above, and within 4 hours either restore the area (s) to within the temperature limit (s) or declare the equipment in l
| |
| 1 the affected area (s) inoperable.
| |
| l SURVEILLANCE REQUIREMENTS 1
| |
| , 4.7.12 The temperature in each of the areas shown in Table 3.7-4 shall be
| |
| ; determined to be within its limit at least once per 12 hours.
| |
| O I
| |
| WOLF CREEK - UNIT 1 3/4 7-37
| |
| | |
| TABLE 3.7-4 AREA TEMPERATURE MONITORING MAXIMUM 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 II 87
| |
| : 8. RHR Pump Room A 119
| |
| : 9. RHR Pump Room B 119
| |
| : 10. CTHT Spray Pump Room A 119
| |
| : 11. CTMT Spray Pump Room B 119
| |
| : 12. Safety Injection Pump Roan A 119
| |
| : 13. Safety Injection Pump Room B 119
| |
| : 14. Centrifugal Charging Pump Room A 119
| |
| : 15. Centrifugal Charging Pump Room B 119
| |
| : 16. Electrical Penetration Room A 101
| |
| : 17. Electrical Penetration Room B 101
| |
| : 18. Component Cooling Water Rt2m A 119
| |
| : 19. Component Cooling Water Raum B 119
| |
| : 20. Diesel Generator Room A 119
| |
| : 21. Diesel Generator Room B 119
| |
| : 22. Control Room 84 O
| |
| WOLF CREEK - UNIT 1 3/4 7-38
| |
| | |
| 3/4.8 ELECTRICAL POWER SYSTEMS O
| |
| '( 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERA 8LE:
| |
| : a. Two physically independent circuits between the offsite transmission network 6nd the Onsite Class 1E Distribution System, and
| |
| : b. Two separate and independent diesel generators, each with:
| |
| : 1) Aschtratedaytankcontainingaminimumvolumeof390 gallons of fur 1,
| |
| : 2) A siparate Fuel Oil Storage System containing a minimum volume of 85,300 gallons of fuel, and
| |
| : 3) A separate fuel transfer pump.
| |
| APPLICA8ILITY: MOD'E 1, 2, 3, and 4.
| |
| Q
| |
| \ )
| |
| ACTION:
| |
| U a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifications 4.8.1.1.1 and 4.8.1.1.2a.4) within 1 hour and at least once per 8 hours thereafter; restore at least two offsite circuits and two diesel generators to OPERA 8LE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours,
| |
| : b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifications 4.8.1.1.1 and 4.8.1.1.2a.4) within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two offsite circuits and two diesel generators to OPERABLE stetus within 72 hours from the time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| : c. With one diesel generator inoperable in addition to ACTION a. or b.
| |
| above, verify that:
| |
| [(
| |
| : 1. All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and WOLF CREEK - UNIT 1 3/4 8-1
| |
| | |
| ELECTRICAL POWER S(SfEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) 2 When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater pump is OPERABLE.
| |
| If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
| |
| : d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Specification 4.8.1.1.2a.4) within 1 hour and at least once per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
| |
| : e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by perform-ing Specification 4.8.1.1.1 within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class 1E Distribution System shall be determined OPERABLE at least once per 7 days by verifying ccrrect breaker alignments and indicated power availability.
| |
| 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
| |
| : a. In accordance with the frequency specified in Tabic 4.8-1 on a STAGGERED TEST BASIS by:
| |
| : 1) Verifying the fuel level in the day tank,
| |
| : 2) Verifying the fuel level in the fuel storage tank, WOLF CREEK - UNIT 1 3/4 8-2
| |
| | |
| ELECTRICAL POWER SYSTEMS
| |
| [V ' SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank,
| |
| : 4) Verifying the diesel starts from ambient condition and accelerates to at least 514 rpm in less than or equal to 12 seconds.* The generator voltage and frequency shall be 1160 + 160 - 420 volts and 60 + 1.2 H: within 12 seconds" after the start signal. The diesel generator shall be started for this test by using one of the following signals:
| |
| a) Manual, or b) Simulated loss-of-offsite power by itself, or c) Safety Injection test signal.
| |
| : 5) Verifying the generator is synchronized, loaded to greater than or equal to 6201 kW in less than or equal to 60 seconds,* operates with a load greater than or equal to 6201 kW for at least 60 minutes, and
| |
| []
| |
| \ /
| |
| : 6) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses. ,
| |
| : b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from the day tanks 7
| |
| : c. At least once per 31 days by checking for and removing accumulated water from the fuel oil' storage tanks;
| |
| : d. By sampling new fuel oil in accordance with ASTM 04057 prior to addition to storage tanks and:
| |
| (1) By verifying in accordance with the tests specified in ASTM 0975-81 prior to addition to the storage tanks that the sample has:
| |
| (a) An API Gravity of within 0.3 degrees at 60*F or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate or an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees;
| |
| *These diesel generator starts from ambient conditions shall be performed only once per 184 days in these surveillance tests and all other engine starts for the purpose of this surveillance testing shall be preceded by an engine qy prelube period and/or other warmup procedures recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.
| |
| WOLF CREEK - UNIT 1 3/4 8-3 l
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| (b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the sup-plier's certification; (c) A flash point equal to or greater than 125*F; and (d) A clear and bright appearance with proper color when tested in accordance with ASTM 04176-82.
| |
| (2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-81 are met when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.
| |
| : e. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM D2276-78, Method A.
| |
| : f. At least once per 18 months, during shutdown, by:
| |
| : 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service,
| |
| : 2) Verifying the diesel generator capability to reject a load of greater than or equal to 1352 kW (ESW pump) while maintaining voltage at 4160 + 160 - 420 volts and frequency at 60 1 5.4 Hz,
| |
| : 3) Verifying the diesel generator capability to reject a load of 6201 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the load rejection,
| |
| : 4) Simulating a loss-of-offsite power by itself, and:
| |
| a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts on the auto-start signal, ener0izes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected shutdown loads through the shutdown sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 4 160 - 420 volts and 60 1 1.2 Hz during this test.
| |
| WOLF CREEK - UNIT 1 3/4 8-4
| |
| | |
| ELECTRICAL POWER SYSTEMS g
| |
| SURVEILLANCE REQUIREMEhTS (Continued) v
| |
| )
| |
| : 5) Verifying that on a Safety Injection test signal without loss-of-offsite power, the diesel generator starts on the auto-start sig-nal and operates on standby for greater than or equal to 5 minutes; and the offsite power source energizes the auto-connected emerg-ency (accident) load through the LOCA sequencer. The generator voltage and frequency shall be 4160 + 160 - 420 volts and 60 +
| |
| 1.2 Hz within 12 seconds after the auto-start signal; the genera-tor steady-state generator voltage and frequency shall be main-tained within these limits during this test;
| |
| : 6) Simulating a loss-of-offsite power in conjunction with a Safety j Injection test signal, and j a) Verifying deenergization of the emergency busses and load j shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected emergency (accident) loads through the LOCA sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with emergency loads. After energiza-A tion, the steady-state voltage and frequency of the emerg-( ) ency busses shall be maintained at 4160 + 160 - 420 volts and V 60 + 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except high jacket coolant temperature, engine overspeed, low lube oil pressure, high crankcase pressure, start failure relay, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safet/
| |
| Injection Actuation signal.
| |
| 1
| |
| : 7) Verifying the diesel generator operates for at least 24 hours. I During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 6821 kW and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 6201 kW. The generator voltage and frequency shall be 4160 + 160 - 420 volts and 60 +
| |
| 1.2 Hz, - 3 Hz within 12 seconds after the start signal; the ,
| |
| steady-state generator voltage and frequency shall be maintained j within 4160 1 160 - 420 volts and 60 1 1.2 Hz during this test. 1 Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2f.6)b)*;
| |
| *If Specification 4.8.1.1.2f.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator j
| |
| , (a) may be operated at 6201 kW for 1 hour or until operating temperature has l l
| |
| V stabilized.
| |
| WOLF CREEK - UNIT 1 3/4 8-5 t
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) _
| |
| : 8) Verifying that the auto-connected loads to each diesel generator do not exceed 6201 kW;
| |
| : 9) Verifying the diesel generator's capability to:
| |
| a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
| |
| : 10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test modo by: (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offr,ite power;
| |
| : 11) Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the installed cross-connection lines; and
| |
| : 12) Verifying that the automatic LOCA and Shutdown sequence timer is OPERABLE with the interval between each load block within i 10% of its design interval.
| |
| : g. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 12 seconds; and
| |
| : h. At least once per 10 years by:
| |
| : 1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution or equivalent, and
| |
| : 2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.
| |
| 4.8.1.1.3 Reports - All diesel generator failures, valid or nonvalid, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days. Reports of diesel generator failures shall include the informa-tion recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revi-sion 1, August 1977. If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.
| |
| WOLF CREEK - UNIT 1 3/4 8-6
| |
| | |
| i l
| |
| i TABLE 4.8-1 i
| |
| ( DIESEL GENERATOR TEST SCHEDULE i NUNBER OF FAILURES IN LAST 100 VALIO TESTS
| |
| * TEST FREQUENCY f
| |
| f,1 At least once per 31 days l 2 At least once per 14 days j 3 At least once per 7 days [
| |
| >4
| |
| _ At least once per 3 days l
| |
| l l
| |
| O :
| |
| ?
| |
| l
| |
| * Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, ;
| |
| Revision 1, August 1977, where the last'100 tests are determined on a per nuclear unit basis. For the purposes of this test schedule, only valid '
| |
| tests conducted after the completion of the preoperational test requirements '-
| |
| of Regulatory Guide 1.108, Revision 1, August 1977, shall be included in the computation of the "Last 100 Valid Tests."
| |
| l {
| |
| I t
| |
| (
| |
| t WOLF CREEK - UNIT 1 3/4 8-7 i
| |
| | |
| ) ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUT 00WN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
| |
| : a. One circuit between the offsite transmission network and the Onsite Class 1E Distribution System, and
| |
| : b. One diesel generator with:
| |
| : 1) A day tank containing a minimum volume of 390 gallons of fuel,
| |
| : 2) A fuel storage system containing a minimum volume of 85,300 gallons of fuel, and
| |
| : 3) A fuel transfer pump.
| |
| APPLICABILITY: MODES 5 and 6.
| |
| ACTION:
| |
| With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactiv-ity changes, movement of irradiated fuel, or crane operation with loads over the spent fuel pool. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.
| |
| SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5)), and 4.8.1.1.3.
| |
| O WOLF CREEK - UNIT 1 3/4 8-8 k
| |
| | |
| f I
| |
| g3 ELECTRICAL POWER SYSTEMS
| |
| 'f
| |
| . V)
| |
| I 3/4.8.2 D.C. SOURCES OPERATIM LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:
| |
| : a. 125-Volt Battery Bank NK11 and NK13, and its associated Full Capacity Chargers NK21 and NK23, and
| |
| : b. 125-Volt Battery Bank NK12 and NK14, and its associated Full Capacity Chargers NK22 and NK24.
| |
| APPLICA8ILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With one of the required battery banks and/or full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERA 8LE status within 2 hours or bo in at least HOT STAN08Y within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:
| |
| : a. At least once per 7 days by verifying that:
| |
| : 1) The parameters in Table 4.8-2 meet the Category A limits, and
| |
| : 2) The total battery terminal voltage is greater than or equal to 130.2 volts on float charge.
| |
| U WOLF CREEK - UNIT 1 3/4 8-9
| |
| | |
| ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:
| |
| : 1) The parameters in Table 4.8-2 meet the Category B limits,
| |
| : 2) There is no visible corrosion at either terminals or connectors, or the cell-to-cell and terminal connection resistance of these items is less than 150 x 10 8 ohm, and
| |
| : 3) The average electrolyte temperature of at least every sixth cell is above 60 F.
| |
| : c. At least once per 18 months by verifying that:
| |
| : 1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
| |
| : 2) The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material,
| |
| : 3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 8 ohm, and
| |
| : 4) The battery charger will supply at least 300 amperes at 130.2 volts for at least I hour,
| |
| : d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status simulated emergency loads for the design duty cycle when the battery is subject to a battery service test;
| |
| : e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and
| |
| : f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
| |
| O WOLF CREEK - UNIT 1 3/4 8-10
| |
| | |
| TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A II) l CATEGORY 8(2)
| |
| PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE I3)
| |
| DESIGNATED PILOT l CONNECTED CELL VALUE FOR EACH CELL i CONNECTED CELL Electrolyte l > Minimum level , > Minimum level Above top of Level ! indication mark, .
| |
| indication mark, plates, and < h" above i and < 4" above and not i maximum level maximum level overflowing indication mark indication mark FloatVoltagel 1 2.13 volts 1 2.13 volts (6) > 2.07 volts I
| |
| i Not more than i
| |
| 0.020 below the
| |
| ! average of all 1 1.195 connected cells l
| |
| Specifi ; > 1.200(5)
| |
| Gravity 4)
| |
| ~
| |
| \ Average of all Average of all '
| |
| l' connected cells connect
| |
| > 1.205 11.195g) cells
| |
| _TA8LE NOTATIONS (1) for any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERA 8LE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days.
| |
| (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERA 8LE provided that the Category 8 parameters are within their allowable values and provided the Category 8 parameter (s) are restored to within limits within 7 days.
| |
| (3) Any Category B parameter not within its allowable value indicates an inoperable battery.
| |
| (4) Corrected for electrolyte temperature and level.
| |
| ($) Or battery charging current is less than 2 amps when on charge.
| |
| (6) Corrected for average electrolyte temperature.
| |
| O WOLF CREEK UNIT 1 3/4 8 11
| |
| | |
| ELECTRICAJ POWER SYSTEMS D.C. SOURCES SHUT 00WN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following D.C. electrical sources shall be OPERABLE:
| |
| : a. 125-Volt Battery Bank NK11 and NK13, and its associated full capacity Chargers NK21 and NK23, or
| |
| : b. 125-Volt Battery Bank NK12 and NK14, and its associated full capacity Chargers NK22 and NK24.
| |
| APPLICABILITY: MODES 5 and 6.
| |
| ACT10N:
| |
| With the required battery bank and/or full capacity charger inoperable, immediately suspend all operations involvina CORE ALTERATIONS, positive reactivity chan005 or movement of irradiated fuel; initiate corrective action to restore the required battery bank and/or full capacity charger to OPERABLE status as soon as possibic.
| |
| O SURVEltLANCE REQUIREMENTS 4.8.2.2 The above required 125 volt battery banks and associated chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.
| |
| O WOLF CREEK - UNIT 1 3/4 8-12
| |
| | |
| rx v g. .
| |
| -p a~ ,, ) ,
| |
| ,,s 4 ag ,. _
| |
| %4 .
| |
| (l t l ,.
| |
| .ELECTPICAL"PbkER SYSTEAS,
| |
| , .n:
| |
| n().
| |
| 3/4;8.3 ONSITE, POWER DISTRIBUTION Us , ,
| |
| OPERATING ,{ g LIMITING CONDITION FOR OPERATION 4
| |
| 3.8.3.1 The following electrical busses shall be energized in the specified manner with tie breakers open between redundant busses within the unit:
| |
| ' a'.7 Division #1 A.C. 5 Emergency Busses consisting of:
| |
| .f C 1) 4160-Volt E:nergency Bus #NB01, and y 2) ,480-Volt Emergency Busses #NG01, NG03 and NG05E.
| |
| . b. , Division #2 A.C. Emergency Busses consisting of:
| |
| : 1) [4160-VolbEmergencyBus#NB02,and
| |
| - 2)' 1480-Volt Emergency Busses #NG02, NG04 and NG06E.
| |
| 4 N).^
| |
| 120-Volt A.C. Vital Bus #NN01 energized from its associated inverter connected to D.C.. Bus #NK01,
| |
| . d. 120-Volt A.C.' Vital Bus #NN02 energized from its associated inverter
| |
| . connec.ted to D.C. Bus #NK02, 7 e. 120-Volt A' C.1 Vital Bus #NNO3 energized from its associated inverter connected td D.C; Bus #NK03, O f. 120 holt A.h.' Vital Bus #NN04 energized from its associated ' inverter
| |
| () connected to D.C. Sus #NK04,
| |
| ( .
| |
| : g. 125-Volt D.C. Bus #NK01 energized from Battery #NK11 and Charger #NK21, s
| |
| q M 125-Vd t 0.C. Bus #NK02 energized from Battery #NK12 and Charger #NK22,
| |
| ,? i. 125-Volt D.C. Bus #NK03 energized from Battery #NK13 and Charger #NK23,
| |
| ~s and
| |
| , t- .
| |
| A[1 { j. 125-Volt D.C. Eus #NK04 energized from Battery #NK14 and Charger #NK24.
| |
| APPLICABILITY: MODES 1, 2, 3, and 4.
| |
| l ,
| |
| l .g <
| |
| ACTION: A r
| |
| l
| |
| > a. With one of'the required divisions of A.C. emergency busses not
| |
| !e 1 fully energized, reenergize the division within 8 hours or be in at
| |
| #y y least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within ithe following 30 bciurs.
| |
| y tic With one A.C. vital bus either not energized from its associated 73 invertW:or with the inverter not connected to its associated D.C.
| |
| '. ~'" bus: (1) reenergize the A.C. vital bus within 2 hours or be in at
| |
| , least HOT: STAOBY within the next 6 hours and in COLD SHUTDOWN within-tl.e following 30 hours, and (2) reenergize the A.C. vital L .I bus from its associated inverter connected to its associated D.C.
| |
| bus withis P4 hours'or be in at least H0T STANDBY within the next t 6 hours anfin COLQ' SHUTDOWN within the following 30 hours.
| |
| id w.
| |
| f'j '
| |
| a ,
| |
| 6 s WOLF CREEK - UNIT 1 3/4 8-13 t .
| |
| L .. -. - - - -
| |
| | |
| ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| : c. With one D.C. bus not energized from its associated battery bank or charger, reenergize the D.C. bus from its associated battery bank and charger within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 8-14
| |
| | |
| ELECTRICAL POWER SYSTEMS A
| |
| ( ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, one of the following divisions of electrical busses shall be energized in the specified manner:
| |
| : a. Division 1, consisting of:
| |
| : 1) 4160-Volt Emergency Bus #NB01, and C
| |
| : 2) 480-Volt Emergency Busses #NG01, NG03 and NG05E, and
| |
| : 3) 120-Volt A.C. Vital Busses #NN01 and NN03 energized from their associated inverter connected to D.C. Busses #NK01 and NK03, and
| |
| : 4) 125-volt D.C. Busses #NK01 and NK03 energized from Batteries
| |
| #NK11 and NK13 and Chargers #NK21 and NK23, or
| |
| : b. Division 2, consisting of:
| |
| : 1) 4160-Volt Emergency Bus #NB02, and
| |
| : 2) 480-Volt Emergency Busses #NG02, NG04 and NG06E, and
| |
| : 3) 120-Volt A.C. Vital Busses #NN02 and NN04 energized from their
| |
| ,A associated inverter connected to D.C. Busses #NK02 and NK04, and '
| |
| : 4) 125-Volt D.C. Busses #NK02 and NK04 energized from Batteries
| |
| #NK12 and NK14 and Chargers #NK22 and #NK24.
| |
| APPLICABILITY: MODES 5 and 6.
| |
| ACTION:
| |
| Without one of the above required divisions of electrical busses energized in the required manner, immediately suspend all operations involving CORE ALTERA-TIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to energize at least one division of the required busses in the specified manner.
| |
| SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
| |
| O WOLF CREEK - UNIT 1 3/4 8-15
| |
| | |
| ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 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 hours, 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 Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out, or removed, or
| |
| : b. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE:
| |
| : a. At least once per 18 months:
| |
| : 1) By verifying 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 verifying that each relay and associated circuit breakers and control circuits function as designed and as specified in Table 3.8-1, and WOLF CREEK - UNIT 1 3/4 8-16
| |
| | |
| J ELECTRICAL POWER SYSTEMS w
| |
| 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 representative sample of at least 10% of each type of lower voltage circuit breakers.
| |
| Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of injecting a current in excess 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 sample of at least 10% of all the circuit breakers of the inoperable type f --
| |
| shall also be functionally tested until no more failures are
| |
| ( found or all circuit breakers of that type have been functionally
| |
| \- tested; and
| |
| : 3) By selecting and functionally testing a representative sample of each type of fuse on a rotating basis. Each representative
| |
| ; sample of fuses shall include at least 10% of all fuses of that type. The functional test shall consist of a nondestructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria. Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. For each fuse found inoperable during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found or all fuses 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.
| |
| WOLF CREEK - UNIT 1 3/4 8-17
| |
| | |
| TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT 13.8-kV Switchgear P-252PA0107 Reactor Coolant Pump B-252PA0110/252PA0101 DPBB01A P-252PA0108 Reactor Coolant Pump B-252PA0110/252PA0101 DPBB01B P-252PA0205 Reactor Coolant Pump B-252PA0211/252PA0202 DPBB01C P-252PA0204 Reactor Coolant Pump B-252PA0211/252PA0202 DPBB01D 480-V Load Center P-125A Fuse at GS01A Hydrogen Recombiner B-52NG0304 SGS01A P-125A Fuse at GS01B Hydrogen Recombiner B-52NG0404 SGS01B P-52PG2102 Pressurizer Backup Through 52PG2111 Heaters B-250 A Fuse P-52PG2202 Pressurizer Backup Through 52PG2211 Heaters B-250 A Fuse P-52NG01TAF1 Containment Cooler B-52NG0108 DSGN01A P-52NG03TAF1 Containment Cooler B-52NG0305 DSGN01C P-52NG02TAF1 Containment Cooler B-52NG0208 DSGN01B P-52NG04TAF1 Containment Cooler B-52NG0405 DSGN01D P-52PG2007 PG20P MCC B-52PG2001 WOLF CREEK - UNIT 1 3/4 8-18
| |
| | |
| i ,3 TABLE 3.8-1 (Continued) l )
| |
| V CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED lN_ UMBER AND LOCATION EQUIPMENT 480-V Load Center (Continued)
| |
| P-52PG2402 at BB02 Pressurizer Heater B-500A' Fuse at 8801 P-52PG2403 at 8802 Pressurizer Heater B-500A Fuse at BB01 P-52PG2404 at BB02 Pressurizer Heater B-500A Fuse at BB01 1
| |
| P-52PG2405 at BB02 Pressurizer Heater B-500A Fuse at BB01 P-52PG2406 at BB02 Pressurizer Heater B-500A Fuse at BB01 P-52PG2407 at BB02 Pressurizer Heater B-500A Fuse at BB01 V 480-V Motor Control Center P-52NG01BDF3 RHR Loop Inlet Iso B-40A Fuse Viv EJHV8701B P-52NG02BHR2 ESW from Ctmt Air B-15A Fuse Coolers Iso Viv EFHV46 4
| |
| P-52NG02BDF2 CCW to Ctmt Iso Viv B-30A Fuse EGHV60
| |
| +
| |
| P-52NG01BHF3 CCW Containment Isolation Valve B-40A Fuse EGHV132 P-52NG01BDF1 PRT to Containment Valve B-15A Fuse BBHV8037A P-52NG01BBR3 Pressurizer Relief Valve B-30A Fuse BBHV8000A P-52PG19NEF5 Reactor Coolant Drain Tank Pump B-100A Fuse DPHB02A P-52PG19NCR3 Lighting Transformer B-150A Fuse at PG19NCR4 XQA26
| |
| (
| |
| WOLF CREEK - UNIT 1 3/4 8-19
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR O
| |
| OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT 480-V Motor Control Center (Continued)
| |
| P-52PG19NCR5 Machine Rm Exhaust Fan B-15A Fuse DCGN04 P-5A Fuse - Total of 4 Fuses Flux Mapping Motor Starters B-52PG19GDF3 SR05A P-5A Fuse - Total of 4 Fuses Flux Mapping Motor Starters B-52PG19GDF6 SR058 P-52NG03CLF2 RCP thermal barrier B-15A Fuse return isolation Vlv BBHV16 P-52NG01BHF4 CCW Containment Isolation Valve B-40A Fuse EGHV130 P-52NG02BBF4 Hydrogen Mixing Fan B-250A Fuse DCGN03B P-52NG02BCF3 Hydrogen Mixing Fan B-250A Fuse DCGN03D P-52NG01BBF4 Hydrogen Mixing Fan B-250A Fuse DCGN03A P-52NG01BCF3 Hydrogen Mixing Fan B-250A Fuse DCGN03C P-52NG02BJF5 CRDM Cooling Fan B-150A Fuse DCGN01B P-52PG20 GAR 2 CRDM Cooling Fan B-150A Fuse DCGN01A P-52PG20NFR3 Pressurizer Cooling Fan B-15A Fuse DCGN05 P-52PG20NEF5 Reactor Coolant B-100A Fuse Drain Tank Pump DPHB028 P-52PG20NEF1 Lighting Transformer B-150A Fuse at PG20NFF1 XQA28 WOLF CREEK - UNIT 1 3/4 8-20 i
| |
| | |
| 4 TABLE 3.8-1 (Continued)
| |
| ~
| |
| , CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT 480-V Motor Control Center (Continued)
| |
| P-52NG02BJF1 Standby Lighting B-40A Fuse P-52NG02BJF2 Standby Lighting B-40A Fuse P-52NG02BHR3 PRT to Containment Vlv B-15A Fuse BBHV8037B P-52NG02BDF1 Pressurizer Relief Valve B-30A Fuse BBHV80008 P-52NG01BJF5 CRDM Cooling Fan B-150A Fuse at NG01BKF3 DCGN01D O
| |
| (j P-52PG19GBR2 B-150A Fuse at PG19GFR1 CRDM Cooling Fan DCGN01C P-52NG02BDF3 Ctat Iso V1v Ret from B-15A Fuse Thrm Barrier Cooling Coil EGHV62 P-52NG02BEF2 Sump to RHR Pump B-60A Fuse Viv EJHV88118 P-52NG02BEF3 Ctmt Recirc Sump B-15A Fuse Iso V1v ENHV7 i
| |
| P-52NG02BGF3 Accumulator Iso Viv B-60A Fuse EPHV8808B l
| |
| P-52NG02BHF2 Accumulator Iso Viv B-60A Fuse EPHV8808D l P-52NG02BFF3 H2 Control System Make-up Air V1v KAHV30 B-15A Fuse P-52NG01BBF3 RC Pump Seal Water B-15A Fuse Iso Viv BGHV8112 P-52NG01BFF3 Sump to RHR Pump B-60A Fuse V1v EJHV8811A
| |
| (
| |
| l WOLF CREEK - UNIT 1 3/4 8-21
| |
| | |
| TABLE 3.8-1(Continued]
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT 480-V Motor Control Center (Continued)
| |
| P-52NG018EF3 Ctmt Recirc Sump B-15A Fuse Iso V1v ENHV1 P-52NG01BGF3 Accumulator Iso Viv B-60A Fuse EPHV8808A P-52NG01BGF2 Accumulator Iso Viv B-60A Fuse EPHV8808C P-52NG01BFF2 Ctmt Air to Aux Bldg B-15A Fuse ESF Filter Iso Viv GSHV20 .
