ML20133M375
| ML20133M375 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 10/10/1985 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | Tucker H DUKE POWER CO. |
| References | |
| NUDOCS 8510250190 | |
| Download: ML20133M375 (553) | |
Text
{{#Wiki_filter:f' 5 October 10, 1985 4
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Docket Nos. 50-413/414 -- Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Tucker:
SUBJECT:
PROOF AND REVIEW OF THE COMBINED CATAWBA UNITS 1 AND 2 TECHNICAL. SPECIFICATIONS Attached is a copy of the Appendix A combined Technical Specifications for the Catawba Nuclear Station, Units 1 and 2, for proof and review. Please review this document and iJentify any statements which do not accurately reflect the Catawba U nits 1 and 2 FSAR (as amended), SER or the "as-built" facility. Your response is necessary by COB October 30, 1985, to support a December 20, 1985 issuance of a Final Draf t version of the combined Catawba Units 1 and 2 Technical Specifications. If you have any questions regarding this matter, please contact the project manager, Kahtan Jabbour, at (3011492-9789. Sincerely, Elinor Adensam, Chief licensing Branch #4 Division of licensing
Enclosure:
As stated DE3Ic" o POIGIN!1 cc: See next page. Certifkd F1 N C-DISTRIBUTION Pocket File KJabbour JPartlow NRC PDR local PDR BGrimes PRC System f.B#4 r/f Edordan EAdensam MDuncan Atty., ELD DI. Jordan, IE ACRS (16) "3IC p 'a /i b !-- LB#4/DL/PM t.B #4/Dk 'LA LB#4/DL/BC /. j' ,/... '. . KJabbour:jl MDd2 ban EAdensam /' 10/cj /85 10/g /8' 10/ /85 / </ R / > 0FFICIAL RECORD COPY l 8510250190 851010 PDR ADOCK 05000413 l P PDR
MODEL: NUREG-1106 NUREG-xxxx TECHNICAL SPECIFICATIONS 4 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 l 50-414 APPENDIX "A" to LICENSE NO. NPF-35 and LICENSE NO. i i I issued by the U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation PROOF & REVIEW COPY QCT 7 1985 l l l l l l l ._.__ _ _ _ - - - - - - -__ _ -
J PROOF & REVIEW COPY . 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 1.4 AX I AL FLUX DI FFERENCE. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 1-1 . 1.5 CHANN E L CA LI B RATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-1 .. 1.6 CHANNEL CHECK................................................. 1-1 -
- 1. 7 CONTAINMENT INTEGRITY......................................... 1-2 '
1.8 CONTROLLED LEAKAGE............................................ 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 RESPONSETIME..................... 1-3 1.13 FREQUENCY N0TATION........................................... 1-3 1.14 IDENTIFIED LEAKAGE........................................... 1-3 4 1.15 MASTER RELAY TEST............................................ 1-3 1.16 MEMBER (S) 0F THE PUBLIC...................................... 1-3
- 1. 1.17 0FFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-3 .
1.18 OPERABLE - OPERABILITY....................................... 1-4 1.19 OPERATIONAL MODE - M00E...................................... 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.................................... 1-4 1.25 RATED THERMAL P0WER.......................................... 1-5 i 1 1.25 REACTOR BUILDING INTEGRITY................................... 1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................ i 1-5 1.28 REPORTABLE EVENT............................................. 1-5 1.29 SHUTDOWN MARGIN.............................................. 1-5 1.30 SITE B00NDARY................................................ 1-5 1.31 SLAVE RELAY TEST............................................. 1-5 CATAWBA - UNITS 1 AND 2 I OCT 7 1985 l
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PROOF & REVIEW COPY OEFINITIONS SECTION PAGE 1.32 SOLIDIFICATION............................................... 1-5 1.33 SOURCE CHECK................................................. 1-6 1.34 STAGGERED TEST BASIS......................................... 1-6 1.35 THERMAL P0WER................................................ 1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 1.37 UNIDENTIFIED LEAKAGE......................................... 1-6 1.38 UNRESTRICTED AREA............................................ 1-6 1.39 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-6 1.40 VENTING...................................................... 1-7 . 1.41 WASTE GAS HOLDUP SYSTEM...................................... 1-7 TAB LE 1. L FREQU ENCY N0TATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-8 .
% TABLE 1.2 OPERATIONALM0 DES....................................... 1-9
.n 1
CATAWBA - UNITS 1 AND 2 II OCT 7 liiB5 i
_ _ - _ ~ . -- .. -- - 4 PROOF & REVIEW COPY SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRcSSURE............................. 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 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 - BASES -- - SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 8 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 8 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS O 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... B 2-3 i CATAWBA - UNITS 1 AND 2 III
- - - - - - - - - - - - - - - - - - - ----- - - - - -~~ " ' - -
PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200*F........................... 3/4 1-1 Shutdown Margin - T,yg 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 BORATION 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 MOVABLE CONTROL ASSEMBLIES Group Height..~........................................... 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00................... 3/4 1-16 Position Indication Systems - Operating. . . . . . . . . . . . . . . . . . 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Bank Insertion Limits............................ 3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION...................................... 3/4 1-22 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 2-4 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)..................... 3/4 2-5 l
q i CATAWBA - UNITS 1 AND 2 IV OCT 7 1985
_- -. . . . - -. _. = PROOF & REVIEW COPY LIMITING CONDITIONS FOR JPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) q AS A FUNCTION OF CORE HEIGHT. 3/4 2-8 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R.................................. 3/4 2-9 FIGURE 3.2-3 REACTOR COOLANT SYSTEM TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN 0PERATION.................................. 3/4 2-11 . 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-12 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-15 TABLE 3.2-1 DNS PARAMETERS........................................ 3/4 2-16 3/4.3 INSTRUMENTATION ^ 3/4.3.1 _ REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 . TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION........... ....... 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 SURVEILLANC REQUIREMENTS......................................E....... 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 3-15 TA8LE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS........................... 3/4 3-27 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i - INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-42 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-51 TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION l FOR PLANT OPERATIONS.. .................................. 3/4 3-52 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-54 l Movable Incore Detectors................................. 3/4 3-55 Seismic Instrumentation.................................. 3/4 3-56 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-57 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-58 l Meteorological Instrumentation........................... 3/4 3-59 CATAWBA - UNITS 1 AND 2 V DCT 7 1985
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PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-60 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................................... 3/4 3-61 Remote Shutdown System................................... 3/4 3-62 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-63 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................ 3/4 3-64 Accident Monitoring Instrumentation...................... 3/4 3-65 TABLE 3.3-10 ACCIDENT HONITORING INSTRUMENTATION.................. 3/4 3-66 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANC REQUIREMENTS......................................E ....... 3/4 3-68 Chlorine Detection Systems............................... 3/4 3-70 Fire Detection Instrumentation........................... 3/4 3-71 TABLE 3.3-11 FIRE DETECTION INSTRUMENTATION ...................... 3/4 3-73 Loose-Part Detection System.............................. 3/4 3-79 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-80 TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION3/4 3-81 TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-83 Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-85 TABLE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/4 3-86 i TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-90 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-93 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND <(;ted CIRCULATION Startup and Power Opera'fi w,, ........................ 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops Filled............................. 3/4 4-5 Cold Shutdown - Loops Not Fi11 ed. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 CATAWBA - UNITS 1 AND 2 VI OCT 7 1985
l l PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 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 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-17 . TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE - Leakage Detection Systems................................ 3/4 4-19
--Operational Leakage...................................... 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. .3/4 .... 4-22 3/4.4.7 CHEMISTRY................................................
3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-25
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TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVE REQUIREMENTS....................................ILLANCE ......... 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMP PR0G RAM. . . . . . . . . . . . . . . . . . . .................
. . . . . . . . . . . . . 3/4 . LE 4-30 AND ANA LYS 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO 16 EFPY................................. 3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIEIS - APPLICABLE UP TO 16 EFPY................................. 3/4 4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - 1 WITH0RAWAL SCHEDULE...................................... 3/4 4-35 Pressurizer.............................................. 3/4 4-36 Overpressure Protection Systems.......................... 3/4 4-37 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-39 , 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-40 i CATAWBA - UNITS 1 AND 2 VII OCT 7 1985 l
l i PR00F & REVIEW COPY
- LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection....................................... 3/4 5-1 Upper Head Injection..................................... 3/4 5-3
- 2 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-5 .
3/4.5.3 ECCS SUBSYSTEMS - T,yg <350*F........................... 3/4 5-9 3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT - Containment Integrity.................................... 3/4 6-1 Contai rment' Lea kage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-2
% TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS............ 3/4 6-5 Containment Air Locks.................................... 3/4 6-8 Internal Pressure........................................ 3/4 6-10 , Air Temperature.......................................... 3/4 6-11 Containment Vessel Structural Integrity.................. 3/4 6-12 Reactor Building Structural Integrity. . . . . . . . . . . . . . . . . . . . 3/4 6-13 Annulus Ventilati,on System............................... 3/4 6-14 Containment Purge Systems................................ 3/4 6-16 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/4 6-18 3/4.6.3 CONTAINMENT ISOLATION VALVES............................. 3/4 6-20 TABLE 3.6-2 CONTAINMENT ISOLATION VALVES.........'................. 3/4 6-22 3/4.6.4 COMBUSTIBLE GAS CONTROL i
i Hydrogen Monitors........................................ 3/4 6-30 j Electric Hydrogen Recombiners............................ 3/4 6-31 Hydrogen Mitigation System............................... 3/4 6-32 3/4.6.5 ICE CONDENSER Ice 8ed.................................................. 3/4 6-33 l Ice Bed Temperature Monitoring System.................... 3/4 6-35 Ice Condenser Doors...................................... 3/4 6-36 Inlet Door Position Monitoring System.................... 3/4 6-38 CATAWBA - UNITS 1 AND 2 VIII OCT 7 1985
PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Divider Barrier Personnel Access Doors and Equipment Hatches...................................... 3/4 6-39 Containment Air Return and Hydrogen Skimmer Systems...... 3/4 6-40 Floor Drains............................................. 3/4 6-42 Refueling Canal Drains................................... 3/4 6-43 Divider Barrier Sea 1..................................... 3/4 6-44 TABLE 3.6-3 DIVIDER BARRIER SEAL ACCEPTABLE PHYSICAL PROPERTIES... 3/4 6-45 3/4.6.6 CONTAINMENT VALVE INJECTION WATER SYSTEM ................ 3/4 6-46 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE
- S a f e ty Va 1 v e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGF NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION...................................... 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxi l ia ry Feedwa te r Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-4 S'pecific Activity........................................ 3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-7 Main Steam Line Isolation Va1ves......................... 3/4 7-8 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-9 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-10 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM............................. 3/4 7-11 3/4.7.5 STANDBY NUCLEAR SERVICE WATER P0ND....................... 3/4 7-12 3/4.7.6 CONTROL ROOM AREA VERTILATION SYSTEM..................... 3/4 7-13 3/4.7.7 AUXILIARY BUILDING FILTERED EXHAUST SYSTEM............... 3/4 7-16 3/4.7.8 SNUBBERS................................................. 3/4 7-18 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST. . . . . . . . . . . 3/4 7-23 3/4.7.9 SEALED SOURCE CONTAMINATION.............................. 3/4 7-24 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System............................ 3/4 7-26 Spray and/or Sprinkler Systems........................... 3/4 7-28 CO 2 Systems.............................................. 3/4 7-30 Fire Hose Stations....................................... 3/4 7-32 CATAWBA - UNITS 1 AND 2 IX OCT 7 1985
PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 FIRE HOSE STATIONS.................................... 3/4 7-33 3/4.7.11 FIRE BARRIER PENETRATIONS................................ 3/4 7-35 3/4.7.12 GROUNDWATER LEVEL........................................ 3/4 7-37 3/4.7.13 STANDBY SHUTDOWN SYSTEM.................................. 3/4 7-39 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-9 TABLE 4.8-2 LOAD SEQUENCING TIMES................................. 3/4 8-10 - Shutdown................................................. 3/4 8-11 3/4.8.2 ~D.C. SOURCES 0perating................................................ 3/4 8-12 TABLE 4.8-3 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-15 Shutdown................................................. 3/4 8-16 3/4.8.3 ONSITE POWER DISTRIBUTION Operating................................................ 3/4 8-17 Shutdown................................................. 3/4 8-18 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices..................................... 3/4 8-19 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES....................................... 3/4 8-21 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 9-7 3/4.9.6 MANIPULATOR CRANE........................................ 3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.......... 3/4 9-9 l l l l OCT 7 1985 CATAWBA - UNITS 1 AND 2 X
i PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-10 l Low Water Leve1.......................................... 3/4 9-11 3/4.9.9 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-12 3/4.9.10 WATER LEVEL - STO RAGE POO L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-13 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM................. 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3,, PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 REACTOR COO LANT L00PS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4
. 10-4 f
3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYS P R0 G RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4
........ . .11-2
. . . . . . . . .I S 00se..................................................... 3/4 11-5 Liquid Radwaste Treatment System......................... 3/4 11-6 Liquid Holdup Tanks...................................... 3/4 11-7 Chemical Treatment Ponds................................. 3/4 11-8 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................ 3/4 11-9 TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM.................................................. 3/4 11-10 Doce - Noble Gases....................................... 3/4 11-13 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form............................. 3/4 11-14 Gaseous Radwaste Treatment System. . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-15 Explosive Gas Mixture.................................... 3/4 11-16 Gas Storage Tanks........................................ 3/4 11-17 CT 7 1985 CATAWBA - UNITS 1 AND 2 XI
PROOF & REVIEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................. 3/4 11-18 3/4.11.4 TOTAL 00SE............................................... 3/4 11-19 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING j i 3/4.12.1 MONITORING PR0 GRAM....................................... 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 S AMPLES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-9 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMP ANALYSIS..........................................LE ....... 3/4 12-10 . 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-13 3/4.12.3 _INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-15 s i OCT 7 1985 CATAWBA - UNITS 1 AND 2 XII
PROOF & REVIEW COPY BASES SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVAB LE CONTROL ASSEMB LIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 ' 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 . 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and REACTOR
~
COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R............................................ B 3/4 2-2 FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL P0WER............................................. B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5
, 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-7 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 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 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 4-7 OCT 7 BBS CATAWBA - UNITS 1 AND 2 XIII
BASES
' PROOF & REVIEW COPY SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS (UNIT 1)................. B 3/4 4-9 TABLE B 3/4.4-2 REACTOR VESSEL TOUGHNESS (UNIT 2)................. 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 CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO 550*F............
NDT B 3/4 4-12 3/4.4.10 ST R U CTU RA L I NTE G R I TY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B
. . 3/4
. . 4-16 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/A.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 .and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 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 DE, PRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-4 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 3/4.6.5 ICE CONDENSER. . . . .a*........................................ B 3/4 6-5 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 7-2 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM.............................. B 3/4 7-3 3/4.7.5 STANDBY NUCLEAR SERVICE WATER P0ND........................ B 3/4 7-3 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM...................... B 3/4 7-3 3/4.7.7 AUXILIARY BUILDING FILTERED EXHAUST SYSTEM................ B 3/4 7-4 I
3/4.7.8 SNUBBERS.................................................. B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS.................................. B 3/4 7-6 3/4.7.11 FIRE BARRIER PENETRATIONS................................. B 3/4 7-7 3/4.7.12 GROUNDWATER LEVEL......................................... B 3/4 7-7 3/4.7.13 STANDBY SHUTDOWN SYSTEM................................... B 3/4 7-8 CATAWBA - UNITS 1 AND 2 XIV OCT 7 1985
BASES PROOF & REVlEW COPY SECTION PAGE 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 ONSITE POWER DISTRIBUTION .......................,......... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE 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 9-2 3/4.9.6 MAN I PU LATO R C RAN E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B
. . 3/4
. . 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING........... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION. . . . . . . . . . .B. .3/4 9-2 3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and STORAGE P00L.............................................. B 3/4 9-3 3/4.9.11 FU'EL HANDLING VENTILATION EXHAUSTSYSTEM.................. B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 i .
3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS..... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......................................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS........................................ B 3/4 11-4 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................. B 3/4 11-8 3/4.11.4 TOTAL 00SE............................................... B 3/4 11-8 OCT 7 1985 CATAWBA - UNITS 1 AND 2 XV
PROOF & REVIEW COPY SECTION PAGE 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... 8 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... 8 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 i I 1 a OCT 7 1985 CATAWBA - UNITS 1 AND 2 XVI
. . ~ , - - - -
. . .- - - - - - ,,-m- - , - ~ - :e. ,.-- ~
r--- - -, --, + , , - - - - , - , ~ , - - - - - - . -
PROOF & REVIEW COPY DESIGN FEATURES SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1.2 LOW POPULATION Z0NE......................................... 5-1 5.1.3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.................... 5-1 FIGURE 5.1-1 EXCLUSION AREA....................................... 5-2 FIGURE 5.1-2 LOW POPULATION 20NE.................................. 5-3 FIGURE 5.1-3 UNRESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE LIQUID EFFLUENTS................................ 5-4 FIGURE 5.1-4 UNRESTRICTED AREA AND SITE BOUNDARY FOR - RADI0 ACTIVE GASEOUS EFFLUENTS............................... 5-5 5.2 CONTAINMENT 5.2.1 CONFIGURATION............................................... 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-6 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................. 5-6 5.3.2 CONTROL ROD ASSEMBLIES...................................... 5-6 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-6 5.4.2 V0LUME...................................................... 5-6 5.5 METEOROLOGICAL TOWER LOCATION................................. 5-6 5.6 FUEL STORAGE 5.6.1 CRITICALITY................................................. 5-7
- 5.6.2 DRAINAGE.................................................... 5-7 5.6.3 CAPACITY.................................................... 5-7 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-7 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.................. 5-8 i
l OCT 7 1985 CATAWBA - UNITS 1 AND 2 XVII
PROOF & REVfEW COPY ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY.............................................. 6-1 6.2 ORGANIZATION................................................ 6-1 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 TA8LE 6.2-1 MINIMUM SHIFT CREW COMPOSITION...................... 6-5 6.2.3 CATAWBA SAFETY REVIEW GR00P............................... 6-6 Function.................................................. 6-6 - Composition............................................... 6-6 Responsibilities.......................................... 6-6 Records................................................... 6-6 6.2.4 SHIFT TECHNICAL ADVIS0R................................... 6-6 6.3 UNIT STAFF QUALIFICATIONS................................... 6-6 6.4 TRAINING.................................................... 6-6 6.5 REVIEW AND AU0IT............................................ 6-7 6.5.1 TECHNICAL REVIEW AND CONTROL ACTIVITIES. . . . . . . . . . . . . . . . . . . 6-7 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSR8)........................ 6-8 Function.................................................. 6-8 l Organization.............................................. 6-9 Review.................................................... 6-10 Audits.................................................... 6-10 Records................................................... 6-11 6.6 REPORTABLE EVENT ACTI0N..................................... 6-12 6.7 SAFETY LIMIT VIOLATION...................................... 6-12 l 6.8 PROCEDURES AND PR0 GRAMS..................................... 6-13 6.9 REPORTING REQUIREMENTS...................................... 6-15 6.9.1 ROUTINE REP 0RTS........................................... 6-15 - Startup Report............................................ 6-15 CATAWBA - UNITS 1 AND 2 XVIII
ADMINISTRATIVE CONTROLS SECTION PAGE Annual Reports............................................ 6-16 Annual Radiological Environmental Operating Report........ 6-16 Semiannual Radioactive Effluent Release Report............ 6-17 Monthly Operating Reports................................. 6-19 Radial Peaking Factor Limi t Report. . . . . . . . . . . . . . . . . . . . . . . . 6-19
.1 6.9.2 SPECIAL REP 0RTS........................................... 6-19 6.10 RECORD RETENTION........................................... 6-19 -
6.11 RADIATION PROTECTION PR0 GRAM............................... 6-21 . 6.12 HIGH RAOIATION AREA........................................ 6-21 6.13 PRU~CESSCONTROLPROGRAM(PCP).............................. 6-22 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM)..................... 6-23 i 6.15 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLIO RA0 WASTE TREATMENT SYSTEMS.......................................... 6-23 l i f I I l I OCT 7 1985 CATAWBA - UNITS 1 AND 2 XIX
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PROOF & RB4EW COPY l SECTION 1 o . DEFINITIONS JUL 181985 0CT 7 1985
l
- 1. 0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
- ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.
ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices. ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functio.7s. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy. AXIAL FLUX DIFFERENCE
- 1. 4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION
- 1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK
- 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 channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
1 CATAWBA - UNITS 1 AND 2 1-1 JUL 181985 OCT 7 1985
PROOF & REVfEW COPY CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: -
- a. All penetrations required to be closed di.fing accident conditions are either:
- 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or .
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification 3.6.1.3, *
- d. The containment leakage rates are within the limits of Specification 3.6.1.2, and
- e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall noc preclude completion of movement of a component to a safe conservative position. 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 ' nixture 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 l Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample. CATAWBA - UNITS 1 AND 2 1-2 JUL 181985 0CT 7 1985
DEFINITIONS PROOF & REVIEW COPY . h ENGINEERED SAFETY FEATURES RESPONSE TIME 4
. 12 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when tne 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.
FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the opera-tion of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or
- c. Reactor Coolant System 'eakage through a steam generator to the Secondary Coolant System.