| |
| P-52NG01BBR2 React Bldg Discharge B-15A Fuse Iso V1v LFFV95 P-52NG02BBF3 RHR Loop Inlet B-40A Fuse Is'o Viv BBPV8702B P-52NG02BCF2 RHR Loop Inlet B-40A Fuse Iso Vlv BBPV8702A P-52NG02BHF3 ESW to Ctmt Air B-15A Fuse Coolers Iso Viv EFHV34 P-52NG01BCF2 ESW to Ctmt Air B-15A Fuse Coolers Iso V1v EFHV33 P-52NG01BDF2 ESW from Ctmt Air B-15A Fuse Coolers Iso Viv EFHV45 P-52NG01BEF2 RHR Loop Inlet Iso Viv B-40A Fuse EJHV8701A P-52NG03CDF4 RCP Thermal Barrier B-15A Fuse CCW Iso Valve BBHV13 P-52NG03CHF1 RCP Thermal Barrier B-15A Fuse CCW Iso Viv BBHV14 P-52PG19NAF4 Reactor Cavity Cooling B-150A Fuse Fan DCGN02A P-52PG19NCF3 Ctmt Atmospheric Control B-60A Fuse System Fan DCGR01A WOLF CREEK - UNIT 1 3/4 8-22
| |
| | |
| ,. p TABLE 3.8-1 (Continued) :
| |
| , . CONTAINMENT-PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 1 PROTECTIVE. DEVICE. POWERED-NUMBER AND LOCATION EQUIPMENT
| |
| ; 480-V Motor Control Center (Continued) i P-52PG19NGF2' RCP A Space Heater F B-40 Fuse ,
| |
| P-52PG19NGF3 RCP B Space Heater B-40 Fuse P-52PG19NEF1' RCP A 011 Lift Pump
| |
| .B-40A Fuse ,
| |
| t
| |
| - P-52PG19NGR3' RCP B Oil Lift Pump
| |
| , B-40A Fuse P-52PG19NFF1 ~ Ctat Normal Sump.
| |
| B-15A Fuse Pump DPLF05A P-52PG19NFF2 Ctmt Normal Sump
| |
| : B-15A Fuse Pump DPLF05C
| |
| , P-52PG19NAF2' Instrument Tunnel
| |
| ! B-25A' Fuse . Sump Pump DPLF07A'
| |
| , P-52NG03CBF4 RCP Thermal. Barrier CCW ,
| |
| ; B-15A Fuse Iso V1v BBHV15 4
| |
| -P-52PG20NCR3-B-150A Fuse Reactor Cavity Cooling Fan DCGN028 P-52PG20NFF4 Ctat Atmospheric Control B-60A Fuse- System Fan DCGR018 l-P-52PG20NBF1 '
| |
| RCP C Space Heater B-40A Fuse P-52PG20NCF1 RCP D Space Heater r B-40A Fuse P-52PG20NFF3 RCP C Oil Lift Pump B-40A' Fuse i 1EPR08C P-3A Fuse Accumulator Tank A Isol V1v l .- RP139 B-3A Fuse _ EPHV8808A 1EPR080 P-3A Fuse Accumulator Tank C Isol V1v
| |
| :. RP139 B-3A Fuse EPHV8808C WOLF CREEK - UNIT 1 3/4 8-23 '
| |
| i
| |
| -m. .. ._1_._. ...1.. . _ . . . . _ . _ _ _ _ _ _ _ . _ _ _ . . . _ . _ _ , _ , . _ _ . . . _ . . _ _ . . _ _ _ . _ , _ _ _ . . . _ _ _ . _ _ . _
| |
| | |
| BBLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT 480-V Motor Control Center (Continued) 4EPR08A P-3A Fuse Accumulator Tank B Isol Vlv RP140 B-3A Fuse EPHV8808B 4EPR08B P-3A Fuse Accumulator Tank D Isol Viv RP140 B-3A Fuse EPHV8808D 1EPK09B P-3A Fuse Accumulator Tank B Vent Viv RL018 B-3A Fuse EPHV89508 IEPK09D P-3A Fuse Accumulator Tank C Vent Viv RL018 B-3A Fuse EPHV8950D 1EPK09F P-3A Fuse Accumulator Tank D Vent Vlv RL018 B-3A Fuse EPHV8950F 4EPK09A P-3A Fuse Accumulator Tank A Vent Viv RL018 B-3A Fuse EPHV8950A 4EPK09C P-3A Fuse Accumulator Tank B Vent Viv RL018 B-3A Fuse EPHV8950C 4EPK09E P-3A Fuse Accumulator Tank C Vent Viv RL018 B-3A Fuse EPHV8950E P-52PG20NFF2 RCP D Oil Lift Pump B-40A Fuse P-52PG20NER2 Ctmt Normal Sump B-15A Fuse Pump DPLF058 P-52PG20NGF4 Ctmt Normal Sump B-15A Fuse Pump DPLF05D P-52PG20NDR2 Instrument Tunnel B-25A Fuse Sump Pump DPLF07B P-52PG1904 Polar Crane B-600A Fuse HKE13 CRDM Control Rod Drive Power P-10A Fuse Gripper Coils (104 fused B-30A Fuse circuits)
| |
| WOLF CREEK - UNIT 1 3/4 8-24
| |
| | |
| - - . . .-- .- .. = _ - - . . .
| |
| t TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR
| |
| ):
| |
| OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED.
| |
| NUMBER AND LOCATION EQUIPMENT i
| |
| CRDM Control Rod Drive Power (Continued)
| |
| P-50A Fuse Lift Coils (52 fused B-150A Fuse circuits)
| |
| Low Voltage Power and Control 5BBG02A P-2A Fuse RCP Oil Lift l PG19NEF1 B-1A Fuse Pump Control
| |
| ; 4GTK038 P-3A Fuse Cont. Mini Purge Isol Damper RLO20 B-3A Fuse GTHZ11 1 4GTK030 P-3A Fuse Cont. Purge Isol Damper j .RL020 B-3A Fuse GTHZ8 1
| |
| < 5BBA01A P-15A Fuse RCP Breaker PA0107 B-15A Fuse Control 5BBA01B P-15A Fuse RCP Breaker PA0108 B-15A Fuse- Control 6BBAdiCP-15AFuse RCP Breaker PA0205 B-15A Fuse Control l' 6BBA010 P-15A' Fuse 4- RCP Breaker PA0204 B-15A Fuse Control 6HBK04B P-3A Fuse RCDT Heat Exchanger return to RCDT Viv HB115 B-3A Fuse HBHV7144 6HBK05A P-3A Fuse RCDT Discharge Viv HB115 B-3A Fuse HBHV7143 1HBK19A P-3A Fuse Containment Isolation Viv RLO21 B-3A Fuse HBHV7176 i 4KAG04A P-2A Fuse Hydrogen Purge Makeup Air NG02BFF3 B-1A Fuse Supply Viv KAHV30 SKCQ15S P-3A Fuse Fire Protection Discharge Viv KC274A B-3A Fuse KCXV261 6KCQ155 P-3A Fuse Fire Protection Discharge Viv KC274A B-3A Fuse KCXV262 WOLF CREEK - UNIT 1- 3/4 8-25
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 5KCQ19X P-3A Fuse Fire Protection Detector KC274A B-3A Fuse KCHPS261-002 6KCQ19Y P-3A Fuse Fire Protection Detector KC274A B-3A Fuse KCHPS262-002 6KES01A P-5A Fuse Fuel Transfer Panel KE124 B-15A Fuse KE124 6KES01A P-5A Fuse Fuel Transfer Panel KE125 B-15A Fuse KE125 SLFG06A P-2A Fuse Containment Normal Sump Pump A PG19NFF1 B-1A fuse DPLF05A SLFG06C P-2A Fuse Containment Normal Sump Pump C PG19NFF2 B-1A Fuse DPLF05C 6LFG06B P-2A Fuse Containment Normal Sump Pump B PG20NER2 B-1A Fuse DPLF05B 6LFG06D P-2A Fuse Containment Normal Sump Pump D PG20NGF4 B-1A cuse DPLF05D 5GNG03A P-5A Fuse CRDM Cooling Fan D Discharge NG01BJF5 B-3A Fuse Isolation Damper GNHZ44 SGNG03C P-5A Fuse CRDM Cooling Fan C PG19GBR2 B-3A Fuse Discharge Isolation Damper GNHZ43 1EMK048 P-3A Fuse SIS Test Line Viv RL017 B-3A Fuse EMHV8824 1EMK04D P-3A Fuse SIS Test Line V1v RL017 B-3A Fuse EMHV8881 4EMK04C P-3A Fuse BIT Test Line Viv RL017 B-3A Fuse EMHV8843 4EMK04E P-3A Fuse SI Test Line V1v RL017 B-3A Fuse EMHV8871 WOLF CREEK - UNIT 1 3/4 8-26
| |
| | |
| _ TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE' DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power-and Control (Continued)
| |
| SEMK05A P-3A Fuse SI Test Line V1v RL017 B-3A Fuse EMHV8889A SEMK05C P-3A Fuse SI Test Line V1v RL017 B-3A Fuse EMHV8889C SEMK05E P-3A Fuse BIT Test Line V1v RL017 B-3A Fuse EMHV8882 6EMK05B P-3A Fuse SI Test Line V1v RL017. B-3A Fuse EMHV8889D 6EMK05D P-3A Fuse SI Test Line V1v RL017 B-3A Fuse EMHV88898 1ENG02A P-2A Fuse Containment Spray Sump Isol V1v
| |
| ['N) NG01BEF3 8-1A Fuse ENHV1
| |
| (
| |
| 4ENG028 P-2A Fuse Containment Spray Sump Isol V1v NG02BEF3 B-1A Fuse ENHV7 1EPG02A P-2A Fuse Accumulator Tank Isolation Valve NG01BGF3 B-1A Fuse EPHV8808A 1EPG028 P-2A Fuse Accumulator Tank Isolation Valve NG01BGF2 B-1A Fuse EPHV8808C 1EPK02C P-3A Fuse Accumulator Tank Isolation Valve RL018 B-3A Fuse EPHV8808B Indication IEMK04A P-3A Fuse SI Test Line V1v RL017 B-3A Fuse EMHV8823 4EJG04B P-2A Fuse RHR to charging /SI pump suctions NG02AFR3 B-1A Fuse EJHV88048 1EJG05A P-2A Fuse RHR Shutdown Suction Line Isol NG01BEF2 B-1A Fuse Valve EJHV8701A SB P-2A Fuse RHR Shutdown Suction Line Isolation NG01BDF3 B-1A Fuse Valve EJHV8701B v
| |
| WOLF CREEK - UNIT 1 3/4 8-27
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 1EJG06A P-2A Fuse Cont Recirc Sump Isolation Valve NG01BFF3 B-1A Fuse EJHV8811A 4EJC06B P-3A Fuse Cont Recirc Sump Isolation Valve NG02BEF2 B-2A Fuse EJHV88118 1EJK07A P-3A Fuse Test Line Isol V1v Hot Leg Inj RL017 B-3A Fuse Line Solenoid EJHCV8825 1EJK07C P-3A Fuse RHR Test Line V1v RL017 B-3A Fuse EJHCV8890A 4EJK078 P-3A Fuse RHR Test Line Viv RL017 B-3A Fuse EJHCV8890B P-1EJY13A 3A Fuse Ctmt Sump Sample Isolation V1v '
| |
| RL011 EJHV21 B-1RLY01G 15A Breaker NG01ACR119 P-4EJY13B 3A Fuse Ctmt Sump Sample Isolation V1v RLO11 EJHV22 B-4RLY01G 15A Breaker NG02ACR140 P-4BMY010 3A Fuse S.G.C Cnt to Nuc Sample Sys V1v RL024 BMHV22 B-4RLY01H 15A Breaker NG02ACR127 P-4BMY02A 3A Fuse S.G.A Tube Sheet Sample V1v RLO24 BMHV35 B-4RLY01H ISA Breaker NG02ACR127 P-4BMY028 3A Fuse S.G.B Tube Sheet Sample Viv RLO24 BMHV36 B-4RLY01H 15A Breaker NG02ACR127 P-4BMY02C 3A Fuse S.G.C Tube Sheet Sample Viv RLO24 BMHV37 B-4RLY01H 15A Breaker NG02ACR127 WOLF CREEK - UNIT 1 3/4 8-28
| |
| | |
| .g TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| P-4BMYO2D 3A Fuse S.G.D Tube Sheet Sample Viv RLO24 . BMHV38
| |
| .B-4RLY01H 15A Breaker NG02ACR127 1BNG03A P-2A Fuse RHR Pump RWST Suction Valve NG01ACR2 B-1A Fuse BNHV8812A 48NG03B P-2A Fuse RHR Pump RWST Suction Valve NG02AFF4 B-1A Fuse BNHV88128 1EFG09A. P-2A Fuse Cont Cooler Isolation Valve NG01BCF2 B-1A Fuse EFHV33 1EFG09C P-2A Fuse Cont Cooler Isolation Valve NG01BDF2 B-1A Fuse EFHV45 4EFG098 P-2A Fuse Cont Cooler Isolation Valve NG02BHF3 B-1A Fuse EFHV34
| |
| '4EFG090 P-2A Fuse Cont Cooler Isolation Valve-NG02BHR2 B-1A Fuse EFHV46 4 EGG 06A P-2A Fuse RC Pump CCW Return Cont Isol V1v NG02BDF2 B-1A Fuse EGHV60 l' 4 EGG 10A P-2A Fuse Cont Isol V1v CCW Return from NG02BDF3 B-1A Fuse RC pump Ther Barr V1v EGHV62 i
| |
| 1 EGG 17A P-2A Fuse EGHV-60 Bypass Valve Cont Isol Viv NG01BHF4 B-1A Fuse EGHV130
| |
| , .1 EGG 17B P-2A Fuse EGHV-62 Bypass Valve Cont Isol.V1v
| |
| ! NG01BHF3 B-1A fuse EGHV132 1EJG04A P-2A Fuse RHR to Charging /SI Pump Suction V1v i
| |
| NG03CNF4 B-1A Fuse EJHV8804A 1BBK35A P-3A Fuse Excess Letdown Path to PRT Isol Viv RL001- .B-3A Fuse BBHV8157A 4BBK358 P-3A Fuse Excess Letdown Path to PRT Isol.Viv t x RL001 B-3A Fuse BBHV8157B WOLF CREEK - UNIT 1 3/4 8-29
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 1BBK37A P-3A Fuse Pressurizer Safety V1v RL021 B-3A Fuse BBHV8010A 1BBK37B P-3A Fuse Pressurizer Safety V1v RLO21 B-3A Fuse BBHV80108 4BBK37C P-3A Fuse Pressurizer Safety V1v RLO21 B-3A Fuse BBHV8010C IBBG39A P-2A Fuse Pressurizer PORV Isolation Valve NG01BBR3 B-1A Fuse BBHV8000A SGRK02A P-3A Fuse Filtration Unit Damper RLO20 B-3A Fuse GRPDZ5 6GRK028 P-3A Fuse Filtration Unit Damper RLO20 B-3A Fuse GRPDZ15 1GTK03A P-3A Fuse Ctmt Minipurge Isol Valve RLO20 B-3A Fuse GTHZ5 P-1GSY010 3A Fuse Hydrogen Analyzer Ctmt Isol Viv RL011 GSHV13 B-1RLY01G 15A breaker NG01ACR119 P-1GSY01E 3A Fuse Hydrogen Analyzer Ctmt Isol V1v RL011 GSHV14 B-1RLY01G 15A Breaker NG01ACR119 P-1GSY10A 3A Fuse Ctmt Atmosphere Monitor Isol Viv RL020 GSHV31 B-1RLY01A 15A Breaker NG01ACR123 P-1GSY108 3A Fuse Containment Atm Monitor Isol V1v RL020 GSHV34 B-1RLY01A 15A Breaker NG01ACR123 P-65QY01A 15A Breaker Loose Parts Simulator PG20GBR238 B-6SQY01A 30A Fuse PG20GBR1 WOLF CREEK - UNIT 1 3/4 8-30
| |
| | |
| t TABLE-3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
| |
| ' PROTECTIVE DEVICE: POWERED-2 NUMBER AND LOCATION, EQUIPMENT ,
| |
| i Low Voltage Power and Control (Continued) ,
| |
| P-6BBY16A 15A Breaker Reactor Coolant Sys Level Alarm i PG20GBR239
| |
| ; B-6BBY16A 30A Fuse
| |
| [
| |
| - PG20GBR1 P-6GNY09F 15A Breaker CRDM Cooling Fan Space Heater j PG20NBR225 B-6GNYO9F 30A-Fuse
| |
| [ . PG20NBR1 :
| |
| l ' P-6GNY09H 15A Breaker CRDM Cooling Fan Space Heater !
| |
| ; PG20NBR226 B-6GNY09H 30A Fuse PG20NBR1
| |
| - P-6GNYO9K-15A Breaker. Reactor Cavity-Cooling Fan
| |
| : PG20NBR223 .
| |
| Space Heater F - B-6GNYO9K 30A Fuse
| |
| : PG20NBR1
| |
| - P-4EPY02C 15A Breaker Accumulator Isolation Valve-
| |
| - NG028AR114 Space Heater
| |
| . B-4EPY02C 30A Fuse
| |
| ; NG02BGR4
| |
| . - P-4EPY020 15A Breaker- Accumulator Isolation Valve NG02BAR115 Space Heater
| |
| ;. B-4EPY020 30A Fuse NG02BGR4
| |
| . - . P-4GNY098 15A Breaker Containment Cooler Fan Space NG02BAR110 Heater
| |
| ; B-4GNYO98 30A Fuse NG02BGR4 P-4GNYO90 15A Breaker Containment Cooler Fan Space
| |
| . NG02BAR111 Heater i: B-4GNY09D 30A Fuse
| |
| , NG02BGR4 I
| |
| i P-4GNYO9M 15A Breaker NG02BAR120 Hydrogen Mixing Fan Space Heater-
| |
| ! B-4GNYO9M 30A Fuse i s NG02BGR4 I
| |
| ' WOLF CREEK - UNIT 1 3/4 8-31 i
| |
| 'PM+t* v e-rue-.u-ep-,.w. e--,.erw wg,e- m.g.%-par ge$ %ew,yes,-, , , . -- ----,---*--c.--==yq.--ewp.----qm-ypr7^TPtr'W-WWW"-'EE'*E'WN6 P-VT''' ' ''T'-'N'W''''Y''"7"F-W '-''T-'W-''7'-T-'-W'*W"
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| P-4GNY09P 15A Breaker Hydrogen Mixing Fan Space NG02BAR121 Heater B-4GNY09P 30A Fuse NG02BGR4 P-1GNY09A 15A Breaker Containment Cooler Fan Space NG018AR119 Heater B-1GNY09A 30A Fuse NG01BER3 P-1GNYO9C 15A Breaker Containment Cooler Fan Space NG01BAR120 Heater B-1GNY09C 30A Fuse NG01BER3 P-1GNY09L 15A Breaker Hydrogen Mixing Fan Space NG01BAR122 Heater B-1GNY09L 30A Fuse NG01BER3 P-1GNY09N 15A Breaker Hydrogen Mixing Fan Space NG018AR123 Heater B-1GNY09N 30A Fuse NG018ER3
| |
| ; P-5GNYO9E 15A Breaker CRDM Cooling Fan Space Heater PG19NHF228 B-5GNY09E 30A Fuse PG19NHF1 i
| |
| P-1EPY02A 15A Breaker Accumulator Iso NG01BAR116 Valve Space Heater B-1EPY02A 30A Fuse NG01BER3 P-1EPY028 15A Breaker Accumulator ISO NG01BAR117 Valve Space Heater B-1EPY028 30A Fuse NG01BER3 P-4GSY10C 3A Fuse Ctmt Atm Monitor Isol Viv RL020 GSHV36 B-4RLY01A 15A Breaker NG02ACR130 WOLF CREEK - UNIT 1 3/4 8-32
| |
| | |
| TABLE 3.8-1 (Continued) b y / CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| P-4GSY10D 3A Fuse Ctmt Atm Monitor Isol Viv RLO20 GSHV39 B-4RLY01A ISA Breaker NG02ACR130 P-5GNY09G 15A Breaker CRDM Cooling Fan Space Heater PG19NHF229 B-5GNY09G 30A Fuse PG19NHF1 P-5GNY09J 15A Breaker Reactor Cavity Cooling Fan Space PG19NHF225 . Heater B-5GNY09J 30A Fuse PG19NHF1 SSFY11AA P-30A Fuse Rod Position Panel SF109A PN0711 B-60A Fuse
| |
| \O SSFY11AB P-30A Fuse Rod Position Panel SF109B PN0710 B' Fuse ILFG08A P-2A Fuse Normal sump Ctmt Isol V1v NG01BBR2 B-1A Fuse LFFV95 5LFG15A P-2A Fuse Instrument Tunnel Sump Pump PG19NAF2 B-1A Fuse DPLF07A
| |
| '6LFG15B P-2A Fuse Instrument Tunnel Sump Pump PG20NDR2 B-1A Fuse DPLF078 1GSG03A P-2A Fuse Hydrogen Purge Ctat Isolation Viv l- NG018FF2 B-1A Fuse GSHV20 I
| |
| I 4BBG39B P-2A Fuse Pressurizer PORV Isol Viv-NG02BDF1 B-1A Fuse BBHV8000B 1BBK40A P-15A Fuse PZR PORV NK5108 B-15A Fuse BBPCV455A 4BBK408 P-15A Fuse PZR PORV NK4421 B-15A Fuse BBPCV456A
| |
| /' 5BGK048 P-3A Fuse Alternate Charging Path Isol Valv
| |
| (,,,)
| |
| / RL001 B-3A Fuse BGHV8147 WOLF CREEK - UNIT 1 3/4 8-33
| |
| , ,- ,,,,,-.-p -. - - , - --n-,,, ,n,,--v - c.
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 6BGK04A P-3A Fuse Normal Charging Path Isol Valv RL001 B-3A Fuse BGHV8146 P-5LFY10A 3A Fuse Containment Cooler Drain Valve RL023 LFLV97 B-5RLY01H 15A Breaker PG19GCR217 P-5LFY10C 3A Fuse Containment Cooler Drain Valve RL023 LFLV99 B-5RLY01H 15A Breaker PG19GCR217 P-6LFY10B 3A Fuse Containment Cooler Drain Valve RLO23 LFLV98 B-6RLY01G 15A Breaker PG20GBR217 P-6LFY10D 3A Fuse Containment Cooler Drain Valve RLO23 LFLV100 B-6RLY01G 15A Breaker PG20GBR217 P-6LFY17A 3A Fuse Refueling Pool Stand Pipe RLO23 Discharge Valve B-6RLY01G 15A Breaker LFLV122 PG20GBR217 i
| |
| P-5LFY20A 15A Breaker Instrument Tunnel Sump Moisture PG19NHF224 Sensor B-5LFY20A 30A Fuse TLVF01 PG19NHF1 P-6LFY20B 15A Breaker Instrument Tunnel Sump Moisture PG20NBR216 Sensor B-6LFY20B 30A Fuse TVLF02 PG20NBR1 P-5SDYO6C 15A Breaker Local Radiation Monitor Power PG19NHF215 Supplies B-SSDYO6C 30A Fuse SPRIA39-42 PG19NHF1 O
| |
| WOLF CREEK - UNIT 1 3/4 8-34
| |
| | |
| ~
| |
| ^
| |
| TABLE 3.'8-1 (Continued)
| |
| CONTAINMENT. PENETRATION CONDUCTOR OVERCURRENT- PROTECTIVE DEVICES-PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)- _
| |
| P-1SJY010 :3A' Fuse ~ Press. Ctat Isol Viv RL011 - .
| |
| SJHV128 B-1RLYO1G~15A Breaker NG01ACR119-P-4SJY01A 3A Fuse Press. Liq /HL 1&3 Sample Clr Viv RL011- SJHV5 B-4RLYO1G 15A Breaker NG02ACR140 1GTK03C P-3A Fuse- Ctat Purge Isol V1v RL020 B-3A' Fuse- GTHZ7
| |
| .P-1GSYO1F 3A Fuse : Hydrogen. Anal Samp Return to Ctat V1v RLO11 GSHV18 B-1RLY01G 15A Breaker-NG01ACR119 P-4GSY01A~3A._ Fuse Hydrogen Anal Ctat Sample'Viv RL011- GSHV4 B-4RLYO1G
| |
| 'NG02ACR140 15A Breaker P-4GSY01B 3A Fuse Hydrogen Anal Ctat Sample Viv RL011' GSHV5 B-4RLY01G 15A Breaker NG02ACR140 P-4GSY01C 3A Fuse Hydrogen and Samp Return to Ctat Viv RLO11' GSHV9
| |
| ;. 'B-4RLY01G 15A Breaker
| |
| :- NG02ACR140 i 5BBG028 P-2A Fuse RCP Oil Lift
| |
| _PG19NGR3 B-1A Fuse Pump Control 6BBG02C P-2A Fuse RCP Oil Lift PG20NFF3 B-1A Fuse Pump Control 6BBG02D'P-2A Fuse RCP Oil Lift PG20NFF2 B-1A Fuse Pump Control 1BBG03A P-2A' Fuse' RCP Thermal Barrier Cooler Isol Viv NG03CDF4 8.1A Fuse BBHV13 o
| |
| L WOLF-CREEKI -UNIT 1 3/4 8-35
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| IBBG038 P-2A Fuse RCP Thermal Barrier Cooler Isol Viv NG03CHF1 B-1A Fuse BBHV14 1BBG03C P-2A Fuse RCP Thermal Barrier Cooler Isol Vlv NG03CBF4 B-1A Fuse BBHV15 1BBG03D P-2A Fuse RCP Thermal Barrier Cooler Isol V1v NG03CLF2 B-1A Fuse BBHV16 5BBK05A P-3A Fuse RCP Seal Water Return V1v RL001 B-3A Fuse BBHV8141A SBBK05B P-3A Fuse RCP Seal Water Return Viv RL001 B-3A Fuse BBHV8141B 6BBK05C P-3A Fuse RCP Seal Water Return Viv RL001 B-3A Fuse BBHV8141C 6BBK05D P-3A Fuse RCP Seal Water Return Vlv RL001 B-3A Fuse BBHV81410 6BBK07A P-3A Fuse PRT Discharge to RCDT V1v RL021 B-3A Fuse BBHV8031 6BBK07B P-3A Fuse Reactor Makeup Water to PRT Viv RL021 B-3A Fuse BBHV8045 6BBK08A P-3A Fuse RCDT Ht. Exch. to PRT V1v HB115 B-3A Fuse BBHV7141 6BBK09A P-3A Fuse RV Flange Leakoff Line to RCDT Viv RLO21 B-3A Fuse BBHV8032 1BBK30C P-3A Fuse Reactor Vessel Head Vent Vlv RLO21 B-3A Fuse EBHV8002A 4BBK300 P-3A Fuse Rx Vessel Head Vent Viv .
| |
| RL021 B-3A Fuse BBHV80028 l I
| |
| 1BBG31A P-2A Fuse PRT Emergency Drain Line Viv !
| |
| NG01BDF1 B-1A Fuse BBHV8037A 4BBG31B P-2A Fuse PRT Emergency Drain Line V1v NG02BHR3 B-1A Fuse BBHV80378 WOLF CREEK - UNIT 1 3/4 8-36
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| .j p) CONTAINMENT PENETRATION CONDUCTOR V
| |
| OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED
| |
| . NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 1EPK02D P-3A Fuse Accumulator Tank D Isol Viv RL018 B-3A Fuse EPHV8808D Indication IEPK02E P-3A Fuse Accumulator Tank A Isol Viv RL018 B-3A Fuse EPHV8808A Indication 1EPK02F P-3A Fuse Accumulator Tank C Isol V1v RL018 B-3A Fuse EPHV8808C Indication 4EPG02C P-2A Fuse Accumulator Tank B Isol Viv NG02BGF3 B-1A Fuse EPHV8808B Indication 4EPG02D P-2A Fuse Accumulator Tank D Isol V1v NG02BHF2 B-1A Fuse EPHV8808D Indication 4EPK02A P-3A Fuse Accumulator Tank A Isol V1v s
| |
| s RL018 B-3A Fuse EPHV8808A Indication
| |
| '\_- 4EPK02B P-3A Fuse Accumulator Tank C Isol V1v RL018 B-3A Fuse EPHV8808C Indication 4EPK02G P-3A Fuse Accumulator Tank B Isol V1v RL018 B-3A Fuse EPHV88088 Indication 4EPK02H P-3A Fuse Accumulator Tank D Isol Viv RL018 B-3A Fuse EPHV8808D Indication SEPK03A P-3A Fuse Accumulator Nitrogen Supply Viv RL018 B-3A Fuse EPHV8875A SEPK03B P-3A Fuse Accumulator Nitrogen Supply V1v l RL018 B-3A Fuse EPHV8875C
| |
| -SEPK03C P-3A Fuse Accumulator Test Line V1v l RL018 B-3A Fuse EPHV8877A 4BBK30B P-3A Fuse- Reactor Vessel Head Vent V1v RL021 B-3A Fuse BBHV8001B SEPK03D P-3A Fuse Accumulator Test line V1v RL018 B-3A Fuse EPHV8877C l
| |
| SEPK03E P-3A Fuse Accumulator Water Fill Line V1v l (''S g_) RL018 B-3A Fuse EPHV8878A WOLF CREEK - UNIT 1 3/4 8-37 l
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| SEPK04A P-3A Fuse Accumulator Water Fill Line V1v RL018 B-3A Fuse EPHV8878C SEPK04B P-3A Fuse Accumulator Test Line Viv RL018 B-3A Fuse EPHV8879A SEPK04C P-3A Fuse Accumulator Test Line Viv RL018 B-3A Fuse EPHV8879C 6EPK04D P-3A Fuse Accumulator Test Line V1v RL018 B-3A Fuse EPHV88798 6EPK04E P-3A Fuse Accumulator Test Line Viv RL018 B-3A Fuse EPHV8879D 6EPK05A P-3A Fuse Accumulator Nitrogen Supply V1v RL018 B-3A Fuse EPHV8875B 6EPK058 P-3A Fuse Accumulator Nitrogen Supply Viv RL018 B-3A Fuse EPHV8875D 6EPK05C P-3A Fuse Accumulator Test Line V1v RL018 B-3A Fuse EPHV88778 6EPK05D P-3A Fuse Accumulator Test Line V1v RL018 B-3A Fuse EPHV88770 6EPK05E P-3A Fuse Accumulator Water Fill Viv RL018 B-3A Fuse EPHV88788 6EPK05F P-3A Fuse Accumulator Water Fill Viv RL018 B-3A Fuse EPHV8878D P-4SJY01B 3A Fuse Press. Vapor. Cont. Iso. Space V1v.
| |
| RL011 SJHV12 B-4RLY01G 15A Breaker NG02ACR140 P-4SJY01C 3A Fuse Accums Sample Cont Isol Viv RLO11 SJHV18 B-4RLY01G 15A Breaker NG02ACR140 0
| |
| WOLF CREEK - UNIT 1 3/4 8-38
| |
| | |
| !. t
| |
| . . '' TABLE-3.8-1 (Continued) ;
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ;
| |
| i PROTECTIVE DEVICE POWERED i- NUpWER AND LOCATION EQUIPMENT
| |
| : 1. .
| |
| i Low Voltage Power and Control (Continued)
| |
| P-5SJYO3B 3A Fuse Accumulator Sample Line Viv RP211 _
| |
| SJHV16 ,
| |
| B-5RPYO90.15A~ Breaker
| |
| ; PG19NHF236 i
| |
| !' P-5SJYO3C 3A Fuse. Accumulator Sample Line Viv
| |
| ! RP211 .
| |
| SJHV17 B-5RPYO90 15A Breaker
| |
| .PG19NHF236 l
| |
| i P-SSJY048 3A Fuse Accumulator Sample-Line Viv i RP211 SJHV14 1 i .B-5RPYO90 15A. Breaker
| |
| * i PG19NHF236 l'
| |
| l P-5SJYO4C 3A Fuse Accumulator Sample'Line Viv l; <
| |
| RP211 SJHV15 m :
| |
| ! B-5RPYO90 15A Breaker !
| |
| PG19NHF236
| |
| ! , P-1SJYO68 3A Fuse HL Sample 3 Viv
| |
| ! .RP332 .
| |
| . SJHV4- ,
| |
| B-1RPY09F 15A Breaker
| |
| : a NG01BAR140 '
| |
| [
| |
| P-4SJYO6A 3A Fuse HL' Sample 1 V1v -
| |
| RP333 SJHV3 l B-4RPYO9F 15A Breaker NG028AR140 l P-5SJYO6C 3A Fuse -
| |
| Press Liquid Space Samp Isol Viv I RP211 SJHV20 B-5RPYO90 15A Breaker '
| |
| PG19NHF236. .
| |
| [ : P-4BMY01A 3A Fuse S.G. A Out to Nuc Sample Sys Viv.
| |
| ; RLO24 BNHV19 [
| |
| B-4RLY01H 15A Breaker
| |
| , NG02ACR127 l _P-4BMY018 3A Fuse S.G. B Out to Nuc Sample Sys V1v
| |
| : RLO24 BNHV20 i- B-4RLYO1H 15A Breaker f NG02ACR127
| |
| ~
| |
| j WOLF CREEK - UNIT 1 3/4 8-39 [
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| P-4BMY01C 3A Fuse S.G. C Out to Nuc Sample Sys Viv RLO24 BMHV21 B-4RLY01H 15A Breaker NG02ACR127 P-5GNY08A 3A Fuse CRDM Cooling Discharge Damper RL020 GNHZ71 B-5RLY01L 15A Breaker PG19GCR230 P-5GNY08C 3A Fuse CRDM Cooling Discharge Damper RL020 GNHZ73 B-5RLY01L 15A Breaker PG19GCR230 P-6GNY08C 3A Fuse CRDM Cooling Discharge Damper RL020 GNHZ72 B-6RLY01J 15A Breaker PG20GBR222 P-6GNY08A 3A Fuse CRDM Cooling Discharge Damper RL020 GNHZ74 B-6RLY01J 15A Breaker PG20GBR222 5BGK10A P-3A Fuse Normal Letdown Isolation Vlv RL001 B-3A Fuse BGLCV459 5BGK10B P-3A Fuse Normal Letdown Isolation Viv RL001 B-3A Fuse BGLCV460 SBGK19A P-3A Fute Pressurizer Spray Vlv RL001 B-3A Fuse BGHV8145 6BGK20A P-3A Fuse Excess Letdown Line Isolation Viv RL001 B-3A Fuse BGHV8143 5BGK35A P-3A Fuse Letdown Orifice Isolation Viv RL001 B-3A Fuse BGHV8149A 5BGK35B P-3A Fuse Letdown Orifice Isolation Vlv RL001 B-3A Fuse BGHV8149B 5BGK35C P-3A Fuse Letdown Orifice Isolation Viv RL001 B-3A Fuse BGHV8149C WOLF CREEK - UNIT 1 3/4 8-40
| |
| | |
| . _ _ _ . , _ . _ , .. _ m . _ _ . - . - - - . . _ . .
| |
| Q 4 .
| |
| v . . ,, ,
| |
| . :p .
| |
| s a f '
| |
| ;/ .TABLf!3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR l' OVERCU.R. RENT PROTECTIVE DEVICES U' PROTECTIVE DEVICE < POWERED l NUMBER.AND-LOCATION te EQUIPMENT-
| |
| .[
| |
| ,7 Low Voltage Power and Control-(Continued)-
| |
| 1BGK36A-P-3A Ftse , , - Letdown Containment Isolation V1v RL001 B-3A Fuse' if BGHV8160 1 BGG 38A P-2A Fus'e ~
| |
| '/ T' Seal Water Ctat Isolation Viv-s-
| |
| S NG01BBF3 B-1A Fuse' *f'q *
| |
| - BGHV8112
| |
| / .
| |
| 1BGK48C P-3A' Fuse ',f ; -
| |
| Excess Letdown /RCS Isolation.V1v u RL001 B-3A' Fuse ,
| |
| s g BGHV8153A
| |
| ^
| |
| +
| |
| 'lBGK480 P-3A F'use / Excess Letdown /RCS Isolation Vlv
| |
| [/:
| |
| v RL001 B-3A Fuse '
| |
| BGHV8154A 4BGK48A P-3A Fuse Excess Letdown /RCS Isolation Viv RL001 B-3A Fuse j BGHV8153B
| |
| ~
| |
| 48GK488-P-3A Fuse s Excess Letdown /RCS Isolation V1v RL001 B-3A Fuse BGHV8154B
| |
| .A ,
| |
| .-4BBG12A P-2A Fuse. RHR Pump Suction Isolation Valve
| |
| '' NG02BCF2 B-1A Fuse < BBPV8702A
| |
| -4BBG128 P-2A Fuse RHR Loop 2 Inlet l'' .
| |
| NG02BBF3 B-1A Fuse-Isolation Valve BBPV8702B
| |
| ~, 1BBK13A P-3A Fuse i Ctat Isol Nitrogen Supply to PRT V1v
| |
| ' ~
| |
| , RLO21 B-3A Fuse < BBHV8026-5BBK14C'P-3A Fuse RCP' Standpipe Makeup Viv y
| |
| RLO21 B-3A Fuse BBLCV180
| |
| ! +
| |
| , 5BBK140:P-3A Fuse y ' RCP Standpipe Makeup Viv RLO2118-3A Fuse p BBLCV181 3
| |
| ,3 6BBK14A P-3A Fuse ' -
| |
| h f RCP Standpipe Makeup Vlv
| |
| 'RLO21 B-3A Fuse,. ,
| |
| BBLCV178-
| |
| ,.A -- 3 ,
| |
| o 'o 6BBK14B P-3A Fuse '.
| |
| RCP Standpipe Makeup V1v L RLO21 B-3A Fuse' BBLCV179 9 & 4' e h 5BBK15B P-3A' Fuse > 4 Reactor Coolant Loops
| |
| 'RLO21 B-3A Fuse a ^ Instrumentation ,
| |
| ! 5BBK15C P-3A Fuse
| |
| * Reactor Coolant Loops RLO21 B-3A Fuse \ 4 ,
| |
| Instrumentation r
| |
| i[ Le[%
| |
| * g9 i
| |
| s.
| |
| g
| |
| ~ WOLF CREEK - UNIT'1, 4I 3/4'8-41
| |
| * .- a .,
| |
| - - - . . , , . . - - l L-- - - . .w ---.-~-*& ----------r - - , - - - - - - - * - , r - -- - -----+-m - ~ -' ~ -' '~ ' ~
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 6BBK15D P-3A Fuse Reactor Coolant Loops RLO21 B-3A Fuse Instrumentation 6BBK15E P-3A Fuse Reactor Coolant Loops RLO21 B-3A Fuse Instrumentation 5BBK19A P-3A Fuse Pressurizer Spray Valve RL002 B-3A Fuse BBPCV455B 5BBK198 P-3A Fuse Pressurizer Spray Valve RL002 B-3A Fuse BBPCV455C 1BBK30A P-3A Fuse Reactor Vessel Head Vent Vlv RLO21 B-3A Fuse BBHV8001A 6GNG03B P-5A Fuse CRDM Cooling Fan B NG02BJF5 B-3A Fuse Discharge Isolation Damper GNHZ42 6GNG03D P-5A Fuse CRDM Cooling Fan A PG20 GAR 2 B-3A Fuse Discharge Isolation Damper GNHZ41 1HBK03A P-3A Fuse RCDT Vapor Space CTMT Isol Viv RL021 B-3A Fuse HBHV7126 6HBK04A P-3A Fuse RCDT Vapor Space CTMT Isol Viv HB115 B-3A Fuse HBHV7127 SEPY078 P-3A Fuse Accumulator Tank Discharge RP043 B-15A CB-1 Valve Position Switch EPHV88080A EPHV8808BA 6EPY07A P-3A Fuse Accumulator Tank Discharge RP044 B-15A CB-1 Valve Position Switch EPHV8808AA EPHV8808CA 6GTY12A P-15A Breaker CTMT Minipurge Exhaust PG20GBR234 B-20A Fuse Isolation Damper GTHZ41 O
| |
| WOLF CREEK - UNIT 1 3/4 8-42
| |
| | |
| TABLE 3.8-l'(Continued)
| |
| ( / CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED g(UMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 6GTY12A P-15A Breaker CTMT Minipurge Exhaust PG20GBR234 B-20A Fuse Isolation Damper GTHZ42 P-SSRY12A SA Breaker CB-1 In Core Neutron Monitoring SR057 6 Path Transfer Motor A B-SSRYO9A 20A Fuse SR02A PG19GCR1 P-5SRY12A SA Breaker CB-2 In Core Neutron Monitoring SR057 15 Path Transfer Motor A B-SSRYO9A 20A Fuse SR03A PG19GCR1 P-SSRY12A SA Breaker CB-3 In Core Neutron Monitoring SR057 6 Path Transfer Motor B B-SSRY09A 20A Fuse SR02B PG19GCR1 P-5SRY12A 5A Breaker CB-4 In Core Neutron Monitoring
| |
| .SR057 15 Path Transfer Motor B-B-5SRY09A 20A Fuse SR03B PG19GCR1 P-SSRY12A SA Breaker CB-5 In Core Neutron Monitoring SR057 6 Path Transfer Motor C B-55RY09A 20A Fuse SR02C PG19GCR1 i P-SSRY12A SA Breaker CB-6 In Core Neutron Monitoring l SR057 15 Path Transfer Motor C B-5SRY09A 20A Fuse SR03C PG19GCR1 P-SSRY12A SA Breaker CB-8 In Core Neutron Monitoring SR057 6 Path Transfer Motor D
| |
| : B-5SRY09A 20A Fuse SR02D PG19GCR1' P-5SRY12A 5A Breaker CB-9 In Core Neutron Monitoring SR057 15 Path Transfer Motor D B-SSRY09A 20A Fuse SR03D PG19GCR1 WOLF CREEK - UNIT 1 3/4 8-43
| |
| | |
| TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR, OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)
| |
| P-SSRYO9A SA Breaker CB-10 In Core Neutron Monitoring SR057 Drive Unit Heaters B-SSRY09A 20A Fuse SR01C & D PG19GCR1 P-5SRY09A SA Breaker CB-11 In Core Neutron Monitoring SR057 Drive Unit Heaters B-55RY09A 20A Fuse SR01A & B PG10GCR1 P-QB95 CKT 12 20A Breaker CB-12 Escape Lock and QB95 Stair Lighting B-QB95 CKT 12 20A Fuse QB95 P-5KFG02V 6A Fuse Reactor Building KF07 Elevator B-5KFG02V 15A Fuse KF07 KF07 P-Loop #9 15A Fuse FU9A P.A. Loop #9 QF076 QF076 B-Loop #9 15A Fuse FU9B QF076 P-Loop #10 15A Fuse FU10A P.A. Loop #10 QF076 QF076 B-Loop #10 15A Fuse FU10B QF076 WEECP1 P-20A Breaker Steam Generator PG20NBR2 B-30A Fuse Instrument Test System WEECP3 P-20A Breaker Steam Generator PG20NBR2 B-30A Fuse Instrument Test System WEECP2 P-20A Breaker Steam Generator PG20NBR2 B-30A Fuse Instrument Test System P-5GRY03A 15A Breaker CACS Charcoal Filter PG19NJF124 Temperature Transmitter S-5GRY03A 30A Fuse GRTT9 PG19NHF1 WOLF CREEK - UNIT 1 3/4 8-44
| |
| | |
| , TABLE 3.8-1 (Continued)
| |
| CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NU6EER AND LOCATION EQUIPMENT 1
| |
| Low Voltage ~ Power and Control (Continued) c P-6GRYO3B 15A Breaker CACS Charcoal Filter FG20NBR242 Temperature Transmitter l S-6GRYO3B 30A Fuse GRTT19
| |
| : PG20NBR1 P-5GRYO4A 15A Breaker Motor Space Heater PG19NHF214 For DCGR01A S-5GRY04A 30A Fuse PG19NHF1 P-6GRY04B 15A Breaker Motor Space Heater PG20NBR212 For DCGR01B S-6GRYO4B 30A Fuse
| |
| , PG20NBR1 P-6 KEG 02A 15A Breaker Fuel Handling PG20GBR224 Manipulator Crane O S-6 KEG 02A 30A Fuse PG20GBR1 HKE01 P-1BBS17A 10A Fuse RHR Loop 1 Inlet Isolation l SB029C Valve Limit Switch S-1BBA17A 20A Fuse BBZS8702AB NN0112 P-4BBS178 10A Fuse RHR Loop 2 Inlet Isolation SB032C Valve S-4BBS178 20A Fuse BBPV87028 NN0412 P-6KFY02A 20A Fuse Personnel Access Hatch TVKF03 Lock Mechanism and Lights S-6KFY02A 30 Breaker TVKF03 TVKF03 i
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| l
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| ; WOLF CREEK - UNIT 1 3/4 8-45 l
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| _ _ .-~.__._... . _ --__ _ . _. _ _.- _ .-_____ _ _ _ _. _ _ .._...
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| | |
| /
| |
| 3/4.9 REFUELING OPERATIONS
| |
| (
| |
| 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:
| |
| : a. A K,ff of 0.95 or less, or
| |
| : b. A boron concentration of greater than or equal to 2000 ppm.
| |
| APPLICABILITY: MODE 6*.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm f of a solution containing greater than or equal to 7000 ppm boron or its equiv-y alent until K,77 is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2000 ppm, whichever is the more restrictive.
| |
| SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
| |
| : a. Removing or unbolting the reactor vessel head, and
| |
| : b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
| |
| 4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours.
| |
| 4.9.1.3 Valves BG-V178 and BG-V601 shall be verified locked closed and secured in position at least once per 31 days.
| |
| l *The reactor shall be maintained in MODE 6 whenever fuel is in the reactor l vessel with the vessel head closure bolts less than fully tensioned or with l ( the head removed.
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| WOLF CREEK - UNIT 1 3/4 9-1 i
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| | |
| REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE each with continuous visual indication in the control room and one with audible indication in the containment and control room.
| |
| APPLICABILITY: MODE 6.
| |
| ACTION:
| |
| : a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes,
| |
| : b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours.
| |
| O SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:
| |
| : a. A CHANNEL CHECK at least once per 12 hours,
| |
| : b. An ANALOG CHANNEL OPERATIONAL TEST withir. 8 hours prior to the initial start of CORE ALTERATIONS, and
| |
| : c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
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| O WOLF CREEK - UNIT 1 3/4 9-2
| |
| | |
| REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least 100 hours.
| |
| ! APPLICABILITY: During movement of irradiated fuel in the reactor vessel.
| |
| , ACTION:
| |
| With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel.
| |
| b SURVEILLANCE REQUIREMENTS 1
| |
| 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel.
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| 4 e
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| +
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| WOLF CREEK - UNIT 1 3/4 9-3
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| | |
| REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
| |
| : a. The equipment door closed and held in place by a minimum of four bolts,
| |
| : b. A minimum of one door in each airlock is closed, and
| |
| : c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
| |
| : 1) Closed by an isolation valve, blind flange, or manual valve, or
| |
| : 2) Be capable of being closed by an OPERABLE automatic containment purge isolation valve.
| |
| APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.
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| ACTION:
| |
| With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building.
| |
| SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment purge isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:
| |
| : a. Verifying the penetrations are in their closed / isolated condition, or
| |
| : b. Testing the containment purge isolation valves per the applicable portions of Specification 4.6.3.2.
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| O WOLF CREEK - UNIT 1 3/4 9-4
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| | |
| (%
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| (j REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling 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.
| |
| I SURVEILLANCE REQUIREMENTS
| |
| ' N~ /
| |
| 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS.
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| WOLF CREEK - UNIT 1 3/4 9-5
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| | |
| REFUELING OPERATIONS 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine shall be used for movement of drive 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 cutof_fs with 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 conditions, and b) Secondary - less than or equal to 150 pounds above the primary overload cutoff.
| |
| : 3) An automatic load reduction trip with a Setpoint of less than or equal to 250 pounds below the suspended weight for wet con-ditions or dry conditions.
| |
| : b. The auxiliary hoist used for latching and unlatching drive rods and thimble plug handling operations having:
| |
| : 1) A minimum capacity 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 within the reactor vessel.
| |
| ACTION:
| |
| With the requirements for refueling machine and/or auxiliary hoist OPERABILITY not satisified, 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 4.9.6.1 The refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior WOLF CREEK - UNIT 1 3/4 9-6
| |
| | |
| .:.a,, + .~. + . , - . - +, - . x. . -
| |
| r O
| |
| REFUELING OPERATIONS j SURVEILLANCE REQUIREMENTS (Continued) t to removal of the reactor vessel head by performing a load test of at least 125%
| |
| of the secondary automatic overload cutoff and demonstrating an automatic load
| |
| : cutoff when the refueling machine load exceeds the Setpoints of Specification
| |
| ! 3.9.6a.2) and by demonstrating an automatic load reduction trip when the load 2
| |
| reduction exceeds the Setpoint of Specification 3.9.6a.3).
| |
| 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to removal of the reactor vessel head by performing a load test of at least 1250 pounds.
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| 1 i
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| i 1
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| i i
| |
| . WOLF CREEK - UNIT 1 3/4 9-7
| |
| | |
| REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY LIMITING CONDITION FOR OPERATION 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 facility.
| |
| ACTION:
| |
| : a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks and physical stops which prevent crane travel with O
| |
| loads in excess of 2250 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.
| |
| O WOLF CREEK - UNIT 1 3/4 9-8
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| | |
| 4 1
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| j .r-REFUELING OPERATIONS
| |
| : j. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION
| |
| ]-
| |
| .HIGH WATER LEVEL
| |
| } LIMITING CONDITION FOR OPERATION
| |
| ,C4?
| |
| 3.9.8.1 At leastLone residual heat removal-(RHR) loop shall be OPERABLE and in operation.*
| |
| APPLICA8ILITY: MODE 6 when the water 1svel above the top of the reactor vessel flange is greater than or equal to 23 feet.
| |
| : i. ACTION:
| |
| i With no'RHR-loop OPERABLE and in operation, suspend all operations involving j' an increase in the reactor decay heat load or a reduction in baron concentra-
| |
| ; tion of the Reactor Coolant System and immediately initiate corrective action i- to return the required RHR loop to OPERABLE and operating status as soon as i possible. -Close all containment penetrations providing~ direct access from the
| |
| ' containment' atmosphere to the outside atmosphere within 4 hours, s
| |
| ; \s I
| |
| i t ,
| |
| 1 .
| |
| J SURVEILLANCE REQUIREMENTS i
| |
| - 4.9.8.1 At least one'RHR loop shall be verified in operation and circulating reactor _ coolant at a flow rate of greater than or ' equal to 2800 gpm at least l once per 12 hours.
| |
| i i
| |
| 4 t
| |
| j *The RHR loop may be removed from operation for up to I hour per 8-hour period
| |
| : during the performance of CORE ALTERATIONS in the vicinity of the reactor j vessel hot legs, i
| |
| 4 i .