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. MEMBER (S) 0F THE PUBLIC 1.16 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally 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 recre-ational, occupational, or other purposes not associated with the plant. OFFSITE DOSE CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (00CM) 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-l mental Radiological Monitoring Prcgram. CATAWBA - UNITS 1 AND 2 1-3 JUL 181985 0CT 7 WB5
.- = _ - - . - - - - - - - -
DEFINITIONS OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or - have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - 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. PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental ' nuclear characteristics of the reactor core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System cr.nponent body, pipe wall, or vessel wall. PROCESS CONTROL PROGRAM 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste. PURGE - FURGING l 1.23 PURGE or PURGING 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 required to purify the confinement. 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. CATAWBA - UNITS 1 AND 2 1-4 m 181985 OCT 7 1985
DEFINITIONS PROOF & REV!BY COPY
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h' RATED THERMAL POWER 1.25 RATED THERMAL POWER shall te a total reactor core hea? **'r.fer rate tio the reactor coolant of 3411 MWt. REACTOR BUILDING INTEGRITY 1.26 REACTOR BUILDING INTEGRITY shall exist when:
- a. Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at least one door shall be closed,
- b. The Annulus Ventilation System is OPERABLE, and
- c. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
REACTOR TRIP SYSTEM RESPONSE TIME ' 1.27 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. REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50. SHUTDOWN MARGIN 1.29 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. SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by licensee. , SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. SOLIDIFICATION 1.32 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. CATAWBA - UNITS 1 AND 2 1-5 JUL 181985 00T 7 1985 l
- ^
DEFINITIONS EW COPY SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:
- 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 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 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 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. ' VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine 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 fodines or particu-lates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. 1 JUL 181985 CATAWBA - UNITS 1 AND 2 1-6 0CT 7 1985 i
PR0OF & REVfEW COPY DEFIFITIONS VENTING 1.40 VENTING shall be the controlled process of discharg1.ig 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 pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process. WASTE GAS HOLDUP SYSTEM 1.41 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases 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, i
't l
i i Jul.181985 CATAWBA - UNITS 1 AND 2 1-7 OCT 7 1985 1
l l PROOF & REWBY COPY l 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. l 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. l e 1 1 1985 CATAWBA - UNITS 1 AND 2 1-8 1985
PROOF & REVIEW COPY , TABLE 1.2 OPERATIONAL MODES . REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K,ff THERMAL POWER
- TEMPERATURE
- 1. POWER OPERATION >0.99 > 5% 1 350*F
- 2. STARTUP 1 0.99 $ 5% 1 350*F
- 3. HOT STANOBY < 0.99 0
, 1 350*F i
- 4. HOT SHUTDOWN < 0.99 0 350*F > T
> 200*F **9 l
- 5. COLD SHUTOOWN < 0.99 0 $ 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.
l i ) { jul i B 1903 CATAWBA - UNITS 1 AND 2 1-9 QGI 7 1985
g_ ~ _ _ _ - . . - l i I .} a I i PROOF & REV1EW COPY i a
- i 4
! SECTION 2.0 < SAFETY LIMITS .
i l AND LIMITING SAFETY SYSTEM SETTINGS ,l 4
- 1 l
l } f jy( 1I l OCT 7N i
- t. - , - , . _ - - - . . - . . - - - . _ - . . . . - - - , - . - . _ _ _ --. . - . _ - - - - - - ,
PROOF & REVIEW COPY g
- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS
- REACTOR CORE 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 pressurizer pressure line, be in HOT STANDBY within 1 hour, and comply with the require-ments 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. ACTION: MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 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. - CATAWBA - UNITS 1 AND 2 2-1 JUL 181985 OCT 7 1985
i l l l l l PROOF & REVIEW COPY l 660
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580 nj.. O 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 IN OPERATION CATAWBA - UNITS 1 AND 2 2-2 000 7 1985 l
PROOF & REVfEW COPY 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 be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for et ch channel in Table 3.3-1. ACTION:
- 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.
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint 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 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 Setpoi.it value.
Equation 2.2-1 Z + R + S 1 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 Allowance) of Table 2.2-1 for the affected channel. CATAWBA - UNITS 1 AND 2 2-3 JUL 181985 OCT 7 1985 l r
Q _ TABLE 2.2.-1
- ! REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS 5
TOTAL SENSOR ALLOWANCE ERROR g FUNCTIONAL UNIT (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE h 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A. " N 2. Power Range, Neutron Flux g a. High Setpoint 7. 5 4.56 0 1109% of RTP* $111.1% of RTP*
- b. Low Setpoint 8.3 4.56 0 125% of RTP* 127.1% of RTP*
- 3. Power Range, Neutron Flux, 1. 6 0.5 0 15% of RTP* with High Positive Rate 16.3% of RTP* with a time constant a time constant 1 2 seconds 1 2 seconds
- 4. Power Range, Neutron Flux, 0.5 High Negative Rate
- 1. 6 0 <S% of RTP* with <6.3% of RTP* with M i time constant a time constant O y M 12 seconds 12 seconds A 5. Intermediate Range, 17.0 8.4 0 Qe 125% of RTP* 131% of RTP*
Neutron Flux 2
- 6. Source Range, Neutron Flux 17.0 10 0 1105 cps $1.4 x 10s cps
- 7. Overtemperature AT 7.2 4.47 2.03 See Note 1 See Note 2 n
- 8. Overpower AT 4.3 1.3 1. 2 See Note 3 See Note 4 O
- 9. Pressurizer Pressure-Low 4.0 2.21 1.5 11945 psig 11938 psig***
- 10. Pressurizer Pressure-High 7. 5 4.96 0.5 12385 psig 12399 psig
- 11. Pressurizer Water Level-High 5.0 2.18 1.5 192% of instrument 193.8% of instrument span span
- 12. Reactor Coolant Flow-Low 2.5 1.77 0.6 190% of loop 189.2% of loop ag design flow ** design flow **
co E 2 tis
*RTP = RATED THERMAL. POWER
** Loop design flow = 96,900 gpm M
T y SI *** Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for, lead d and 1 second for lag. Channel calibration shall ensure that these time constants are adjusted to these values.
g TABLE 2.2-1 (Continued) Y REACTOR T' RIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR ALLOWANCE ERROR E FUNCTIONAL UNIT (TA) Z (S) _ TRIP SETPOINT ALLOWABLE VALUE d 13. Steam Generator Water s
> a. Unit 1 17 14.2 1. 5 >17% of span >15.3% of span from 5 Level Low-Low Trom 0% to 30% 0% to 30% RTP*
m RTP* increasing increasing linearly linearly to to >53.2% of span
> 54.9% of span from 30% to 100% RTP*
Trom 30% to 100% RTP*
- b. Unit 2 17 14.2 1.5 >17% of narrow >15.3% of narrow y range span range span ,
u,
- 14. Undervoltage - Reactor 8.57 0 1.0 >77% of bus ->76% (5016 volts)
Coolant Pumps voltage (5082 volts) with a 0.7s response time
- 15. Underfrequency - Reactor 4.0 0 1.0 >56.4 Hz with a ->55.9 Hz Coolant Pupps D.2s response time
- 16. Turbine Trip l
- a. Low Control Valve EH Pressure N.A. N.A. N.A. >550 psig >500 psig o
h
'rt
- b. Turbine Stop Valve N.A. N.A. N.A. >1% open >1% open %
Closure - _o g 17. Safety Injection Input N.A. N.A. N.A. N.A. N.A. h ' H from ESF g
~
a E sn
-?
"RTP = RATED THERMAL POWER '
9 TABLE 2.2-1 (Continued) f REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS 5
. TOTAL SENSOR
- c. ALLOWANCE ERROR
$ FUNCTIONAL UNIT (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE
- 18. Reactor Trip System
- Interlocks E
- a. Intermediate Range N.A. N.A. N.A. 11 x 10 80 amps 16 x 10 12 amps N
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" $12.2% of RTP*
l o
- 2) P-13 input N.A. N.A. N.A. $10% RTP* Turbine $12.2% RTP* Turbine 7 Impulse Pressure o
m Impulse Pressure O Equivalent Equivalent
- c. Power Range Neutron N.A. N.A. N.A. -<48% of RTP" <50.2% of RTP*
Flux, P-8 - o
- d. Power Range Neutron N.A. N.A. N.A.
Q 169% of RTP* $70% of RTP* Flux, P-9 Q
- e. Power Range Neutron N.A. N.A. N.A 110% of RTP* 17.8% of RTP*
8o Flux, P-10 -<
- f. Power Range Neutron N.A. N.A. N.A. $10% of RTP*
Flux, Not P-10
$12.2% of RTP*
- g. Turbine Impulse Chamber N.A. N.A. N.A. $10% RTP* Turbine Pressure, P-13
$12.2% RTPa Turbine Impulse Pressure Impulse Pressure .
Equivalent Equivalent ob 19. Reactor Trip Breakers N.A. N.A. N.A N.A. N.A. O o r-M
- 20. Automatic Trip and N.A. N.A. N.A. N. A. N.A. M N
on Interlock Logic T H
*RTP = RATED THERNAL POWER
i c) TABLE 2.2-1 (Continued) TABLE NOTATIONS h E NOTE 1: OVERTEMPERATURE AT AT fI l Il g S) < AT, {Kg -K 2 [^ [T (1 , 3) - T'] + K3 (P - P') - fg(AI)} d Where: AT = Measured AT by RTD Manifold Instrumentation; 1 * = Lead-lag compensator on measured AT; 1 o N
=
It. T2 Time constants utilized in lead-lag compensator for AT, r2 =8s, m tz = 3 s; 2 o 0 1 = Lag c spensator on measured AT; 1+T 5 3 m r3 = Time constant utilized in the lag compensator for AT, T3 = 0; m n $ AT, = Indicated AT at RATED THERMAL POWER; @ c2 Kg = 1.411; o tz K3 = 0,02401/*F; ' 1 + t.S = The function generated by the lead-lag compensator for T 1 + TsS dynamic compensation; **U
= = 28 s, T4, 13 Time constants utilized in the lead-lag compensator for T,yg, 1 4 ts = 4 s; T = Average temperature *F; 1 =
Lag compensator on measured Tavg; ggr
--e 1+rS g
y $ Ta
=
Time constant utilized in the measured T,,g lag compensator, t. = 0; b i j
n TABLE 2.2-1 (Continued) h TABLE NOTATIONS (Continued) 5
> NOTE 1: (Continued) e T' 5 590.8 F (Nominal T,yg all owed by Safety Analysis);
d K3 = 0.001189; P = Pressurizer pressure, psig; . m P' = 2235 psig (Nominal RCS operating pressure); g o" . S = Laplace transform operator, s 1; So and f t(AI) is a function of the indicated difference between top and bottom detectors of the " power-range neutron ion chambers; with gains to be selected based on measured instrument $ response during plant STARTUP tests such that: g (i) For qt 9bbetween -43% and -6.5%, ft (AI) = 0, where qt and q are b percent RATED 8 THERMAL POWER in the top and bottom halves of the core respectively, and qt*Ab is E total THERMAL POWER in percent of RATED THERMAL POWER; (ii) For each percent that the magnitude of q t ~9 b is more negative than -43%, the AT Trip Setpoint shall be automatically reduced by 2% of its value at RATED THERMAL POWER; and (iii) For each percent that the magnitude of q t 9bis more positive than -6.5%, the AT Trip Setpoint shall be automatically reduced by 1.641% of its value at RATED THERMAL POWER. NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.4%. - O. N 5 4#1 1
l c) TABLE 2.2-1 (Continued) (a TABLE NOTATIONS (Continued) E
> NOTE 3: OVERPOWER AT e
g AT (1 + r,5) ( 1 ) $ I I ) I I q un (1 + T25) (1 + TsS) < AT0 (K* - K* I(1 + T 5) (1 + TsS) T - K. [T (1 + t.5)I - T"] 7 2
- f (AI)}
pa g Where: AT = As defined in Note 1, o
'1 '
N 7
= As defined in Note 1, 3 8n Ts. T2 = As defined in Note 1, go 1 ::o
= As defined in Note 1, 3
]
N E T3 = As defined in Note 1, kn o AT = As defined in Note 1, "T3 o -< K. = 1.0704,
=
K3 0.02/*F for increasing average temperature and O for decreasing average temperature,
S =
3 The function generated by the rate-lag controller for T,yg dynamic compensation, T, = Time constant utilized in the rate-lag controller for T,yg, t, = 10 s, , O r- = As defined in Note 1, 1 + T.S g n
$ Ts = As defined in Note 1, ,h n
* "Q O
i i 9 TABLE 2.2-1 (Continued)
# TABLE NOTATIONS (Continued)
E y NOTE 3: (Continued) E K. = 0.001707/*F for T > 590.8'F and K. = 0 for T 5 590.8*F, U T = As defined in Note 1, w o T" = Indicated T avg at RATED THERMAL POWER (Calibration temperature for t.T m instrumentation, 5 590.8*F), S = As defined in Note 1, and m f 2(AI) = 0 for all AI. g o n m NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by M 4 more than 2.6%. O h 2 o O
--, r-6-*
8
a __ ..__r -
, g a w --* d' %-# 1 DRAFF PROOF & RENIEW COPY -
BASES FOR SECTION 2.0 . SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l l l A'L 181985 ! OCT 7 1985
~ BRAFT PROOF & REYfEW COPY 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. -
) i e pt 18 385 OCT 7 1985
1 2.1 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. i Operation above the upper boundary of the nucleate 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 WRB-1 correlation. The WRB-1 DNB correlation 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 DNB design basis is as follows: there must be at least a 95% probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95% probability with 95% confidence that DNB will not occur when the minimum DNBR is at the DNBR limit. In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. JUL 181985 CATAWBA - UNITS 1 AND 2 B 2-1 l OCT 7 1985
hff 2.1 SAFETY LIMITS PROOF & RETSE3Y COPY BASES N This curve is based on a nuclear enthalpy rise hot channel factor,
- F AH, f 1.49 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: N FaH = 1.49 [1 + 0.3 (1-P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f 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 with core Safety Limits. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor vessel, pressurizer, and the Reactor Coolant System piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum 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 Reactor Coolant System is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation. JUL 18 E85 CATAWBA - UNITS 1 AND 2 B 2-2 OCT 71385 l _ --
- 2. 2 LIMITING SAFETY SYSTEM SETTINGSPR00F & REVIEW COPY 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 option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combina-tion 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, I + R + S < TA, the interactive eff(cts 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 Set-point.
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 factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. i ' The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess met of the Allowable Value exhibits the behavior that the racA has not its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drif t, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. 1 CATAWBA - UNITS 1 AND 2 B 2-3 JUL 18 h25 OCT 7 1985
l LIMITING SAFETY SYSTEM SETTINGS BASES PR0OF & REVIEW COPY h{ 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 varianles, 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 accident 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 Rance, 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 all rod ejection accidents. The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a 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 applicable design limit DNBR value for each fuel type. CATAWBA - UNITS 1 AND 2 JUL 1b naoo B 2-4 g 7
LIMITING SAFETY SYSTEM SETTINGS jQ' 8ASES PROOF & REVIEW COPY ni i Intermediate and Source Range, Neutron Flux
- The Intermediate and Source Range, Neutron Flux trips provide core pro-tection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.
These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. a Reactor trip at about 10s The Source Range channels will initiate counts per second unless manually blocked when P-6 becomes active or automatically blocked when P-10 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. Overtemperature AT The Overtemperature AT trip provides core protection to prevent DN8 for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water and includes dynamic compensation for piping dela temperature detectors, (2) pressurizer pressure, and ys from the core (3) axial power to the loop distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.2-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors..the Reactor trip is automatically reduced according to the notations in Table 2.2-1. Overpower aT The Overpower AT 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 loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226,
" Reactor Core Response to Excessive Secondary Steam Releases."
I i JUL 18 hW) CATAWBA - UNITS 1 AND 2 B 2-5 0CT 7 1985
LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent l 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, is 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 level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, is automatically reinstated by P-7. Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection and prevents DNB by citigating 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 P-7 an automatic Reactor trip will occur on low reactcr coolant flow in more than one loop and below P-7 the trip function is automatically blocked. CATAWBA - UNITS 1 AND 2 B 2-6 Jul.18 F5 00T 719b3
LIMITING SAFETY SYSTEM SETTINGS BASES PROOF & Ram COPY RAf[ Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat resulting sink from in of loss thenormal event of a sustained steam /feedwater flow mismatch feedwater. The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System. Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide 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 undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For 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% of RATED THERMAL POWER; with a turbine impulse chamber pressure at approxi-mately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7. Turbine Trio 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 69% 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. CATAWBA - UNITS 1 AND 2 8 2-7 JUL I 81985 0CT 7 1985
i
.l 1
LIMITING SAFETY SYSTEM SETTINGS BASES PROOF & REYiEW COPY h[ _ 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 trip (i.e. , prevents premature block of Source Range trip), deener-gizes the high voltage to the detectors. On decreasing power, Source restored. voltage Range Level trips are automatically reactivated and high P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactcr coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. are automatically blocked.On decreasing power, the abnve listed trips 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 above listed trips. P-9 On increasing Turbine trip. power P-9 automatically enables Reactor trip on trip on TurbineOn decreasing power, P-9 automatically blocks Reactor trip. P-10 C,., increasing power P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the voltage Source Range trip and deenergizes the Source Range high power. On decreasing power, the Intermediate Range trip and Provides input to Power the Low Setpoint P-7. Range trip are automatically reactivated. P-13 Provides input to P-7. CATAWBA - UNITS 1 AND 2 JUL 18 B85 8 2-8 OCT 7 1985
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y . . PROOF & REVfEW COPY e 0 i SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i I t I l l i t l JUL 181985 0CT 7 1985
PROOF & REVIEW COPY 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 Opration 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 it, placing theasunit in a MODEin:in which the specification does not apply by applicable, 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 SHUTDOWN withir the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. specifications. Exceptions to these requirements are stated in the individual 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 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. stated in the individual specifications. Exceptions to these requirements are CATAWBA UNITS 1 AND 2 3/4 0-1 OCT 7 1985
l APPLICABILITY PROOF & REVIEW COPY BRAFT LIMITING CONDITION FOR OPERATION
)
3.0.5 Limiting Conditions for Operation including the associated ACTION requirements as follows: shall apply to each unit individually unless otherwise indicated a. Whenever the Limiting Condition for Operation refers to systems or components which are shared by botn units, the ACTION requirements will apply to both units simultaneously. This will be indicated in the ACTION section; b. Whenever the Limiting Condition for Operation applies to only one unit, this will be identified in the APPLICABILITY section of the specification; and c. Whenever certain portions of a specification contain operating para-meters, setpoints, etc. , which are different for each unit, this will be identified in parentheses or footnotes. (For example, "... flow rate of 54,000 cfm (Unit 1) or 43,000 cfm (Unit 2)...") l CATAWBA UNITS 1 AND 2 3/4 0-2 JUL 181985 0CT 7 1985
APPLICABILITY PROOF & REVIEW COPY 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 saximum 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 are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting 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);
CATAWBA UNITS 1 AND 2 3/4 0-3 JUL I 81985 OCT 7 1985
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 Adaenda shall be applicable as follows in these Technical Specifications:
ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days - Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days
- c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities;
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
l l CATAWBA UNITS 1 AND 2 3/4 0-4 JUL 181985 OCT 7 1985
3/4.1 REACTIVITY CONTROL SYSTEMS PROOF & REVIEW COPY 3/4.1.1 BORATION CONTROL g SHUTDOWN MARGIN - T avg >200*F ' LIMITING CONDITION FOR OPERATION 1 e 1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% ak/k for f.m 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 4.1.1.1.1 to 1.3% ak/k: The SHUTDOWN MARGIN shall be determined to be greater than or equal 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; c. When in MODE 2 with K,ff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted 3.1.3.6; control rod position is within the limits of Specification critical d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.le. below, with the control banks at the maximum inservice limit of Specification 3.1.3.6; and
*See Special Test Exceptions Specification 3.10.1.
JUL 181985' CATAWBA - UNITS 1 AND 2 3/4 1-1 OCT 7 1985
REACTIVITY CONTROL SYSTEMS PROOF & REV1EW COPY , i SURVEILLANCE REQUIREMENTS (Continued)
- e. When in MODE 3 or 4, at least once per 24 hours by consideration o'f 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 be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. 4.1.1.1.3 At least once per 18 months, the Reactor Makeup Water pumps shall ' be demonstrated OPERABLE by verifying a total combined flow rate of less than or equal to 240 gpm and a flow rate of less than or equal to 120 gom for each pump. 4.1.1.1.4 At least once per 31 days, while in MODE 4, one Reactor Makeup Water pump shall be demonstrated inoperable by verifying that the motor circuit breaker is secured in the open position. i CATAWBA - UNITS 1 AND 2 3/4 1-2 JUL 181985 OCT 7 1985
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yg i 2004 PROOF & REVIEW COPY h 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 4.1.1.2.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% 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 entrippable, the SHUTDOWN 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.
4.1.1.2.2 At least once per 18 months, each Reactor Makeup Water pump shall be demonstrated OPERABLE by verifying a flow rate of less than or equal to 120 gpm. At least once per 31 days, one Reactor Makeup Water pump shall be demonstrated inoperable by verifying that the motor circuit breaker is secured in the open position. CATAWBA - UNITS 1 AND 2 3/4 1-3 g18 l{8g 7
REACTIVITY CONTROL SYSTEMS PROOF & REVIEW COPY DRAFT 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 all the rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition; and
- b. Less negative than -3.7 x 10 4 Ak/k/*F for all the rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.
APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only#. Specification 3.1.1.3b. - MODES 1, 2, and 3 only#. 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 subseq'uent 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 Speci*ication 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 K,ff greater than or equal to 1. #See Special Test Exceptions Specification 3.10.3.
l l l CATAWBA - UNITS 1 AND 2 3/4 1-4 JUL 181985 0CT 71985
PROOF & RE\1EW COPY 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 Specifi-cation 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
-2.8 x 10 4 Ak/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -2.8 x 10 4 ak/k/*F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification 3.1.1.3b., at least once per 14 EFPD during the remainder of the fuel cycle. .
i
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i CATAWBA - UNITS 1 AND 2 3/4 1-5 JUL 181985 00T 7 1985
PROOF & REVIEW COPY 5 REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION . 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be greater than or equal to 551*F. APPLICABILITY: MODES 1 and 2#*. ACTION: With a Reactor Coolant System operating loop temperature (T,yg) less than 551'F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System Temperature (Tavg) shall be determined to 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,yg is less than 561'F with the T,yg-Tref Deviation Alarm not reset.
*With K,ff greater than or equal to 1. #See Special Test Exceptions Specification 3.10.3.
CATAWBA - UNITS 1 AND 2 3/4 1-6 JUL 181985 0CT 7 1985
, REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS PROOF & REVIEW COPY FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE source: and capable of being powered from an OPERABLE emergency power a.
A flow path from the boric acid tank via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a. is OPERABLE, or b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b. is OPERABLE. ' APPLICABILITY: MODES 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 OPERABLE: At least one of the above required flow paths shall be demonstrated a. At least once per 7 days by verifying that the temperature of the heated portion of the flow path is greater than or equal to 65 F when a flow path from the boric acid tanks is used, and b. 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. i a i CATAWBA - UNITS 1 AND 2 3/4 1-7 JUL 181985 OCT 7 1985
REACTIVITY CONTROL SYSTEMS _ FLOW PATHS - OPERATING PR OF & REVIEW COPY LIMITING CONDITION FOR OPERATION - 3.1.2.2 be OPERABLE:At least two* of the following three boron injection flow paths shall a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System. APPLICABILITY: MODES 1, 2, 3, and 4. 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 lea HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% 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 OPERABLE: At least two of the above required flow paths shall be demonstrated a. At least once per 7 days by verifying that the temperature of the flow path from the boric acid tanks is greater than or equal to 65*F when it is a required water source; b. 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; c. At least once per 18 months during thutdown by verifying that each . automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and d. At least once per 18 months by verifying that the flow path required by Specification Coolant System. 3.1.2.2a. delivers at least 30 gpm to the Reactor
*0nly one boron injection flow path is required to be OPERABLE whenever the temperature than or equal of to one or more of the Reactor Coolant System cold legs is less 285*F.
CATAWBA - UNITS 1 AND 2 3/4 1-8 JUL 181985 OCT 71985
~
l 1 REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN PROOF & REVIEW COPY
%[
LIMITING CONDITION FOR OPERATION
- 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.
APPLICABILITY: MODES 5 and 6. ACTION: With no 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 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying that a differential pressure across the pump of greater than or equal to 2380 psid is developed when tested pursuant to Specification 4.0.5. 4.1.2.3.2 All 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, or that the discharge of each charging pump has been isolated from the Reactor Coolant System by at least two isolation valves with the power removed from the valve motor operators. l i CATAWBA - UNITS 1 AND 2 JUL 181985 1 3/4 1-9 OCT 7 1985
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING PROOF & REVfEW COPY gg' LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two* charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a 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 COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying that a differential pressure across each pump of greater than (" equal to 2380 psid is developed when tested pursuant to Specification 4.0 5. 4.1.2.4.2 All charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable at least once per 31 days whenever the temperature of one of more of the Reactor Coolant System cold legs is less than or equal to 285*F by verifying that the motor circuit breakers are secured in the open position or that the discharge of each charging pump has been isolated from the Reactor Coolant System by at least two isolation valves with power removed from the valve motor operators.
*A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the Reactor Coolant System cold legs is less than or equal to 285*F.
l i l CATAWBA - UNITS 1 AND 2 3/4 1-10 JUL 181985 OCT 7 1985
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTOOWN PROOF & REVIEW COPY hk 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 5100 gallons,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65*F.
- b. The refueling water storage tank with:
- 1) A minimum contained borated water volume of 26,000 gallons,
- 2) A minimum boron concentration of 2000 ppm, and
- 3) A minimum solution temperature of 70 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 l 3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the refueling water storage tank temperature when it is the source of borated water and the outside air temperature is less than 70*F.
CATAWBA - UNITS 1 AND 2 3/4 1-11 OCT 7 1985 l l
REACTIVITY CONTROL SYSTEMS / fly ggpy BORATED WATER SOURCES - OPERATING , LIMITING CONDITION FOR OPERATION l 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE ! as required by Specificatiar 3.1.2.2:
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water volume of 19500 gallons,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65*F.
- b. The refueling water storage tank with:
- 1) A contained borated water volume of at least 363,513 gallons, -
1
- 2) A minimum boron' concentration of 2000 ppm, -
- 3) A minimum solution temperature of 70 F, and
- 4) A maximum solution temperature of 114 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, restore the 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 refueling water storage tank 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. l CATAWBA - UNITS 1 AND 2 3/4 1-12 OCT 7 1985 l
REACTIVITY CONTROL SYSTEMS PROOF & REVIEW COPY SURVEILLANCE REQUIREMENTS 4.1.2.6 Each 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 refueling water storage tank temperature when the outside air temperature is either less than 70*F or greater than 100 F. .
i l \ l JUL 181565 CATAWBA - UNITS 1 AND 2 3/4 1-13 OCT 7 1985 i
+-, , - .~~ , - - + . - - - m, - .,
4 _ REACTIVITY CONTROL SYSTEMS PROOF & ggTW COPY 3/4.1.3 MOVA8LE PONTROL ,ASSEM8 LIES GROUP HEIGHT I LIMITING CONDITION FOR OPERATION t 3.1. 3.1
- positicned Allwithin full-length shutdown and control rods shall be OPERABLE and demand position. 12 steps (indicated position) of their group step counter
} ADPLICABILITY: MODES 1* and 2*. ACTION: ' a. With one or more full-length rods inoperable due to being innovable
^
as a result of excessive friction or mechanical interference or 4 known to be untrippable, determine that the SHUTDOWN MARGIN require-l ment of Specification 3.1.1.1 is satisfied within 1 hour and be in - HOT STAN08Y within 6 hours. < I b. With more than one full-length rod inoperable or misaligned from the group step counter demand position by more than 112 steps (indicated position), be in HOT STAND 8Y within 6 hours. i c. With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its i group step counter demand height by more than 112 steps (indicated position), POWER OPERATION may continue provided that within 1 hour: i 1. The rod is restored requirements, or to OPERABLE status within the above alignment 1 2. i The rod is declared inoperable and the remainder of the rods in i the group with the inoperable rod are aligned to within i 12 steps i of the inoperable rod while raintaining the rod sequence and 4 I insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuance to Specification'3.1.3.6 during subse-quent operation, or i ! 3. The rod is declared inoperable and the SHUTDOWN MARGIN require-t ment of Specification 3.1.1.1 is satisfied. POWER OPERATION - may then continue provided that: I a) A reevaluation of each accident analysis of Table 3.1-1 is i i 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 i determined at least once per 12 hours;
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
CATAWBA - UNITS 1 AND 2 3/4 1-14 M 'O OCT 7 1985
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) ' c) A power distribution map is obtained from the movable N incore detectors and FO(Z) and F are verified to be AH 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. SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group damand 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. CATAWBA - UNITS 1 AND 2 3/4 1-15 JUL 181985 ; OCT 7 1985
PROOFg ggggy (gpy TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0PERABLE 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) 1 CATAWBA - UNITS 1 AND 2 JUL 181985 3/4 1-16 0CT 7 1985 x
i l PROOF & REVIEW COPY REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING l 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 inoperable 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 nonindicacing 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 within 8 hours,
- 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. CATAWBA - UNITS 1 AND 2 3/4 1-17 JUL 181985 l OCT 7 1985
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION . 3.1.3.3 One digital rod position indicator (excluding demand position indication) 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. The Reactor Trip System breakers can be closed in order to perform this surveillance. i l
*With the Reactor Trip System breakers in the closed position.
#See Special Test Exceptions Specification 3.10.5 i
CATAWBA - UNITS 1 AND 2 3/4 1-18 M I0 OCT 7 1985
i REACTIVITY CONTROL SYSTEh5 PROOF & REVIEW COPY 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 3.3 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 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 THERMAL POWER is restricted to less than or equal to 66% of RATED 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 on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and
- c. At least once per 18 months.
CATAWBA - UNITS 1 AND 2 3/4 1-19 JUL 18 bdb OCT 7 1985
l REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT' PROOF & REYIEW copy 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 surveil-lance testing pursuant to Specification 4.1.3.1.2, within I 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 to be fully withdrawn: a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or 0 during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3. #With K,ff greater than or equal to 1. .
1 l CATAWBA - UNITS 1 AND 2 3/4 1-20 JUL 181985 OCT 7 1985
REACTIVITY CONTROL SYSTEMS CONTROL BANK INSERTION LIMITS PROOF & REVfEW COPY - LIMITING CONDITION FOR OPERATION . 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1. APPLICABILITY: MODES 1* and 2*#. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2: i a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figure, or
- c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank 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 hours.
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
FWith K,7f greater than or equal to 1. l CATAWBA - UNITS 1 AND 2 3/4 1-21 JUL 181985 0CT 7 1985
PROOF & REVIEW COPY DRAFT (FULLY \ WITHDRAWN) ' 228 - ,. _;t20% h2M e. 220 -
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[ 40 _ f!BANKD! g_- _ 20 '- 0 ~~ , 0 10 20 30 40 50 60 70 (FULLY 80 90 100 INSERTED) RELATIVE POWER (PERCENT OF RATED THER FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 1 22 OCT 71985
~
3/4.2 PCWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE PROOF & REVIEW COPY 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. t 5% for Cycle 1 core average accumulated burnup of less than or equal to 5000 MWD /MTU;
- b. + 3%, -9% for Cycle 1 core average accumulated burnup of greater than 5000 MWD /MTU; and
- c. +3%, -12% for subsequent cycles,
- i j The indicated AFD may deviate outside the above required target level 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 I hour during the previous 24 hours.
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 does not exceed 1 hour during the previous 24 hours. APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.* ACTION: I
- 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 target band limits, or
- 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
With the indicated AFD outside of the above required target band for b. more than 1 hour of cumulative penalty deviation times 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: i ! 1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and
*See Special Test Exceptions Specification 3.10.2.
CATAWBA - UNITS 1 AND 2 3/4 2-1 JUL 181985 OCT 7 1985
.__. . - - ~ _ __ - -
l POWER DISTRIBUTION LIMITS
' 00F&
Copy I LIMITING CONDITION FOR OPERATION - ' 1
- ACTION (Continued) i 2. The Power Range Neutron Flux" - High Setpoints to less than or
; equal to 55% of RATED THERMAL POWER within the next 4 hours. ;
i l c. With the indicated AFD outside of the above required target band j for more than 1 hour of cumulative penalty deviation time during l 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. The provisions of Specification 3.0.4 are not applicable for 4 operation up to 50% of RATED THERMAL POWER.
! SURVEILLANCE REQUIREMENTS i ! 4.2.1.1 The indicated AFD shall be determined to be within its limits during l POWER OPERATION above 15% of RATED THERMAL POWER by: I j a. Monitoring the indicated AFD for each OPERA 8LE excore channel:
- 1) At least once per 7 days when the AFD ' Monitor Alarm is OPERABLE, and i 2) At least once per hour for the first 24 hours after restoring j the AFD Monitor Alarm to OPERABLE status.
- b. Monitoring and logging the indicated AFD for each OPERABLE excore i
! channel at least once per hour for the first 24 hours and at least j once per 30 minntes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
- c. The provisions of Specification 4.0.4 are not applicable.
! 4.2.1.2 The indicated AFD shall be considered outside of its target band when i two or more OPERA 8LE 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: l a. One minute penalty deviation for each 1 minute of POWER OPERATION l outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and {
- Surveillance testing of the Power Range Neutron Flux Channel may be perfomed I pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained )
within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours ' operation may be accumulated with the AFD outside of the above required target i band during testing without penalty deviation. CATAWBA - UNITS 1 ANO 2 3/4 2-2 OCT 7 885
DRAFT POWER DISTRIBUTION LIMITS PROOF & REWEW COPY SURVEILLANCE REQUIREMENTS (Continued)
- b. One-half minute penalty deviation for each 1 mincte 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 , det?r:nined 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 provi-sions of Specification 4.0.4 are not applicable. i CATAWBA - UNITS 1 AND 2 3/4 2-3 ML 181985 0CT 7 1985
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[ji' 'it - 50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (61) % FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF l RATED THERMAL POWER l l JUL 181985 CATAWBA - UP!ITS 1 AND 2 3/4 2-4 OCT 7 1985
POWER DISTRIBUTION LIMITS PROOF & REVfEW COPY 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be limited by the following relationships: F0 (Z) 5 [2.32] [K(Z)] for P > 0.5 P - Fq (Z) 1 [4.64] [K(Z)] for P $ 0.5 Where: P _ THERMAL POWER , and RAIED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: MODE 1. ACTION: With Fq (Z) exceeding its limit:
- a. Reduce THERMAL POWER at least 1% for each 1% q F (Z) exceeds the limit within 15 minutes and similarly reduce tne 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 q (Z) exceeds the limit, and
- 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 Fq (2) is demonstrated through incore mapping to be within its limit.
- CATAWBA - UNITS 1 AND 2 3/4 2-5 JUL 181985 00T 7 1985
J POWER DISTRIBUTION LIMITS PROOF & REVfEW COPY SURVEILLANCE REQUIREMENTS k 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: 9
- 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 xy 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 xy computed (Fx) btained in Specification 4.2.2.2b.,
above to: The F
- 1) xy limits for RATED THERMAL POWER x(FRTP) for the appropriate measured core planes given in Specification 4.2.2.2e. and f.,
below, and
- 2) The relationship:
Ff=FRTP x x [1+0.2(1-P)], l Where F is the limit for fractional THERMAL PO)fER operation express as a function of F P and P is the fraction of RATED THERMAL POWER at which F xy was measured. d. xy according to the following schedule: Remeasuring F C
- 1) When F x
is greater than the F xRTP limit for the appropriate measured core plane but less than the F relationship, additional C powergistributionmapsshallbetaken dF *Y compared to F*RTP xy either: l and F Y l a) Within 24 hours after exceeding by 20% of RATED THERMAL l POWER or greater, t'.e THERMAL POWER at which F C was xy last determined, or b) At least once per 31 EFPD, whichever occurs first. I 1 i l OCT 7 1985 , CATAWBA - UNITS 1 AND 2 3/4 2-6 1 l
POWER DISTRIBUTION LIMITS PROOF & REVfEW COPY DRAFT SURVEILLANCE REQUIREMENTS (Continued)
- 2) When the F x is less than or equal to the F xRTP limit for the appropriate measured core plane, additional power distributian maps shall be taken and F compared to F and F at least x
once per 31 EFPD. e. xy liuits for RATED THERMAL POWER x(FRTP) shall be provided for The F all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifica-tion 6.9.1.9;
- f. The F 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 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%
and 74.9 2 2%, inclusive, and
- 4) Core plane regions within i 2% of core height (1 2.88 inches) about the bank demand position of the Bank "0" control rods.
- g. With F excceding F ', the effects of F xy xy on Fq (Z) shall be evaluated to determine if 9F (Z) is within its limits.
4.2.2.3 When Fq (Z) is measured for other than Fxy determinations, an overall measured qF (Z) shall be obtained from a power distribution map and increased by 3% to acc. for manufar.turing tolerances and further increasef by 5% to account for measurement uncertainty. . t r CATAWBA - UNITS 1 AND 2 3/4 2-7 M OCT 7 1985
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0 2.0 4.0 6.0 8.0 10.0 12.0 CORE HEIGHT (FT) FIGURE 3.2-2 K(Z) - NORMALIZE 0 F g(Z) AS A FUNCTION OF CORE HEIGHT CATAWBA - UNITS 1 AND 2 3/4 2-8 JUL 181985 OCT 7 1985 l l
POWER DISTRIBUTION LIMITS PROOF & REVIEW COPY 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System 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 F AH a* , R = 1.49 [1.0 + 0.3 (1.0 - P)]
- b. THERMAL POWER , and P _- RATED THERMAL POWER
- c. Fh=MeasuredvaluesofFhobtainedbyusingthemovableincore detectors to obtain a power distribution map. The measured valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 includes penalties for undetected feedwater venturi fouling of 0.1% and for measurement uncertainties of 2.1% for flow and 4%
forincoremeasurementofFh. APPLICABILITY: MODE 1. ACTION: With the combination of Reactor Coolant System 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 Reactor Coolant System 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 Reactor Coolant System total flow rate comparison that the combination of R and Reactor Coolant System 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.
CATAWBA - UNITS 1 AND 2 3/4 2-9 OCT 7 1985
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION PROOF & REVIEW COPY h[ 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 OPERA-TION may proceed provided that the combination of R and indicated Reactor Coolant System total flow rate are demonstrated, through incore flux mapping and Reactor Coolant System 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 Reactor Coolant System total flow rate determined by process computer readings or digital voltmeter measurement 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 Reactor Coolant System total flow ate 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 Specifica-tion 4.2.3.2, is assumed to exist. 4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected . to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement. 4.2.3.5 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months. JUL 18 bob CATAWBA - UNITS 1 AND 2 3/4 2-10 OCT 7 1985 i _
M007 PENALTIES FOULING AND MEASUREMENT UNCERTAINTIES OF 2.1% FOR OF 0.1% FOR UNDETECTED FEEDWATER VE Copy FLOW AND 4.0% FOR INCORE MEASUREMENT OF Fh ARE INCLUDED IN THIS FIGURE.
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. __ _ .._ _ ._ ..__. _= =..p _.. __=_ .. =. . -._.=_=___=__=.=_j..___=_=.._.
. -- s
;_z = . = -
41*00 - - - - - - - - - - ~ ~ ~ --
' ' ~ ' - - - - - " - - ' ' - ' -- F'- ~ -
- =, -- =:E .f. . = . ;.: - . ;:. . .g[.'.: , 3.[ EE.:[:.E 'n j :Q.,;i;t.:::t t.:f-J.__;:- .:
E . _ _ . . . . . . _ _ . . . . ,. _ _ _ _ _
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.0 =:t . =. ..m ._- .
% -E ACCEPTABLE OPERATION REGION ' _-- .f.~,: UNACCEPTABLE OPERATION -
C 40.50 " ~ ~ - - ' - - - ' ' - - W :_ . . _.s== r. .= . . . : :.:. s . .
.. : s . -"."# :3 .==u...=-
- t.. n.rt-
. , _ - . . , . = _ . _ :--
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.....,,,,3,_f...[.
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. . _ . -- .i _=_ ._- q_ __-
_:. .- 0 a . : ::=. g n .= :=c. . , .:::.__ . r- . . +. +: e =;;s:in=s=+. a. n =-p_.i+sia e = m us e _ .r c;. p. _ p
- _i='_p=}. . : .#... p ::..l:, t: :p .: u U_:" 3.:E.__ L-- r :: ?.~.:: ri:Vp =;.j .
4 a .
- "r.
...-._:=.:......
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. r. .. f : . . . . I . . .= --
g . ;fM-'r :-t_..z:iz_: :FL.. '
- n.': --
; 9=.-y.1_E j -i : l-:. _
-j g:. '
>- .== ^._~ , _ . == 39. 61
--- a. -- - r h' . . . : .. l n::
E 39.50 2=
..."--: '- ^ ^ ~ ~ -- - _.-
_.=a u. _ . ' _ . ' _ . . . . ' _ . . -~
"-="~"="*~"b...'..'F.....
W -= ACCEPTABLE FOR P 5 98 %RTP --' y N :=! _ . _ ..
... . . _ . . _ _ . . . . : -- + :--: --L :2:=- : _ Er hf. ; !: - : j L.:
v3 .
. , __- 39.214 . . _ _ . . - . :. _ . . - - j. .h..v ij s_.=. e:5- .. ..
.- - . "" : :-- .-- - .:, .: = . e := :: . .u = = . ._: :: =
w! z "~- ACCEPTABLE FOR P $ 96 %RTP
~ ~## '~~~'---
- . a:==; _"_y_-..=_~.==-_:- : .:. :r_u '._ .-' - '*t ===r ~ ".=. s -
. 5==
= :c. c:
$.39.00 C
- .: .4 .=:: .
-.. . . _ _ : 5 38.818 ;_Ei. :i=3::;; ;:t:i".qq -;g
- -0o ##E
-s - - -- -
-==
*# -+_:*E. :-. '1 ~ tin" 'A~# ' 1.E ^-.d- :--
g ----ACCEPTABLE FOR P S 94 %RTP,':--- . :- 2.: : {= :.:-.u.. . t :, ,.- . :, ;! r:. q.- ::_=-. .= .. lr::: . . . .- C 38.50 . - , - = _ . - - - - - - - -
;_. ~ -- -
2 = 38.422 U -----
.d ..n ..
. .- l- = _.-. :. =.:: :. .: .
g, .._. --
- - r - :. . . .: *-~_y.. -s:.
W _ .1 : e
'-- .._** .i.e.:.:..
-r
..1__
g ... = ACCEPTABLE FOR P S 92 %RTP,~ .:-.. . .
. =__2 r. ' _ .- --l -. ". .. ! .,. .- . f . ...r. . .
m .. :.-_. _:.- ___...==_1."-'---+-+.=h.-
~~
-4 38.00 ._., i. 3 8. 0 2 6 - .v_m= - ::er ;. 7.
nt c : "---
.: .:=
- . =-
-.._.~.t... - r. :..:." .: ,:_... ='_=:-
, . ..- .,...c .. - : : r-- h.
.=
.61.2. 4 ACCEPTABLE FOR P - 90 %RTP. Ti:v,- ~ ~ P . 'T se n t e. .
, ,.r_s_.
_ . .rEi 1.:_..=__._. .,e :: :r.- . , n .e- - T - : :".' _ ; :. 3. . i ...+ . .p . i .:. . . .i_:.. l
.r" : r;.
..:2: !.:.:..- .. r_ _ nt ; :_ n .
37.50 .
- i 1 .l37.630 _.
. ;- a.: . :=-
.. l :r e- iu :=h- .
.6.. .t
, n -
__w __
.:. _ t _._ .,..__..,.
-; .y c.:q ;i..-W ._....._.4,..._.[_._.._._._.,
6.-- . -- ; n 6 , ; .i m e i .
.. .l
. i 0.94 ^ 0.96 0.98 1.00 1.02 1.04 1.06 R = h/1 A011 + 0.3(1.P))
FIGURE 3.2-3 REACTOR COOLANT SYSTEM TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN OPER CATAWBA - UNITS 1 AND 2 3/4 2-11 OCT 7 1985 l [ __ - _ . _ _ _
POWER DISTRIBUTION LIMITS PROOF & REVIEW COPY 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02 above 50% of RATED THERMAL POWER. APPLICABILITY: MODE 1.* ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
- 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. Within 2 hours either:
a) Reduce the QUADRANT POWER TILT RATIO to within its liuit, 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 i 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 until verified scceptable at 95%
j or greater RATED THERMAL POWER.
*See Special Test Exceptions Specification 3.10.2.
CATAWBA - UNITS 1 AND 2 3/4 2-12 dUI 10 IUOU OCT 7 1985
POWER DISTRIBUTION LIMITS PROOF 8 REVIEW COPY LIMITING CONDITION FOR OPERATION ACTION (Continued)
- b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdewn 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% 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 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 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 shutdu.tn 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 i l b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. l l JUL 18 IWS l CATAWBA - UNITS 1 AND 2 3/4 2-13 OCT 7 1985
-4 POWER DISTRIBUTION LIMITS PROOF & RE11EVV COPY LIMITING CONDITION FOR OPERATION 4
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 REQUIREKENTS 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.