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| ; WOLF CREEK - UNIT 1 3/4 9-9 r
| |
| i t-
| |
| , , . , - , , - . - - _ . . , - . - - . , , , , - - - - , , - __ ,, -. , -___-,,._,_,_-,.,,n.m.-.,-
| |
| | |
| REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, ~m and at least one RHR loop shall be in operation.*
| |
| APPLICABILITY: MODE 6 when the water level above the top of the reactor vessel flange is less than 23 feet.
| |
| ACTION:
| |
| : a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
| |
| : b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating l reactor coolant at a flow rate of greater than or equal to 2800 gpm at least l once per 12 hours.
| |
| l l
| |
| 1
| |
| * Prior to initial criticality, the RHR loop may be removed from operation for up to 1 hour per 2-hour period during the performance of CORE ALTERATIONS in i
| |
| the vicinity of the reactor vessel hot legs.
| |
| O WOLF CREEK - UNIT 1 3/4 9-10
| |
| | |
| [
| |
| 4 i!
| |
| REFUELING OPERATIONS ,
| |
| -3/4.9.9 CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION i ;
| |
| j 3.9.9 The Containment Ventilation System shall be OPERABLE.
| |
| i 1
| |
| ; APPLICA8ILITY. During CORE ALTERATIONS or movement of irradiated fuel within '
| |
| the containment. l i
| |
| i ~ ACTION:
| |
| : a. With the Containment Ventilation System inoperable, close each of i the purge valves providing direct access from the containment l atmosphere to the outside atmosphere.
| |
| i i b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| I
| |
| {- SURVEILLANCE REQUIREMENTS
| |
| 'i i j 4.9.9 The Containment Ventilation System shall be demonstrated OPERA 8LE <
| |
| ; within 100 hours prior to the start of and at least once per 7 days during '
| |
| i CORE ALTERATIONS by verifying that containment purge isolation occurs on i manual initiation and on a High Radiation test signal from each of the containment radiation monitoring instrumentation channels.
| |
| 4 4
| |
| I i
| |
| ?
| |
| t I
| |
| l l
| |
| k I
| |
| WOLF CREEK - UNIT I 3/4 9-11 i
| |
| | |
| REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.
| |
| APPLICABILITY: During movement of fuel assemblies within the containment when the fuel assemblies being moved are irradiated.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies within the reactor vessel.
| |
| SURVEILLANCE REQUIREMENTS O
| |
| 4.9.10.1 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies.
| |
| O WOLF CREEK - UNIT 1 3/4 9-12
| |
| | |
| (( )
| |
| ~
| |
| REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL CONTROL RODS LIMITING CONDITION FOR OPERATION 3.9.10.2 At least 23 feet of water shall be maintained over the top of the irradiated fuel assemblies within the reactor pressure vessel.
| |
| APPLICABILITY: During movenent of control rods within the reactor pressure vessel while in MODE 6.
| |
| ACTION:
| |
| With the requirements of the above specification not satisfied, suspend all operations involving movement of control rods within the pressure vessel.
| |
| SURVEILLANCE REQUIREMENTS 4.9.10.2 The water level shall be determined to be at least its minimum
| |
| ; required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of control rods within the reactor vessel.
| |
| I s
| |
| i l
| |
| 1 WOLF CREEK - UNIT 1 3/4 9-13
| |
| | |
| REFUELING OPERATIONS 3/4.9.11 WATER LEVEL-STORAGE P0OL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
| |
| APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool.
| |
| ACTION:
| |
| : a. With the requirements of the above specificatien not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| O SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the spent fuel pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pool.
| |
| O WOLF CREEK - UNIT 1 3/4 9-14
| |
| | |
| A ,
| |
| REFUELING OPERATIONS 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 Spent fuel assemblies stored in Region 2 shall be subject to the following conditions:
| |
| : a. The combination of initial enrichment and cumulative exposure shall be within the acceptable domain of Figure 3.9-1, and
| |
| : b. No spent fuel assemblies shall be placed in Region 2, nor shall any storage location be changed in designation from being in Region 1 to being in Region 2, while refueling operations are in progress.
| |
| APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool.
| |
| ACTION:
| |
| : a. With the requirements of the above specification not satisfied, h
| |
| \
| |
| suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move the non-complying fuel assemblies to Region 1. Until these requirements of the above specification are satisfied boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppm at least once per 8 hours,
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.9.12 The burnup of each spent fuel assembly stored in Region 2 shall be ascertained by analysis of its burnup history and independently verified, prior to storage in Region 2. A complete record of such analysis shall be kept for the time period that the spent fuel assembly remains in Region 2 of the spent fuel pool.
| |
| !(
| |
| i WOLF CREEK - UNIT 1 3/4 9-15
| |
| | |
| O 44, 0(,0 -. *
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| p ti. 1+
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| :: ! t: n: i -
| |
| 40,000 ".
| |
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| "; . :- :t:: 1:u t+-::
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| t:-
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| 3 36,000 :' .' .".r !ii;+iii
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| m ,. .... , .. @ti ... i.i.E:.!.it.' i. i a: 32,000 .... . .. . .. . . .. . . . . : . : . : :.::: .::. ..
| |
| 3 :lli !U! .. i ACCEPTABLE FOR.. : " tl o iffi ti!! !!!i*IIREGIO.N,2 STO.RA,GE,i :j .i.- i.
| |
| : a. 28,000 . . . .
| |
| x .i .I !jp Mf: :.' j . . .
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| W ..! * * *
| |
| -MI w ::t:
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| > 24,000 ,
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| |
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| |
| . :a. :. : : :
| |
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| s o 16,000
| |
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| |
| !!F
| |
| ! . 7 .
| |
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| |
| !!!! Uli INOT ACCEPTdBLE'FOR E y i ! !!!! I
| |
| ~
| |
| iil: iii! ;i' i. 'iiji
| |
| " REGION .2 STORAGE .
| |
| 2 12,000 . . . !...:
| |
| .:r. -:
| |
| p.:.
| |
| :.:M a.!: ..g a:::: .:s u. .. .j .r.. : . -
| |
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| |
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| |
| ::t: p:t; . :r j:t
| |
| :tg: ..{. ::.: ..li n- g ::.
| |
| f i- .n t (f) .: '
| |
| ::, :..g:- ''
| |
| < 8.000 :: .a ::
| |
| :.- p:- :a:
| |
| .n; ;
| |
| ::.p: :::_
| |
| * :ip:
| |
| :. an :/: =::.
| |
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| |
| " fii w 1!! !!!! !!!: -
| |
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| |
| 3 4,000
| |
| '"' ~"'
| |
| S .f .:.. i p=.
| |
| g zu: ==
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| . . .n y . .
| |
| 0 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 FUEL ASSEMBLY INITIAL ENRICHMENT, w/o U-235 l
| |
| l FIGURE 3.9-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 I
| |
| l I
| |
| O WOLF CREEK - UNIT 1 3/4 9-16
| |
| | |
| (
| |
| ~
| |
| (3 ) REFUELING OPERATIONS
| |
| -(O 3/4.9.13 EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.9.13 Two independent Emergency Exhaust Systems shall be OPERABLE.
| |
| APPLICABILITY: Whenever irradiated fuel is in the spent fuel pool.
| |
| ACTION:
| |
| : a. With one Emergency Exhaust System inoperable, fuel movement within the fuel storage areas or crane operation with loads over the fuel storage areas may proceed provided the OPERABLE Emergency Exhaust
| |
| ' System is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.
| |
| : b. With no Emergency Exhaust System OPERABLE, suspend all operations involving movement of fuel within the fuel storage areas or crane i operation with loads over the fuel storage areas until at least one Emergency Exhaust System is restored to OPERABLE status.
| |
| : c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.9.13 The above required Emergency Exhaust Systems shall be demonstrated
| |
| .0PERABLE:
| |
| : a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least
| |
| , 10 continuous hours with the heaters operating;
| |
| ^
| |
| : b. At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system, by:
| |
| : 1) Verifying that the Emergency Exhaust System satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers and uses the test procedure guidance in Regulatory Positions C.5.a, j C.S.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 9000 cfm 110% at > 7.2 inches W.G. ,
| |
| . (dirty filter);
| |
| WOLF CREEK - UNIT 1 3/4 9-17
| |
| | |
| REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)
| |
| : 2) Verifying, within 31 days after removal, that a laboratory analy-sis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revisison 2, March 1978, for a methyl iodide penetration of lEss than 1%; and
| |
| : 3) Verifying a system flow rate of 9000 cfm 110% at > 7.2 inches W.G. (dirty filter) during system operation when tested in accordance with ANSI N510-1980.
| |
| : c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Posi-tion C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;
| |
| : d. At least once per 18 months by:
| |
| : 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than or equal to 7.2 inches Water Gauge while operating the system at a flow rate of 9000 cfm 110%.
| |
| : 2) Verifying that on a Fuel Building Exhaust Gaseous Radioactivity-High test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks and isolates the normal fuel building exhaust flow to the auxiliary / fuel building exhaust fan;
| |
| : 3) Verifying that the system maintains the Fuel Building at a nega-tive pressure of greater than or equal to 1/4 inches Water Gauge relative to the outside atmosphere during system operation; and
| |
| : 4) Verifying that the heaters dissipate 37 1 3 kW when tested in accordance with ANSI N510-1975. ,
| |
| : e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers in accordance with ANSI N510-1975 (however Prerequisite Testing, Sections 8 and 9 shall be in accordance with ANSI N510-1980) for a DOP test aerosol while operating the system at a flow rate of 9000 cfm 110%; and
| |
| : f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% for HEPA filters and 0.05% for charcoal adsorbers in accord-ance with ANSI N510-1975 (however Prerequisite Testing, Sections 8 and 9 shall be in accordance with ANSI N510-1980) for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 9000 cfm 110%.
| |
| WOLF CREEK - UNIT 1 3/4 9-18
| |
| | |
| [
| |
| ( 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip 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 reactivity equivalent available for trip insertion, 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 Specification 3.1.1.1 is restored.
| |
| \
| |
| : b. With all full-length control rods fully inserted and the reactor subcritical 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 Specification 3.1.1.1 is restored.
| |
| SURVEILLANCE REQUIREMENTS l
| |
| 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours.
| |
| 4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position t within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits l
| |
| of Specification 3.1.1.1.
| |
| I
| |
| {
| |
| l WOLF CREEK - UNIT 1 3/4 10-1 I
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
| |
| : a. The THERMAL POWER is maintained less than or equa! to 85% of RATED THERMAL POWER, and
| |
| : b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2, below.
| |
| APPLICABILITY: MODE 1.
| |
| ACTION:
| |
| With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:
| |
| : a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
| |
| : b. Be in HOT STANDBY within 6 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL F0WER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
| |
| 4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours during PHYSICS TESTS:
| |
| : a. Specifications 4.2.2.2 and 4.2.2.3, and
| |
| : b. Specification 4.2.3.2.
| |
| O WOLF CREEK - UNIT 1 3/4 10-2
| |
| | |
| y\
| |
| i }
| |
| y,/ SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6, may be suspended during the performance of PHYSICS TESTS provided:
| |
| : a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
| |
| : b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and
| |
| : c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 541 F.
| |
| APPLICA8ILITY: MODE 2.
| |
| ACTION:
| |
| ,/ a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
| |
| : b. With a Reactor Coolant System operating loop temperature (T,yg) less than 541 F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.
| |
| SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%
| |
| of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
| |
| 4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS.
| |
| 4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to 541 F at least once per 30 minutes during PHYSICS TESTS.
| |
| f (3) v WOLF CREEK - UNIT 1 3/4 10-3
| |
| | |
| SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of the following requirements may be suspended:
| |
| : a. Specification 3.4.1.1 - During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided:
| |
| : 1) The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and
| |
| : 2) The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
| |
| : b. Specification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least three reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE.
| |
| APPLICABILITY: During operation below the P-7 Interlock Setpoint or performance of hot rod drop time measurements.
| |
| ACTION:
| |
| : a. With the IHERMAL POWER greater than the P-7 Interlock Setpoint during the performance of startup and PHYSICS TESTS, immediately open the Reactor trip breakers,
| |
| : b. With less than the above required reactor coolant loops OPERABLE during performance of hot rod drop time measurements, immediately place two reactor coolant loops in operation.
| |
| SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS.
| |
| 4.10.4.2 Each Intermer'iate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS.
| |
| 4.10.4.3 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours prior to initiatior. of the hot rod drop time measure-ments and at least once per 4 hours during the hot rod drop time measurements by verifying correct breaker alignments and indicated power availability.
| |
| WOLF CREEK - UNIT 1 3/4 10-4
| |
| | |
| ( SPECIAL TEST EXCEPTIONS V;
| |
| 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full-length shutdown and control rod drop time measurements provided only one shutdown or control bank is withdrawn from the fully inserted position at a time.
| |
| APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements and during surveillance of digital rod position indicators for OPERABILITY.
| |
| ACTION:
| |
| With the Position Indication System inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.
| |
| SURVEILLANCE REQUIREMENTS i
| |
| '' 4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:
| |
| : a. Within 12 steps when the rods are stationary, and
| |
| : b. Within 24 steps during rod motion.
| |
| WOLF CREEK - UNIT 1 3/4 10-5
| |
| | |
| I ,
| |
| i h\
| |
| y/ 3/4.11 RADI0 ACTIVE EFFLUENTS
| |
| .. t ,
| |
| : '. 3/4.11.1 LIQUID FiFLUENTS CONCENTRATION '
| |
| '/
| |
| LIMITING CONDITION FOR OPERATION 3.11.1/i'Thhconcentrationofradioactivematerialreleasedinliquideffluents to UNRESTRICTED AREAS (see Figure 5.1-4) shall be limited to the concentrations V specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcurie /ml total activity.
| |
| . APPLICABILITY: At all times.
| |
| ACTION:
| |
| With'the concentration of radioactive material released in liquid effluents to
| |
| , UNRESTRICTED AREAS exceeding the above limits, immediately restore the concen-tration to within the above limits.
| |
| I v
| |
| SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and, analysis program of Table 4.11-1.
| |
| 4.11.1.1.2 The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.
| |
| 4 4
| |
| =
| |
| , ~E v
| |
| s i
| |
| l WOLF CREEK - UNIT 1 3/4 11-1
| |
| ,,~
| |
| * f.
| |
| | |
| i l
| |
| l TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM !
| |
| l l
| |
| LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (1) .
| |
| TYPE ANALYSIS I FREQUENCY FREQUENCY
| |
| ( ))
| |
| : 1. Batch Waste P P
| |
| -7 Release Each Batch Each Batch Principal Gamma 5x10 ;
| |
| Tanks (2) Emitters (3)
| |
| -6 I-131 1x10
| |
| : a. Waste ,
| |
| Monitor !
| |
| -5 Tank P M Dissolved and 1x10 One Batch /M Entrained Gases (Gamma Emitters)
| |
| : b. Secondary Liquid 1x10
| |
| -5 Waste Moni- P M H-3 tor Tank Each Batch Composite (4) g Gross Alpha 1x10
| |
| -8 P Q Sr-89, Sr-90 5x10 Each Batch Composite (4)
| |
| -6 Fe-55 1x10
| |
| -7
| |
| : 2. Continuous W Principal Gamma 5x10 Releases (5) Daily (6) Composite (4) Emitters (3) l Grab Sample
| |
| -6 l I-131 1x10 Steam Generator
| |
| -5 Blowdown M M Dissolved and 1x10 Entrained Gases Grab Sample (Gamma Emitters) 1x10
| |
| -5 M H-3 Daily (6) Composite (4) -
| |
| _7 Grab Sample Gross Alpha 1x10
| |
| -8 Q Sr-89, Sr-90 5x10 Daily (0) Composite (4) -6 Grab Sample Fe-55 1x10 WOLF CREEK - UNIT 1 3/4 11-2
| |
| | |
| o
| |
| (# TABLE 4.11-1 (Continued)
| |
| TABLE NOTATIONS (1)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
| |
| For a particular measurement system, which may include radiochemical separation:
| |
| LLD =
| |
| E V 2.22 x IOG Y exp (-AAt)
| |
| Where:
| |
| LLD = the "a priori" lower limit of detection (microCuries per unit mass or volume),
| |
| s = the standard deviation of the background counting rate or of (od i t e counting rate of a blank sample as appropriate (counts per minute),
| |
| E = the counting efficiency (counts per disintegration),
| |
| V = the sample size (units of mass or volume),
| |
| 2.22 x 108 = the number of disintegrations per minute per microcurie, Y ' the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (s 1), and at = the elapsed time between the midpoint of sample collection and the time of counting (s).
| |
| Typical values of E, V, Y, and at should be used in the calculation.
| |
| It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
| |
| (2)A batch release is the discharge of liquid wastes of a discrete volume.
| |
| Prior to sampling for analyses, each batch shall be isolated, and p then thoroughly mixed by a method described in plant procedures to assure representative sampling.
| |
| WOLF CREEK - UNIT 1 3/4 11-3
| |
| | |
| TABLE 4.11-1 (Continued)
| |
| TABLE NOTATIONS (Continued)
| |
| (3)The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.
| |
| (4)A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids releasec. Prior to analysis, all samples taken for the composite shall be throughly mixed in order for the composite samples to be representative of the effluent release.
| |
| (5)A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.
| |
| (6) Samples shall be taken at the initiation of effluent flow and at least once per 24 hours thereafter while the release is occurring. To be representative of the liquid effluent, the sample volume shall be proportioned to the effluent stream discharge volume. The ratio of sample volume to effluent discharge volume shall be maintained constant for all samples taken for the composite sample.
| |
| O WOLF CREEK - UNIT 1 3/4 11-4
| |
| | |
| i i
| |
| , RADIOACTIVE EFFLUENTS i
| |
| DOSE-
| |
| )
| |
| LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive i 4- materials in-liquid effluents released, from each unit, to UNRESTRICTED AREAS ;
| |
| i (see Figure 5.1-4) shall be limited:
| |
| : a. During any calendar quarter to less than or equal to 1.5 mrems to l the whole body and to less than or equal to 5 mrems to any organ, '
| |
| and
| |
| : b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.
| |
| . APPLICABILITY: At all times.
| |
| ACTION:
| |
| 4 a. - With the calculated dose from the release of radioactive materials I.
| |
| in liquid. effluents exceeding any of the above limits, prepare and i submit to the Commission within 30 days, pursuant to Specification ;
| |
| j 6.9.2, a Special Report that identifies the cause(s) for exceeding
| |
| ) the limit (s) and defines the corrective actions that have been-taken
| |
| (. . to reduce the releases and the proposed corrective actions to be ,
| |
| taken to assure that subsequent releases w;11. be in compliance with the above limits. This Special Report shall Liso include: (1) the-results of radiological analyses of the drinking water source, and ;
| |
| (2) the radiological impact on finished drinking water supplies with 4
| |
| regard to the requirements of 40 CFR Part 141, Clean Drinking Water
| |
| , Act.*
| |
| : b. The provisions of Specifications 3.0.3 and.3.0.4 are not applicable.
| |
| 2.
| |
| I SURVEILLANCE REQUIREMENTS
| |
| . 4.11.1.2 Cumulative dose contributions from liquid effluents for the current' i
| |
| : calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCf! at least once per 31 days.
| |
| 1
| |
| !', "The requirements of ACTION a.(1) and (2) are applicable only if drinking water ,
| |
| i supply is taken from the receiving water body within 3 miles of the plant dis- '
| |
| charge. In the case of river-sited plants this is 3 miles downstream only.
| |
| WOLF CREEK - UNIT 1 3/4 11-5
| |
| - a- en ,- ,,_>-c- e ,, - -r-,e--,s%<.. w,-. .m -
| |
| - . . - , - -_. .-,_ c rnv-- e
| |
| | |
| RADI0 ACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-4) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2, a Special Report that includes the following information:
| |
| : 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
| |
| : 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
| |
| : 3. Summary description of action (s) taken to prevent a recurrence.
| |
| l
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| 1 SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being l fully utilized.
| |
| 4.11.1.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2.
| |
| 1 WOLF CREEK - UNIT 1 3/4 11-6
| |
| | |
| RADI0 ACTIVE EFFLUENTS LIQUID HOLDUP TANKS l
| |
| 1 LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 150 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.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| , a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank O, contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release
| |
| . Report, pursuant to Specification 6.9.1.7.
| |
| 'b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS l'
| |
| l 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added and within 7 days following any addi-tion of radioactive material to the tank.
| |
| t
| |
| !O l
| |
| WOLF CREEK - UNIT 1 3/4 11-7
| |
| _ _ . - - - _ _ . _ _ _ _ _ _ _ _ ~ _ . - _ - _
| |
| | |
| RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
| |
| : a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and
| |
| : b. For Iodine-131 and 133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).
| |
| SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due.to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.
| |
| 4.11.2.1.2 The dose rate due to Iodine-131 and 133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
| |
| O WOLF CREEK - UNIT 1 3/4 11-8
| |
| | |
| O O TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 2
| |
| h MINIMUM LOWER LIMIT OF i
| |
| , SAMPLING ANALYSIS TYPE OF DETECTION (LLD) 1) j GASEOUS, RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS ,
| |
| (pCi/ml)
| |
| ~ P P Principal Gamma Emitters (2) 1x10
| |
| -4
| |
| [ 1. Waste Gas Decay Each Tank Each Tank Tank Grab Sample
| |
| -4
| |
| : 2. Containment Purge Ea h PURGE (3) Eac PURGE (3) Principal Gamma Emitters (2) 1x10 or Vent Grab Sample M H-3 (oxide) 1x10'0 g(3) -4
| |
| ! 3. Unit Vent M(3) (4)
| |
| Principal Gamma Emitters (2) 1x10 w Grab -6 Sample M(4) H-3 (oxide) 1x10 j } -4 y 4. Spent Fuel Building M I) M Principal Gamma Emitters (2) 1x10 4 Vent Grab Sample -6 M(5) H-3 (oxide) 1x10
| |
| -4 1 5. Radwaste Building M Principal Gamma Emitters (2) 1x10 j Vent Grab Sample M 1
| |
| y(7) -12
| |
| : 6. All Release Types Continuous (6)(8) I-131 1x10 as listed in 1., Charcoal Sample I-133 1x10 10 g(7) Principal Gamma Emitters (2) 1x10
| |
| -11 above Continuous (6)(8) i Particulate Sample
| |
| -11 Continuous (6)(8) M Gross Alpha 1x10 Composite Particulate Sample
| |
| -11 Continuous (6)(8) Q Sr-89, Sr-90 1x10 Composite Particulate Sample I
| |
| | |
| TABLE 4.11-2 (Continued)
| |
| TABLE NOTATIONS (1)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
| |
| For a particular measurement system, which may include radiochemical separation:
| |
| 4' S b
| |
| LLD =
| |
| E V 2.22 x 108 Y exp (-AAt)
| |
| Where:
| |
| LLD = the "a priori" lower limit of detection (microCuries per unit mass or volume),
| |
| sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, (counts per minute),
| |
| E = the counting efficiency (counts per disintegration),
| |
| V = the sample size (units of mass or volume),
| |
| 2.22 x 108 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (s 1), and At = the elapsed time between the midpoint of sample collection and the time of counting (s).
| |
| Typical values of E, V, Y, and At should be used in the calculation.
| |
| It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
| |
| O WOLF CREEK - UNIT 1 3/4 11-10
| |
| | |
| TABLE 4.11-2 (Continued)
| |
| TABLE NOTATIONS (Continued)
| |
| (2)The principal gamma emitters for which the LLD specification applies include the following radionucl. ides: Kr-87, Kr-88, Xe-133, .
| |
| < Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7, in the format outlined in Regulatory Guide 1.21, Appendix 8, Revision 1, June 1974.
| |
| (3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within 1 hour period.
| |
| (4) Tritium grab samples shall be taken and analyzed at least once per 24 hours when the refueling cana1 is flooded.
| |
| (5) Tritium grab samples shall be taken and'an~alyzed at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool. Grab samples need to be taken only when
| |
| : spent fuel is in the spent fuel pool.
| |
| i (6)The ratio of the sample flow rate to the sampled stream flow rate shall i be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.
| |
| (7) Samples shall be changed at least once per 7 days and analyses shall be.
| |
| 4 completed within 48 hours after changing, or after removal from sampler.
| |
| For unit vent, sampling shall also be performed at least once per 24 hours i
| |
| for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period, and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LLDs may be increased i by a factor of 10. This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3, and (2) the nob.le gas monitor shows that effluent activity has not increased more than a factor of 3.
| |
| (8)Continous sampling of the spent fuel building exhaust needs to be performed only when spent fuel is in the spent fuel pool.
| |
| l l WOLF CREEK - UNIT 1 3/4 11-11 i
| |
| | |
| RADI0 ACTIVE EFFLUENTS DOSE - N0BLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE B0UNDARY (see Figure 5.1-3) shall be limited to the following:
| |
| : a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
| |
| : b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
| |
| O WOLF CREEK - UNIT 1 3/4 11-12 ;
| |
| | |
| (R RADI0 ACTIVE EFFLUENTS DOSE - 10 DINE-131 AND 133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE B0UNDARY (see Figure 5.1-3) shall be limited to the following:
| |
| : a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
| |
| : b. During any calendar year: Less than or equal to 15 mrems to any organ.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the calculated dose from the release of Iodine-131 and 133, b tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS i
| |
| 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per l 31 days.
| |
| a WOLF CREEK - UNIT 1 3/4 11-13
| |
| | |
| RADI0 ACTIVE EFFLUENTS GASE0US RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed:
| |
| : a. 0.2 mrad to air from gamma radiation, or
| |
| : b. 0.4 mrad to air from beta radiation, or
| |
| : c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:
| |
| : 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
| |
| : 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
| |
| : 3. Summary description of action (s) taken to prevent a recurrence.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE B0UNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized.
| |
| 4.11.2.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and WASTE GAS HOLDUP SYSTEM shall be considered OPERABLE by meeting Specification 3.11.2.1 and 3.11.2.2 or 3.11.2.3.
| |
| WOLF CREEK - UNIT 1 3/4 11-14
| |
| | |
| l h
| |
| g RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE 1
| |
| LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen 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.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 3% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours.
| |
| : b. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM
| |
| . greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately suspend all additions of waste gases 3 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 Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 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
| |
| . monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11.
| |
| i i
| |
| {
| |
| l' i
| |
| l WOLF CREEK - UNIT 1 3/4 11-15
| |
| | |
| RADI0 ACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 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 Curies of noble gases (con-sidered as Xe-133 equivalent).
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and, within 48 hours, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radicactive material contained in each gas storage tank shall be determined to be within the above limit 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.
| |
| O WOLF CREEK - UNIT 1 3/4 11-16
| |
| | |
| /9 RADI0 ACTIVE EFFLUENTS
| |
| (" )
| |
| 3/4.11.3 SOLID RADI0 ACTIVE WASTES LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes-shall be solidified or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the disposal site.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the Solid Waste System as necessary to prevent recurrence.
| |
| : b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence.
| |
| [] c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.11.3 SOLIDIFICATION of at-least one representative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludget, spent resins, evaporator bottoms, boric acid solutions and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:
| |
| : a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM,
| |
| ; and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of i the batch may then be resumed using the alternative SOLIDIFICATION I
| |
| parameters determined by the PROCESS CONTROL PROGRAM; l
| |
| : b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each
| |
| ! consecutive batch of the same type of wet waste until at least three l consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste; and G c. With the installed equipment incapable of meeting Specification f
| |
| i 3.11.3 or declared out-of-service, restore the equipment to operable l status or provide for contract capability to process wastes as j
| |
| necessary to satisfy all applicable transportation and disposal requirements.
| |
| WOLF CREEK - UNIT 1 3/4 11-17
| |
| | |
| RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b.,
| |
| 3.11.2.3a., or 3.11.2.3b., calculations should be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.
| |
| 4.11.4.2 Cumulative dose contributions from direct radiation from the units i and from radwaste storage tanks shall be determined in accordance with the i methodology and parameters in the ODCH. This requirement is applicable only j under conditions set forth in ACTION a. of Specification 3.11.4.
| |
| WOLF CREEK - UNIT 1 3/4 11-18
| |
| | |
| s 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING
| |
| ,3 l 1 3/4.12.1 MONITORING PROGRAM
| |
| , \j LIMITING CONDITION FOR OPERATION 3.12.1 The Radislogical Environmental Monitoring Program shall be conductea as specified in Table 3.12-1.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence,
| |
| : b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for. exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual
| |
| ,o dose
| |
| * to a MEMBER OF THE PUBLIC is less than the calendar year limits
| |
| ( ) of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than one V of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
| |
| , concentration (1) . concentration (2) . . . . > 1. 0 reporting level (1) reporting level (2)
| |
| When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
| |
| * to A MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of L plant effluents; however, in such an event, the condition shall be l reported and described in the Annual Radiological Environmental l Operating Report required by Specification 6.9.1.6.
| |
| : c. With the availability of milk or fresh leafy vegetable samples from one or more of the sample locations required by Table 3.12-1 not practicable or possible, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological
| |
| ; Environmental Monitoring Program given in the ODCM.
| |
| i
| |
| *The methodology and parameters used to estimate the potential annual dose to O a MEMBER OF THE PUBLIC shall be indicated in this report.
| |
| WOLF CREEK - UNIT 1 3/4 12-1 L
| |
| | |
| 1 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION ACTION (Continued)
| |
| The specific locations from which saniples were unavailable may then be deleted from the monitoring program. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.
| |
| : d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.
| |
| l WOLF CREEK - UNIT 1 3/4 12-2
| |
| | |
| - __ ._ _ ~ . . - - .-
| |
| /m -
| |
| t '
| |
| ; TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM j Q NUMBER OF
| |
| ; E REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY l AND/0R SAMPLE SAMPLE LOCATIONS (1) COLLECTION FREQUENCY OF ANALYSIS C
| |
| h 1. Direct Radiation (2) 40 routine monitoring stations Quarterly. ' Gamma dose quarterly.
| |
| t g either with two or more dosi-l meters or with one instrument for measuring and recording i dose rate continuously, placed as follows:
| |
| An inner ring of stations, one in each meteorological sector in the
| |
| . ca general area of the SITE BOUNDARY s
| |
| An outer ring of stations, one in M each meteorological sector in d, the 6- to 8-km (3 to 5 mile) range from the site j The balance of the stations to be placed in special i interest areas such as population centers, nearby.
| |
| , residences, schools, and in one or two areas to serve l as control stations.
| |
| l 1
| |
| .l I
| |
| l 1
| |
| f i
| |
| | |
| TABLE 3.12-1 (C:ntinued)
| |
| RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Q
| |
| m NUMBER OF REPRESENTATIVE
| |
| * EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE SAMPLE LOCATIONS (1) COLLECTION FREQUENCY OF ANALYSIS 5
| |
| w
| |
| : 2. Airborne
| |
| ~ Radioiodine and Samples from five locations Continuous sampler Radioiodine Cannister:
| |
| Particulates operation with sample I-131 analysis weekly.
| |
| Three samples from close to collection weekly, or the three SITE BOUNDARY locations, more frequently if in different sectors, of the required by dust Particulate Sampler:
| |
| highest calculated annual average loading. Gross beta radioactivity ground level D/Q. analysis following filter change;(4) and w One sample from the vicinity gamma isotopic analysis (5) 1 of a community having the highest of composite (by
| |
| ~ calculated annual average ground- location) quarterly.
| |
| 7 level D/Q.
| |
| One sample from a control loca-tion, as for example 15 to 30 km (10to20 mile)distantandingy least prevalent wind direction.
| |
| : 3. Waterborne
| |
| : a. Surface One sample upstream.(6) Monthly grab sample Gamma isotopic analysis (5) and sample downstream, monthly. Composite for tritium analysis quarterly.
| |
| : b. Ground Samples from one or two source Quarterly. Gamma isotopic (5) and tritium only if likely to be affected.[8) ,
| |
| analysis quarterly.
| |
| : c. Drinking One sample of each ,of one to Composite Jample I-131 analysis on each three of the n M st water over2-weeRfseriod(7) composite when the dose supplies that codld be when I-131 analysis calculated for the consump-affected by its discharge. is performed, monthly tion of the water is gjater l composite otherwise. than 1 mrem per year. Com-1 One sample from a control posite for gross beta ag location. gamma isotopic analyses monthly. Composite for tritium analysis quarterly.
| |
| O O O
| |
| | |
| _ , m fs
| |
| ) (g/ .)
| |
| TABLE 3.12-1 (Continued)
| |
| Ei g; RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM k NUMBER OF g REPRESENTATIVE
| |
| , EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY
| |
| : c. AND/0R SAMPLE SAMPLE LOCATIONS (7) COLLECTION FREQUENCY OF ANALYSIS i'i
| |
| * 3. Waterborne
| |
| * (Continued)
| |
| , d. Sediment One sample from downstream area Semiannually. Gamma isotopic analysis (5) from with existing or potential semiannually.
| |
| shoreline recreational value.
| |
| : 4. Ingestion
| |
| : a. Milk Samples from milking animals Semimonthly when Gamma isotopic (5) and I-131 m in three locations within 5 km animals are on analysis semimonthly when
| |
| } (3 mile) distance having the pasture, monthly at , animals are on pasture; highest dose potential. If there y other times. monthly at other times.
| |
| 'y are none, then, one sample from
| |
| ! m milking animals in each of three areas between 5 to 8 km (3 to 5 mile) distant where doses arecalculategg be greater than 1 mrem per yr One sample from milking animals at a control location 15 to 30 km (10 to 20 mile) distant and in the least prevalent wind direction.
| |
| : b. Fish One sample of 1 to 3 recreationally Sample in season, or Gamma isotopic analysis (5) important species in vicinity of semiannually if they on edible portions.
| |
| plant discharge area. are not seasonal.
| |
| One sample of similar species in l areas not influenced by plant j discharge.
| |
| I
| |
| | |
| TABLE 3.12-1 (Continued) 5 Q; RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM h NUMBER OF W REPRESENTATIVE
| |
| , EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY c AND/0R SAMPLE SAMPLE LOCATIONS (1) COLLECTION FREQUENCY OF ANALYSIS 5
| |
| * 4. Ingestion H
| |
| (Continued)
| |
| : c. Food One sample of each principal At time of harvest.(10) Gamma isotopic analyses (5)
| |
| Products class of food products from any on edible portion.
| |
| area that is irrigated by water in which liquid plant wastes have been discharged.
| |
| w Samples of 1 to 3 different kinds Monthly when Gamma isotopic (5) and I-131
| |
| } of available broad leaf vegetation available. analysis.
| |
| ~ grown nearest each of two different r? offsite locations of highest m predicted annual average ground-level D/Q if milk sampling is not performed.
| |
| Control sample of each of the Monthly when Gamma isotopic (5) and I-131 similar broad leaf vegetation grown available. analysis.
| |
| 15 to 30 km (10 to 20 mile) distant in the least prevalent wind direction if milk sampling is not performed.
| |
| O O O
| |
| | |
| TABLE 3.12-1 (Continued) n TABLE NOTATIONS (v)
| |
| (1) Specific parameters of distance and direction sector from the centerline of one unit, and additional description where pertinent, shall be provided for each and every sample location in Table 3.12-1 in a table and figure (s) in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavail-ability, malfunction of automatic sampling equipment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunc-tion, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operation Report. pursuant to Specification 6.9.1.6. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supportina information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location (s) for obtaining samples.
| |
| O (2)0ne or more instruments, such as a pressurized ion chamber, for measuring and Q recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet' are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., some sectors will be over water so that the number of dosimeters may be reduced accordingly.
| |
| The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.
| |
| (3)The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.
| |
| (4) Airborne particulate sample filters shall be analyzed for gross beta radio-activity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
| |
| A t]
| |
| WOLF CREEK - UNIT 1 3/4 12-7
| |
| | |
| TABLE 3.12-1 (Continued)
| |
| TABLE NOTATIONS (Continued)
| |
| (5) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
| |
| (6)The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone.
| |
| (7)A composite sample is one in which the quantity (aliquot) of liquid sampled is constant over the sampling period and in which the method of sampling employed results in a specimen that is representative of the liquid concen-trate. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.
| |
| (8) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.
| |
| (9)The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.
| |
| (10)If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.
| |
| O WOLF CREEK - UNIT 1 3/4 12-8
| |
| | |
| , /m ,x TABLE 3.12-2 8
| |
| Q; REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES c, .- .
| |
| A REPORTING LEVEL 4 9 * $, ,? 4
| |
| ' WATER AIRBORNE PARTICULATE FISH I MILK FOOD PRODUCTS
| |
| @ ANALYSIS (pCi/2) OR GASES (pCi/m3 ) (pCi/kg, t) (pCi/E) (pCi/kg, wet) p -
| |
| ~ H-3 20,000* -
| |
| Mn-54 1,000 30,000,j.i. .&
| |
| Fe-59 400 10,00 4 4
| |
| Co-58 1,000 3 000 .
| |
| -, 1 ..
| |
| Co-60 300 40,000 Zn-65 300
| |
| . $ ,000 '#
| |
| m 4 Zr-Nb-95 400 $f .
| |
| I-131 2 0.9 -
| |
| 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 20 2,000 70 2,000 50 j Ba-La-140 200 300
| |
| *For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCI/E may be used.
| |
| I
| |
| | |
| TABLE 4.12-1 O
| |
| 9 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (1)
| |
| LOWER LIMIT OF DECTECTION (LLD)(2,3)
| |
| WATER AIRBORNE PARTICULATE FISH MILK . FOOD PRODUCTS SEDIMENT E ANALYSIS (pCi/f) OR GAS (pCi/m3 ) (pCi/kg, wet) (pCi/E) (pCi/kg, wet) (pCi/kg, dry)
| |
| H Gross Beta 4 0.01 w
| |
| H-3 2000*
| |
| Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 k' -
| |
| Zn-65 30 260
| |
| [
| |
| m Q Zr-Nb-95 15 -
| |
| t I-131 1(4) 0.07 .k. ,1 60 Cs-134 15 0.05 130 15 60 -
| |
| 150 Cs-137 18 0.06 150 18 80 .,s 180 Ba-La-140 15 15
| |
| *If no drinking water pathway exists, a value of 3000 pCi/1 may be used.
| |
| ?
| |
| O O O
| |
| | |
| A r TABLE 4.12-1 (Continued)
| |
| (y {
| |
| t O) .
| |
| <?
| |
| +
| |
| g dl TABLE NOTATIONS (1)This list does not mean that only these nuclides are to be considered.
| |
| Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operat ng_ Report pursuant to Specification 6.9.1.6.
| |
| (2) Required detection"ce,csbilities for thermoluminescent dosimeters used for environmental mea /urdments shall be in accordance with the recommenda-
| |
| .tionsofRegulatoryGu[de4,13, Revision 1, July 1977.
| |
| 6
| |
| -(3)TheLLDisdefined,fbrpurposesofthesespecifications,asthesmallest concentration of radioactive material in a sample that will yield a net count, above syste'm background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation
| |
| . represents a "real" signal.
| |
| For a particular measurement system, which may include radiochemical separation: ,
| |
| 4.66 s b
| |
| ; LLD =
| |
| $ E V 2.22 Y exp (-Aat)
| |
| Where:
| |
| v LLD = the "a priori" lower limit of detection (picoCuries per unit mass or volume),
| |
| s = the stanf$rd deviation of the background counting rate or of b
| |
| the counting rate of a blank sample as appropriate (counts per minute),s a
| |
| E =stne counting efficiency (counts per disintegration),
| |
| V = the sample size (units of mass or vcbme.
| |
| 2.2,2 =-the number of disintegrations ty
| |
| * a o.e per picocurie, Y = the fractional radiochemical yield, when applicable,
| |
| +
| |
| A = the radioactive decay constant for the particular radionuclide
| |
| 'f, (s 1), and ,
| |
| At = the elapsed time Letween sample collection, or end of the sample collection period, and time of counting (s).
| |
| Typical values of E, V, Y, and At should be used in the calculation.
| |
| r WOLF CREEK - UNIT 1 3/4 12-11 d, , - -- ._. _ __ _ _ _ _ ._
| |
| | |
| TABLE 4.12-1 (Continued)
| |
| TABLE NOTATIONS (Continued)
| |
| It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of 4 measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
| |
| Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable.
| |
| In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
| |
| (4)LLD for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.
| |
| O O
| |
| WOLF CREEK - UNIT 1 3/4 12-12
| |
| | |
| 3/4.12.2 LAND USE CENSUS n'
| |
| r i LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden
| |
| * of greater than 50 m2 (500 ft2) producing broad leaf vegetation.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.
| |
| : b. With a Land Use Census identifying a location (s) that yields a calcu-lated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new location (s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was con-ducted. Pursuant to Specification 6.14, submit in the next Semiannual O)
| |
| (v Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with information supporting the change in sampling locations.
| |
| : c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
| |
| * Broad leaf vegetation sampling of available vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 3.12-1, Part 4.c. shall be followed, including l
| |
| analysis of control samples.
| |
| l C
| |
| WOLF CREEK - UNIT 1 3/4 12-13
| |
| | |
| 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 3.12-1.
| |
| APPLICABILITY: At all times.
| |
| ACTION:
| |
| : a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
| |
| : b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
| |
| SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM.
| |
| O A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
| |
| l l
| |
| l l
| |
| l O
| |
| l WOLF CREEK - UNIT 1 3/4 12-14 I
| |
| | |
| 4
| |
| \
| |
| 4 e
| |
| d BASES FOR SECTIONS 3.0 AND 4.0
| |
| , LIMITING CONDITIONS FOR OPERATION AND i
| |
| SURVEILLANCE REQUIREMENTS i
| |
| (
| |
| 4 I
| |
| f
| |
| | |
| O NOTE The BASES contained in succeeding pages summarize ;
| |
| the reasons for the Specifications in Sections 3.0 l and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.
| |
| j O
| |
| 1 l
| |
| I l
| |
| 9
| |
| | |
| g
| |
| ! ) 3/4.0 APPLICABILITY v
| |
| BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4. In the event of a disagreement between the requirements stated in these Technical Specifications and that stated in an applicable Federal Regulation or Act, the requirements stated in the applicable Federal Regulation or Act, shall take precedence and shall be met.
| |
| 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.
| |
| 3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.
| |
| 3.0.3 The specification delineates the measures to be taken for those circum-stances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specification 3.5.2 requires two independent ECCS subsystems to be OPERABLE and provides explicit ACTION requirements if one ECCS subsystem is inoperable. Under the requirements of Speci-fication 3.0.3, if both the required ECCS subsystems are inoperable, within 1 hour measures must be initiated to place the unit in at least HOT STANDBY within the next
| |
| (~N 6 hours, and in at least HOT SHUTDOWN within the following 6 hours. As a further
| |
| (
| |
| V ) example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one Containment Spray System is inoperable. Under the requirements of Specification 3.0.3 if both the required Containment Spray Systems are inoperable, within I hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours, in at least HOT SHUTDOWN within the following 6 hours, and in COLD SHUTDOWN within the subsequent 24 hours. It is acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of-service time to that period for operation in subsequent lower OPERATIONAL MODE (S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperabil-ity is discovered but the times provided for achieving a mode reduction are not applicable if the inoperability is discovered in a mode lower than the applicable mode. For example, if the Containment Spray System was discovered to be inoperable while in STARTUP, the ACTION Statement would allow up to 156 hours to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours rather than the allowed 78 hours, 140 hours would still be available before the plant would be required to be in COLD SHUTDOWN. Hcwever, if this system was discovered to be inoperable while in HOT STANDBY, the 6 hours provided to achieve HOT STANDBY would not be additive to the time available to achieve COLD SHUTDOWN so that the total allowable time is reduced from 156 hours to 150 hours.
| |
| 3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with: (1) the full complement of required systems, equipment, or components OPERABLE, and (2) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allow-(Q) v able deviations and out-of-service provisions contained in the ACTION statements.
| |
| WOLF CREEK - UNIT 1 B 3/4 0-1 t.
| |
| | |
| APPLICABILITY BASES The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded.
| |
| Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications.
| |
| 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements.
| |
| Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification.
| |
| 4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities.
| |
| The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval.
| |
| 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under these criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveillance Requirements. Items may be determined inoperable during use, during surveillance tests or in accordance with this specifica-tion. Therefore, ACTION statements are entered when the Surveillance Require-ments should have been performed rather than at the time it is discovered that the tests were not performed.
| |
| 4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MODE or other WOLF CREEK - UNIT 1 B 3/4 0-2
| |
| | |
| lp ii 1 APPLICABILITY l 'R J l
| |
| BASES applicable conditioo. The intent of this provision is to ensure that surveil-lance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.
| |
| Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.
| |
| 4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications.
| |
| This specification includes a clarification of the frequencies for performing
| |
| ,o the inservice inspection and testing activities required by Section XI of the
| |
| ( ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification A is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
| |
| Under the terms of this specification, the more restrictive requirements of the Technical ~ Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to 1 week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable.
| |
| l O_
| |
| l WOLF CREEK - UNIT 1 B 3/4 0-3
| |
| | |
| I 3/4.1 REACTIVITY CONTROL SYSTEMS
| |
| 'J BASES 3/4.1.1 80 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
| |
| SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. The most restrictive condition occurs at E0L, with T avg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% ak/k is required to control the reactivity transient.
| |
| Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg x' less than 200 F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% ak/k SHUTDOWN MARGIN provides adequate protection.
| |
| 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.
| |
| The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.
| |
| a WOLF CREEK - UNIT 1 B 3/4 1-1
| |
| | |
| REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)
| |
| The most negative MTC value equivalent to the most positive moderator d:nsity 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 t"mperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.1 x 10 4 Ak/k/ F. The MTC value of
| |
| -3.2 x 10 4 ok/k/ F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of
| |
| -4.1 x 10 4 Ak/k/ F.
| |
| 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 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 OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT NDT temperature.
| |
| 3/4.1.2 BORATION SYSTEMS The Boration Systems ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) 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 inoper-able. 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 capability require-ment occurs at E0L from full power equilibrium xenon conditions and requires 17,658 gallons of 7000 ppm borated water from the boric acid storage tanks or 83,745 gallons of 2000 ppm borated water from the RWST. With the RCS average temperature less than 350 F, only one boron injection flow path is required.
| |
| WOLF CREEK - UNIT 1 8 3/4 1-2
| |
| | |
| 7-( ) REACTIVITY CONTROL SYSTEMS O
| |
| BASES B0 RATION SYSTEMS (Continued)
| |
| 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 Baron 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 opera-tion 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% 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 14,076 gallons of 2000 ppm borated
| |
| ,w water from the RWST.
| |
| l i
| |
| () The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
| |
| The limits on contained water volume and baron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the olution 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 reactivity control while in MODE 6.
| |
| 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-Lained, and (3) the potential effects of rod misalignment on associated acci-dent 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 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of O) i
| |
| 'd 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 picked for verification of agreement with demanded position.
| |
| WOLF CREEK - UNIT 1 B 3/4 1-3
| |
| | |
| REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued)
| |
| 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 maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T avg greater than or equal to 551 F and with all reactor coolant pumps operating 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 hours with more frequent verifications required if an automatic monitoring channel is inoperable.
| |
| These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
| |
| 1 O
| |
| WOLF CREEK - UNIT 1 B 3/4 1-4
| |
| | |
| (n) 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (b) limiting
| |
| .the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.
| |
| The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
| |
| F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local 9 heat flux on the surface of a fuel rod at core elevation Z divided
| |
| ' by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F
| |
| H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.
| |
| i 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound 9
| |
| envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.
| |
| Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL
| |
| ; POWER levels are obtained by multiplying the RATED THERMAL POWER value by l the appropriate fractional THERMAL POWER level. The periodic updating of l the target flux difference value is necessary to reflect core burnup considerations.
| |
| u WOLF CREEK - UNIT 1 B 3/4 2-1
| |
| | |
| POWER DISTRIBUTION LIMITS l BASES 1
| |
| AXIAL FLUX DIFFERENCE (Continued) l Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution l sufficiently to change the envelope of peaking facters which may be reached I on a subsequent return to RATED THERMAL POWER (with the AFD within the target l band) provided the time duration of the deviation is limited. Accordingly, a 1 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance.
| |
| Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour and 2 hours, respectively.
| |
| Figure B 3/4.2-1 shows a typical monthly target band.
| |
| 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit.
| |
| Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:
| |
| : a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, from the group demand position,
| |
| : b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6, WOLF CREEK - UNIT 1 B 3/4 2-2
| |
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| -30% -20% -10% 0 +10% + 20% +30%
| |
| l lNDICATED AXIAL FLUX DIFFERENCE
| |
| /
| |
| n \ FIGURE B 3/4.2-1 t
| |
| ( .-
| |
| ) TYPICAL IllDICATED AXIAL FLUX DIFFERE'lCE VERSUS TiiERf1AL POWER WOLF CREEK - Uti!T 1 8 3/4 2-3
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| | |
| POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
| |
| : c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and
| |
| : d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
| |
| F will be maintained within its limits provided Conditions a. through H
| |
| : d. above are maintained. As noted on Figure 3.2-3, RCS flow rate and F H may be " traded off" against one another (i.e., a low measured RCS flow rateisacceptableifthemeasuredFhisalsolow)toensurethatthecalcu lated DNBR will not be below the design DNBR value. The relaxation of F H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
| |
| gascalculatedinSpecification3.2.3andusedinFigure3.2-3, accounts for F 3g less than or equal to 1.49. This value is used in the various accident analyses where F H influences parameters other than DNBR, e.g., peak clad tem perature, and thus is the maximum "as measured" value allowed.
| |
| Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins, totaling 9.1% DNBR, completely offset any rod bow penalties. This margin includes the following:
| |
| : 1) Design limit DNBR of 1.30 vs. 1.28,
| |
| : 2) Grid spacing (K ) f 0.046 vs. 0.059, s
| |
| : 3) Thermal Diffusion Coefficent of 0.038 vs. 0.059,
| |
| : 4) DNBR Multiplier of 0.86 vs. 0.88, and
| |
| : 5) Pitch Reduction.
| |
| The applicable values of rod bow penalties are referenced in the FSAR.
| |
| When an qF measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
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| O WOLF CREEK - UNIT 1 B 3/4 2-4
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| | |
| g POWER DISTRIBUTION LIMITS BASES HEAT FLUX tt0T CHANNEL-FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
| |
| The Radial' Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (z), remains within its limit. The RTP)qas provided in the Radial Peaking F
| |
| xy limit for RATED THERMAL POWER (F x Factor Limit Report per Specification 6.9.1.9 was determined from expected power control manuevers over the full range of burnup conditions in the core.
| |
| When.RCS flow rate and F H are measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-3. Measurement errors of 2.1% for RCS total flow rate and 4% for F H have been allowed for in determination of the design DNBR value.
| |
| The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venture which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, an inspection is perfnrmed of the feedwater venture each refueling outage.
| |
| The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3. This surveillance also provides adequate monitoring to detect any core crud buildup.
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| 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
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| Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
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| The limit of 1.02, at which corrective ACTION is required, provides DN8 and linear heat generation rate protection with x y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
| |
| The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such ACTION does not correct the tilt, the margin for uncertainty on F qis reinstated by reducing
| |
| , the maximum allowed power by 3% for each percent of tilt in excess of 1.
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| WOLF CREEK - UNIT 1 8 3/4 2-5
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| | |
| POWER DISTRIBUTION LIMITS BASES QUADRANT POWER TILT RATIO (Continued)
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| For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
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| 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The indicated T avg valve of 592.5 F and the indicated pressurizer pressure value of 2220 psig correspond to analytical limits of 595 F and 2205 psig respectively, with allowance for measurement unce-tainty.
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| The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
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| O WOLF CREEK - UNIT 1 B 3/4 2-6
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| p 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
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| The OPERABILITY of these systems is required to provide the overall reliability,-redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
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| The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
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| t' The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.
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| To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, aiid the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent' span, is the statistical summation of errors assumed in the analysis exc.luding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for the actuation.
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| R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either WOLF CREEK - UNIT 1 B 3/4 3-1
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| | |
| BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.
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| The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances tf these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
| |
| The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
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| The Engineered Safety Features Actuation System senses selected plant para-meters and determines whether or not predetermined limits are being exceeded.
| |
| If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) Feed-water System isolates, (4) the emergency diesel generators start, (5) contain-ment spray pumps start and automatic valves position, (6) containment isolates, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.