- c. The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when abose 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 full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. OCT 7 1985 CATAWBA - UNITS 1 AND 2' 3/4 2-14 c g _ _
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS PROOF & REVIEW COPY h 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 .u
- 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 tneir limits at least once per 12 hours. l JUL 181565 CATAWBA - UNITS 1 AND 2 3/4 2-15 OCT 71983 j
I PROOF & REVIEW COPY TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS Four Loops in Operation Average Temperature Meter average - 4 channels: < 592*F
- 3 channels: 5592*F Computer average - 4 channels: 1 593*F
- 3 channels: 1 593*F Pressurizer Pressure '
Meter average - 4 channels: 1 2225.9 psig*
- 3 channels: > 2230 psig*
Computer average psig*
- 4 channels: 1 2221
- 3 channels: 1 2224 psig*
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 2-16
3/4.3 INSTRUMENTATION PROOF & REVIEW COPY 3/4.3.1 KEACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks Table of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMEhTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERA 8LE by the performance of 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.
I i i I JUL 18 G85 CATAWBA - UNITS 1 AND 2 3/4 3-1 > OCT 7 1965
TABLE 3.3-1 9 REACTOR TRIP SYSTEM INSTRUMENTATION 5 2* MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHA'NELS N TO TRIP OPERABLE MODES ACTION Q
- 1. Manual Reactor Trip 2 1 2 1, 2 1 2 1 2 3*, 4*, 5* 10 3E
- 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#
t' High Negative Rate y
- o
{ 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 8n
- 6. Source Range, Neutron Flux 2*
- a. Startup 2 1 2 2## 4 :D
- b. Shutdown 2 1 2 3,4,5 5 $
m
- 7. Overtemperature AT ,
E Four Loop Operation 4 2 3 1, 2 6# Qv
- 8. Overpower AT "
Four Loop Operation 4 2 3 1, 2 6#
- 9. Pressurizer Pressure-Low 4 2 3 1 6#
oI ae- c:::3
* :::t3 .
a A . ts gw% .
TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION E MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Q
- 10. Pressurizer Pressure-High 4 2 3 1, 2 6#
[ , E 11. Pressurizer Water Level-High 3 2 2 1 7#
" 12. Reactor Coolant Flow-Low
- a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 7#
any oper- each oper-ating loop ating loop
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 7#
below P-8) two oper- each oper-Ru ating loops ating loop Y
- 13. Steam Generator Water 4/sta 2/sta gen 3/sta gen 1, 2 6# $
O Level--Low-Low gen in any each o operating operating 'n sta gen sta gen 9* - o
- 14. Undervoltage-Reactor Coolant 4-1/ bus 2 3 1 6# E Pumps (Above P-7) g
- 15. Underfrequency-Reactor Coolant 4-1/ bus Pumps (Acove P-7) 2 3 1 6# Qu
- 16. Turbine Trip
- a. Low Control Valve EH Pressure 4 2 3 1#### 7#
- b. Turbine Stop Valve Closure 4 4 1 1#### 11#
1
- 17. Safety Injection Input D
@ from ESF 2 1 2 1, 2 9 M
^p e
T
-3 n H isI LT1 =
TABLE 3.3-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION h MINIMUM i TOTAL NO. CHANNELS CHANNELS APPLICABLE e FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i"i d 18. Reactor Trip System Interlocks
- a. Intermediate Range -
3 Neutron Flux, P-6 2 1 2 2## 8 5 m 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 Flux, P-8 4 2 3 1 8 T.
- d. Power Range Neutron 4 2 3 1 8 Y Flux, P-9 s
- e. Pcwer Range Neutron Flux, P-10 (Above P-10) 4 2 3 1##### 8
- f. Power Range Neutron Flux, Not P-10 4 3 3 4 1###, 2 8 :0 a
- g. Turbine Impulse Chamber o M
Pressure, P-13 2 1 2 1 8 8'
- o
- 19. Reactor Trip Breakers 2 1 2 2
1, 2 . 9 s - o 1 2 3*, 4*, 5* 10 g
- 20. Automatic Trip and Interlock 2 1 y Logic 2 1 2
2 1, 2 3*, 4*, 5* 9 10 Qo 5 u,
PROOF & REVIEW COPY TABLE 3.3-1 (Continued) 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 pruvisions 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.
####Above the P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
#####Above 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 at least 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 chunnel is placed in the tripped condition within 1 hour,
- b. The Minimum Channels OPERABLE requirement is mei.; 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 POWER 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 increasing THERMAL POWER above the P-6 Setpoint; or
- b. Above the P-6 (Intermediate Range Neutron Fluv. Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER. 0CT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-5 l
v e, w PROOF & REVISV COPY
- i TABLE 3.3-1 (Continued)
ACTIONSTATEMENTS(Continueg ACTION 4 - With the number of OPERABLE channels one less than the Minimum Chant.els OPERABLE requirement, suspend all operations involving positive reactivity changes. ACTION 5 .With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor trip breakers, suspend all operations involving positive reactivity changes and verify Valves NV-231, NV-237, NV-241, and NV-244 are closed and secured in position within the next hour. 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 nev.t required ANALOG CHANNEL OPERA-TIONAL TEST provided the ir. operable 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 status light (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 i ; to OPERABLE status within 48 hours or open the Reactor trip - l breakers within the next hour. i ACTION 11 - With the number of OPERABLE channels less than the Total Number I of Channels, operation may cnntinue provided the inoperable l channels are placed in the tripped condition within 1 hour. I CATAWBA - UNITS 1 AND 2 3/4 3-6 M q g5 ocT
n TABLE 3.3-2 2-
$ REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E
FUNCTIONAL UNIT RESPONSE TIME E
] 1. Manual Reactor Trip N.A.
- 2. Power Range, Neutron Flux 5 0.5 second*
E
- 3. Power Range, Neutron Flux, High Positive Rate N.A.
- 4. Power Range, Neutron Flux, High Negative Rate E
5 0.5 second* 8n
- 5. Intermediate Range, Neutron Flux N.A. Ro o
y 6. Source Range, Neutron Flux N.A. Q Y w
- 7. Overtemperature aT $ 4 seconds * <
a
- 8. Overpower AT 5 4 seconds
- 9. Pressurizer Pressure-Low < 2 seconds
- 10. Pressurizer Pressure-High 5 2 seconds ,
+' 11. Pressurizer Water Level-High N.A.
C
-gll1 N
M * ' l
-* r-
- Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion
, e of the channel shall be measured from detector output or input of first electronic component in channel. J
-a co ,
- ww i
l
TABLE 3.3-2 (Continued) 9 5;'
-REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E
2 FUNCTIONAL UNIT RESPONSE TIME E
] 12. Low Reactor Coolant Flow
- a. Single Loop (Above P-8) < 1 second E b. Two Loops (Above P-7 and below P-8) i 1 second o -
" 13.
l Steam Generator Water Level-Low-Low $ 2.0 seconds
- 14. Undervoltage-Reactor Coolant Pumps 5 1.5 seconds w
- 15. Underfrequency-Reactor Coolant Pumps 5 0.6 second g o
- 16. Turbine Trip '
W *
- a. Low Control Valve EH Pressure N.A. @ -
Y b. Turbine Stop Valve Closure N.A. < m __ rn
- 17. Safety Injection Input from ESF N.A. I
- 18. Reactor Trip System Interlocks o N.A. y
- 19. Reactor Trip Breakers N.A.
- 20. Automatic Trip and Interlock Logic N.A. g M
- D*
n -e V M C0 w w
n TABLE 4.3-1
$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Ei
' TRIP ANALOG ACTUATING MODES FOR E CHANNEL DEVICE WHICH U CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
[ g 1. Manual Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3*, 4*, 5*
- 2. Power Range, Neutron Flux
- a. High Setpoint S D(2, 4), M N.A. N. A. 1, 2 M(3, 4), o Q(4, 6),
R(4, 5) o
- b. Low Setpoint R(4) M N.A. N.A. 1###, 2 S
[ R 3. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 $ [ High Positive Rate y
- 4. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 High Negative Rate o
- 5. Intermediate Range, R(4, 5) S/U(1),M N
S N.A. N.A. 1###, 2 Neutron Flux
- 6. Source Range, Neutron Flux 5 R(4, 5) S/U(1),M(9) N.A. N.A. 2##, 3, 4, 5
- 7. Overtemperature AT S R(12) M N.A. N.A. 1, 2
- 8. Overpower AT S R H N.A. N.A. 1, 2 O M
- 9. Pressurizer Pressure-Low S R M N.A. N.A. 1 D
- 10. Pressurizer Pressure-High S R M N.A. N.A. 1, 2 4_
- 11. Pressurizer Water Level-High 5 R M N.A. N.A. ,1
] 12. Reactor Coolant Flow-Low S R H N.A. N.A. 1 a
4
n TABLE 4.3-1 (Continued) '
$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 ' TRIP ANALOG ACTUATING MODES FOR E CHANNEL DEVICE WHICH Q
- CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED s
E 13. Steam Generator Water Level- S R M N.A. N.A. 1, 2 Low-Low
- 14. Undervoltage - Reactor Coolant N.A. R N.A. M N.A. 1 Pumps
- 15. Underfrequency - Reactor N.A. R N.A. M N.A. 1 Coolant Pumps R 16. Turbine Trip '
* ::o
- a. Low Control Valve EH O T Pressure N.A. R N.A. S/U(1,10) N.A. 1# n
, b. Turbine Stop Valve N.A. R N.A. S/U(1,10) N.A. 1# Closure
- 17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 Q ,
ESF
- 18. Reactor Trip System Interlocks 8
- a. Intermediate Range Q
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. c 1 ,
l c. Power Range Neutron i e Flux, P-8 N.A. R(4) M(8) N.A. N.A. 1 l Q d. Low Power Range Neutron
- ~ Flux, P-9 N.A. R(4) M(8) N.A. N.A. 1
- w .
l $, u 4
n TABLE 4.3-1 (Continued)
@ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS s"
' TRIP ANALOG ACTUATING MODES FOR E CHANNEL DEVICE WHICH Z CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
[ E 18. Reactor Trip System Interlocks (Continued)
- e. Power Range Neutron Flux, P-10 (Above P-10) N.A. R(4) M(8) N.A. N.A. 1####
- f. Power Range Neutron Flux, Not P-10 (8elow P-10) N.A. R(4) M(8) N.A. N.A. 1###, 2 R
- g. Turbine Impulse Chamber Pressure, P-13 N.A. R M(8) N.A. N.A. I h 19. Reactor Trip Breaker N.A. N.A. N.A. M(7,11) N.A. 1, 2, 3*, 4*, 5*
- 20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) 1, 2, 3*, 4*, 5*
Logic O
~ - . ~S .
i - , w e
TABLE 4.3-1 (Continued) TABLE NOTATIONS PROOF & REVIEW Only if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal. '
# Above P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint. -
## Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
### Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
#### Above P-10 (Low 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 ,
~ Specification 4.0.4 are not applicable for entry into MODE 2 or 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 applicable 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 manufacturer'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 provisions 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 I interlock is in the required state by observing the permissive status light. l (9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verifi- , cation that permissives P-6 and P-10 are in their required state for I existing plant conditions by observation of the permissive status light. l Monthly surveillance shall include verification of the Boron Dilution l Alarm Setpoint of less than or equal to one-half decade (square root of 10) i above background. (10) Setpoint verification is not applicable. (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. l (12) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate. CATAWBA - UNITS 1 AND 2 3/4 3-12
INSTRUMENTATION DRAFT
#80g 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION /
LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3 ACTION:
- a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value 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 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.
sn.1g M CATAWBA - UNITS 1 AND 2 3/4 3-13 OCT 7 1985
INSTRUMENTATION SURVEILLANCE REQUIREMENTS PROOF & REVIN copy 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of tha Engineered Safety Features Actuation System 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 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 per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. l JUL 1b CATAWBA - UNITS 1 AND 2 3/4 3-14 7gg l 1 . _
TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIM E
- p i
MINIMUM c TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w 1. Safety Injection (Reactor g Trip, Phase "A" Isolation, iB Feedwater Isolation, Control to Room Area Ventilation Opera-tion, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Operation, Auxiliary Building Filtered Ventilation Exhaust Operation, Emergency Diesel Generator Operation, Component R
- Cooling Water, Turbine Trip, and Nuclear Service Water y Operation) un
- a. Manual Initiation 2 1 2 1,2,3,4 18
- b. Automatic Actuation 2 1 2 1,2,3,4 14 3
o Logic and Actuation O Relays m go
- c. Containment 3 2 22 Pressure-High 2 1,2,3 15* q g
- d. Pressurizer 4 2 3 1, 2, 3# 19* o.
Pressure-Low O S H e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15* Low in any steam line u
TABLE 3.3-3 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION Ri MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
]
g 2. Containment Spray g a. Manual Initiation 2 1 2 1,2,3,4 18 N
- b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays
- c. Containment Pressure- 4 2 3 1,2,3 16 High-High R 3. Containment Isolation s
w a. Phase "A" Isolation h 1) Manual Initiation 2 1 2 1,2,3,4 18
- 2) Automatic Actuation 2 1 2 1,2,3,4 14 o Logic and Actuation o Relays Safety Injection
[
- 3) See Item 1. above for all Safety Injection initiating functions and ::o requirements. g
- b. Phase "B" Isolation (Nuclear Service Water Operation) $
O
- 1) Manual Initiation 2 1 2 1,2,3,4 18 O N
- 2) Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays Q c)
S r h 3) Containment Pressure-High-High 4 2 3 1,2,3 16 D D cm
.J W
- I
TABLE 3.3-3 (Continued) 9 U ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5 HINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE Z w FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION -
- 3. Containment Isolation (Continued)
- c. Purge and Exhaust Isolation
- 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 4
- 3) Safety Injection See Ites 1. above for all Safety Injection initiating functions and Y requirements.
0
- 4. Steam Line Isolation ,
- a. Manual Initiation
- 1) System 2 1 2 1,2,3 22
- 2) Individual 1/ steam line 1/ steam line 1/ operating 1,2,3 23 steam line m:o
- b. Automatic Actuation 2 1 2 1,2,3 21 o Logic and Actuation Relays go
- c. Containment Pressure- 4 2 3 1,2,3 16 "
High-High g g '$ ti d. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15* 2::p $ Low any steam o eg g - w
TABLE 3.3-3 (Continued) S g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5 2-8 MINIMUM c TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP h, u OPERABLE MODES ACTION H 4. Steam Line Isolation (Continued)
- e. Steam Line Pressure - 3/ steam line 2/ steam line 2/ steam Ifne 3## 15*
m Negative Rate-High in any steam line
- 5. Feedwater Isolation
- a. Automatic Actuation 2 1 2 1, 2 27 Logic and Actuation g Relays i
y b. Steam Generator 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2 19* N
- Water Level- in each oper- 2
- in any oper- o High-High (P-14) ating stm.
gen. ating sta. gen. p Q c. avg -Low (P-4 T 2 Interlock) 4 2 3 1, 2 19* g
~
C3
- d. Doghouse Water <
Level-High 2 1 2 1, 2 27 8
- e. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
k
- 6. Turbine Trip D
u a. Manual Initiation 1 1 1 1,2 25 1 c. M C F b. Automatic Actuation O F Logic and Actuation Relays 2 1 2 1,2 27 w a. v .
i TABLE 3.3-3 (Continued) 9 g ENGINEERED SAFETY FEATURES "CTUATION SYSTEM INSTRUMENTATION 5 2 MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
]
- 6. Turbine Trip (Continued)
N o c. Steam Generator Water N Level-High-High (P-14) 4/sta. gen. 2/sta. gen. 3/sta. gen. 1,2 19* in any in each operating stm. operating sta. gen. gen.
- d. Trip of All Main w Feedwater Pumps 2/ pump 1/ pump 1/ pump 1,2# 25 1
w e. Reactor Trip (P-4) 2 2 2 1,2,3 22
- f. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
g o
- 7. Containment Pressure Control O
3 System 3
- a. Start Permissive 4/ train 2/ train 3/ train 1, 2, 3, 4 19* [
- b. Termination 4/ train 2/ train 3/ train 1, 2, 3, 4 19*
- 8. Auxiliary Feedwater 8o
- a. Manual Initiation 1/ train 1/ train 1/ train 1, 2, 3 26 ,
- b. Automatic Actuation Logic 2 1 2 1,2,3 21 g e and Actuation Relays E Oeco .
N u,
TABLE 3.3-3 (Continued) 9 5;! ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTR'JMENTATION 2 MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
]
- 8. Auxiliary Feedwater (Continued)
N C'
- c. Sta. Gen. Water Level-
" Low-Low
- 1) Start Motor-Driven. Pumps 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 19*
in any opera- in each ting stm. gen. operating sta. gen. R* 2) Start Turbine-Y Driven Pump 4/sta. gen. 2/sta. gen. 3/sta. gen. 1,2,3 19* o
@ in any two in each a operating operating c3 stm. gen. stm. gen. '"
- d. Safety Injection- M Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions $
and requirements. g
- e. Loss-of-Offsite Power- 8 m
Start Motor-Driven
- I Pumps and Turbine-Driven
' . Pump 6-3/ bus 2/ bus 2/ bus 1, 2, 3 15* either bus ' S f. Trip of All Main D
~ Feedwater Pumps- D on Start Motor-Driven Pumps 2/ pump 1/ pump 1/ pump 1, 2# 25
- l[ .
m
o l TABLE 3.3-3 (Continued) ' 2-
$ ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5
2-e MINIMUM c: TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 8. Auxiliary Feedwater (Continued) 2 g. Auxiliary Feedwater 5 Suction Pressure-low
- 1) CAPS 5220, 5221, 6-3/ pump 2/ pump 2/ pump 1, 2, 3 15 522?
- 2) CAPS 5230, 5231, 5232 6-3/ pump 2/ pump 2/ pump 1,2,3 15
- 9. Containment Sump Recirculation t'
- a. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation T
N Relays .@
- b. Refueling Water Storage o o
Tank Level-Low 4 2 1,2,3,4 Coincident With 3 16 [
- o Safety Injection See Item 1. above for all Safety Injection initiating functions and Q requirements. Si
- 10. Loss of Power
- a. 4 kV Bus Undervoltage- 3/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 15* O o
Loss of Voltage X
- b. 4 kV Bus Undervoltage-
. Grid Degraded Voltage 3/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 d5*r .
- 11. Control Room Area '
E Ventilation Operation
- a. Automatic Actuation Logic w and Actuation Relays 2 1 2 All 24 5,
u
TABLE 3.3-3 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMFNTATION 5 2-MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE
] MODES ACTION
- 11. Control Room Area g Ventilation Operation (Continued)
- b. Loss-of-Offsite Power 3 2 2 1,2,3 IS*
- c. Safety Injection See Item 1. above for all Safety Injections initiating functions and i-requirements.
- 12. Containment Air Return and Hydrogen Skimmer Operation 5 a. Manual Initiation 2 1 2 1,2,3,4 18 m b. Automatic Actuation Logic m N
- o and Actuation Relays 2 1 2 1,2,3,4 14 Sn
- c. Containment Pressure- 9' '
High-High 4 2 3 1,2,3 16 = m
- 13. Annulus Ventilation Operation Ri
- E
- a. Manual Initiation 2 1 2 1,2,3,4 18 8u
- b. Automatic Actuation Logic "
and Actuation Relays 2 1 2 1,2,3,4 14
- c. Safety Injection Sea Item 1. above for all Safety Injection initiating functions gg and requirements.
n g 14. Nuclear Service Water Operation h Eg zB a. Manual Initiation 2 1 2 1,2,3,4 18
- b. Automatic Actuation Logic ,
and Actuation Relays 2 1 2 1,2,3,4 21 1 i i ,
n TABLE 3.3-3 (Continued) f
$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E
MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE ACTION
] MODES
- 14. Nuclear Service Water Operation g (Continued) m c. Loss-of-Offsite Power 3 2 2 1,2,3 15*
- d. Containment Spray See Item 2. above for all Containment Spray initiating functions and requirements.
- e. Phase "B" Isolation See Item 3.b. above for all Phase "B" Isolation initiating functions and requirements.
- f. Safety Injection See Item 1. above for all Safety Injection initiating functions y and requirements.
w g. Suction Transfer-Low A Pit Level 2 1 2 1,2,3,4 21 O
- 15. Emergency Diesel Generator
- Operation (Diesel Building o
,o Ventilation Operation, Nuclear Service Water Operation) 3 3
- a. Manual Initiation 2 1 2 1,2,3,4 18 $
- b. Automatic Actuation Logic and Actuation Relays 2 1 2 1,2,3,4 21 g
- c. Loss-of-Offsite Power 3 2 2 1,2,3,4 15* Q
- d.
- Shfety Injection See Ites 1. above for all Safety Injectior. initiating functfons' -
and requirements. E
" 16. Auxiliary Building Filtered Exhaust Operation N
- a. Manual Initiation 2 1 2 , 1,2,3,4 18 E
- b. Automatic Actuation Logic and Actuation Relays 2 1 2 1,2,3,4 21
TABLE 3.3-3 (Continued) s' . h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E. 2 MINIMUM ' E TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
]
- 16. Auxiliary Building Filtered g Exhaust Operation (Continued) m c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
- 17. Diesel Building Ventilation Operation
- a. Manual Initiation 2 1 2 1,2,3,4 18 w b. Automatic Actuation Logic
> and Actuation Relays 2 1 2 1,2,3,4 21 Y c. Emergency Diesel Generator T Operation See Item 15. above for all Emergency Diesel Generator Operation initiating functions and requirements.
- o
- 18. Engineered Safety Features Actuation System Interlocks @
m '
- a. Pressurizer Pressure, 3 2 2 1,2,3 20 P-11 $
- b. Pressurizer Pressure, 3 2 2 1,2,3 20 Fi not P-11 E
- c. Low-Low T,yg, P-12 4 2 3 1,2,3 20 r
Q
' Reactor Trip, P-4 i
- d. 2 2 2 1,2,3 22 1
8*
- e. Steam Generator Water 4/sta. gen. 2/sta. gen. 3/sta. 1,2,3 20 Level, P-14 in any gen. in 4
operating each [ sta. gen. operating y sta. gen.
PROOF g REVIEW COPY TABLE 3.3-3 (Continued) TABLE NOTATIONS
# Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) setpoint.
M 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.
ACTION STATEMENTS ACTION 14 - 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 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 and exhaust valves are maintained closed. ACTION 18 - With the number of OPERABLE channels one less than +.he Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT 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: i l a. The inoperable channel is placed in the tripped condition within 1 hour, and
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for sur-veillance testing of other channels per Specification 4.3.2.1.