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| O WOLF CREEK - UNIT 1 8 3/4 3-2
| |
| | |
| fm
| |
| ( ) INSTRUMENTATION v
| |
| BASES Engineered Safety Features Actuation System Interlocks The Engineered Safety Features Actuation System interlocks perform the following functions:
| |
| P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped.
| |
| Reactor not tripped prevents manual block of Safety Injection.
| |
| P-11 On increasing pressure P-11 automatically reinstates safety injection actuation on low pressurizer pressure and low steamline pressure and automatically blocks steamline isolation on negative steamline pressure rate. On decreasing pressure; P-11 allows the manual block of Safety Injection on low pressurizer pressure and low steamline pressure and allows steamline isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI.
| |
| 3/4.3.3 MONITORING INSTRUMENTATION V 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) suffi-cient 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 they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and acutation of Emergency Exhaust or Control Room Emergency Ventilation Systems.
| |
| 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum compicment 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.
| |
| ForthepurposeofmeasuringF(Z)orFhafullincorefluxmapisused. q Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in V' recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Neutron Flux channel is inoperable.
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| WOLF CREEK - UNIT 1 0 3/4 3-3
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| | |
| r INSTRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This 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.
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| 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental 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 with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.
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| 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION 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 is lost and is consistent with General Design Criteria 3 and 19 and Appendix R of 10 CFR Part 50.
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| 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables 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.
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| 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection System ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," Revision 1, January 1977.
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| WOLF CREEK - UNIT 1 B 3/4 3-4
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| p
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| ( INSTRUMENTATION
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| \
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| BASES 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that both ade-quate warning capability is available for the prompt detection of fires and that Fire Suppression Systems, that are actuated by fire detectors, will discharge extinguishing agents in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
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| Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.
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| Consequently, the minimum number of operable fire detectors must be greater.
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| The loss of detection capability for Fire Suppression Systems, actuated by fire detectors represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that (Q)
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| U provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.
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| 3/4.3.3.9 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate 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 Primary System of Light-Water-Cooled Reactors," May 1981.
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| 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm /
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| Trip Setpoints for these instruments shall be calculated and adusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
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| k WOLF CREEK - UNIT 1 B 3/4 3-5
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| i INSTRUMENTATION BASES 3/4.3.3.11 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be adjusted to values calculated in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitor used to show compli-ance with the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10-6 pCi/cc are measurable.
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| 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed pro-tection 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 from excessive turbine overspeed is required.
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| O WOLF CREEK - UNIT 1 8 3/4 3-6
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| A
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| ( 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 operation and maintain DNBR above 1.30 during 311 normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours.
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| 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.
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| In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.
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| In MODE 5 with reactor coolant loops not filled, a single RHR loop provides
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| ( sufficient heat removal capability for removing decay heat; but single failure V considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
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| 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.
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| The restrictions on starting a reactor coolant pump in MODES 4 and 5 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 Arpendix 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.
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| 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. The relief capacity of a single safety valve is adequate to relieve any overpressure ccndition which could occur during shutdown. In the event that no safety valve 3 are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief (q capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.
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| WOLF CREEK - UNIT 1 B 3/4 4-1
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| 3/4.4 REACTOR COOLANT SYSTEM BASES SAFETY VALVES (Continued)
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| During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
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| 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.
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| 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.
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| 3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The reqcirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.
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| 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 design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
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| Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
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| 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on,a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
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| O WOLF CREEK - UNIT 1 0 3/4 4-2
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| g REACTOR COOLANT SYSTEM
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| ( )
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| v STEAM GENERATOR (Continued)
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| Unscheduled inservice inspections are performed on each steam generator following; 1) reactor to secondary tube leaks; 2) seismic occurrence greater 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 equivalent to a steam generator safety valve failing open; to ensure that steam generator tubes retain sufficient integrity for continued operation. Transients less severe than these do not require inspections because the resulting stresses are well within the stress criteria established by Regulatory Guide 1.121 which unplugged steam generator tubes must be capable of withstanding.
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| 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 limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during
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| '(3i plant operation would be limited by the limitation of steam generator tube leakage between the reactor Coolant System and the Secondary Coolant System n _) (reactor-to-secondary leakage = 500 gallons per day per steam generator).
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| 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.
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| 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.
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| Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
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| Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may
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| ,, result in a requirement for analysis, laboratory examinations, tests, addi-
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| / T tional eddy-current inspection, and revision of tFe Technical Specifications,
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| ( ) if necessary.
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| WOLF CREEK - UNIT 1 0 3/4 4-3
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| f 9 REACTOR C00LANF SYSTEMS BASES 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 boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
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| 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gmss failure of the pressure boundary. There-fore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
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| 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 tb a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.
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| 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 event of a main steam line rupture or under LOCA condtions.
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| 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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| 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.
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| 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
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| ,/
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| (
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| WOLF CREEK - UNIT 1 B 3/4 4-4
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| ( ) REACTOR COOLANT v BASES OPERATIONAL LEAKAGE (Continued) 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.
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| 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.
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| 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion s
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| V) studies show that operation may be continued with contaminant concentration levels in excess of the' Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.
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| The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
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| 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE B0UNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific. site parameters of the Wolf Creek Generating Station, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
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| O O
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| WOLF CREEK - UNIT 1 8 3/4 4-5
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| REACTOR COOLANT BASES SPECIFIC ACTIVITY (Continued)
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| The ACTION statement permitting POWER OPERATION to continue for limited 1ime periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 800 hours per year (approximately 10% of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. The reporting of cumulative operating time over 500 hours in any 6 month consecutive period with greater than 1 microcurie / gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800-hour limit.
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| Reducing T avg to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.
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| The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective ACTION. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
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| 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressurc changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G:
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| : 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
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| : a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and O
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| WOLF CREEK - UNIT 1 B 3/4 4-6
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| p M'
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| (Q REACTOR COOLANT SYSTEM BASES P_RESSURE/ TEMPERATURE LIMITS (Continued)
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| : 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.
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| : 2. :These limit lines shall be calculated periodically using methods provided below.
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| : 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70 F.
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| : 4. The pressurizer heatup and cooldown rates shall not exceed 100 F/h and 200 F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583 F.
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| : 5. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements
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| /^\ of ASME Boiler and Pressure Vessel Code, Section XI.
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| 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.cnde.
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| Heatup and cooldown limit curves are calculated using the most limiting value of the' nil-ductility reference temperature, RTNDT, at the end of 16 effective full power years (EFPY) of service life. The 16 EFPY service life period is chosen such that the limiting RT NDT at the 1/4T location in the core region is greater than the RT NDT f the limiting unirradiated material. The selection of such a limiting RT NDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.
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| The reactor vessel materials have been tested to determine their initial RTNDT;.the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by either Regulatory Guide 1.99, Revision 1, " Effects of NDT Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials" or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNDT at the end of 16 EFPY as well as adjustments
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| \ for possible errors in the pressure and temperature sensing instruments.
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| WOLF CREEK - UNIT 1 B 3/4 4-7
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| REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
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| Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NDT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure.
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| Allowable pressure-temperature relationships for various heatup and cooldown rates are calcul,ted using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A.
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| The general method for calculating heatup and cocldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.
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| In the calculation procedures a semi elliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.
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| The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabil. ties of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARINDT, c rresponding to the end of the period for which heatup and cooldown curves are generated.
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| The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K IR' IR is obtained from the reference for the metal temperature at that time. K fracture toughness curve, defined in Appendix G to the ASME Code. The K curve is given by the equation: IR K
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| IR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160H (1)
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| WOLF CREEK - UNIT 1 B 3/4 4-8
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| REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
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| Where: K IR is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RT Thus, NDT.
| |
| the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
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| CKgg + kit < KIR (2)
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| Where: Kyg = the stress intensity factor caused by membrane (pressure) stress, K
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| 7g = the stress intensity factor caused by the thermal gradients, KIR = a function of temperature relative to the RTNDT of the material as provided by the Code, C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.
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| \ At any time during the heatup or cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradiets through the vessel wall are calculated and then the corresponding thermal stress intensity factor, kit, f r the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
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| C00LDOWN For the calculation of the allowable pressure versus coolant temperature -
| |
| during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thern.al gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
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| The use of the composite curve in the cooldown analysis is necessary because
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| ! control of the cooldown procedure is based on measurement of reactor coolant I
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| temperature, whereas the limiting pressure is actually dependent on the material
| |
| ! temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID.
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| This condition, of course, is not true for the steady-state situation. It l
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| follows that at any given reactor coolant temperature, the AT developed l
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| l l
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| WOLF CREEK - UNIT 1 B 3/4 4-9 l
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| l
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| TABLE B 3/4.4-1 5 REACTOR VESSEL TOUGHNESS G
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| n 50 FT-LB m
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| ASME T
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| 35 Mil AVG. UPPER SHELF RT COMP MATERIAL Cu P NDT Temp NDT NMWD MWD
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| , COMPONENT CODE TYPE (%) (%) ( F)_ ( F) ( F) (FT-LB) (FT-LB)
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| E Closure Head Dome R2516-1 A533B, CL.1 0.12 0.010 -40 60 0 112 ----
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| Q Closure Head Torus R2515-1 A533B, CL.1 0.11 0.009 -20 <40 -20 119 ----
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| - Closure Head Flange R2504-1 A508, CL.2 ----
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| 0.013 20 <80 20 139 ----
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| Vessel Flange R2501-1 A508, CL.2 ----
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| 0.012 20 <80 20 102 ----
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| Inlet Nozzle R2502-1 A508, CL.2 ----
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| 0.010 -20 <40 -20 147 ----
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| Inlet Nozzle R2502-2 A508, CL.2 ----
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| 0.009 -20 <40 -20 137 ----
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| Inlet Nozzle R2502-3 A508, CL.2 0.11 0.010 -20 <40 -20 156 ----
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| Inlet Nozzle R2502-4 A508, CL.2 0.11 0.010 -30 <30 -30 156 ----
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| Outlet Nozzle R2503-1 A508, CL.2 ----
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| 0.006 -10 <50 -10 126 ----
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| Outlet Nozzle R2503-2 A508, CL.2 ----
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| 0.009 0 <60 0 129 ----
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| c, Outlet Nozzle R2503-3 A508, CL.2 ----
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| 0.007 0 <60 0 136 ----
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| Outlet Nozzle R2503-4 A508, CL.2 ----
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| 0.007 0 <60 0 114 ----
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| * Nozzle Shell R2004-1 A533B, CL.1 0.05 0.010 -40 70 10 118 ----
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| * Nozzle Shell R2004-2 A5338, CL.1 0.04 0.011 -40 80 20 121 ----
| |
| Nozzle Shell R2004-3 A5338. CL.1 0.04 0.008 -50 60 0 133 ----
| |
| Inter. Shell R2005-1 A5338, CL.1 0.04 0.008 -20 <40 -20 127 156 Inter. Shell R2005-2 A533B, CL.1 0.04 0.007 -30 40 -20 127 143 Inter. Shell R2005-3 A533B, CL.1 0.05 0.007 -30 40 -20 135 164 Lower Shell R2508-1 A5338, CL.1 0.09 0.009 -40 60 0 87 118 Lower Shell R2508-2 A533B, CL.1 0.06 0.008 -10 70 10 100 127 Lower Shell R2508-3 A533B, CL.1 0.07 0.008 -20 100 40 86 127 Bottom Head Torus R2517-1 A533B, CL.1 0.11 0.010 -80 30 -30 92 ----
| |
| Bottom Head Dome R2518-1 A533B, CL.1 0.12 0.009 -60 0 -60 154 ----
| |
| Inter. & Lower Shell G2.06 SAW 0.04 0.006 -50 <10 -50 150 ----
| |
| Long. Weld Seams Inter. to Lower Shell E3.16 SAW 0.05 0.007 -50 10 -50 98 ----
| |
| Girth Weld Seam NMWD - Normal to Major Working Directions MWD - Major Working Directions e G 4
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| | |
| . I l l
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| 2 - - - - - - - - - - - - - - - - - - - - --- L-
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| 1/4T 1.3 X 10'' n/cm' 10'' -
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| w h 3/4T 2.9 X 10'8 n/cm 2 2
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| O k
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| 3 18
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| ge e e 91em *h@ . N =1 -4 ee.Mi 10'#
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| 0 10 20 30 40 50 EFFECTIVE FULL POWER (YEARS)
| |
| FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION EFFECTIVE FULL POWER LIFE WOLF CREEK - UNIT 1 8 3/4 4-11
| |
| | |
| b n 400 -
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| UPPER LIMIT o ' '
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| 0.30% COPPER BASE,0.25% WELD' i [ ; , i 7 . . .
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| 4 9_.+.,
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| 7-e 0.25% COPPER BASE,0.20% WELD, I i ,
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| . 4l t. I 4i
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| -l I 6 t 6 i i I LOWER LIMIT
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| , 100 ._ j_ .4_
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| f h,g i . _.. _ . _ . _ __
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| .._-_{-..
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| I- -+- '
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| I- 2
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| ! - - - -j-- --- f -- q--
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| $ 80 - -
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| 4
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| i !. .i_.
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| f)_ 0.20% COPPER BASE,0.15% WELD !
| |
| 60 p-i - l -p- I R
| |
| f,i 0.15% COPPER B ASE,0.10% WELD i
| |
| ._. . . g _a
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| . p y' .
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| : 4 0.10% COPPER BASE,0.05% WELD E
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| N l -
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| l
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| . .e .-. _ . . _ -
| |
| 40 . .. ..
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| ..l _._. . . . . . . _ -.. ...._-..n_..
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| , , .{ l 20
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| ! ! ! l !l l l l , t 4 s a 19 2 4 s a 20 18 2 10 10 10 FAST NEUTRON FLUENCE (N/CM*, E >1 MeV)
| |
| FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RTNDT FOR REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550 F G e
| |
| | |
| 7 i \
| |
| (v ) REACTOR COOLANT SYSTEM BASES-C00LDOWN (Continued) during cooldown results in a higher value of K at the 1/4T location for IR finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in K IR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.
| |
| The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.
| |
| HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions f'_N, as well as finite heatup rate conditions assuming the presence of a 1/4T
| |
| ( <
| |
| defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K IR f r the 1/4T crack during heatup is lower than the K f r the 1/4T crack during steady-state IR conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K IR 's for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
| |
| The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
| |
| (O j Rather, each heatup rate of interest must be analyzed on an individual basis.
| |
| WOLF CREEK - UNIT 1 B 3/4 4-13
| |
| | |
| 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 pressuce 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.
| |
| Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
| |
| Although the pressurizer 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 performed in accordance with the ASME Code requirements.
| |
| 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 Aopendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less tnan 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 a ove the RCS cold leg tempera-tures, or (2) the start of a centrifugal charginh pump and its injection into a water solid RCS.
| |
| RHR RCS suction isolation valves 8701A and B are interlocked with an "A" train wide range pressure transmitter and valves 8702A and B are interlocked with a "B" train wide range pressure transmitter. Removing power from valves 8701B and 8702A, prevents a single failure from inadvertently isolating both RHR suction relief valves while maintaining RHR isolation capability for both RHR flow paths.
| |
| 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 8 3/4 4-14
| |
| | |
| A.
| |
| ( I REACTOR COOLANT SYSTEM v
| |
| BASES COLD OVERPRESSURE The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50 F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; (3) instrument uncertainties; and 4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of both Safety Injection pumps and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary coolant temperature is more than 50 F above reactor coolant temperature. Exceptions to these requirements are acceptable as described below.
| |
| Operation above 350*F but less than 375 F with only one centrifugal charging pump OPERABLE and no Safety Injection pumps OPERABLE is allowed for up to 4 hours, f'~'s\ As shown by analysis LOCAs occurring at low temperature, low pressure conditions
| |
| \s_-) can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time duration and the condition of having only one centrifugal charging .
| |
| pump OPERABLE is allowed and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed. +
| |
| Operation below 350*F but greater than 325 F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours. During low pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure - High are blocked. In normal conditions a single failure of the ESF actuation circuitry will result in the starting of at -
| |
| most one train of Safety Injection (one centrifugal charging pump, and one Safety Injection pump). For temperatures above 325 F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORVs without exceeding Appendix G limit. Given the short time duration that this condition is allowed and the low probability of a single failure causing in overpressure event during this time, the single failure of a PORV is not assumed.
| |
| Initiation of both trains of Safety Injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents.
| |
| Although COMS is required to be OPERABLE when RCS temperature is less than 368 F, operation with all centrifugal charging pumps and both Safety Injection pumps OPERABLE is acceptable when RCS temperature is greater than 350 F. Should fsg an inadvertent Safety Injection occur above 350 F, a single PORV has sufficient I j capacity to relieve the combined flow rate of all pumps. Above 350 F two RCPs
| |
| \d and all pres are safety valves are required to be OPERABLE. Operation of an WOLF CREEK - UNIT 1 8 3/4 4-15
| |
| | |
| REACTOR COOLANT SYSTEM BASES COLD OVERPRESSURE (Continued)
| |
| RCP eliminates the possibility of a 50 F difference existing between indicated and actual RCS temperature as a result of heat transport effects. Considering instrument uncertainties only, an indicated RCS temperature of 350 F is suffi-ciently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pre-e rizer safety valves provide acceptable and redundant overpressure protection.
| |
| The Maximum Allowed PORV 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 Part 50, Appendix H and in accordance with the schedule in Table 4.4-5.
| |
| 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).
| |
| Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Aadenda through Summer 1975.
| |
| 3/4.4.11 REACTOR COOLANT SYSTEM VENTS 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 Re--tor 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, " Clarification of TMI Action Plan Requirements," November 1980.
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| O WOLF CREEK - UNIT 1 8 3/4 4-16
| |
| | |
| --~ d g
| |
| \ 3/4.5 EMERGENCY CORE COOLING SYSTEMS (V
| |
| BASES 3/4.5.1- ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.
| |
| The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the' safety analysis are met.
| |
| The accumulator power operated isolation valves are considered to be
| |
| " operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.
| |
| - The limits for operation with an accumulator inoperable for any reason
| |
| \
| |
| :(m except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened,- the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.
| |
| 3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.
| |
| Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak. cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.
| |
| With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
| |
| U WOLF CREEK - UNIT 1 8 3/4 5-1
| |
| | |
| EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)
| |
| The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety Injec-tion pumps except the required OPERABLE charging pump to be inoperable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or RHR suction relief valve.
| |
| The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The Surveillance Requirements for leakage testing of ECCS check valves ensures that a failure of one valve will not cause an intersystem LOCA. The Surveillance Requirements to vent the ECCS pump casings and accessible, i.e., can be reached without personnel hazard or high radiation dose, discharge piping ensures against inoperable pumps caused by gas binding or water hammer in ECCS piping.
| |
| 3/4.5.5 REFUELING WATER STORAGE TANK The OPERAEILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for inject-ion by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.
| |
| The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
| |
| 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.
| |
| O WOLF CREEK - UNIT 1 B 3/4 5-2
| |
| | |
| 3/4.6 CONTAINMENT SYSTEMS
| |
| ( \
| |
| l BASES 3/4.6.1- PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage 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.
| |
| 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations 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, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L a r 0.75 L t, as applicable, during performance of the periodic
| |
| .p) test to account for possible degradation of the containment leakage barriers between leakage tests.
| |
| L) For reduced pressure tests, the leakage characteristics yielded by measurements L and L shall establish the maximum allowable test leakage tm am rate Ltof n t more than La ('tm Lam). In the event L tm lam is. greater than 0.7, Ltshall be specified as equal to La (P t/Pa The surveillance testing for measuring leakage rates are consistent with -
| |
| the requirements 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.
| |
| i
| |
| ,O I
| |
| 2 WOLF CREEK - UNIT 1 B 3/4 6-1
| |
| | |
| CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSUPE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from 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 psig. The limit of 1.5 psig for initial positive containment pressure will limit the total pressure to 49.5 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, whsther by fixed or portable instruments, prior to determining the average air temperature.
| |
| 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained in accordance with safety analysis requirements for the life of the facility. Structural integrity is required to ensure that the contain-ment will withstand the maximum pressure of 50 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 exterior surfaces of the contain-ment, and the Type A leakage test are sufficient to demonstrate this capability.
| |
| The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of proposed Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Pre-stressed 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 engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition ,
| |
| of the concrete (especially at tendon anchorages), the inspection procedure, l the tolerance on cracking, the results of the engineering evaluation and the corrective actions taken. I WOLF CREEK - UNIT 1 B 3/4 6-2 l
| |
| l
| |
| | |
| CONTAINMENT SYSTEMS
| |
| { h V BASES 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 36-inch containment purge supply and exhaust isolation valves are required 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 purge 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 hours 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 af 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 (O
| |
| . year regardless of the allowable hours.
| |
| Leakage integrity tests with a maximum allowable leakage rate for ccatainment purge supply and exhaust supply valves will provida early indication of resilient material seal degradation and will allow opportunity for repair before gross leak-age failures could develop. The 0.60 La leakage limit of Specification 3.6.1.2.b.
| |
| 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 LOCA or steam line break. The pressure reduction and resultant lower containment leakage 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 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 Na0H D' 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 8 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 con'ained solution volume limit includes an allowance for solution not usable ber ause of tank discharge line location or other physical characteristics. The educator flow test of 52 gpm with RWST water is equivalent to 40 gpm NaGb 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 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 have been appropriately adjusted. However, the allowable out-of-service time 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 atinosphere or pressurization of the containment and is consistent with the requirements of G0C54 thru 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time linits specified for those isolation valves designed to close auto-matically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
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| 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of con-trolling the expected hydrogen generation associated with: (1) zirconium-water 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-gen Control Systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident," Revision 2, November 1978.
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| Adequate mixing of the containment atmosphere following a LOCA is ensured by natural circulation without reliance on a hydrogen mixing systems. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.
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| WOLF CREEK - UNIT 1 B 3/4 6-4
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| 7
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| ( ') 3/4.7 PLANT SYSTEMS G
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| BASES 3/4.7.1 TURBINE CYCLE' 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
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| The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, (1971 Edition). The total relieving capacity for all valves on all of the steam lines is 18.23 x 108 lbs/h which is 115% of the total secondary steam flow of 15.85 x 106 lbs/h at 102% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.
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| {K STARTUP and/or POWER OPERATION is allowable with safety valves inoperable V) t within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases:
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| For four loop operation:
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| SP = (X) - (Y)(V) x (109).
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| Where:
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| SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, X = Total relieving capacity of all safety valves per steam line in lbs/ hour, f'N O
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| WOLF CREEK - UNIT 1 B 3/4 7-1
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| PLANT SYSTEMS BASES SAFETY VALVES (Continued)
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| Y = Maximum relieving capacity of any one safety valve in lbs/ hour.
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| 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss-of-offsite power.
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| Testing of each electric motor-driven auxiliary feedwater pump on a fixed orifice recirculation flow and ensuring a discharge pressure of greater than or equal to 1535 psig verifies the capability of each pump to deliver a total feed-water flow at the pump discharge of 575 gpm and creating pressure of 1221 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow at the pump discharge of 1145 gpm and creating a pressure of 1221 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System tempera-ture to less than 350 F when the RHR System may be placed into operation.
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| 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 4 hours with steam discharge to the atmosphere concurrent with total loss-of-offsite power and then a cooldown to 350 F at 50 F per hour. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
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| 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.
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| This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.
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| O WOLF CREEK UNIT 1 B 3/4 7-2
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| p I
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| (G 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.
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| 3/4.7.2 STEAM GENERATCR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RTNDT f 60 F and are sufficient to prevent brittle fracture.
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| [ \
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| 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses. Each independent CCW loop contains two 100% capacity pumps and, therefore, the failure of one pump does not affect the OPERABILITY of that loop.
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| 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM The OPERABI!.ITY of the Essential Service Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment dt..ing normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analysis.
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| 3/4.7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level and temperature ensure
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| -that sufficient cooling capacity is available either to: (1) provide normal cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits.
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| v WOLF CREEK - UNIT ) B 3/4 7-3
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| PLANT SYSTEMS BASES ULTIMATE HEAT SINK (Continued)
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| The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply from the Essential Service Water pumps to safety-related equipment without exceeding its design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants," March 1974.
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| 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the Control Room Emergency Ventilation System ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions. Operation of the system with the heaters operating to maintain low humidity using automatic control for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the charcoal adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50. ANSI N510-1975 and N510-1980 will be used as procedural guides for surveillance testing. Surveillance testing provides assurance that system and component performances continue to be in accordance with performance speci-fications for Wolf Creek Unit 1, including applicable parts of ANSI N509-1976.
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| 3/4.7.7 EMERGENCY EXHAUST SYSTEM The OPERABILITY of the Emergency Exhaust System ensures that radioactive materials leaking from the ECCS equipment within the pump room following a LOCA are filtereo prior to reaching the environment. Operation of the system with the heaters operating to maintain low humidity using automatic control for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the charcoal adsorbers and HEPA filters. The cpera-tion of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1975 and N510-1980 will be used as procedural guides for surveillance testing.
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| O WOLF CREEK - UNIT 1 B 3/4 7-4
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| 77 (V ) PLANT SYSTEMS BASES 3/4.7.8 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads.
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| 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 visual inspection purposes. Therefore, each snubber type may be grouped for inspection in accordance with accessibility.
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| A list of individual snubbers with detailed information of snubber loca-tion 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 Committee.
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| The determination shall be based upon the existing radiation levels and the ps expected time to perform a visual inspection in each snubber location as well i as other factors associated with accessibility during plant operations (e.g.,
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| \ 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.
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| The visual inspection 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 of inoperable 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%)
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| may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter in'spection interval will override the previous schedule.
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| The acceptance criteria are to be used in the visual inspection 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 shall not be determined OPERABLE via functional testing. Since the visual U
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| WOLF CREEK - UNIT 1 B 3/4 7-5
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| PLANT SYSTEMS BASES SNUBBERS (Continued) inspections are augmented by functional testing program, the visual inspection need not be a hands on inspection, but shall require visual scrutiny sufficient 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.
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| To provide assurance of snubber functional reliability, one of three functional testing methods are used with the stated acceptance criteria:
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| : 1. Functionally test 10% of a type of snubber with an additional 10%
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| tested for each functional testing failure, or
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| : 2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or
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| : 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.