CATAWBA - UNITS 1 AND 2 3/4 3-25 JUL 181985 OCT 7 1985 -
TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) ACTION 20 - With less than the Minimum Channels OPERABLE, withih 1 hour determine by observation of the associated permissive status light (s) that the interlock is in its required statie 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 SHUTDOW within the following 6 hours; however, one channel may be bypassed for up to 2 hours fo 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 SHUTDOWN within the following 6 hours. ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channeis, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.4. ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE, restore the inoperable channel to OPERABLE status within 48 hours, or initiate and maintain operation of the Control Room Area Ventilation System with flow through the HEPA filters and carbon adsorbers. ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours. ACTION 26 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirertent, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTOOWN within the following 6 hours. 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. 0CT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-26
TABLE 3.3-4 n h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 2 SENSOR E TOTAL ERROR
] FUNCTIONAL UNIT ALLOWANCE (TA) Z , (S) TRIP SETPOINT ALLOWABLE VALUE
- 1. Safety Injection (Reactor Trip,
@ Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation. Emergency Diesel Generator Operation, R
Component Cooling Water, Turbine Trip, and Nuclear Y Service Water Operation) O
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. o and Actuation Relays ~
-O a
o
- c. Containment Pressure-High 8.2 0.71 1. 5 $ 1.2 psig 1 1.4 psig
{
- d. Pressurizer Pressure-Low 16.1 14.4 1. 5 1 1845 psig 1 1839 psig
- e. Steam Line Pressure-Low 4.6 1.31 1.5 1 725 psig 1 694 psig* g
- 2. Containment Spray ' ' 0 0
o k a. Manual Initiation N.A. N.A. N.A. N.A. N.A. y b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. g and Actuation Relays , El
- c. Containment Pressure-High-High 12.7 0.71 1.5 1 3 psig 5 3.2 psig
TABLE 3.3-4 (Continued) 9
# ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E
- SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE G
- 3. Containment Isolation s
@; a. Phase "A" Isolation o
- 1) Manual Initiatton N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- 3) Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
R#
- b. Phase "B" Isolation (Nuclear Service Water Operation)
T
~"
y 1) Manual Initiation N.A. N.A. N.A. N.A. N.A. 9a
- o
- 2) Automatic Actuation Logic and Actuation N.A. N.A. N.A. N.A. N.A. Q Ri Relays E o
- 3) Containment Pressure- 12.7 0.71 1.5 1 3 psig 5 3.2 psig -
High-High ."o
- c. Purge and Exhaust Isolation C3 g
- 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
o I2! g r-- 2) Automatic Actuation N.A. N.A. N.A. N.A. N.A. Logic and Actuation
- a Relays h 3) Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
i f l TABLE 3.3-4 (Continued) bh 2-ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 6, 3 SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE m 1 4. Steam Line Isolation g a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
@ b. Automatic Actuation Logic N.A.
t3 N.A. N.A. N.A. N.A. m and Actuation Relays
- c. Containment Pressure-High-High 12.7 0.71 1.5 5 3 psig 5 3.2 psig
- d. Steam Line Pressure - Low 4.6 1.31 1.5 1 72b psig 1 694 psig"
- e. Steam Line Pressure- 8.0 0.5 0 5 100 psi 5 122.8 psi **
Negative Rate - High
- 5. Feedwater Isolation g) R a. Automatic Actuation Logic N.A.
- N.A. N.A. N.A. N.A.
Actuation Relays
- b. Steam Generator Water go up Level-High-High (P-14) E,j
. 1. Unit 1 5.4 2.18 1.5 i
5 82.4% of 5 84.2% of narrow narrow range range instrument rg i instrument span -- span
- 2. Unit 2 Sj
- 9. 7 2.18 1. 5 1 78.1% of narrow range 1 79.9% of narrow range instrument []
,o instrument span -<
span ca c. T,yg-Low 4. 0 1.12 1. 2 1 564 F > 562"F 54 d. Doghouse Water Level-High 1.0 0 0.5 11 inches 12 inches
.a above 577' above 577' floor level floor level ff e. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
___ _ _._.____.__ _ _ _. -__-- _ _ _ . e
TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ( 5 SENSOR i TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z. (S) TRIP SETFOINT ALLOWABLE VALUE
- 6. Turbine Trip
] s a. Manual Initiation N.A. N.A. N.A. N.A. N.A. I b. Automatic Actuation N.A. N.A. N.A. N.A. N.A. Logic and Actuation Relays
- c. Steam Generator Water Level-High-High (P-14)
- 1. Unit 1 5.4 2.18 1.5 5 82.4% of 5 84.2% of narrow narrow range range instrument instrument span span R
- 2. Unit 2 9.7 2.18 1.5 < 78.1% of < 79.9% of narrow iiarrow range range instrument
{ o instrument span span
- d. Trip of All Main N.A. N.A. N.A. N.A. N.A.
Feedwater Pumps
- e. Reactor Trip (P-4) N.A. N.A. N.A. N.A. N.A. $
o
- f. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- 7. C'ntainment Pressure Control Q System , @
- a. Start Permissive N.A. N.A. N.A. 5 0.4 psid 5 0.45 psid ['
- b. Termination N.A. N.A. N.A. > 0.3 psid -> 0.25 psid u o -<
o 8. Auxiliary Feedwater
- a. Manual Initiation N.A N.A. N.A. N.A N.A.
$ b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
y and Actuation Relays
j 3 TABLE 3.3-4 (Continued) 2-g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS h SENSOR ERROR
, TOTAL ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE c- FUNCTIONAL UNIT
- 8. Auxiliary Feedwater (Continued)
- c. Steam Generator Water 2 Level - Low-Low z
o 1) Unit 1 17 14.2 1.5 > 17% of span > 15.3% of N Trom 0% to span from 0% to 30% RTP 30% RTP increasing increasing linearly to linearly to > 53.2% of span
> 54.9% of Trom 30% to 100%
span from 30% RTP to 100% RTP
- 2) Unit 2 14.2 1. 5 > 17% of > 15.3% of narrow
$ 17 narrow range range instrument w
J, span span
- d. Safety Injection See Item 1. above for all Safety injection Setpoints and Allowable Values.
- e. Loss-of-Offsite Power N.A. N.A. N.A. > 3500 V > 3200 V l
- Trip of All Nain Feedwater h
f. Pumps N.A. N.A. N.A. N.A. N.A.
- g. Auxiliary Feedwater Suction Q
, Pressure-Low go
- 1) CAPS 5220, 5221, 5222 N.A. N.A. N.A. > 10.5 psig > 9.5 psig 20
- 2) CAPS 5230, 5231, 5232 N.A. N.A. N.A. > 6.2 psig > 5.2 psig g l
o 9. Containment Sump Recirculation - E o o
- a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. Q and Actuation Relays h b. Refueling Water Storage N.A. N.A. N.A. > 177.15 inches > 162.4 inches
& Tank Level-Low i Coincident With Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
O TABLE 3.3-4 (Continued) .' 9 s ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE Z
- 10. Loss of Power s
3; a. 4 kV Bus Undervoltage-Loss N.A. N.A. N.A. > 3200 V
-> 3500 V of Voltage -
ro
- b. 4 kV Bus Undervoltage- N.A. N.A. N.A. -> 3685 V > 3611 V Grid Degraded Voltage
- 11. Control Room Area Ventilation Operation R a. Automatic Actuation Logic
- and Actuation Relays N.A. , N.A. N.A. N.A. N.A.
Y M b. Losr of-Offsite Power N.A. N.A. N.A. > 3500 V > 3200 V
- c. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- 12. Containment Air Return and '
- o Hydrogen Skimmer Operation g
o vi
]
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A. so o
- b. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N. A. g O c. Containment Pressure- 12.7 0.71 e e
- 1. 5 -< 3 psig -< 3.2 psig o High-High @
-a -<
'l
TABLE 3.3-4 (Continued) ,
> ;u $ ENGINEERE0 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS - 0 0
E
' SENSOR 9' TOTAL ERROR m E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE Q .-e
[ 13. Annulus Ventilation Operation o g a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- 14. Nuclear Service Water Operation y a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
Y b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. ti and Actuation Relays
- c. Loss-of-Offsite Power N.A. N.A. N.A. 1 3500 V 2 3200 V
- d. Containment Spray See Item 2. above for all Containment Spray Setpoints and Allowable Values.
- e. Phase "B" Isolation See Item 3.b. above for all Phase "B" Isolation Setpoints and Allowable Values.
- f. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- g. Suction Transfer-Low Pit Level NA NA NA 1El. 554.4 ft. gel. 552.9 ft.
- 15. Emergency Diesel Generator SIE Operation (Diesel Building
" Ventilation Operation, Nuclear h J Service Water Operation) $$ a. Manual Initiation N.A N.A. N.A. N.A. N.A.
i i , i i I TABLE 3.3-4 (Continued) S 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE i p -
- 15. Emergency Diesel Generator Operation (Diesel Building 3;
Ventilation Operation, Nuclear Service Water Operation) (Continued) ' i N } b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. . i and Actuation Relays 1
- c. Loss-of-Offsite Power N.A. N.A. N.A. -> 3500 V -> 3200 V
! d. Safety Injection See Item. 1 above for all Safety Injection Setpoints and Allowable Values. R
- 16. Auxiliary Building Filtered Y Exhaust Operation
, Y
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A. -
i
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- c. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
- 17. Diesel Building Ventilation Operation g
- a. Manual Initiation N. A. N.A. N.A. N.A.
, N. A,. ,
- e. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
o and Actuation Relays 4 -
- c. Emergency Diesel Generator See Item 15. above for all Emergency Diesel Generator Operation '
l
-a Operation Setpoints and Allowable Values. Q n
- N. %
s
. t
> TABLE 3.3-4 (Continued)
$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E
SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIi' SETPOINT a ALLOWABLE VALUE
- 18. Engineered Safety Features
[ Actuation System Interlocks 2 [ a. Pressurizer Pressure, P-11 N.A. N.A. N.A. 1955 psig 11944 psig
- b. Pressurizer Pressure, not P-11 N.A. N.A. N. A. 1955 psig <1966 psig
- c. Low-Low T,yg, P-12 N.A. N.A. N.A. 1553*F 1551*F
- d. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.
R
- e. Steam Generator Level, P-14 See Item 5. above for all Steam Generator Water Level Trip Setpoints and Allowable Values.
l^* u R M M
%_ e,
- w. -<
? MM
- 1 TABLE 3.3-4 (Continued)
TABLE NOTATIONS PROOFg #EV/gWC
- Time constants utilized in the lead-lag controller for Steam Line '
Pressure-Low are 12 > 50 seconds and T2 5 5 seconds. Channel calibration 1 shall ensure that these time constants are adjusted to these values. l
**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. CATAWBA - UNITS 1 AND 2 3/4 3-36 M 10 00T 7 1985
TABLE 3.3-5 ON/fW ' COPP ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. Manual 2nitiation ,
- a. Safety Injection (ECCS) N.A.
- b. Containment Spray N.A.
- c. Phase "A" Isolation N.A.
- d. Phase "B" Isolation N.A.
- e. Purge and Exhaust Isolation N.A.
- f. Steam Line Isolation N.A. ,
- g. Diesel Building Ventilation Operation N.A.
- h. Nuclear Service Water Operation N.A.
- i. Turbine Trip N.A.
I
- j. - Component Cooling Water N.A.
- k. Annulus Ventilation Operation N.A.
- 1. Auxiliary Building Filtered N.A.
Exhaust Operation
- m. Reactor Trip N.A.
- n. Emergency Diesel Generator Operation N.A.
- o. Containment Air Return and Hydrogen Skimmer Operation N.A. .
- p. Auxiliary Feedwater N.A.
- 2. Containment Pressure-High
- a. Safety Injection (ECCS) $ 27(1)/12(3)
- 1) Reactor Trip _2
- 2) Feedwater Isolation -
<7
- 3) Phase "A" Isolation (2) $ 18(3)/28(4) l 4) Purge and Exhaust Isolation <6
- 5) Auxiliary Feedwater(5) N.A.
- 6) Nuclear Service Water Operation 1 65(3)/76(4)
- 7) Turbine Trip N.A.
- 8) Component Cooling Water 5 65(3)/76(4) ,
- 9) Emergency Diesel Generator Operation 5 11 .
i 10) Control Room Area Ventilation l Operation N.A. CATAWBA - UNITS 1 AND 2 3/4 3-37 OCT 7 1985
l TABLE 3.3-5 (Continued) PRoorg R&lEW Cop ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 2. Containment Pressure-High (Continued)
~
. 11) Annulus Ventilation Operation 1 23
- 12) Auxiliary Building Filtered N.A.
Exhaust Operation
- 13) Containment Sump Recirculation N.A.
- 3. Pressurizer Pressure-Low
- a. Safety Injection (ECCS) < 27(1)/12(3)
. 1) Reactor Trip <2 -
- 2) Feedwater Isolation <7
- 3) Phase "A" Isolation (2) 18(3)/28(4)
- 4) Purge and Exhaust Isolation <6
- 5) Auxiliary Feedw'ater(5) N.A.
- 6) Nuclear Service Water Operation 1 65(3)/76(4)
- 7) Turbine Trip N.A.
- 8) Component Cooling Water 1 65(3)/76(4)
- 9) Emergency Diesel Generator Operation i 11
- 10) Control Room Area Ventilation N.A.
1 Operation
- 11) Annulus Ventilation Operation i 23
- 12) Auxiliary Building Filtered N.A.
Exhaust Operation
- 13) Containment Sump Recirculation N.A.
- 4. Steam Line Pressure-Low
- a. Safety Injection (ECCS) _< 12(3)/22(4)
- 1) Reactor Trip _2
- 2) .Feedwater Isolation <7
- 3) Phase "A" Isolation (2) h18(3)/28(4)
- 4) Purge and Exhaust Isolation <6
- 5) Auxiliary Feedwater(5) < 60
- 6) Nuclear Service Water Operation 65(3)/76(i)
- 7) Turbine Trip N.A.
- 8) Component Cooling Water i 65(3)/76(4)
- 9) Emergency Diesel Generator Operation i 11 CATAWBA - UNITS 1 AND 2 3/4 3-38 OCT 7 1985
TABLE 3.3-5 (Continued) PRoopE Sfp/EW ENGINEERED SAFETY FEATURES RESPONSE TIMES DP}' INITIATING SIGNAL AND FUNCTION RESPONSE TIME JN SECONDS
- 4. Steam Line Pressure-Low (Continued) -
- 10) Control Room Area Ventilation N.A.
Operation
- 11) Annulus Ventilation Operation < 23
- 12) Auxiliary Building Filtered Exhaust Isolation N.A.
- 13) Containment Sump Recirculation N.A.
~ b. Steam Line Isolation _7
- 5. Containment Pressure-High-High
- a. Containment Spray < 45 Phase "B" Isolation b.
_< 65(3)/76(4) Nuclear Service Water Operation N.A.
- c. Steam Line Isolation _7
- d. Containment Air Return and Hydrogen -< 600 Skimmer Operation
- 6. Steam Line Pressure - Negative Rate-High
- Steam Line Isolation _7
- 7. Steam Generator Water Level-High-High
- a. Turbine Trip <3
- b. Feedwater Isolation _7
- 8. T -l
- avg Feedwater Isolation N.A.
! 9. Doghouse Water Level-High Feedwater Isolation N.A.
- 10. Start Permissive
- Containment Pressure Control System N.A.
l .
- 11. Termination .
Containment Pressure Control System N.A. OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-39
f & REVIEW Copy TABLE 3.3-5 (Continued) i ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 12. Steam Generator Water Level-Low-Low
- a. Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Turbine-Driven Auxiliary Feedwater Pump 5 60
- 13. Loss-of-Offsite Power
- a. Motor-Driven Auxiliary Feedwater Pumps 5 60
- b. Turbine-Driven Auxiliary Feedwater Pumps 1 60 ,
- c. Control Room Area Ventilation Operation N.A.
- d. Emergency Diesel Generator Operation 5 11 ~
- 1) Diesel Building Ventilation Operation N.A.
- 2) Nuclear Service Water Operation 5 65(3)/76(4)
- 14. Trip of All Main Feedwater Pumps
- a. Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Turbine Trip N. A.
- 15. Auxiliary Feedwater Suction Pressure-Low Auxiliary Feedwater (Suction Supply < 15(6)
Automatic Realignment)
- 16. Refueling Water Storage Tank Level-Low Coincident with Safety Injection Signal (Automatic Switchover to Containment Sump)
$ 60
- 17. Loss of Power
- a. 4 kV Bus Undervoltage - -< 8.5 Loss of Voltage
- b. 4 kV Bus Undervoltage- < 600 Grid Degraded Voltage
- 18. Suction Transfer-Low Pit Level Nuclear Service Water Operation N.A.
l 1 CATAWBA - UNITS 1 AND 2 3/4 3-40 JUL 18 BBS j OCT 71985
TABLE 3.3-5(Continuep TABLE NOTATIONS PRO 0f&REvlE (1) Diesel generator starting and sequence loading delay, included. Response time limit includes opening of valves to establish aafety Infbetion path and attainment of discharge pressure for centrifugal charging _ pumps, Safety Injection and residual heat removal pamps. (2) Valves KC305B and KC3158 are exceptions to the response times listed in the table. The following response times in seconds are the required velues for these valves for the initiating signal and function indicated:
- 2. d < 30( )
i 3.d 4.d 7 7 3030((3)/40(4) 3)/40(4)
(3) ' Diesel generator starting and sequence loading delays not included.
Offsite power available. Response time limit includes opening of valves , to establish Safety Injection path and attainment of discharge pressure for centrifugal charging punips. (4) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish Safety Injection path and attainment of discharge pressure for centrifugal charging pumps. (5) Response time for motor-driven auxiliary feedwater pumps on all Safety Injection signals shall be less than or equal to 60 seconds. Response time limit includes attainment of discharge pressure for auxiliary feedwater pumps. (6) Response time includes a 5-second delay. o l l QCT 7 1985
- CATAWBA - UNITS 1 AND 2 3/4 3-41 l
l l
o TABLE 4.3-2 f 9 5;f ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS TRIP
, E ANALOG ACTUATING MODES Q CHANNEL DEVICE MASTER SLAVE FOR WHICH
- CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE CHANNEL FUNCTIONAL UNIT CHECK _CALJ PRATIO_N TEST _ TEST LOGIC TEST TEST TEST IS REQUIRED
- 1. Safety Injection (Reactor Trip,
" Phase "A Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary
- Feedwater-Motor-Driven Pump, Purge and Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered 4:' Exhaust Operation, Emergency
- Diesel Generators Operation, Y Component Cooling Water, A Turbine Trip, and Nuclear i Service Water Operation)
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
- b. Automatic Actua- N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 tion Logic and Actuation Relays .3 I c. Contafnment S R H N.A. N.A. N.A. N.A. 1, 2, 3 b Pressure-High [
- d. Pressurizer S R M N.A. N.A. N.A. N.A. 1, 2, 3 :o Q
Pressure-Low e e - 1,2,3 .Q
- e. Steam Line S R M N.A. N.A. N.A. N.A.
o Pressure-Low n O 2. Containment Spray % a a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
@ b. Automatic Actua- N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4
"' tion Logic and Actuation Relays .
TABLE 4.3-2 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS
. TRIP c- ANALOG ACTUATING MODES 5 CHANNEL DEVICE MASTER SLAVE FOR WHICH d CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
~ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED R 2. Containment Spray (Continued) m c. Containment S R H N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-High-High
- 3. Containment Isolation
- a. Phase "A" Isolation R*
- 1) Manual Initia- N.A. N.A. N.A. R N.A. N.A. N.A 1,2,3,4 tion E 2) Automatic Actua- N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 8
M tion Logic and Actuation Relays
'O 3) Safety Injection . See Item 1. above for all Safety Injection Surveillance Requirements. %
- b. Phase "B" Isolation (Nuclear Service Water Operation) 8 1) Manual Initia- !!. A. N.A. N.A. R N.A. N.A. N.A 1, 2, 3, 4 y tion
- 2) Automatic Actua- N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 tion Logic and ,
Actuation Relays
@"E 3) Containment S R M N.A. N.A. N.A. N.A. 1,2,3
'; Pressure-High-High ,
- c. Purge and Exhaust Isolation
- 1) Manual Initia- N.A. N.A. N.A. R -
N.A. N.A. N.A. 1, 2, 3, 4 tion i
TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS _
" TRIP
' ANALOG ACTUATING MODES
; E CHANNEL DEVICE MASTER SLAVE FOR MIICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE Q CHANNEL FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
]
3E 3. Containment Isolation m (Continued)
- 2) Automatic Actua- N.A. N.A. H.A. N.A. M(1) M(1) Q 1,2,3,4 tion Logic and Actuation Relays
- 3) Safety Injection See Item 2. above for all Safety Injection Surveillance Requirements. o
, w m 1 4. Steam Line isolation Y a. Manual II itiation 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 W 2
Logic and Actuation Relays $
- c. Containment S R M N.A. N.A. N.A. N.A. 1, 2, 3 @
Pressure-High- n High Q
- d. Steam Line S R H N.A. N.A. N.A. N.A. 1, 2, 3
- Pressure-Low
- e. Steam Line Pressure- S R H N.A. N.A. N.A. N.A. 3 Negative Rate-High g S. Feedwater Isolation lllllQ l
! oU a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2 I 3 Logic and Actuation
. Relays
- a l
aB i
TABLE 4.3-2 (Continued) 9 '-ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Hi
" TRIP
' ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH E OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE Z CHANNEL CHANNEL CHANNEL
- TEST LOGIC TEST TEST TEST IS REQUIRED FUNCTIONAL UNIT CHECK CALIBRATION TEST s
g 5. Feedwater Isolation (Continued) M N.A. N.A. N.A. N.A. 1, 2 m b. Steam Generator S R Water Level-High- m m g High (P-14) o m
- c. T,yg-Low (P-4 N.A. N.A. N.A. N.A. 1, 2 P Interlock) S R M m
- d. Doghouse Water m level-High N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2 $
- e. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. @
n T o 3 6. Turbine Trip
- N.A. N.A. N.A. N.A. N.A. 1, 2
- a. Manual Initiation N.A. R N.A. N.A. M(1) M(1) Q 1, 2 Automatic N.A. N.A.
- b. O Actuation Logic and Actuation Relays W M N.A. N.A. N.A. N.A. 1, 2
- c. Steam Generator S R i Water Level-High-High (P-14)
N.A. N.A. N.A. N.A. 1, 2
- d. Trip of All Main N.A. N.A. R Feedwater Pumps N.A. N.A. N.A. N.A. N.A. 1, 2, 3
- e. Reactor Trip (P-4) N.A. R o c--
Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. g jE f. y 7. Containment Pressure
- - Control System E8 mm a. Start Permissive S R M N.A. N.A. N.A. N.A. 1, 2, 3, 4 H N.A. ~ N.A N.A. N.A. 1,2,3,4
- b. Termination S R
l n TABLE 4.3-2 (Continued) D ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP E ANALOG ACTUATING MODES Z CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED E
- 8. Auxiliary Feedwater
- a. Phnual 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
- c. Steam Generator S R H N.A. N.A. N.A N.A 1,2,3 y Water Level-Low-Low y w d. Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements. $
h e. Loss-of-Offsite N.A. R N.A. M(3) N.A. N.A. N.A 1,2,3 Power
- f. Trip of Al1 Main N.A. N.A. N.A. R N.A. N.A. N.A 1, 2 Q Feedwater Pumps
- g. Auxiliary Feedwater n
Suction Pressure- o Low N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 4
- 9. Containment Sump Recirculation
- a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 c Logic and Actuation Relays
-e ~lES_
- b. Refueling Water S R M N.A. N.A. N.A. N.A. 1, 2, 3, 4
% Storage Tank Level -
E n; Q Low Coincident With
$m Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. E 3:n
n TABLE 4.3-2 (Continued) 4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Z CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST [ IS REQUIRED g 10. Loss of Power to a. 4 kV Bus - N.A. R N.A. M (2) N.A. N.A. N.A. 1, 2, 3, 4 Undervoltage-Loss of Voltage
- b. 4 kV Bus N.A. R N.A. M (2) N.A. N.A. N.A. 1, 2, 3, 4 Undervoltage-Grid OegradedVolpage w 11. Control Room Area Ventilation Operation o D a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q All $
w logic and Actuation g Relays Q p
- b. Loss-of-Offsite N.A. R N.A. R N.A. N.A. N.A. 1, 2, 3 :e Power
- c. Safety Injection Q
See Item 1. above for all Safety Injection Surveillance Requirements. Q
- 12. Containment Air Return o and Hydrogen Skimmer Operation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 i b. Automatic N.A. N. A. N.A. N.A. M(1) M(1) Q 1,2,3,4
- Actuation Logic and Actuation Relays 'C g,E c. Containment S R M N.A. N.A. N.A. N.A. 1,2,3 w-cm Pressure-High-High
*- 13. Annulus Ventilation .
Q$ Operation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
TABLE 4.3-2 (Continued) O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Z CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED [ R 13. Annulus Ventilation Operation (Continued)
" b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Actuation Logic and Actuation Relays
- c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. E
- 14. Nuclear Service Water Operation
; 1 a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,49' Y
i!;
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4h -
and Actuation @ Relays a
- c. Loss-of-Of fsite N.A. R N.A. M(3) N.A. N.A. N.A. 1, 2, 3 Power g
- d. Containment Spray See Item 2. above for all Containment Spray Surveillance Requirements.
- e. Phase "B" Isolation See Item 3.b. above for all Phase "B" Isolation Surveillance Requirements.
- f. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- g. Suction Transfer-g Low Pit Level S R R N.A. N.A. N.A. N.A. 1, 2, 3' QE 15. Emergency Diesel
- Generator Operation N oa (Diesel Building sg Ventilation Operation, .
0? vi Nuclear Service Water Operation) i a. Manual Initiatidn N.A. N.A. N.A. R - N.A. N.A. N.A. 1,2,3,4
~
TABLE 4.3-2 (Continued) O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH
, Q CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE l FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST
[ TEST IS REQUIRED R 15. Emergency Diesel
, Generator Operation j (Diesel Building ! Ventilation Operation l
Nuclear Service Water Operation) (Continued)
- b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Actuation Logic w and Actuation j 1 Relays Y c. Loss-of-Of fsite N.A. R N.A. M(2) N.A. N.A. N.A. 1, 2, 3, 4
$; Power
- d. Safety Inicction See Item 1. above for all Safety Injection Surveillance Requirements.
- 16. Auxiliary Building Filtered Exhaust Operation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
- b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Actuation Logic and Actuation Relays
- c. Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirembnti %-
- k. %
~
isis w h f I
. T
\ l
TABLE 4.3-2 (Continued)
% ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Z CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
[ E 17. Diesel Building Ventilation Operation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4
- b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Actuation Logic and Actuation Relays w c. Emergency Diesel See Item 15. above for all Emergency Diesel Generator Operation Surveillance ,
i Generator Operation Requirements. g o
, 18. Engineered Safety Features n Actuation System Interlocks Ro
- a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure P-11 g
- b. Pressurizer N.A. R M N.A. N.A. N.A. N.A 1,2,34 Pressure, not P-11 w g
- c. Low-Low T,yg, P-12 N.A. R M N.A. N.A. N.A. N.A. 1,2,3 2
- d. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3
- e. Steam Generator S R H N.A. M(1) M(1) Q 1,2,3 Water Level, P-14 '
, $$3 w r-
- TABLE NOTATIONS U % cc @U (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
u' O (2) Monthly testing shall consist of voltage sensor relay testing excluding -
- actuation of load shedding diesel start, and time - delay timers.
, (3) Monthly testing shall consist of relay testing excluding final actuation of the pumps or valves.
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONIT3 RING 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 operations for the MODES and at the frequencies shown in Table 4.3-3. 4 l l JUL 1B N CATAWBA - UNITS 1 AND 2 3/4 3-51 OCT 7 1985 I -
t TABLE 3.3-6 k RAOIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS E
, MINIMUM c- CHANNELS CHANNELS APPLICABLE ALARM / TRIP
, ?5 FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION d g 1. Containment 5 a. Containment Atmosphere - High 1 1 All *** 30 m Gaseous Radioactivity (Low Range - EMF-39)
- b. Reactor Coolant System Leakage
- Detection i
i 1) Particulate Radioactivity i (Low Range - EMF-38) N.A. 1 1,2,3,4 N.A. 33 R 2) Gaseous Radioactivity
- * (Low Range - EMF-39) N.A. 1 1,2,3,4 N.A. 33 h 2. Fuel Storage Pool Areas
] o
- a. High Gaseous Radioactivity T i (Low Range - EMF-42) 1 1 **
$ 1.7x10 4 pCi/ml 34 8'
- b. Criticality-Radiation Level (Fuel Bridge - Low Range - h 1 EMF-15, 2 EMF-4) 1 1
- 1 15 mR/h 32 h o
! 3. Control Room O o o M o Air Intake-Radiation Level - 1/ intake 2 (1/in- All '~
< 1.7x10 4 pCi/mi 31 .
High Gaseous Radioactivity take) y (Low Range - EMF-43 A & B) IB g 4. Auxiliary Building Ventilation 1 1 All $ 1.7x10 4 pCi/ml 35 High Gaseous Radioactivity
; (Low Range - EMF-41) ; 5. Component Cooling Water System (EMF-46 A&B) 1 1 Al'1 1 1x10 8 pC1/ml 36
N00F & Rgyjgy(gpY TABLE 3.3-6 (Continued) TABLE NOTATIONS With fuel in the fuel storage pool areas. With irradiated fuel in the fuel storage pool areas. ' Trip Setpoint concentration value (pCi/ml) is to be established such that the actual submersion dose rate would not exceed 2 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 parsmeters in the ODCM during containmant 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 30 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge and
- exhaust valves are maintained closed.
ACTION 31 - With the number of operable channels one less than the Minimum i Chaanels OPERABLE requirement, withir. I hour isolate the affected i Control Room Ventilation System intake from outside air with flow through the HEPA filters and carbon adsorbers. ACTION 32 - 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 storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel building. ACTION 33 - Must satisfy the ACTION requirement for Specification 3.4.6.1. ACTION 34 - With the number of OPERA 3LE channels less than the Minimum Channels OPERABLE requirement, operatioi, may continue provided the Fuel Handling Ventilation Exhaust System is operatirig and discharging through the HEPA filters and carbon adsorbers. Otherwise, suspend all operations involving fuel movement in the fuel building. ACTION 35 - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, operation may continue provided the Auxiliary Building Filtered Exhaust System is operating and discharging through tie HEPA f*lter and carbon adsorbers. - ACTION 36 - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, operation may continue for up to 30 days provided that, at least once per 12 hours, grab samplet are collected and analyzed for radioactivity (gross gamma) at a lower limit of detection of no more than 10 7 pCi/ml. OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-53
4 TABLE _4.3-3 h RADIATION MONITORING INSTRUMENTATION FOR PLANT
$ OPERATIONS SURVEILLANCE REQUIREMENTS ANALOG E CHANNEL MODES FOR WHICH Z
CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRE 0 s g 1. Containment
- a. Containment Atmosphere - High N
Gaseous Radioactivity (Low Range - EMF-39) S R M All
- b. Reactor Coolant System S R H 1, 2, 3, 4 Leakage Detection (Low Range - EMF-38 and Low Range - EMF-39) w A 2. Fuel Storage Pool Areas
- a. High Gaseous Radioactivity **
Y S R M
- W (Low Range - EMF-42)
- b. Criticality-Radiation Level S R M .m (Fuel Bridge - Low Range - -
1 EMF-15,2 EMF-4) o
- 3. Control Room Air Intake Radiation Level -
High Gaseous Radioactivity - S R M All Q rn (Low Range - EMF-43 A & B) W
- 4. Auxiliary Building Ventilation 8o -
. x -
8 H High Gaseous Radioactivity (Low Range - EMF-41) S R M All [ 5. Component Cooling Water System S R M All g (EMF-46 A&B) , TABLE NOTATIONS
- With fuel in the fuel storage pool area.
** With irradiated fuel in the fuel storage pool areas.
i
. INSTRUMENTATION COPY MOVABLE INCORE DETECTORS l
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 equipment to map these thimbles.
APPLICABILITY: When the Movable Incore Detection System is used for: ,
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoririg the QUADRANT POWER TILT RATIO, or
- c. MeasurementofFh,F(Z)andF q xy, ACTION:
With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. 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 irradiating each detector used and determining the acceptability of its voltage curve for: '
- a. Recalibration of the Excore Neutron Flux Detection system, or
- b. Monitoring the QUADRANT POWER TILT RATIO, or
- c. MeasurementofFh,F(Z),andF g xy*
JUL 181985 ! CI 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-55 l i
INSTRUMENTATION PROOF & REVIEW COPYl 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 CALI-BRATION an'd ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-4. 4.3.3.3.2 Each of the above accessible seismic monitoring instruments actcated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of l the vibratory ground motion. Data retrieved from the triaxial time-history accelerograph shall include a post-event CHANNEL CALIBRATION obtained by actuation of the internal test and calibrate function immediately prior to , removing data. CHANNEL CALIBRATION shall be performed immediately after insertion of the new recording media in the triaxial time-history accelero-graph recorder. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety. 1 1 JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 3-56 00T 7 1985
/
i PROOF & REVIEW COPY TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUU.NTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Time-History Accelerographs
- a. 1MIMT 5070 (Remote Sensor A) -1 g to + 1 g 1 Containment Base Slab
- b. 1MIMT 5080 (Remote Sensor B) -1 g to + 1 g 1 Containment Vessel Elev 619'5"
- c. 1MIMT 5090 (Starter Unit) 0.005 g to 0.05 g 1 Containment Base Slab
- 2. Triaxial Peak Accelerographs
- a. 3MIMT 5010 - Containment Bldg. O g to + 2 g 1 Elev 613'8 9/16"
- b. 1MIMT 5020 - Containment Bldg. O g to + 2 g 1 Elev 567'2 "
- c. 1MIMT 5030 - Auxiliary Bldg. O g to + 2 g 1 Elev 543'
- 3. Triaxial Seismic Switch IMIMT 5000 - Containment 0.025 g to 0.25 g 1*
Base Slab
- 4. Triaxial Response-Spectrum Recorders
- a. 1MIMT 5040 - Containment 0 to 34 g at 1*
Base Slab 2 to 25 Hz
- b. 1MIMT 5050 - Containment Bldg. O to 34 g at 1 l
Elev 579'3 " 2 to 25 Hz l c. IMIMT 5060 - Auxiliary Bldg. O to 34 g at 1 Elev 577' 2 to 25 Hz i l
*With reactor control room indication.
JUL 18 285 CATAWBA - UNITS 1 AND 2 3/4 3-57 OCT 7 1985
FR00F & REVIEW COPY s-
. TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG ~
CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelerographs
- a. 1MIMT 5070 (Remote Sensor A) M* R SA Containment Base Slab
- b. 1MIMT 5080 (Remote Sensor B) M* R SA Containment Vessel Elev 619'5"
- c. IMIMT 5090 (Starter Unit) N.A. R SA Containment Base Slab
- 2. Triaxial Peak Accelerographs
- a. 1MIMT 5010 - Containment Bldg. N.A. R N.A.
Elev 613' 8 9/16"
- b. 1MIMT 5020 - Containment Bldg. N.A. R N.A.
Elev 567' 2 "
- c. 1MIMT 5030 - Auxiliary Bldg. N.A. R N.A.
Elev 543'
- 3. Triaxial Seismic Switch IMIMT 5000 - Containment M R SA Base Slab **
- 4. Triaxial Response-Spectrum Recorders
- a. IMIMT 5040 - Containment Base M R SA Slab **
- b. 1MIMT 5050 - Containment Bldg. N.A. R t ..
Elev 579' 3 "
- c. 1MIMT 5060 - Auxiliary Blag. N.A. R N.A.
Elev 577' - l l
*Except seismic trigger. <
- With reactor control room hdications. i JUL 181985 !
CATAWBA - UNITS 1 AND 2 3/4 3-58 OCT 7 sus l
I l l INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION REWEW COPY h LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE. 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.
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 operations at the frequencies shown in Table 4.3-5. ' l JUL 18 885 CATAWBA - UNITS 1 AND 2 3/4 3-59 OCT 7 1985 l _ . ._ -
PRoopg REVIEW Copy TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE
- 1. Wind Speed
- a. Meteorological Tower Nominal Elev. 661'10" 1
- b. Meteorological Tower Nominal Elev. 768'10" 1
- 2. Wind Direction
- a. Meteorological Tower Nominal Elev. 661'10" 1
- b. Meteorological Tower Nominal Elev. 768'10" 1 ,
- 3. Air Temperature - AT Meteorological Tower Nominal Elev. 768'10"-661'10" 1 I
1 JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 3-60 OCT 71385
TABLE 4.3-5
&lEWCopy hfff[
METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Wind Speed
- a. Nominal Elev. 661' 10" D SA .
- b. Nominal Elev. 768' 10" D SA
- 2. Wind Direction
- a. Nominal Elev. 661' 10" D SA -
- b. Nominal Elev. 768' 10" D SA
- 3. Air Temperature - AT Nominal Elev. 768'10" - 661'10" D SA
=
1 8 1985 CATAWBA - UNITS 1 AND 2 3/4 3-61 7 1985 E _. - -
INSTRUMENTATION F & REVIEW Copy REMOTE SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION - 3.3.3.5 The Remote Shutdown monitoring instrumentation channels given in Table 3.3-9 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, or be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6. 0 OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-62
n TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION 5 MINIMUM READOUT TOTAL NO. CHANNELS E INSTRUMENT LOCATION OF CHANNELS OPERABLE A
- 1. Reactor Trip Breaker Indication Reactor Trip 1/ trip breaker 1/ trip breaker Switchgear
[ 2
- 2. Reactor Coolant Loop A&B Hot Leg Auxiliary Shutdown 2 1*
Temperature Control Panel
- 3. Reactor Coolant Loop A&B Cold Auxiliary Shutdown 2 1*
Leg Temperature Control Panel
- 4. Pressurizer Pressure Auxiliary Shutdown 2 1 Control Panel w
- 5. Pressurizer Level Auxiliary Shutdown 2 1 o Control Panel 4
w a o
- 6. Steam Generator Pressure Auxiliary Feedwater 1/ steam generator 1/ steam generator
- Pump Motor Control 8*
Panels A&B $ 5
- 7. Steam Generator Level Auxiliary Feedwater Pump Motor Control 1/ steam generator 1/ steam generator g Panels A&B 8u
- 8. Auxiliary Feedwater Flow Rate Auxiliary Feedwater 1/ steam generator 1/ steam generator Pump Motor Control e Panels A&B ,
OE [ oo m -
- Channel required to be OPERABLE during operation from remote shutdown panel. D 1
e-
i i TABLE 4.3-6 9 5;! REMOTE SHUTDOWN MONITORING INSTRUMENTATION g SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL E INSTRUMENT CHECK CALIBRATION Q [ 1. Reactor Trip Breaker Indication M N.A. g 2. Reactor Coolant Loop A&B Hot Leg Temperature M R
- 3. Reactor Coolant Loop A&B Cold Leg Temperature M R
- 4. Pressurizer Pressure M R
- 5. Pressurizer Level M R I
- 6. Steam Generator Pressure M R Y
- 7. Steam Generator Level M R :%6 a
o R 8. Auxiliary Feedwater Flow Rate M R '" h 20 a
=
X n - m M% u, w .
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION PROOF & REVIEW C0PE 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, or be in at least HOT STANDBY within the - next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels except the unit vent-high-high range area monitor, the steam relief valve exhaust radiation monitor, the containment atmosphere-high range radiation monitor, and the reactor coolant radiation level lest than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTOOWN withir. the following 6 hours.
~
- c. With the number of OPERABLE Channels for the unit vent-high-high range area monitor, or the steam relief valve exhaust radiation monitor, or the containment atmosphere-high range radiation monitor, or the reactor coolant radiation level less than required by the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s) within 72 hours, and either restore the inoperable channel (s) 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 the 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 operations at the frequencies shown in Table 4.3-7. OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-65 l
TABLE 3.3-10 9 '
# ACCIDENT MONITORING INSTRUMENTATION 5
TOTAL MINIMUM NO. OF CHANNELS E INSTRUMENT CHANNELS OPERABLE Z
- 1. Containment Pressure 2 1
[ g 2. Reactor Coolant Outlet Temperature - TH0T (Wide Range) 2 1 N
- 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 R
- 7. Steam Generator Water Level - Narrow Range 2/ steam generator 1/ steam generator T
- 4 8 8. Refueling Water Storage Tank Water Level 2 1 O o
n
- 9. Auxiliary Feedwater Flow Rate 2/ steam generator 1/ steam generator b
- o
- 10. Reactor Coolant System Subcooling Margin Monitor 1 1 Q
- 11. PORV Position Indicator
- 2/ Valve 1/ Valve k c>
- 12. PORV Block Valve Position Indicator ** O 2/ Valve 1/ Valve o
- 13. Pressurizer Safety Valve Position Indicator 1/ Valve 1/ Valve EN
"# 14. Containment Sump Water Level (Wide Range) 2 1
N C7 y .a e
TABLE 3.3-10 (Continued) ACCIDENT MONITORING INSTRUMENTATION h TOTAL MINIMUM NO. OF CHANNELS e INSTRUMENT CHANNELS OPERABLE l 3 d 15. In Core Thermocouples 4/ core quadrant 2/ core quadrant w 3, 16. Unit Vent - High-High Range Area Monitor (EMF-54) N.A. 1 5 i m 17. Steam Relief Valve Exhaust Radiation Monitor (1 EMF-26, 27, 28 or 29 and 2 EMB-10, 11, 12 or 13) N.A. 1
- 18. Containment Area - High Range Radiation Monitor (EMF-53 A or B) N.A. 1
! 19. Reactor Vessel Water Level 2 1
- 20. Reactor Coolant Radiation Level (EMF-48) N.A. 1 R
Y 0 j TABLE NOTATIONS
- Not applicable if the associated block valve is in the closed position. o
** Not applicable if the associated block valve is in the closed po,sition and power is removed. M o
Go m s ~ o E 8
-?
i ~ t
t TABLE 4.3-7 9 Y ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
# CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Containment Pressure M R
}
" M R
- 2. Reactor Coolant Outlet Temperature - TH0T (Wide Range) z
- 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R
- 4. Reactor Coolant Pressure - Wide Range M R
- 5. Pressurizer Water Level M R
- 'o
- 6. Steam Line Pressure M R @n R 7. Steam Generator Water Level - Narrow Range M R eo
- m
- 8. Refueling Water Storage Tank Water Level M R Q'
- 9. Auxiliary Feedwater Flow Rate M R
- 10. Reactor Coolant System Subcooling Margin Monitor M R
- 11. PORV Position Indicator M R
- 12. PORV Block Valve Position Indicator M R
- 13. Pressurizer Safety Valve Position Indicator M R
- 14. Containment Sump Water Level (Wide Range) M R M
ec 3:m NN SEE
] -
u, CD 5' ,
TABLE 4.3-7 (Continued) 9 g ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E 8 CHANNEL CHANNEL INSTRUMENT (Continued) CHECK CALIBRATION E y 15. In Core Thermocouples M R w N 16. Unit Vent - High-High Range Area Monitor (EMF-54) M R 5 17. Steam Relief Valve Exhaust Radiation Monitor (1 EMF-26, 27, N 28 and 29, 2 EMF-10, 11, 12 and 13) M R
- 18. Containment Area - High Range Radiation Monitor (EMF-53 A&B) M R*
- 19. Reactor Vessel Water Level M R
- 20. Reactor Coolant Radiation Level (EMF-48)
, M it A Y
- b Do
- o
- CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector , Q for range decades above 10R/h and a one point calibration check of the detector below 10R/h with an installed or portable gamma source.
b a
=4 u,
INSTRUMENTATION EN00f & REVIEW CDPy 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 ona Chlorine Deter. tion System inoperable, restore the inoperable system to OPERABLE sta;us within 7 days or within the next 6 hours initiate and raintain operation of the Control Room Area Ventilation System with flow through the HEPA filters and charcoal adsorbers.
- b. With both Chlorine Detection Systems inoperable, within 1 hour initiate and maintain (peration of the Control Room Area Ventilation System wit'i flow through the HEPA filters and charcoal adsorbers.
- c. The provisions of Specification 3.0.4 are not applicable.
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. l i l $ l CATAWBA - UNIT 1 3/4 3-70 7 1985 l l l -. . - - _ _ .
INSTRUMENTATION PROOF & REVIEW COPY FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 As a minimum, the fire detection instrumentation for eactr 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 1 hour establish a fire watch patrol to inspect the zone (s) w#th the inoperable instrument (s) at least once per hour, unless the i
instrcnent(s) is located inside the containment, then inspect that containment zone at least once per 8 hours or monitor the containment ai. 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 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 +he contdo-ment, 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 smoke detection or flame detection instru-ments 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. Detectors which are not accessible during plant operation shall be demon-strated 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 Each of the above required heat detection instruments shall be demon-strated OPERABLE as follows: .