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| Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. Duncan.
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| Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.
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| The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.
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| 3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and 5pecial Nuclear Material sources will not exceed allowable intake values.
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| WOLF CREEK - UNIT 1 B 3/4 7-6
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| N Ny PLANT SYSTEMS BASES SEALED SOURCE CONTAMINATION (Continued)
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| Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
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| 3/4.7.10 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the Fire Suppression Systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located. The Fire Suppression System consists of the water system, spray, and/or sprinklers, Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility Fire Protection Program.
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| - In the event that portions of the Fire Suppression Systems are inoperable,
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| [m (j} alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.
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| When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.
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| The Surveillance Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying either the weight or the level of the tanks. Level measurements are made by either a U.L., or F.M. approved method, or by ultrasonic measurement corrected for temperature using equipment calibrated to standards traceable to NBS. The term, " simulated fire" test signal, is interpreted to mean actuation of an auto-matic Fire Protection System by any of the release mechanisms provided, e.g.,
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| fire detectors, hand pull stations, fusible line/ mechanical, manual, hydro /
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| mechanical, etc.
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| In the event the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.
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| 3/4.7.11 FIRE BARRIER PENETRATIONS The functional integrity of the fire barrier penetrations ensures that fires will be confined or adequately retarded from spreading to adjacent portions of the facility. This design feature minimizes the possibility of a t single fire rapidly involving several areas of the facility prior to detection
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| [Q of and the extinguishing of the fire. The fire barrier penetrations are a passive element in the facility fire protection program and are subject to periodic inspections.
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| WOLF CREEK - UNIT 1 B 3/4 7-7
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| PLANT SYSTEMS BASES FIRE BARRIER PENETRATIONS (Continued)
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| Fire barrier penetrations, including cable penetration barriers, fire doors and dampers are considered functional when the visually observed condition is the same as the as-designed condition. For those fire barrier penetrations that are not in the as-designed condition, an evaluation shall be performed to show that the modification has not degraded the fire rating of the fire barrier penetration.
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| During periods of time when a barrier is not functional, either: (1) a continuous fire watch is required to be maintained in the vicinity of the affected barrier, or (2) the fire detectors on at least one side of the affected barrier must be verified OPERABLE and an hotrly fire watch patrol established, until the barrier is restored to functional status.
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| 3/4.7.12 AREA TEMPERATURE MONITORING 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 13 F.
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| O WOLF CREEK - UNIT 1 B 3/4 7-8
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| 4
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| ,m
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| ( b 3/4.8 ELECTRICAL POWER SYSTEMS V
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| BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION The'0PERABILITY of the A.C. and D.C power sources and associated distri-bution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C.
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| power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.
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| The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are con-sistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C.
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| power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. The A.C. and 0.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical O)
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| V Power Sources", December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, sub-systems, trains, components and devices, that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term verify as used in this context means to administrative 1y check by examining logs or other information to determine if certain components are out-of-service for main-tenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.
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| The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that:
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| (1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient instrumentation and control capability are available for monitoring and maintaining the unit status.
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| The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, g August 1977, and 1.137, " Fuel-0il Systems for Standby Diesel Generators,"
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| > > Revision 1, October 1979.
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| V Wolf Creek - unit 1 B 3/4 8-1
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| ELECTRIC POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUfION (Continued)
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| The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129,
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| " Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear PowerfPlants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
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| Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests cnsures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.
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| Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, cnsures the OPERABILITY and capability of ths battery.
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| Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than
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| ! the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.
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| O WOLF CREEK - UNIT 1 B 3/4 8-2
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| l n
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| ( h ELECTRICAL POWER SYSTEMS
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| \J BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected
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| .by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.
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| The Surveillance Requirements applicable to lower voltage circuit breakers and fuses provide assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker and/or fuse. Each manufacturer's molded case and metal case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within any manufacturer's brand of circuit breakers and/or fuses, it is necessary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for surveillance purposes, q
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| f \
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| b n
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| V Wolf Creek - unit 1 B 3/4 8-3
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| /D i i 3/4.9 REFUELING OPERATIONS V
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| BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
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| (1) -the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of
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| > no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations. The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portions of the Reactor Coolant System. This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses.
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| 3/4.9.2 INSTRUMENTATION 3
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| The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
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| /\ 3/4.9.3 DECAY TIME N' '] The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time
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| , has elapsed to allow the radioactive decay of the short. lived fission products.
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| This decay time is consistent with the assumptions used in the safety analyses.
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| 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be l restricted from leakage to the environment. The OPERABILITY and closure
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| ! restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
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| l The OPERABILITY of this system ensures the containment purge penetrations l will be automatically isolated upon detection of high radiation levels within l containment. The OPERABILITY of this system is required to restrict the l
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| release of radioactive materials from the containment atmosphere to the environment.
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| l i
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| 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling fs station personnel can be promptly informed of significant changes in the
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| (,m) facility status or core reactivity conditions during CORE ALTERATIONS.
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| l WOLF CREEK - UNIT 1 8 3/4 9-1
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| REFUELING OPERATIONS BASES 3/4.9.6 REFUELING MACHINE 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.
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| 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool areas ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
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| 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 incident and prevent boron stratification.
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| 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.
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| 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.
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| 3/4.9.9 CONTAINMENT VENTILATION SYSTEM The OPERABILITY of this system ensures that the containment purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
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| O WOLF CREEK - UNIT 1 B 3/4 9-2
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| I BASES 7s
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| \ \
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| ~
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| ' \}
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| 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the-safety analysis.
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| 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE The restrictions placed on spent fuel assemblies stored in Region 2 of the spent fuel pool ensure inadvertent criticality will not occur.
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| 3/4.9.13 EMERGENCY EXHAUST SYSTEM The limitations on the Emergency Exhaust System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through
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| ' the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.
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| Operation of the system with the heaters operating to maintain low humidity with automatic control for at least 10 continuous hours in a 31-day period is
| |
| - suf ficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
| |
| The OPERABILITY of this system and the resulting iodine removal capacity are
| |
| /''N consistent with the assumptions of the safety analyses. ANSI N510-1975 and N510-1980 will be used as procedural guides for surveillance testing.
| |
| (v) r 1
| |
| 1 4 l
| |
| WOLF CREEK - UNIT 1 B 3/4 9-3
| |
| | |
| ) 3/4.10 SPECIAL TEST EXCEPTIONS
| |
| 'J BASES 3/4.10.1 SHUTDOWN MARGIN 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 permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.
| |
| 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS 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) measure control rod worth, and (2) determine the reactor stability index and damping factor under xenon oscillation conditions.
| |
| 3/4.10.3 PHYSICS TESTS
| |
| / \
| |
| ( '
| |
| ) This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the RCS T 3yg slightly 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 instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 which in turn may cause the RCS T avg to fall slightly below the minimum temperature of Specification 3.1.1.4.
| |
| 3/4.10.4 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain STARTUP and PHYSICS TESTS while at low THERMAL POWER levels.
| |
| 3/4.10.5 POSITION INDICATION SYSTEM-SHUTDOWN This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements.
| |
| p v
| |
| WOLF CREEK - UNIT 1 B 3/4 10-1
| |
| | |
| . /m
| |
| \ 3/4.11 RADI0 ACTIVE EFFLUENTS (Q
| |
| BASES 3/4;11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of '
| |
| radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B,~ Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
| |
| The required detection capabilities for radioactive materials in liquid
| |
| } waste-samples are tabulated in terms of the lower limits of detection (LLDs).
| |
| (/ Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
| |
| 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose' calculation methodology and parameters in the 00CM implement the require-ments in Section III. A of Appendix I which specify that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appro-
| |
| /G priate pathways is unlikely to be substantially underestimated. The equations t
| |
| v)
| |
| WOLF CREEK - UNIT 1 B 3/4 11-1
| |
| | |
| RADI0 ACTIVE EFFLUENTS BASES DOSE (Continued) specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.
| |
| 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part
| |
| : 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.
| |
| 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.
| |
| Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
| |
| 3/4.11.2 GASE0US EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS.
| |
| The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable WOLF CREEK - UNIT 1 B 3/4 11-2
| |
| | |
| ,O
| |
| \ RADI0 ACTIVE EFFLUENTS BASES DOSE RATE (Continued) assurance that radioactive material discharged in gaseous effluents will not result .in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE B0UNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above back-ground to a MEMBER OF THE PUBLIC at or beyond the SITE B0UNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/
| |
| year to the skin. These release rate limits also restrict, at all times the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year.
| |
| The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs).
| |
| Detailed discussion of the LLD, and other detection limits can be found in
| |
| ( HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits
| |
| \' for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
| |
| 3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Section. II.8, III. A and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance _ Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive materials in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and p) i Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled N _,/
| |
| WOLF CREEK - UNIT 1 8 3/4 11-3
| |
| - _-~ - - _ -_-_. -
| |
| | |
| RADIOACTIVE EFFLUENTS BASES DOSE - NOBLE GASES (Continued)
| |
| Reactors," Revision 1," July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon either the historical average or real time atmospheric conditions.
| |
| 3/4.11.2.3 DOSE - 10 DINE-131 AND 133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.
| |
| The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111,
| |
| " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.
| |
| These equations also provide for determining the actual doses based upon either the historical average or real time conditions. The release rate specifications for Iodine-131 and 133, tritium, and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDAPY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.
| |
| 3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low a3 is reasonably achievable." This WOLF CREEK - UNIT 1 B 3/4 11-4
| |
| | |
| RADI0 ACTIVE EFFLUENTS BASES GASEOUS RADWASTE TREATMENT SYSTEM (Continued) specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. .The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
| |
| 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of 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 radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.,
| |
| 3/4.11.2.6 GAS STORAGE TANKS L
| |
| The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. 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 LOUNDARY will not exceed 0.5 rem.
| |
| This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981.
| |
| 3/4.11.3 SOLID RADI0 ACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.
| |
| 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The O specification requires the preparation and submittal of a Special Report t
| |
| ) whenever the calculated doses due to releases of radioactivity and the radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ,
| |
| .except the thyroid, which shall be limited to less than or equal to 75 mrems.
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| WOLF CREEK - UNIT 1 B 3/4 11-5
| |
| | |
| RADI0 ACTIVE EFFLUENTS BASES TOTAL DOSE (Continued)
| |
| For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and from outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been cor-rected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not appif in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifica-tions 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying oct any opera-tion that is part of the nuclear fuel cycle.
| |
| O WOLF CREEK - UNIT 1 B 3/4 11-6
| |
| | |
| (m\)
| |
| \
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| 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING v
| |
| BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this speci-fication provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.
| |
| The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required O)
| |
| (
| |
| V by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
| |
| Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determinat.on - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
| |
| 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census. Information that will provide the best results, such as door-to-door survey, aerial survey, or consulting with local agricultural authorities, shall be used. This census satisfies the re-quirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m 2 provides assurance that significant exposure pathways via leafy vegetables will be identfied and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year)
| |
| (v; of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
| |
| WOLF CREEK - UNIT 1 B 3/4 12-1
| |
| | |
| RADIOLOGICAL ENVIRONMENTAL MONITORING BASES LAND USE CENSUS (Continued)
| |
| To determine this minimum garden size, the following assumptions were made:
| |
| (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2 ,
| |
| 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.
| |
| O e
| |
| WOLF CREEK - UNIT 1 B 3/4 12-2
| |
| | |
| , \
| |
| w i
| |
| J i SECTION 5.0 ,
| |
| 6 4
| |
| i i DESIGN FEATURES l
| |
| J 4
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| 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA
| |
| , 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.
| |
| LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.
| |
| MAPS DEFINING UNRESTRICTED AREAS AND SITE B0UNDARY FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1-3 and 5.1-4. The definition of UNRESTRICTED AREA used in implementing the Radiological Effluent Technical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, estab-lished at or beyond the SITE B0UNDARY, is utilized in the Limiting Conditions Cg For Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a.
| |
| ()
| |
| 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
| |
| : a. Nominal inside diameter = 140 feet,
| |
| : b. Nominal inside height = 205 feet,
| |
| : c. Nominal thickness of concrete walls = 4 feet,
| |
| : d. Nominal thickness of concrete dome = 3 feet,
| |
| : e. Nominal thickness of concrete base slab = 10 feet,
| |
| : f. Nominal thickness of steel liner = 0.25 inch, and
| |
| : g. Net free volume = 2.5 x 106 cubic feet.
| |
| DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 60 psig and a temperature of 320 F.
| |
| O V
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| WOLF CREEK - UNIT 1 5-1
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| ACCESS ROAD o lina'** ' ~
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| WOLF CREEK METERS COOLING LAKE Note:
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| " a Meteorological Tower 4 \ 1. The exclusion-restricted g^y';E A \ .'
| |
| j e-area is a 1200 meter radius
| |
| \ f Circle Centered around Unit 1 Containment.
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| FIGURE 5.1-1 o'" '~^ b exctusroN aRen
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| '% WOLF CPEEK COOLING LAKE
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| WOLF CREEK GENERATING STATION UNIT NO. I FIGURE 5.1-2 O'' LOW POPULATION ZONE WOLF CREEK - UNIT 1 5-3
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| rigure $.1-3 j Boundary for Gaseous Effluents wolf CattR COOLpeG LAKE '
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| * than*'. !
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| nonru / wei . coo Wam c WOLF CREEK GENERATING STATION UNIT NO,8 Figure 5.1 4 Boundary for Liquid Effluents b) h \ was.ooo
| |
| %\ v _7 Nou'E No.1 m e. on. /
| |
| > tow tevet ourter a no<r- C " " '
| |
| Ei,,.,
| |
| oo== sinuctuac =/j WOLF CREEK - UNIT 1 5-5 scouo.r.s sm ,
| |
| n i.ooo l
| |
| | |
| DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 Ine core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enrichment of 3.10 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.50 weight percent U-235.
| |
| CONTROL R0D ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rods shall be hafnium, clad with stainless steel tubing.
| |
| 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
| |
| : a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
| |
| : b. For a pressure of 2485 psig, and
| |
| : c. For a temperature of 650 F, except for the pressurizer which is 680 F.
| |
| VOLUME 5.4.2 The total volume of the Reactor Coolant System, including pressurizer and surge line, is 12,135 + 100 cubic feet at a nominal T avg f 557 F.
| |
| 5.5 METEOROLOGICAL TOWER LOCATION
| |
| : 5. 5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
| |
| O WOLF CREEK - UNIT 1 5-6
| |
| | |
| r i i DESIGN FEATURES l_ ' J
| |
| -5.6 FUEL STORAGE CRITICALITY
| |
| : 5. 6.1.1 The spent fuel storage racks are designed and shall be maintained with:
| |
| : a. LA k,77 equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 2.6%
| |
| Ak/k for uncertainties as described in Section 4.3 of the FSAR.
| |
| This is based on new fuel with an enrichment of 3.50 weight percent U-235 in Region 1 and on spent fuel with combination of initial enrichment and discharge exposures, shown in Figure 5.6-1, in Region 2; and
| |
| : b. A nominal 9.14 inch center-to-center distance between fuel assemblies i
| |
| placed in the storage racks.
| |
| '5.6.1.2 The k for spentfuelstof$herac.newfuelforthefirstcoreloadingstoreddryinthe ks shall not exceed 0.98 when aqueous foam moderation is
| |
| ;n assumed, q, DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 2040 feet.
| |
| CAPACITY l 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1344 fuel assemblies.
| |
| l 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
| |
| L l
| |
| l
| |
| /* .
| |
| f l%
| |
| WOLF CREEK - UNIT 1 5-7 I
| |
| | |
| O 44,000 y 40,000 ---- - - - - - - - - - - - - - - - - - - - - - - - - - -
| |
| g .....,._.; . . . , . . , _ . u, .
| |
| 36,000 ..j.. ::q 323 . .;_ j. :.g3... .
| |
| _- } . :=: :._.9.. -
| |
| ..w..- :._.... -:.--_.
| |
| g ._.4_
| |
| cc 32,000 . . .: . . . . . . . . -- ---
| |
| 3
| |
| * T':T;f ACCEPTABLE FOR ... .. .2 ;_ _ _.._ T-! ~.J J..~ REGION 2 STORAGE -
| |
| 2 28,000 .
| |
| 4~
| |
| -; . . .. : ; . .. --"+----
| |
| d ~; , ,- pgdn.13-- . .
| |
| . { _ . .__.
| |
| "2 li;. -l--F : ! - - -
| |
| p . _ . ._ -~:. . ! .[ t
| |
| .J .
| |
| 20,000 ---- -- - - - - - - - - - -
| |
| g .2 . _ u: .J . . _ . _
| |
| o o 16,000 _
| |
| 3 ;_ __ _ NOT ACC'EPTABLE FOR!_.
| |
| ._. _..I
| |
| ~
| |
| .' . REGl0N.2 STORAGE
| |
| $ 12,000 ..
| |
| --., :- p ._ . . .: :
| |
| ^
| |
| y 8,000
| |
| ] _g,. ; g.. ,
| |
| d fl. [.
| |
| 2 4,000 ''
| |
| ' ~ ~
| |
| L- .1 ..:..-.-. .
| |
| 0 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 FUEL ASSEMBLY INITIAL ENRICHMENT, w/o U-235 FIGURE 5.6-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 WOLF CREEK - UNIT 1 5-8
| |
| | |
| s ;, . o E -
| |
| Q TABLE 5.7-1 n1 5 COMPONENT CYCLIC OR TRANSIENT LIMITS-E i
| |
| c: CYCLIC OR DESIGN CYCLE
| |
| {
| |
| ~
| |
| COMPONENT TRANSIENT LIMIT OR TRANSIENT.
| |
| Reactor Coolant System 200 heatup cycles at < 100*F/h Heatup cycle - T avg from 5 200*F and 200 cooldown cycles at to 2 550*F.
| |
| < 100*F/h. Cooldown cycle - T avg fra 3 550 F to 1 200*F.
| |
| 200 pressurizer cooldown cycles Pressurizer cooldown cycle at $ 200 F/h. temperatures from 1 650*F to 5 200*F.
| |
| 80 loss of load cycles, without 2 15% of RATED THERMAL POWER to immediate Turbine or Reactor trip. 0% of RATED THERMAL POWER.
| |
| 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical A.C. electrical power. ESF Electrical. System.
| |
| 80 cycles of loss of flow in one. Loss'of only.one reactor reactor coolant loop. coolant pump.
| |
| 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER.
| |
| 10 auxiliary spray acutation Spray water temperature differential cycles. > 320 F.
| |
| 50 leak tests. Pressurized to 1 2485 psig.
| |
| 5 hydrostatic pressure tests. Pressurized to.1 3106 psig.
| |
| Secondary Coolant System 1 large steam line break. Break in a > 6-inch steam line.
| |
| 5 hydrostatic pressure tests. Pressurized to 1 1350 psig.
| |
| A
| |
| | |
| A 4.em-n4-._m.J_.m ,aem-..s-, - ewa m4ame.4m,_M4 -w-.me..e.a_m.m ._.1- b 4 A_._ea--m.4.a.a .-m maa.ew__--. -e a sm 6 m.-M-A-45de AAh.4 et4 m _-.W4 _ -.4. __a_
| |
| 4 i
| |
| 4
| |
| =
| |
| 2 t
| |
| l I
| |
| i ,
| |
| I l
| |
| SECTION 6.0 ADMINISTRATIVE CONTROLS r
| |
| 6 y
| |
| 9 l
| |
| I l ,
| |
| i.
| |
| i.
| |
| l l
| |
| L
| |
| '-t-_.
| |
| | |
| ')~J l
| |
| ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall Unit operation and shall delegate in writing the succession to this responsibility during his absence.
| |
| 6.1. 2 The Supervising Operator, under the Shift Supervisor, shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President-Nuclear shall be reissued to all station personnel on an annual basis.
| |
| 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1.
| |
| UNIT STAFF 6.2.2 The Unit organization shall be as shown in Figure 6.2-2 and:
| |
| 7 a. Each on duty shift shall be composed of at least the minimum shift
| |
| ( ])
| |
| N _ ,-
| |
| crew composition shown in Table 6.2-1;
| |
| : b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the Unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Operator shall be in the control room;
| |
| : c. An individual from the Health Physics Group *, qualified in radiation protection procedures, shall be on site when fuel is in the reactor;
| |
| : d. ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
| |
| : e. A site Fire Brigade of at least 5 members
| |
| * shall be maintained onsite at all times. The Fire Brigade shall not include the Shift Supervisor, and the two other members of the minimum shift crew necessary for safe shutdown of the Unit and any personnel required for other essential functions during a fire emergency; and r
| |
| *May be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action l(])
| |
| l /
| |
| v is taken to fill the required positions, WOLF CREEK - UNIT 1 6-1
| |
| | |
| ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)
| |
| : f. Administrative procedures shall be developed and implemented to limit the working hours of Unit Staff who perform safety-related functions; e.g., Senior Operators, Operators, Health Physicists, Auxiliary operators, and key maintenance personnel.
| |
| The amount of overtime worked by Unit Staff members performing safety related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. R2-12).
| |
| O e
| |
| WOLF CREEK - UNIT 1 6-2
| |
| | |
| m f \
| |
| -S-T x ang
| |
| (----
| |
| m g
| |
| DNWAN OF TE BOAM
| |
| ----FOR MTTERS OF AN RWEDIATE NATURE THE 8
| |
| n VICE PRES 10ENT . NUCLEAR CORWUMCATES i
| |
| i M I l OMECTLY MTH TE PIES 0ENT AND CHAllWAN r' I 0F THE BOARD.
| |
| A GROUP g
| |
| * VICE PRESIDENT c I TEDNCAL SENCES I E I
| |
| " l
| |
| ! ~ l l \
| |
| l L _ _ __ VICE PRES 0ENT
| |
| ; NUCLEAR NUCLEAR SAFETY REVIEW CORNETEE WSRO
| |
| . I I O OMECTOR DIRECTOR ENGINEERmG AND OMECTOR TEDWCAL SEmHS OUAUTY
| |
| .hllCLEAR i
| |
| I I I I I I I i PROCURE g y W WAGER M M GER MMGER W M GER SUR.
| |
| MANAGER gggg OUAUTY E NAGER 1 & NTERIALS NUCLEAR OPERATIONS NUCLEAR PLANT MANAGEMENT GUAUTY ADRNMSTRATIVE ASSURANCE CONSTRUCT 0N SUPERINTENDEF.= TRANNG SUPPORT ENGINEEENG SYSTEMS CONTROL SERVICES WCGS I
| |
| PLANT MANAGER MANAGER MANAGER OUAUTY 1 NUCLEAR ASSLM NCE l WCGS SERVICES 910ME OFFICO 4
| |
| ; FIGURE 6.2-1 ;
| |
| 0FFSITE ORGANIZATION i
| |
| 1
| |
| | |
| s k
| |
| ,, .r.
| |
| b -
| |
| G; n
| |
| @ Plant Manager E
| |
| l '
| |
| E Z ..
| |
| g -Supt. of Technical -Supt. of Regulato.y, Quality
| |
| -Supt. of Mair:tenance -Supt. of Operations Support -Supt. of Plant Support and Administrative Services
| |
| -Maint Services Supv -Ops Coord - Ops # -Reactor Engineering -Emergency Planning
| |
| -Fire Protection.
| |
| -Mai6a Support Supv -Instrument & Centrols -Document Control
| |
| -Shift Supervisor -Secutiry.
| |
| -Mechanical l-C "*
| |
| -Electrical -Supervising Opr -Results Engineering -Administration
| |
| -Reactor Opr -Chemistry
| |
| -Station Operator e -Utility Helper -Health Physics
| |
| * 5 -Ops Coordinator -
| |
| l Planning and Projects
| |
| -Engr / Specialist
| |
| -Surveillance Coord
| |
| *For technical matters of an immediate nature the respective individual reports directly to the Plant Manager.
| |
| #This position requires an SRO License.
| |
| FIGURE 6.2-2 UNIT ORGANIZATION e e e
| |
| | |
| TABLE 6.2-1 V)
| |
| (
| |
| MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6 SS 1 1^
| |
| SR0 1 None R0 2 1 50 4 1 STA 1** None CHM 1 None SS -
| |
| Shift Supervisor with a Senior Operator _ license on Unit 1 SR0 -
| |
| Individual with a Senior Operator license on Unit 1 R0 -
| |
| Individual with an Operator license on Unit 1 S0 -
| |
| Station Operator STA - Shift Technical Advisor CHM - Chemistry Personnel The Shift Crew Composition may be one less than the minimum requirements of pS's Table 6.2-1 for a_ period of time not to exceed 2 hours in order to accommodate 5 ) unexpected absence of on-duty shift crew members provided immediate action is
| |
| *x d taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
| |
| During any absence of the Shift Supervisor from the control room while the unit is in MODE 1, 2, 3, or 4, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the Unit is in MODE 5 or 6, an individual with a valid Operator license (other than the Shift Technical Advisor) shall be designated to assume the control room command function.
| |
| *0ne SRO, either Shift Supervisor or Supervising Operator.
| |
| **The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift
| |
| ' Supervisor or the individual with a Senior Operator license meets the qualifications for the STA as required by the NRC.
| |
| f'\
| |
| G WOLF CREEK UNIT 1 6-5
| |
| | |
| ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)
| |
| FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, NRC issuances, industry advisories, REPORTABLE EVENTS and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving plant safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities or other means of improving plant safety to the Manager Nuclear Safety.
| |
| COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field.
| |
| RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of plant activities to provide independent verification
| |
| * that these activities are performed correctly and that human errors are reduced as much as practical.
| |
| RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to Manager Nuclear Safety.
| |
| 6.2.4 SHIFT TECHNICAL ADVISOR The Shift Technical Advisor (STA)** shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering and plant analysis with regard to the safe operation of the Unit.
| |
| 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the Unit Staff shall meet or exceed the minimum qualifications of ANSI /ANS 3.1-1978, except for the Site Health Physicist who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 for a Radiation Protection Manager. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all '3censees.
| |
| *Not responsible for sign-off function.
| |
| **The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Supervisor or the individual with a Senior Operator license meets the qualifications for the STA as required by the NRC.