- a. For nonrestorable spot-type detectors, at least two detectors out of every hundred, or fraction thereof, shall be removed every 5 year::
and functionally tested. For each failure that occurs on the detectors removed, two additional detectors shall be removed and tested; and l CATAWBA - UNIT 1 3/4 3-71
PROOF & REVIEW COPY INSTRUMENTATION FIRE DETECTION INSTRUMENTATION Els SURVEILLANCE REQUIREMENTS (Continued)
- b. For restorable spot-type heat detectors, at least one detector on each signal initiating circuit shall be demonst-ated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Different detectors shall be selected for each 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.
4.3.3.8.3 The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. CATAWBA - UNIT 1 3/4 3-72 0 l
i TABLE 3.3-11 ' h FIRE DETECTION INSTRUMENTS - s. b FIRE MINIMUM INSTRUMENTS OPERABLE
- 5 ZONE DESCRIPTION LOCATION SM0KE FLAME HEAT FUNCTION **
d p 1 R.H.R. Pump IB GG-53 E1.522 + 0 1 0 1 A 2,. 2 R.H.R. Pump 1A FF-53 E1.522 + 0 1 0 1 A 5 3 Cont. Spray Pump IB GG-54 E1.522 + 0 3 0 3 A u 4 Cont. Spray Pump 1A GG-55 E1.522 + 0 2 0 2 A 5 R.H.R. Pump 2B GG-61 E1.522 + 0 1 0 1 A 6 R.H.R. Pump 2A FF-61 E1.522 + 0 1 0 1 A 7 Cont. Spray Pump 28 GG-60 E1.522 + 0 3 0 3 A 8 Cont. Spray Pump 2A GG-59 E1.522 + 0 2 0 2 A 9 Aux. F. W. Pumps BB-51 E1.543 + 0 14 0 12(6) A(B) 10 Mech. Pene. Room JJ-52 E1.543 + 0 3 0 3 A w 11 Corridor / Cables NN-51 E1.543 + 0 6 0 6 A i 12 Recip. Chg. Pump JJ-53 E1.543 + 0 1 0 1 A w 13 Safety Inj Pump IB HH-53 E1.543 + 0 1 0 1 A 4, 14 Safety Inj Pun lA GG-53 E1.543 + 0 1 0 1 A
- 15 Cent. Chg. Pump 1B JJ-54 E1.543 + 0 2 0 2 A 16 Cent. Chg. Pump 1A JJ-55 E1.543 + 0 2 0 2 A 17 Aisles / Cables KK-56 E1.543 + 0 18 0 18 A
! 18 Aisles / Cables EE-55 E1.543 + 0 6 0 6 A 19 AFW Pumps (Unit 2) 88-63 E1.543 + 0 14 0 12(6) p A(B) 21 Aisles / Cables NN-61 E1.543 + 0 6 0 6 A m 22 Recip. Chg. Pump JJ-60 E1.543 + 0 o 23 Safety Inj. Pump 28 1 0 1 A o HH-60 E1.543 + 0 1 0 1 A 24 Safety Inj. Pump 2A GG-60 E1.543 + 0 1 0 1 A h 25 Cent. Chg. Pump 28 JJ-59 E1.543 + 0 2 0 2 A
- 26 27 Cent. Chg. Pump 2A Aisles / Cables JJ-58 KK-59 E1.543 + 0 E1.543 + 0 20 2 0 0
2 20e eA A $ n, , 28 Aisles / Cables EE-58 El.543 + 0 6 0 6 A I g 29 SW Gear Equip. Room AA-50 E1.560 + 0 7 0 0 A O o H 30 Elect. Pene. Room CC-50 E1.560 + 0 8 0 0 A N 31 Corridor / Cables EE-53 El.560 + 0 5 0 5 A
" 32 Corridor / Cables KK-52 E1.560 + 0 8 0 8 A l g 33 Corridor / Cables NN-54 E1.560 + 0 70 0 10 A
TABLE 3.3-11 (Continued) < 9 y FIRE DETECTION INSTRUMENTS 5 b FIRE MINIMUM INSTRUMENTS OPERABLE
- 5 ZONE DESCRIPTION LOCATION SH0KE FLAME HEAT FUNCTION **
d s 34 Aisles / Cables JJ-56 E1.560 + 0 14 0 14 A
> 35 Motor Control Centers GG-56 E1.560 + 0 2 0 2 A 5 36 Cable Trcy Access FF-56 E1.568 + 0 2 0 2 A m 37 Equip. Batteries 00-55 E1.554 + 0 5 0 4 A 38 Equip. Batteries CC-55 E1.554 + 0 5 0 4 A 39 Battery Room CC-56 E1.554 + 0 17 0 0 A 41 SW Gear Equip. Room AA-64 E1.560 + 0 7 0 0 A 42 Elect. Pene. Room CC-65 E1.560 + 0 8 0 0 A 43 Corridor / Cables FF-61 E1.560 + 0 5 0 5 A 45 Aisles / Cables NN-60 E1.560 + 0 13 0 13 A w 46 Aisles / Cables HH-59 E1.560 + 0 13 0 13 A -
1 47 Motor Control Center GG-58 E1.560 + 0 2 0 2 A w 48 Cable Tray Access FF-58 E1.560 + 0 2 0 2 A i 4 49 Equip. Batteries 00-60 E1.560 + 0 5 0 4 A
- 50 Equip. Batteries CC-60 E1.560 + 0 5 0 4 A 51 Battery Room CC-59 E1.560 + 0 17 0 0 A 53 SW Gear Equip. Room AA-49 E1.577 + 0 7 0 0 A 54 Aisles / Cables CC-50 E1.577 + 0 10 0 0 A :3 55 Aisles / Cables E1.577 + 0 56 Aisles / Cables NN-52 PP-55 E1.577 + 0 13 9
0 0 9 13 A A 8 m 57 Aisles / Cables LL-55 E1.577 + 0 1L 0 11 A h 58 Aisles / Cables HH-55 E1.577 v u u 0 21 A = 59 Motor Control Center EE-54 E1.577 + 0 2 0 2 A Q 60 Cable Room CC-56 E1.574 + 0 18 0 15 A Ri 62 SW Gear Equip. Room AA-64 E1.577 + 0 7 0 0 A E 63 64 Elect. Pene. Room Aisles / Cables CC-64 PP-62 E1.577 + 0 E1.577 + 0 10 9 0 0 0, 9 c A g A o 65 Aisles / Cables PP-59 E1.577 + 0 16 0 16 A " 66 Aisles / Cables LL-59 E1.577 + 0 11 0 11 A o 67 Aisles / Cables HH-59 E1.577 + 0 21 0 21 A S 68 Motor Control Center FF-60 E1.577 + 0 2 0 2 A 69 Cable Room CC-59 E1.577 + 0 18 0 15 A w w
TABLE 3.3-11 (Continued) f
# FIRE DETECTION INSTRUMENTS 5
e c: FIRE MINIMUM INSTRUMENTS OPERABLE
- ZONE DESCRIPTION LOCATION SM0KE FLAME HEAT FUNCTION **
sa 71 Elect Pene. Room CC-51 E1.594 + 0 10 0 0 A
> 72 Control Room CC-56 E1.594 + 0 23 0 6 A
$ 73 Vent. Equip. Room FF-56 E1.594 + 0 9 0 0 A m 74 Aisles / Cables LL-56 E1.594 + 0 25 0 25 A 76 Aisles / Cables PP-54 E1.594 + 0 15 0 15 A 71 Elect. Pene. Room BB-63 E1.594 + 0 11 0 0 A 8 61 Control Room 88-59 E1.594 + 0 22 0 6 A 81 Ven. Equip. Room FF-58 E1.594 + 0 12 0 0 A 82 Aisles / Cables KK-58 E1.594 + 0 28 0 ' 27 A 84 Aisles / Cables NN-58 E1.594 + 0 17 0 17 A u, 89 Fuel Pool Area #1 PP-50 E1.605 + 10 19 7 19 A 1 90 Fuel Pool Area (Unit 2) PP-64 E1.605 + 10 19 7 19 A u, 128 UHI Bldg. HH-44 E1.550 + 0 2 3 2 A 4
129 Fuel Pool Purge Room NN-50 E1.631 + 6 6 0 6 A 130 UHI Bldg. (Unit 2) HH-71 E1.594 + 0 2 3 2 A 131 Reactor Bldg. 0*-45* Be1. E1.565 + 3 4 0 0 A 132 Reactor Bldg. 45*-90* Be1. E1.565 + 3 3 0 0 A 133 Reactor Bldg. 90*-135* Be1. E1.565 + 3 4 0 0 A @$ 134 Reactor Bldg. 135*-180 Be1. E1.565 + 3 5 0 0 A M 135 Reactor Bldg. 180*-225* Be1. E1.565 + 3 4 0 0 A 8* 136 Reactor Bldg. 270*-315* Be1. E1.565 + 3 3 0 0 A -t 137 Reactor Bldg. Reactor Bldg. 315*-O' 0*-45* Be1. E1.565 + 3 Be1. E1.586 + 3 8 0 0 A '@ 138 139 Reactor B1dg. 45 -90 Bel. E1.586 + 3 6 4 0 0 0 0 A A 140 Reactor Bldg. 90 -135* Be1. E1.565 + 3 141 Reactor Bldg. 135 -180 Be1. E1.586 + 3 3 8 0 0 0
- 0. r A A 8 T .
142 Reactoe Bldg. 180 -225* Be1. E1.586 + 3 5 0 0 A 143 Reactor Bldg. 315*-0 Be1. E1.586 + 3 5 0 0 A 8 144 Reactor Bldg. 0 -45 Be1. E1.593 + 2 14 0 0 A 145 Reactor Bldg. 45*-90 Be1. E1.593 + 2 17 0 0 A 146 Reactor Bldg. 90-135 Be1. E1.593 + 2 11 0 0 A 147 Reactor B1dg. 135*-180* Be1. E1.593 + 2 ID 0 0 A N, u
TABLE 3.3-11 (Continued) w FIRE DETECTION INSTRUMENTS s2-FIRE MINIMUM INSTRUMENTS OPERABLE
- E ZONE DESCRIPTION LOCATION SM0KE FLAME HEAT FUNCTION **
w 148 Reactor Bldg. 180*-225* Be1. E1.593 + 2\ 2 0 0 A 149 Reactor Bldg. 315*-0* Be1. E1.593 + 2\ 7 0 0 A E 150 Reactor Bldg (Unit 2) 0*-45* Be1. E1.565 + 3 4 0 0 A 151 Reactor Bldg (Unit 2) 45*-90* Be1. E1.565 + 3 3 7 19 A 152 Reactor Bldg (Unit 2) 90 -135* Be1. E1.565 + 3 4 3 2 A 153 Reactor Bldg (Unit 2) 135"-180* Be1. E1.565 + 3 5 0 0 A 154 Reactor Bldg (Unit 2) 180*-225* Be1. E1.565 + 3 3 0 0 A 155 Reactor Bldg (Unit 2) 270 -315* Be1. E1.565 + 3 4 0 0 A 156 Reactor Bldg (Unit 2) 315"-0* Be1. E1.565 + 3 6 0 0 A 157 Reactor Bldg (Unit 2) 0*-45* Be1. E1.586 + 6 6 0 0 A w 158 Reactor Bldg (Unit 2) 45*-90* Be1. E1.586 + 6 4 0 0 A i 159 Reactor Bldg (Unit 2) 90*-135* Be1. E1.586 + 6 3 0 0 A w 160 Reactor Bldg (Unit 2) 135*-180* Be1. E1.586 + 6 8 0 0 A O 161 Reactor Bldg (Unit 2) 180*-225* Be1. E1.586 + 6 5 0 0 A 162 Reactor Bldg (Unit 2) 315*-0* Be1. E1.586 + 6 5 0 0 A 163 Reactor Bldg (Unit 2) 0*-45* Be1. E1.593 + 2 13 0 0 A 164 Reactor Bldg (Unit 2) 45*-90* Be1. E1.593 + 2 17 0 0 A g 165 Reactor Bldg (Unit 2) 90*-135* Be1. E1.593 + 2\ 13 0 0 A 0 o 166 Reactor Bldg (Unit 2) 135*-180* Be1. E1.593 + 2 10 0 0 A 0 167 Reactor Bldg (Unit 2) 180*-225* Be1. E1.593 + 2\ 2 0 0 A n 168 Reactor Bldg (Unit 2) 315*-0* Be1. E1.593 + 2 7 0 0 A 9* 169 RCP-1A Reactor Bldg. E1.593 + 2 0 0 1 A D 170 RCP-1B Reactor Bldg. El.593 + 2\ 0 0 1 A <:- 171 RCP-1C Reactor Bldg. E1.593 + 2 0 0 1 A 172 RCP-ID Reactor Bldg. E1.593 + 2 0 0 1 A 173 RCP 2A 45 Be1. E1.593 + 2 0 0 1 A O O ' 174 RCP 28 135* Be1. E1.593 + 2\ 0 0 1 A u 175 RCP 2C 225* Be1. E1.593 + 2 0 0 1 A " c3 S 176 RCP 20 315* Be1. E1.593 + 2 0 0 1 A 177 Filter Bed Unit IB Reactor Bldg.Be1. E1.565 + 3 2 0 2 A m 178 Filter Bed Unit IA Reactor Bldg.Be1. E1.565 + 3 2 0 2 A g 179 Filter Bed Unit 2A Reactor Bldg.Be1. E1.565 + 3 2 0 2 A . OI l
TARLE 3.3-11 (Continued) FIRE DETECTION INSTRUMENTS s>
' MINIMUM INSTRUMENTS OPERABLE
- i FIRE E ZONE DESCRIPTION LOCATION SM0KE FLAME HEAT FUNCTION **
Z 180 Filter Bed Unit 2B Reactor Bldg.Be1. E1.565 + 3 2 0 2 A
" Annulus E1.561 + 0 0 0 1 A 181a R 181b Annulus E1.583 + 0 0 0 1 A 8 181c Annulus E1.604 + 0 0 0 1 A
" E1.629 + 5 0 0 1 181d Annulus
- A 181e Annulus E1.649 + 5 0 0 1 A
, 181f Annulus E1.664 + 0 0 0 1 A 182a Annulus E1.561 + 0 0 0 1 A J 182b Annulus E1.583 + 0 0 0 1 A I 182c Annulus E1.604 + 0 0 0 1 A
- u. 182d Annulus E1.629 + 5 0 0 1 A i 182e Annulus E1.649 + 5 0 0 1 A
- u. 182f Annulus El.664 + 0 0 0 1 A
! 4 212 Aisles / Cables GG-57 E1.522 + 0 2 0 2 A l 213 Aux. Batt. Room AA-55 E1.554 + 0 4 0 4 A 214 Aux. Cont. Pwr. Batt. AA-59 E1.560 + 0 4 0 4 A i 215 D/G Corridor 88-45 E1.556 + 0 3 0 3 A i 216 D/G Corridor AA-45 E1.556 + 0 2 0 2 A 217 D/G Corridor (Unit 2) 3 0 3 A ::o 218 D/G Corridor (Unit 2) CC-71),E1.560+0 BB-71 E1.560 + 0 2 0 2 A @ 219 Mech. Peno. Room HH-52 E1.577 + 0 6 0 6 A m 220 Mech. Pene. Room JJ-62 E1.577 + 0 6 0 6 A Ro 222 Airlock Access JJ-51 E1.605 + 10 1 0 1 A mo 224 Airlock Access (Unit 2) JJ-63 E1.605 + 10 1 0 1 e A Q
~
225 RN Pump Structure West Section E1.600 + 0 8 0 8 A 226 RN Pump Structure East Section E1.600 + 0 8 0 8 A - i o 231 Reactor Bldg. (Unit 1) 260*-303* E1.668 + 10 10 0 0 A o D M 232 Reactor Bldg. (Unit 2) 260"-303* Be1. E1.668 + 10 10 0 0 A o FF-53 E1.543 + 0 1(Duct) 0 0 A -< 184 HVAC Duct for Rooms 331 and 332 w 185 HVAC Duct for Rooms 203, 205 M-60 E1.543 + 0 1(Duct) 0 0 A g 205A, 206A, 206B, 207 and j
- u. 209A .
i 4 y l
TABLE 3.3-11 (Continued) \ n D FIRE DETECTION INSTRUMENTS s> '
' MINIMUM INSTRUMENTS OPERABLE
- FIRE E LOCATION SM0KE FLAME HEAT FUNCTION **
ZONE DESCRIPTION
-e
- HVAC Duct for Rooms 301, NN-60 El. 560 + 0 1(Duct) 0 0 A 186
" 302, 305, and 307 E RFIA Diesel Generator 1A EE-41 E1. 556 + 0 0 0 0(10) A(B)
RF1B Diesel Generator 18 AA-41 E1. 556 + 0 0 0 0(10) A(B)
" EE-72 E1. 556 + 0 0 0 0(10) A(B)
RF2A Diesel Generator 2A RF2B Diesel Generator 28 AA-72 El. 556 + 0 0 0 0(10) A(8)
*The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate tests.
** Function A: Early warning fire detection and notification only.
Function B: Actuation of fire suppression system and early warning and notification. { Y ' 5 . a U
,8 9e m
, 4 a-E
- i. .
g
- ~ 4 w
1
INSTRUMENTATION PRO
~
LOOSE-PART DETECTION SYSTEM W COPY LIMITING CONDITION FOR OPERATION 3.3.3.9 The Loose-Part Detection System shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION:
- a. With one or more Loose-Part Detection System channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to 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 Systems 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 i
- c. A CHANNEL CALIBRATION at least once per 18 months.
l i l I l l CATAWBA - UNITS 1 AND 2 3/4 3-79 0 01 7 BBS
l INSTRUMENTATION PROOF & REVIEW C RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITTON 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 OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. -
-ACTION: .
- a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint lest conservative than required by the above specification, immediately suspend the releaae of radioactive liquid effluents monitored by the affected channel, or declare the channel .
inoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. 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, and 3.0.4. are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATICN, and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-8. OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-80
TABLE 3.3-12 ' 9
$ RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION i 5
' MINIMUM CHANNELS E OPERABLE ACTION 4
INSTRUMENT w R 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release
- a. Waste Liquid Discharge Monitor (Low Range - EMF-49) I per station 40
- b. Turbine Building Sump Monitor (Low Range - EMF-31) 1 42
- c. Steam Generator Water Sample Monitor (Low Range - EMF-34) 1 43 R
- 2. Continuous Composite Samplers and Sampler Flow Monitor Conventional Waste Water Treatment Line 1 per station 42 T
$ 3. Flow Rate Measurement Devices
- a. Waste Liquid Effluent Line 1 per station 41
- b. Conventional Waste Water Treatment Line 1 per station 41
- c. Low Pressure Service Water Minimum Flow Interlock 1 per station 41 R
co M
~
s a k
- a 8
e l 4 . t i
1 i TABLE 3.3-12 (Continued) PROOF & R&gW COPY ACTION STATEMENTS l ACTION 40 - With the number of channels OPERABLE less than required u by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that p.rior 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: '
- 1. The discharge line valving, and
- 2. The manual portion of the computer input for the release.
rate calculations performed on the computer, or the entire release rate calculations if such calculations are performed manually. Otherwise, suspend release of radioactive effluents via this pathway. ACTION 41 - 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 curves generated in place may be used to estimate flow. ACTION 42 - 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 radioactivity at a lower limit of detection of no . more than 10 7 microcurie /ml:
- a. At least once per 12 hours when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
ACTION 43 - With the number of channels OPERABLE less than required by the
- Minimum Channels OPERABLE requirement, gaseous effluent releases via the atmospheric vent valves (off normal mode) may continue pro-
- vided grab samples of steam generator water are analyzed for radio-activity for up to s0 days at a lower limit of detection of no more than 10 7 microcurie /ml
, a. At least once per 12 hours when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOM EQUIVALENT I-131.
! OCT 7 1985 l CATAWBA - UNITS 1 AND 2 3/4 3-82
TABLE 4.3-8 S g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
' ANALOG CHANNEL E CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST ]
- 1. Radioactivity Monitors Providing R Alarm and Automatic Termination o of Release m
- a. Waste Liquid Discharge Monitor (Low Range - 0 P R(2) Q(1)
EMF-49)
- b. Turbine Building Sump Monitor (Low Range - D 11 R(2) Q(1)
EMF-31) R c. Steam Generator Water Sample Monitor D M R(2) Q(1)
* (Low Range - EMF-34)
- 2. Continuous Composite Samplers and Sampler Flow Monitor Conventional Waste Water Treatment Line D(3) N.A. R N.A.
, 3. Flow Rate Measurement Devices
- a. Waste Liquid Effluent Line D(3) N.A. R N.A.
- b. Conventional Waste Water Treatment Line D(3) N.A. R N.A. O
- c. Low Pressure Service Water Minimum Flow D(3) N.A. R Q O
' Interlock ' '
[- B
< w w
- O O
o M
TABLE 4.3-8 (Cont:nued) PROOF & rey /EW CO TABLE NOTATIONS (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation' occur if any of the following conditions exist: ,
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure (Alarm only), or
- c. Instrument indicates a downscale failure (Alarm only).
(2) The initial CHANNNEL CALIBRATION shall be performed using one or more of
... the reference standards certified by the National Bureau of Standards (NQS)
' or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(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 OCT 719B6 CATAWBA - UNITS 1 AND 2 3/4 3-84
1 INSTRUMENTATION PROOF & W W C RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ! LIMITING CGNDITION 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 00CM. 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 ti.e ACTION, or explain in the 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 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 operations at the frequencies shown in Table 4.3-9. i I 001 7 wSS CATAWBA - UNITS 1 AND 2 3/4 3-85 1
TABLE 3.3-13 n
$ RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION 5
MINIMUM CHANNELS E INSTRUMENT OPERABLE APPLICABILITY ACTION Q [ 1. WASTE GAS HOLDUP SYSTEM 3E a. Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release (Low Range - EMF-50) 1 per station
- 45
- b. Effluent System Flow Rate Measuring Device 1 per station
- 46
- 2. WA.C' " GAS HOLDUP SYSTEM Explosive Gas
{ Mr , ing System
- a. .lydrogen Monitors 1/ train per station **
51
- b. Oxygen Monitors 2/ train per station **
49
- 3. Condenser Evacuation System Noble Gas 1 1,2,3,4 47 Activity Monitor (Low Range - EMF-33)
- 4. Vent System
- a. Noble Gas Activity Monitor (Low Range - EMF-36) 1
- 47 87 Ro
- b. , Iodine Sampler (EMF-37) 1
- 50e m i
g .