| |
| WOLF CREEK - UNIT 1 6-6
| |
| | |
| /%
| |
| V) i ADMINISTRATIVE CONTROLS 6.4 TRAINING' 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5 of ANSI /ANS 3.1-1978 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the ISEG or another plant group.
| |
| 6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC) ,
| |
| FUNCTION 6.5.1.1 The PSRC shall function to advise the Plant Manager on all matters related to nuclear safety.
| |
| COMPOSIT'AN p 6.5.1.2 The PSRC shall be composed of the:
| |
| Chairman: Plant Manager Member: Superintendent of Operations Member: Superintendent of Technical Support Member: Superintendent of Maintenance Member: Instrument and Control Supervisor Member: Reactor Engineering Supervisor Member: Health Physicist Member: Chemist Member: Results Engineering Supervisor Member: Superintendent of Plant Support Member: Superintendent of Regulatory, Quality and Administrative Services ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PSRC Chairman to serve on a temporary basis; however no more than two alternates shall participate as voting members in PSRC activities at any one time.
| |
| [
| |
| (
| |
| WOLF CREEK - UNIT 1 6-7
| |
| | |
| ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.1.4 The PSRC shall meet at least once per calendar month and as convened by the PSRC Chairman or his designated alternate.
| |
| QUORUM 6.5.1.5 The quorum of the PSRC necessary for the performance of the PSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.
| |
| RESPONSIBILITIES 6.5.1.6 The PSRC shall be responsible for:
| |
| : a. Review of: (1) all procedures required by Specification 6.8 and changes thereto, (2) all programs required by Specification 6.8 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety;
| |
| : b. Review of all proposed changes, tests and experiments which may involve an unreviewed safety question as defined in Section 50.59, 10 CFR;
| |
| : c. Review of all proposed changes to Technical Specifications or the Operating License;
| |
| : d. Review of all safety evaluations performed under the provision of Section 50.59(a)(1),10 CFR, for changes, tests and experiments;
| |
| : e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence to the Director, Nuclear Operations, and to the Nuclear Safety Review Committee (NSRC);
| |
| : f. Review of all REPORTABLE EVENTS;
| |
| : g. Review of reports of operating abnormalities, deviaM ons from expected performance of plant equipment and of unanticipated deficiencies in the design or operation of structures, systems or components that affect nuclear safety;
| |
| : h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman, NSRC;
| |
| : i. Review of the plant Security Plan and implementing procedures and shall submit recommended changes to the NSRC;
| |
| : j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the NSRC; WOLF CREEK - UNIT 1 6-8
| |
| | |
| .n
| |
| -( ) ADMINISTRATIVE CONTROLS V
| |
| RESPONSIBILITIES (Continued)
| |
| : k. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL and the Radwaste Treatment Systems, and
| |
| : 1. Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Plant Manager and to the Nuclear Safety Review Committee.
| |
| 6.5.1.7 The PSRC shall:
| |
| : a. Recommend in writing to the Plant Manager appioval or disapproval of items considered under Specification 6.5.1.6a. through d. above,
| |
| : b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6a. through e. above constitutes an unreviewed safety question, and
| |
| : c. Provide written notification within 24 hours to the Director Nuclear Operations and the Nuclear Safety Review Committee of disagreement between the PSRC and the Plant Manager; however, the Plant Manager
| |
| [\ss)/'
| |
| shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.
| |
| RECORDS
| |
| : 6. 5.1. 8 The PSRC shall maintain w'ritten minutes of each PSRC meeting that, at a minimum, document the results of all PSRC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Director Nuclear Operations and the Nuclear Safety Review Committee.
| |
| 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE (NSRC)
| |
| FUNCTION 6.5.2.1 The NSRC shall function to provide independent review and audit of designated activities in the areas of:
| |
| : a. Nuclear power plant operations,
| |
| : b. Nuclear engineering,
| |
| : c. Chemistry and radiochemistry,
| |
| : d. Metallurgy,
| |
| ' /'~'N
| |
| \ _, / e. Instrumentation and control,
| |
| : f. Radiological safety, WOLF CREEK - UNIT 1 6-9
| |
| | |
| ADMINISTRATIVE CONTROLS FUNCTION (Continued)
| |
| : g. Mechanical and electrical engineering, and
| |
| : h. Quality assurance practices.
| |
| The NSRC shall report to and advise the Vice President-Nuclear on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.
| |
| COMPOSITION 6.5.2.2 The NSRC shall be composed of at least the following:
| |
| Chairman: Manager Nuclear Services Member: Director Engineering and Technical Services Member: Manager Quality Assurance (Home Office)
| |
| Member Director Nuclear Operations Member: Manager Licensing and Radiological Services Member. Vice President-Engineering Member: Manager Nuclear Safety Additional members and Vice Chairman may be appointed by the Chairman.
| |
| ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NSRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSRC activities at any one time.
| |
| CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NSRC Chairman to provide expert advice to the NSRC.
| |
| MEETING FREQUENCY 6.5.2.5 The NSRC shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter.
| |
| QUORUM 6.5.2.6 The quorum of the NSRC necessary for the performana of the NSRC review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least two-thirds of the NSRC members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the Unit.
| |
| WOLF CREEK UNIT 1 6-10
| |
| | |
| ['' ADMINISTRATIVE CONTROLS REVIE_W' 6.5.2.7 The NSRC shall be responsible for the review of:
| |
| : a. The safety evaluations for: (1) changes to procedures, equipment, systems or facilities, and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question;
| |
| : b. Proposed changes to procedures, equipment, systems, or facilities which involve an unreviewed safety question as defined in Section 50.59, 10 CFR;
| |
| : c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR;
| |
| : d. Proposed changes to Technical Specifications or this Operating License;
| |
| : e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;
| |
| : f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety;
| |
| : g. All REPORTABLE EVENTS;
| |
| : h. All recognized indications of an unanticipated deficiency in some aspect of design of operation of structures, systems, or components that could affect nuclear safety; and
| |
| : i. Reports and meeting minutes of the PSRC.
| |
| AUDITS 6.S.2.8 Audits of Unit activities shall be performed under the cognizance of the NSRC. These audits shall encompass:
| |
| : a. The conformance of Unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months;
| |
| : b. The performance, training and qualifications of the entire Unit Staff at least cnce_per 12 months;
| |
| : c. The results of actions taken to correct deficiencies occurring in N unit equipment, structures, systems, or method of operation that affect nuclear safety at least once per 6 months; WOLF CREEK - UNIT 1 6-11 1
| |
| | |
| ADMINISTRATIVE CONTROLS AUDITS (Continued)
| |
| : d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months;
| |
| : e. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA or ISEG personnel;
| |
| : f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year;
| |
| : g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;
| |
| : h. The ODCM and implementing procedures at least once per 24 months;
| |
| : i. The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months;
| |
| : j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months;
| |
| : k. The Emergency Plan and implementing procedures at least once per 12 months;
| |
| : 1. The Security Plan and implementing procedures at least once per 12 months; and
| |
| : m. Any other area of Unit operation considered appropriate by the NSRC or the Vice President-Nuclear.
| |
| RECORDS 6.5.2.9 Records of NSRC activities shall be prepared, approved, and distributed as indicated below:
| |
| : a. Minutes of each NSRC meeting shall be prepared, reviewed by partici-pating members and forwarded to the Vice President-Nuclear within 14 days following each meeting;
| |
| : b. Reports of reviews encompassed by Specification 6.5.2.7 above, shall be prepared, reviewed by participating members and forwarded to the Vice President-Nuclear within 14 days following completion of the review; and WOLF CREEK - UNIT 1 6-12
| |
| | |
| ,m v() ADMINISTRATIVE CONTROLS RECORDS (Continued)
| |
| : c. Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the Vice President-Nuclear and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization. ,
| |
| 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
| |
| : a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 of 10 CFR Part 50, and
| |
| : b. Each REPORTABLE EVENT shall be reviewed by the PSRC and submitted to the NSRC and the Vice President-Nuclear.
| |
| 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
| |
| : a. The NRC Operations Center shall be notified by telephone as soon as
| |
| ./S possible and in all cases within 1 hour. The Vice President-Nuclear
| |
| \ and the NSRC shall be notified within 24 hours;
| |
| : b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PSRC. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective ACTION taken to prevent recurrence;
| |
| : c. The Safety Limit Violation Report shall be submitted to the Commission, the NSRC and the Vice President-Nuclear within 14 days of the violation; and
| |
| : d. Critical operation of the Unit shall not be resumed until authorized by the Commission.
| |
| 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained
| |
| ! covering the activities referenced below:
| |
| : a. The applicable procedures recommended in Appendix A, of Regulatory Guide 1.33, Revision 2, February 1978;
| |
| : b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Sec-tion 7.1 of Generic Letter No. 82-33;
| |
| : c. Security Plan implementation; V d. Emergency Plan implementation; WOLF CREEK - UNIT 1 6-13
| |
| | |
| ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : e. Process Control Program implementation;
| |
| : f. 0DCM implementation; and
| |
| : g. Quality Assurance Program implementation for effluent and environ-mental monitoring.
| |
| Major Procedures, supported by appropriate Minor Procedures (such as checkoff lists, operating instructions, data sheets, alarm responses, etc.), shall be provided for the above activities.* A Major Procedure is a procedure which con-trols safety related activities, and establishes one or more basic controls, overall responsibilities, authority assignments or administrative and opera-tional ground rules at the Wolf Creek plant. Major Procedures are written to meet the requirements of ANSI N18.7-1976/ANS 3.2 and generally are supported by Minor Procedures which provide delineation of details such as for valve lineups, calibration procedures, operating instructions, data sheets, alarm responses, and other procedures identified as " supporting." Major Procedures require sig-nature approval in all cases by the Plant Manager or a Call Superintendent in his absence. A Minor Procedure is a procedure which controls safety-related activities in support of a Major Procedure. It addresses a specific topic or sub-topic established by its ' parent' Major Procedure, expanding on it by pro-viding working level instructions. Minor Procedures are not permitted to con-tradict requirements contained in their governing Major Procedure. Minor Procedures require signature approval by the Plant Manager, or a Call Super-intendent in his absence, only at Revision '0.'
| |
| 6.8.2 Approval of Procedures
| |
| : a. All Major Procedures of the categories listed in Specification 6.8.1 and modifications to the intent thereof shall be reviewed by the PSRC and approved by the Plant Manager prior to implementation and reviewed periodically as set forth in Administrative Procedures.
| |
| : b. Minor Procedures (checkoff lists, operating instructions, data sheets, alarm responses, chemistry and analytical procedures, technical instructions, special and routine maintenance procedures, laboratory manuals, etc.) shall, prior to initial use, be approved by the PSRC or a Subcommittee thereof.
| |
| : c. Corporate Emergency Plan implementing procedures shall be reviewed by appropriate corporate and plant personnel and approved by the Vice President-Nuclear as set forth in General Procedures.
| |
| *With the exception of Corporate Emergency Plan implementing procedures.
| |
| Corporate Emergency Plan implementing procedures shall be provided but shall not be designated as major or minor procedures.
| |
| WOLF CREEK - UNIT 1 6-14
| |
| | |
| l
| |
| ,q ADMINISTRATI'VE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.3 Ch'anges to Procedures
| |
| : a. Temporary changes to Major Procedures, of the categories listed in Specification 6.8.1 which do not change the intent or generate an unreviewed safety question of the original or subsequent approved procedure, may be made provided such changes to operating procedures are approved by the Shift Supervisor (SR0 licensed) and one of the Call Superintendents. For temporary changes to Major Procedures under the jurisdiction of Maintenance, Instrumentation and Control, Reactor Engineering, Chemistry, or Health Physics which do not change the intent or generate an unreviewed safety question, changes may be made upon approval of the Cognizant Group Leader and a Call Superintendent.
| |
| All temporary changes to Major Procedures (made by a Call Super-intendent and either a Cognizant Group Leader or the Shift Supervisor) shall subsequently be reviewed by the PSRC and approved by the Plant Manager within 14 days, except that temporary changes to Major Pro-cedures made during a refueling outage may be reviewed and approved at any time prior to initial criticality of the reload core. All permanent changes to Major Procedures shall be made in accordance
| |
| _(9
| |
| ( /
| |
| with Specification 6.8.2.a.
| |
| : b. All temporary or permanent changes to Minor Operating Procedures (checkoff lists, alarm responses, data sheets, operating instructions, etc.) shall be approved by the Shift Supervisor, and shall be subse-quently reviewed and approved by the Operations PSRC Subcommittee.
| |
| All temporary or permanent changes to other Minor Procedures under the jurisdiction of Maintenance, Instrumentation and Control, Reactor Engineering, Chemistry, or Health Physics, shall be approved by a Cognizant Group Leader and shall be subsequently reviewed and approved by the appropriate PSRC Subcommittee.
| |
| : c. Temporary chang - to Corporate Emergency Plan implementing procedures may be made pro ided that: (1) the intent of the original procedure is not altered, (2) the change is approved by the Emergency Planning Coordinator, and (3) the change is documented, reviewed by appropriate Corporate and plant personnel and approved by the Vice President-Nuclear within 14 days of the implementation.
| |
| 6.8.4 The following programs shall be established, implemented, and maintained:
| |
| : a. Reactor Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The (G
| |
| D) systems include the appropriate portions of the Containment Spray System, Safety Injection System, Chemical and Volume Control System, RHR System, and the Nuclear Sampling System (PASS only). The program shall include the following:
| |
| WOLF CREEK UNIT 1 6-15
| |
| | |
| ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
| |
| : 1) Preventive maintenance and periodic visual inspection requirements, and
| |
| : 2) Integrated leak test requirements for each system at refueling cycle intervals or less,
| |
| : b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
| |
| : 1) Training of personnel,
| |
| : 2) Procedures for monitoring, and
| |
| : 3) Provisions for maintenance of sampling and analysis equipment.
| |
| : c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:
| |
| : 1) Identification of a sampling schedule for the critical variables and control points for these variables,
| |
| : 2) Identification of the procedures used to measure the values of the critical variables,
| |
| : 3) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,
| |
| : 4) Procedures for the recording and management of data,
| |
| : 5) Procedures defining corrective action for all off-control point chemistry conditions, and
| |
| : 6) A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
| |
| : d. Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:
| |
| : 1) Training of personnel,
| |
| : 2) Procedures for sampling and analysis, and
| |
| : 3) Provisions for maintenance of sampling and analysis equipment.
| |
| WOLF CREEK - UNIT 1 6-16
| |
| | |
| i s
| |
| I i ADMINISTRATIVE CONTROLS
| |
| 'J 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.
| |
| i STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
| |
| 6.9.1.2 The Startup Report shall address each of the tests identified in the Final Safety Analysis Report FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective ACTIONS that were required to obtain satisfactory operation shall also be described. Any additional specific details required e's in license conditions based on other commitments shall be included in this
| |
| ( ) report.
| |
| LJ 6.9.1.3 Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption of commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.
| |
| ANNUAL REPORTS 6.9.1.4 Annual Reports covering the activities of the Unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.
| |
| 6.9.1.5 Reports required on an annual basis shall include:
| |
| : a. Tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrems/yr and their associated man rem exposure according to work and job functions,* e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance
| |
| ,_s
| |
| ~
| |
| (describe maintenance), waste processing, and refueling. The dose
| |
| [ \
| |
| V
| |
| *This tabulation supplements the requirements of S20.407 of 10 CFR Part 20.
| |
| WOLF CREEK - UNIT 1 6-17
| |
| | |
| ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.
| |
| In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work func-tions; and
| |
| : b. Documentation of all challenges to the PORVs or safety valves.
| |
| ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
| |
| : 6. 9.1. 6 Routine Annual Radiological Environmental Operating Reports covering the operation of the Unit during the previous calendar year shall be submitted by May 1 of each year. The initial report shall be submitted by May 1 of the year following initial criticality.
| |
| The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activites for the report period, including a comparison with preoperational studies, with operational controls and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the Land Use Census required by Specification 3.12.2.
| |
| The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
| |
| The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps
| |
| * covering all sampling locations keyed to a table giving distances and directions from the centerline of the reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective actions being taken if the specified program is not being performed as required by Specification 3.12.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by Specification 3.12.1 and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environmental sample measure-ments that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to Specification 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
| |
| *0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
| |
| WOLF CREEK - UNIT 1 6-18
| |
| | |
| / ~'N ADMINISTRATIVE CONTROLS
| |
| \ )
| |
| 1 V
| |
| i SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.9.1.7 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the Unit during the previous 6 months of operation shall be sub-mitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.
| |
| The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the Unit as outlined in Regulatory Guide 1.21, " Measuring, Eval-uating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix 8 shall be supplemented with three additional categories:
| |
| class of solid waste (as defined by 10 CFR Part 60), type of container (e.g.,
| |
| LSA, Type A, Type B, large Quantity), and SOLIDIFICATION agent or absorbent (e.g., cement, urea formaldehyde).
| |
| The Semiannual Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an annual summary of nourly meteorological data collected over the previous year. This annual summary may T be either in the form of an hour-by-hour listing on magnetic tape of wind speed, s ) wind direction, atmospheric stcbility, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the Unit or Station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figures 5.1-3 and 5.1-4) during the report period
| |
| , using historical average atmosphere conditions. All assumptions used in making these assessments, i.e. , specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as deter-mined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance wi+h the methodology and parameters in the ODCM.
| |
| The Semiannual Radioactive Effluent Release Report to be submitted within 4 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent
| |
| , pathways and direct radiation, for the previous calendar year to show conformance
| |
| *In lieu of submission with the Semiannual Radioactive Effluent Release Report, l the licensee has the option of retaining this summary of required meteorological l data on site in a file that shall be provided to the NRC upon request.
| |
| WOLF CREEK - UNIT 1 6-19 l
| |
| | |
| ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT REPORT (Continued) with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.
| |
| The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the report-ing period.
| |
| The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PCP and the ODCM, pursuant to Specifica-tions 6.13 and 6.14, respectively, as well as any major changes to Liquid, Gas-eous, or Solid Radwaste Treatment Systems, pursuant to Specification 6.15. It shall also include a listing of new locations for dose calculations identified by the Land Use Census, pursuant to Specification 3.12.2.
| |
| The Semiannual Radioactive Effluent Release Reports shall also include the following information: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.
| |
| MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.
| |
| RADIAL PEAKING FACTOR LIMIT REPORT RTP 6.9.1.9 The F xy limits for RATED THERMAL POWER (Fxy ) shall be provided to the NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, Attention: Chief, Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington,.D.C. 20555 for all core planes containing Bank "D" control rods and all unrodded core planes and the plot of predicted (F q T ,pRel) vs Axial Core Height with the limit envelope at least 60 days prior to each cycle initial criticality unless otherwise approved by the Commission by letter. In addition, in the event that the limit should change requiring a new submittal or an amended submittal to the Peaking Factor Limit Report, it shall be submitted 60 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. Any information needed to support F
| |
| xy will be by request from the NRC and need not be included in this report.
| |
| WOLF CREEK - UNIT 1 6-20
| |
| | |
| / ' ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for ehch report.
| |
| 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
| |
| 6.10.1 The following records shall be retained for at least 5 years:
| |
| : a. Records and logs of unit operation covering time interval at each power level;
| |
| : b. Records and logs of_ principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety;
| |
| : c. All REPORTABLE EVENTS;
| |
| : d. Records of surveillance activities, inspections, and calibrations h required by these Technical Specifications;
| |
| [O e. Records of changes made to the procedures required by Specification 6.8.1;
| |
| : f. Records of radioactive shipments;
| |
| : g. Records of sealed source and fission detector leak tests and results; and
| |
| : h. Records of annual physical inventory of all sealed source material of record.
| |
| 6.10.2 The following records shall be retained for the duration of the Unit Operating License:
| |
| : a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report;
| |
| : b. Records of new and irradiated fuel inventory, f uei transfers and assembly burnup histories;
| |
| : c. Records of radiation exposure for all individuals entering rad'ation control areas; U
| |
| WOLF CREEK - UNIT 1 6-21
| |
| | |
| ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued)
| |
| : d. Records of gaseous and liquid radioactive material released to the environs;
| |
| : e. Records of transient or operational cycles for those Unit components identified in Table 5.7-1;
| |
| : f. Records of reactor +ests and experiments;
| |
| : g. Records of training and qualification for current members of the Unit Staff;
| |
| : h. Records of in-service inspections performed pursuant to these Technical Specifications;
| |
| : i. Records of Quality Assurance activities required by the QA Manual;
| |
| : j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59;
| |
| : k. Records of meetings of the PSRC and the NSRC;
| |
| : 1. Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance records;
| |
| : m. Records of secondary water sampling and water quality; and
| |
| : n. Records of analysis required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.
| |
| 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations inoilving personnel radiation exposure.
| |
| 6.12 HIGH RADIATION AREA 6.12.1 Pursuant to Paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the
| |
| " control device: or " alarm signal" required by Paragraph 20.203(c)(2), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radiation is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radia-tion source or from any surface which the radiation penetrates shall be barri-caded and conspicuously posted as a high radiation area and entrance thereto j shall be controlled by requiring issuance of a Radiation Work Permit (RWP).
| |
| WOLF CREEK - UNIT 1 6-22
| |
| | |
| [] ADMINISTRATIVE CONTROLS D
| |
| HIGH RADIATION AREA (Continued)
| |
| Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protec-tion procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
| |
| : a. A radiation monitoring device which continuously indicates the radiation dose rate in the area, or
| |
| : b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them, or
| |
| : c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring
| |
| /9 device, who is responsible for providing positive
| |
| ! ) control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Site Health Physicist in the RWP.
| |
| 6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Supervisor / Supervising Operator on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allow-able stay time for individuals in that area. In lieu of the stay time specifi-cation of the RWP, direct or remote (such as closed-circuit TV cameras) contin-uous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being per-formed within the area.
| |
| For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR con-tainment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.
| |
| o WOLF CREEK UNIT 1 6-23
| |
| | |
| ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP.shall be approved by the Commission prior to implementation.
| |
| 6.13.2 Licensee-initiated changes to the PCP:
| |
| : a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
| |
| : 1) Sufficiently detailed information to tot' ally support the rationale for the change wthout benefit of additional or supplemental information;
| |
| : 2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
| |
| : 3) Documentation of the fact that the change has been reviewed and found acceptable by the PSRC.
| |
| : b. Shall become effective upon review and acceptance by the PSRC.
| |
| 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
| |
| 6.14.2 Licensee-initiated changes to the ODCM:
| |
| : a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which change (s) was made effective. This submittal shall contain:
| |
| : 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s);
| |
| : 2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and
| |
| : 3) Documentation of the fact that the change has been reviewed and found acceptable by the PSRC.
| |
| : b. Shall become effective upon review and acceptance by the PSRC.
| |
| 6.15 MAJOR CHANGES TO LIQUID, GASE0US, AND SOLID RADWASTE TREATMENT SYSTEMS
| |
| * 6.15.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):
| |
| * Licensees may choose to submit the information called for in this specification as part of the annual FSAR update.
| |
| WOLF CREEK - UNIT 1 6-24
| |
| | |
| gm ADMINISTRATIVE CONTROLS
| |
| ( )i MAJOR CHANGES TO LIQUID, GASE0US, AND SOLID RADWASTE TREATMENT SYSTEMS (continued)
| |
| : a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PSRC. The discussion of each change shall contain:
| |
| : 1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;
| |
| : 2) Sufficient detailed information.to totally support the reason for the change without benefit of additional and supplemental information;
| |
| : 3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems.
| |
| : 4) An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;
| |
| : 5) An evaluation-of the change, which shows-the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;
| |
| : 6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
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| : 7) An estimate of the exposure to plant operating personnel as a result of the change; and
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| : 8) Documentation of the fact that the change was reviewed and found acceptable by the PSRC.
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| : 2. Shall become effective upon review and acceptance by the PSRC.
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| WOLF CREEK - UNIT 1 6-25
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| cN i REPOR r Nu-4E R I as e, ., r,oC - - N. . . -,
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| g,ica- = u s. NvCtun uout,.1Oa, CO E*,"S$'- BIBLIOGRAPHIC DATA SHEET NUREG-1136 f
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| r- SEE sNSTRUCTIONS ON THE REVERSE
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| ) 3 itTLE AN0 5ueTITLE 3 LEAVE OLANE f
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| Q Technical Specification for Wolf Creek Generating Station, Unit No.1~ . oA1E RE,ORr COoPLEno oON1- . EAR l
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| . Aur,. Oms, June 1985 6 DAYE REPORT ISSUED MONTH VtaR June 1985 7, PERF ORMING ORGANi2 AllON NAME AND MAIUNG ADDRESS isacker to Comt 8 PROJECT.T ASE WORK ONil NUMSER Division of Licensing 0ffice of Nuclear Reactor Regulation *''' "6"'''"""*'"
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| U.S. Nuclear Regulatory Comission ,
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| Washington, DC 20555 10 SPONSORsNG ORG ANI2 AT60N N AME AND M AILING ADDRESS finc&de le Coper 11a TYPE OF REPORT Technical Sane as . 7. above. ,,,,,,,co,,,,,,,,,,,,,,,,,,,,
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| 12 SUPPLEMENT ARY NOTES Docket No. STN 50-482 - Appendix A to License No. NPF-42 1.1 A85TR ACT f200 = ores or 'essJ
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| '(3
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| /
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| L/
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| ) The Wolf Creek Generating Station, Unit No.1 Technical Specifications were prepared by the U.S. Nuclear Regulatory Comission to set forth the limits, operating conditions, and other requirements applicable to a nuclear reactor facility as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public.
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| : i. oOcu.ENT AN ALvsi. . . .E mORos.otsCR.PrOR i. *v,Agigtv Unlimited 16 SECURITY CLASSIFICAYsON ITnn oneer
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| . ,oENT,. ERs OPEN ENoEo nRos Unc1 assi fied
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| ., \ s rn.s recorrs f
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| Uncl assi fied 17 NuNSER Of PAGES 18 PRICE
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| }
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| UNITED STATES covaru ctass u4st NUCLEAR REGULATORY COMMISSION e sra WASHINGTON, D.C. 20555 u's* ''c'* "*'
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| 4 wass o c.
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| PERMIT 44 G 87 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 0
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| 0
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| .}}
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