- c. Particulate Sampler (EMF-35) 53 1
- 50 g
.a d. Flow Rate Monitor 1
- 46 g tz N
& e. Sampler Flow Rate Monitor 1 ,
46 " i
a TABLE 3.3-13 (Continued) f 9 g RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 5 MINIMUM CHANNELS 3 INSTRUMENT OPERABLE APPLICABILITY ACTION Z
- 5. Containment Purge System y Noble Gas Activity Monitor - Providing '
o Alarm and Automatic Termination of Release 1 *** 48 N (Low Range - EMF-39) '
- 6. Containment Air Release and Addition System Noble Gas Activity Monitor - Providing 1 ,
45 Alarm (Low Range - EMF-39) R. Y 5 8n 9e 4
- . e e h ,
i W 0 @
- x Vo
TABLE 3.3-13 (Continued) PROOF & REVIEW COPY TABLE NOTATIONS At all times except when the isolation valve is closed and, locked. During WASTE GAS HOLDUP SYSTEM operation.
~
*** At all times.
ACTION STATEMENTS ACTION 45 - 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 either:
- a. Vent system noble gas activity monitor providing -
alarm and automatic termination of release (Low Range - EMF-36) has at least one channel OPERABLE, or
- b. At least two independent samples of the tank's contents are analyzed, and at least two technically qualified members of the facility staff independently verify:
- 1. The discharge valve lineup, and
- 2. The manual portion of the computer input for the release rate calculations performed on the computer, or the entire release rate calcula-tions if such calculations are performed manually.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. ACTION 47 - 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 48 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 3-88
i TABLE 3.3-13 (Continued) PROOF & REVIEW CO TABLE NOTATIONS ACTION 49 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, ' operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours. With both channels inoperable, operation 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 50 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days
. '~ provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.
ACTION 51 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner. l l CATAWBA - UNITS 1 AND 2 3/4 3-89 OCT 7 1985
TABLE 4.3-9 f
$ RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
ANALOG CHANNEL MODES FOR WHICH E CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED
]
- 1. WASTE GAS HOLDUP SYSTEM 1
- a. Noble Gas Activity Monitor -
Providing Alarm and Automatic Termination of Release P P R(3) Q(1) (Low Range - EMF-50)
- b. Effluent System Flow Rate P N.A. R N.A.
- Measuring Device R
- 2. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System Y
8 a. Hydrogen Monitors D N.A. Q(4) M **
- b. Oxygen Monitors N.A. M **
D Q(5)
- 3. Condenser Evacuation System Noble Gas Activity Monitor D M R(3) N.A. 1, 2, 3, 4 (Low Range - EMF-33) g D
- 4. Vent System o n
- a. Nob *le Gas Activity Monitor ' *' 8*
D M R(3) Q(2) (Low Range - EMF-36) 2 C E b. Iodine Sampler (EMF-37) W N.A. N.A. N.A. a h rn
~ , 8
$ 4
TABLE 4.3-9 (Continued) ,, S s RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
!x5
" ANALOG
' CHANNEL MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE E CHECK CALIBRATION TEST IS REQUIRED CHECK Zw INSTRUMENT
- 4. Vent System (Continued)
E N.A. N.A. N.A. *
- c. Particulate Sampler (EMF-35) W
- d. Flow Rate Monitor D N.A. R N.A.
- e. Sampler Flow Rate Monitor 0 N.A. R N.A.
- 5. Containment Purge System R Noble Gas Activity Monitor -
- Providing Alarm and Automatic ***
Y Tarmination of Release D P R(3) Q(1)
$ (Low Range - EMF-39)
- 6. Containment Air Release and Addition System Noble Gas Activity Monitor- D P R(3) Q(1)
Providing Alarm (Low Dange - EMF-39) 8 m
* ' , ,. Go N
O h m a W o . 8
=
N N w : __m
TABLE 4.3-9 (Continued) l TABLE NOTATIONS At all times except when the isolation valve is closed and lo,cked. During WASTE GAS HOLDUP SYSTEM operation.
*** At all times. .
(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure (Alarm only), or
~
~ c. Instrument indicates a downscale failure (Alarm only).
(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annuciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using stardards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) The CHANNEL CALIBRATION shall include the use of standard gas samples in accordance with the manufacturer's recommendations. In addition, a standard gas sample of nominal four volume percent hydrogen, balance nitrogen, shall be used in the calibration to check linearity of the hydrogen analyzer. (5) The CHANNEL CALIBRATION shall include the use of standard gas samples in accordance with the manufacturer's recommendations. In addition, a standard gas sample of nominal four percent oxygen, balance nitrogen, shall be used in the calibration to check linearity of the oxygen analyzer. l - OCT 7 eos CATAWBA - UNITS 1 AND 2 3/4 3-92
INSTRUMENTATION F& g&lEW COPY 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 control valve per high pressure turbine steam line inoperable and/or with one intermediate stop valve or one intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam line(s) 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 while in MODE 1 and while in MODE 2 with the turbine operating, 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) Four high pressure turbine control valves,
- 3) Six low pressure turbine intermediate stop valves, and
- 4) Six low pressure turbine intercept valves.
- b. At least once per 31 days while in MODE 1 and while in MODE 2 with the turbine operating, by direct observation of the movement of each of the above valves through one complete cycle from the running position,
- c. At least once per 18 months by performance of a CHANNEL GALIBRATION on the Turbine Overspeed Protection Systems, and .
i d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or corrosion. OCT 7 isos CATAWBA - UNITS 1 AND 2 3/4 3-93 i
t 3/4.4 REACTOR COOLANT SYSTEM . 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION P400rg R&ty y CO 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 STANDBY 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. i
*See Special Test Exceptions Specification 3.10.4.
Y** 3g( 1 B CATAWBA - UNITS 1 AND 2 3/4 4-1 g7 7 1985
REACTOR COOLANT SYSTEM bS R$/Sy COPY 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 reacto'r 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 tne next 12 hours,
- b. With only one reactor coolant loop in operation and the reactor trip breakers in the closed position, restore at least two loops to opera-tion 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 baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loops 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 water level to be greater than or equal to 12% at least once 1 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 satucation temperature. 1 CATAWBA - UNITS 1 AND 2 3/4 4-2 l
\
l PROOF & REVIEW COPY 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 8 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. Residual Heat Removal Loop A, and
- f. Residual Heat Removal Loop B.
APPLICABILITY: MODE 4. ACTION:
- a. With less than the above required reactor coolant and/or residual heat removal loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible;
- if the remaining OPERABLE loop is a residual heat removal loop, be in COLD SHUTDOWN within 24 hours.
- b. With no reactor coolant or residual heat removal 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 residual heat removal 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. **A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 285 F unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures.
g 13B5 CATAWBA - UNITS 1 AND 2 3/4 4-3 M 1 es6 g
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), 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 water level to be greater than or equal to 12% at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or residual heat removal loop shall be verified in operation and circulating reactor coolant at least once per 12 hours. 4 j e i l I CATAWBA - UNITS 1 AND 2 3/4 4-4 1985 . 00T
PRG0F & REVIEW COPY REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED 1 LIMITING CONDITION FOR OPERATION ' 3.4.1.4.1 At least one residual heat removal loop shall be OPERABLE and in operation *, and either:
- a. One additional residual heat removal loop shall be OPERABLE #, or
- b. The secondary side water level of at least two steam generators shall be greater than 12%.
APPLICABILITY: MODE 5 with reactor coolant loops filled ##. ACTION: , ,
- a. With one of the residual heat removal loops inoperable and with less
' than the required steam generator level, immediately initiate correc-tive action to return the inoperable residual heat removal loop to OPERABLE status or restore the required steam generator level as soon as possible.
- b. With no residual heat removal 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 residual heat removal loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one residual heat removal loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. t
*The residual heat removal pump may be deenergized for up to 1 hour provided: .
(1) no operations are permitted that xould cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is main-tained at least 10*F below saturation temperature. i i
# One residual heat removal loop may be inoperable for up to 2 hours for surveillance testing provided the other residual heat removal loop is OPERABLE and in operation.
" A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 285*F unless the secondary water tempercture of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.
CATAWBA - UNITS 1 AND 2 3/4 4-5 Jt ,T 1 8 1985 7 ed5
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION __ 3.4.1.4.2 Two residual heat removal loops shall be OPERABLE # and at least one residual heat removal loop shall be in operation.* APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:
- a. With less than the above required residual heat removal loops OPERABLE, immediately initiate corrective action to return tha required residual heat removal loops to OPERABLE status as soon as possible.
- b. With no residual heat removal 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 residual heat removal loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one residual heat removal loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
# One residual heat removal loop may be inoperable for up to 2 hours for surveillance testing provided the other residual heat removal loop is OPERABLE and in operation.
- The residual heat removal 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 main-tained at least 10*F below saturation temperature. g hbh CATAWBA - UNITS 1 AND 2 3/4 4-6 1 BBS
REACTOR COOLANT SYSTEM F & R&lW COPY 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code cafety valve shall be OPERABLE with a lift setting of 2485 psig i 1%.* APPLICABILITY: MODES 4 and 5. ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE resi-dusi heat removal loop into operation in the shutdown cooling mode. SURVEILLANCE REQUIREMENTS 1
- 4.4.2.1 No additional requirements other than those required by Specification 4.0.5.
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
JUL 1 001 7 985 CATAWBA - UNITS 1 AND 2 3/4 4-7
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM 1 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 SHUTOOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5.
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
1 i g BBS CATAWBA - UNITS 1 AND 2 3/4 4-9 'l 800
J l l PROOF & REVIEW COP REACTOR COOLANT SYSTEM j 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION j j 3.4.3 The pressurizer shall be OPLRABLE with a water volume of less than or equal to 1656 cubic feet and at least two groups of pressurizer heaters each ! having a capacity of at least 150 kW. 4 APPLICABILITY: MODES 1, 2, and 3. ACTION: 1
- a. With one group of pressurizer heaters inoperable, restore at least i two groups to OPERABLE status within 72 hours or be in at least HOT ,
STANOBY 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 STAND 8Y with the Reactor trip breakers open within 6 hours and in HOT SHUT-DOWN within the following 6 hours.
l ) SURVEILLANCE REQUIREMENTS i i 4.4.3.1 The pressurizer water volume 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 l heaters shall be verified by energizing the heaters and measuring circuit ' j current at least once per 92 days. 4.4.3.3 The emergency power supply for the pressurizer heaters shall be , demonstrated CPERABLE at least once per 18 months by manually transferring
- power from the normal to the emergency power supply and energizing the i heaters.
i i M g$ CATAWBA - UNITS 1 AND 2 3/4 4-9 00T
.._,.____.-__m__,..__-_.-_._...-_,.___.-.m.__._.
PROOF & REVIEW CO' 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 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 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
l
- c. With more than one PORV(s) inoperable due to causes other than exces-sive seat leakage, within 1 hour either restore 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 SHUTOOWN 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 the ACTION b. or c. above, as appropriate, for the isolated PORV(s).
- e. The provisions of Specification 3.0.4 are not applicable.
g 1B I l CATAWBA - UNITS 1 AND 2 3/4 4-10 QCT t l
00f & R&lEW Copy REACTOR COOLANT SYSTEM 50RVEILLANCE 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:
- a. Performance of a CHANNEL CALIBRATION, and
- b. Operating the valve through one complete cycle of full travel.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4. 4.4.4.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by:
- a. Manually transferring motive and control power from the normal to the emergency power supply, and
- b. Operating the valves through a complete cycle of full travel.
l g \366 CATAWBA - UNITS 1 AND 2 3/4 4-11 1 7N
PROOF g.pgyggy ggpy REACTOR COOLANT SYSTEM n 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 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. 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, ir.spection 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 Sp Q cation 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
- a. W ere 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; t, . The first sample of tubes selected for each inservice inspection ' (subsequent to the preservice inspection) of each steam generator shall include: 1 2n g\ q s85 CATAWBA - UNITS 1 AND 2 3/4 4-12 GCI
PROOFg gEVIEW COPY REACTOR COOLANT SYSTEM . DRAFT SURVEILLANCE REQUIREMENTS (Continued) -
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
- 2) Tubec 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 1 C-1 Less than 5% of the .otal 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. I
%Yo$
O t q BBS CATAWBA - UNITS 1 AND 2 3/4 4-13 ~ 0 01
1
\
REACTOR COOLANT SYSTEM PROOF & gg IEW COPY 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 preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
- b. If the 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 not less than 12 or more than 24 months. The increase in inspection fre-quency shall apply until the subsequent inspections satisfy the cri-teria of 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 leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or
- 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.
OCT 7B0 CATAWBA - UNITS 1 AND 2 3/4 4-14 i
REACTOR COOLANT SYSTEM 00F & REVIEW COPY 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; 1
- 2) Degradation means a service-induced cracking, wastage, wear or j general corrosion occurring on either inside or outside of a tube; l
, 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 dffected 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 or repaired by an accept-able method such as sleeving. It is equal to 40% of the nominal tube wall thickness; Unserviceable describes the condition of a tube if it leaks or 7) 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 4.4.5.3c., above;
- 8) Tube Inspection means an inspection of the steam generator tube 1 from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and 4
)
q BBS OCT CATAWBA - UNITS 1 AND 2 3/4 4-15 4
REACTOR COOLANT SYSTEM FS R&lEW COPP SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Inspection means an inspection of the fu'll 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 expected to be used during subsequent inservice inspections.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair 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 o.f 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 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 per:etration for each indication of an imperfection, and
- 3) Identification of tubes plugged or repaired.
- 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 plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
OCT 7# , CATAWBA - UNITS 1 AND 2 3/4 4-16 l
_ . - _ . ~ _ - . - _ _ _- . _ _ - - . I ! Table 4.4-1 i ! n MINIMUM NUMBER OF STEAM GENERATORS TO BE h INSPECTED DURING INSERVICE INSPECTION 5 i e E Preservice Inspection No Yes
'a No. of Steam Generators per Unit Four Four ca I
g First Inservice Inspection All Two j N Second & Subsequent Inservice Inspections
~
Onel One2 i 1 ! TABLE NOTATIONS j 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
! w 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 4 A generators may be found to be more severe than those in other steam generators. ! j
" Under such circumstances the sample sequence shall be modified to inspect the ; most severe conditions. ; 2. Each of the other two steam generators not inspected during the first inservice l inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.
i _t'
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3
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- O
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- g o>
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Tab,. a.<-2 STEAM GENERATOR TUBE INSPECTION PRoop g REVIEW COPY IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION 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. 5 Tubes per S.G. C-2 Plug
- defective tubes C-1 None N.A. N.A.
and inspect additional 25 tubes in this S.G. C-2 Pluga defective tubes C-1 Mone and inspect additional 45 tubes in this S. G. C-2 Plug
- defective tubes Perfore action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N.A. N.A.
sample
- C-3 Inspect all tubes in All other this S.G., pluga de- 5.G.s are None N.A. N.A.
fective tubes and C-1 " inspect 25 tubes in each other S. G. Some S.G.s C-2 but no Perform action for M.A. N.A. additional C-2 result of second Notification to NRC 5.G. are sample pursuant to $50.72 C-3 (b)(2) of 10 CFR Part 50 Additional Inspect all tubes in S.G. is C-3 each S.G. and plug
- defective tubes.
Notification to NRC N.A. N.A. pursuant to $50.72 (b)(2) of 10 CFR 1 Part 50 S=3N g Where N is the number of steam generators in the unit, and n is the number of steam generators inspected n during an inspection ) "Where plugging is indicated, sleeving or other approved repair methods are also acceptable. I CATAWBA - UNITS 1 AND 2 3/4 4-18 OGI
PRoogg REACTOR COOLANT SYSTEM N COPY 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION I 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- a. The Containment Atmosphere Gaseous Radioactivity Monitoring System,
- b. The Containment Floor and Equipment Sump Level and Flow Monitoring Subsystem, and c.
Either the Containment Ventilation Unit Condensate Drain Tank Level Monitoring Subsystem or the Containment Atmosphere Particulate Radioactivity Monitoring System. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANOBY within the next 6 hours and in COLD
! SHUTDOWN within the following 30 hours.
l l 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 Floor and Equipment Sump Level and Flow Monitoring Subsystem-performance of CHANNEL CALIBRATION at least once per 18 months, and
- c. Containment Ventilation Unit Condensate Drain Tank Level Monitoring i
Subsystem performance of CHANNEL CALIBRATION at least once per 18 months. i l gg 1N l CATAWBA - UNITS 1 AND 2 3/4 4-19 l -- . - - _ . _ - - _ _ . - - . _ _- ___ . - - - - _ _ -
PROOF & REVIEW REACTOR COOLANT SYSTEM l l OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION J 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE, q b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. 1 gpm total reactor-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig, and i
- f. I gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig i 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 BOUNDARY LEAKAGE, be in at least h0T STANOBY 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 BOUNDARY 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, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by i use of at least two closed manual or deactivated automatic valves,
- or be in at least HOT STANDBY within the next 6 hours and in COLD .
SHUTDOWN within the following 30 hours. i 1Q i j CATAWBA - UNITS 1 AND 2 3/4 4-20 Jb uvi 165 l 1 _ . _ _ . - - - __--
PROOF & REVIEW COPY 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 radioactivity monitor at least once per 12 hours;
- b. Monitoring the containment floor and equipment sumps inventory and discharge at least once per 12 hours;
- c. Measurement of the CONTROLLED LEAXAGE to 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 i
- e. Monitoring the reactor head flange leakoff 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, I
- b. Prior to entering MODE 2 whenever the plant 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.
The provisions of Specification 4.0.1 are not applicable for entry into MODE 3 or 4. 1 i i
\M
$b q\M CATAWBA - UNITS 1 AND 2 3/4 4-21 0 01
-- . ~ - . . ,- .. . -- . --- - . . - - - - - - - -- - - _ _ _ _ , _ _ __ _
i l PRoop& R&IEW CD TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVy VALVE NUMBER FUNCTION
~
N159 Accumulator Discharge NI60 Accumulator Discharge NI70 Accumulator Discharge NI71 Accumulator Discharge NI81 Accumualtor Discharge NI82 Accumulator Discharge , NI93 Accumulator Discharge i NI94 Accumulator Discharge
.,NI124 Safety Injection (Hot Leg) .
NI125 Residual Heat Removal (Hot Leg) NI126 Safety Injection (Hot Leg) NI128 Safety Injection (Hot Leg) NI129 Residual Heat Removal (Hot Leg) NI134 Safety Injection (Hot Leg) i NI156 Safety Injection (Hot Leg) NI157 Safety Injection (Hot Leg) NI159 Safety Injection (Hot Leg)
; NI160 Safety Injection (Hot Leg) i NI165 Safety Injection / Residual Heat Removal (Cold Leg)
NI167 Safety Injection / Residual Heat Removal (Cold Leg) i NI169 Safety Injection / Residual Heat Removal (Cold Leg) NI171 Safety Injection / Residual Heat Removal (Cold Leg) NI175 Safety Injection / Residual Heat Removal (Cold Leg)
; NI176 Safety Injection / Residual Heat Removal (Cold Leg)
NI180 Safety Injection / Residual Heat Removal (Cold Leg) NI181 Safety Injection / Residual Heat Removal (Cold Leg) i NI248 Upper Head Injection i NI249 Upper Head Injection
- NI250 Upper Head Injection l l NI251 Upper Head Injection l NI252 Upper Head Injection i
NI253 Upper Head Injection J } 1 CATAWBA - UNITS 1 AND 2 3/4 4-22
1 PRoop OV Copy 1
, TABLE 3.4-1 (Continued) l REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES i VALVE NUMBER FUNCTION
, ND1Ba RHR ND2A* RHR ND368*l ND37A* RHR RHR l
.* Testing per Specification 4.4.6.2.2d. not applicable due to positive indicatton i of valve position in Control Room.
- 1. Leakage rates less t.'han or equal to 1.0 gpm are considered acceptable.
- 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded j
! the rate determined by the previous test by an amount that reduces '
the margin between measured leakage rate and the maximum permissible
; rate of 5.0 gpm by 50% or greater.
4
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm
]
are considered unacceptable if the latest measured rate exceeded the
. rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate j of 5.0 gpm by 50% or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
i l i i. q ess ocT 1 i CATAWBA - UNITS 1 AND 2 3/4 4-23
PROOF & REVIEW REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2. APPLICABILITY: At all times. ACTION: MODES 1, 2, 3, and 4:
- a. With any one or more chemistry 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; and
- 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 SHUTOOWN 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 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 , l 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 JUL 181985 CATNdBA - UNITS 1 AND 2 3/4 4-24 OCT 7 1985
TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen
- 1 0.10 ppm i 1.00 ppm Chloride 1 0.15 ppm i 1.50 ppm Fluoride 1 0.15 ppm i 1.50 ppm
- Limit not applicable with T,yg less than or equal to 250*F.
1 l i 1 CATAWBA - UNITS 1 AND 2 3/4 4-25 JUL 181985 l OCT 7 1985 j J
t PROOF g.RgyIEW CO TABLE 4.4-3 REACTOR COOLANT SYSTEM r CHEMISTRY LIMITS 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 l
*Not required with T,yg less than or equal to 250*F i
l [ E jut 181985 < CATAWBA - UNITS 1 AND 2 3/4 4-26 00T 7 bo5 L. - _ ___ . _ _ _ _ _. _ - - - . _ - _ _ . _ . _ . . . _ . _ . . . _ . _ _ _ _ _ _ _ _ _ . _ _ _ . - _ . . ~ . _ . _ . - _
REACTOR COOLANT SYSTEM W COPY 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. i 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 Allow-i able 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 cumulative operating time under these circumstances does not exceed 800 hours in any consecutive 12-month period. The provisions of 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 EQUIV-ALENT I-131 exceeding 500 hours in any consecutive 6-month period, prepa're 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 coolant 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 T less than 500'F within 6 hours; and avg .
- d. With,the gross specific activity of the reactor coolant greater than 100/E microCuries per gram of gross radioactivity, be in at least HOT STANDBY with T,yg less than 500*F within 6 hours.
- "With T,yg greater than or equal to 500 F.
CATAWBA - UNITS 1 AND 2 3/4 4-27 JUL 181985 0CT 7 1985 l .
i l PROOfg pgy REACTOR COOLANT SYSTEM - 8 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 1004 micro-curies per gram of gross radioactivity, perform the sampling and analy-sis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits. In the next Annual Report, pursuant to Specification 6.9.1.4, submit the results of the specific activity analyses together with the following information:
- a. Reactor power history starting 48 hours prior to the sample in 1 which the limit was exceeded;
- b. Results of: (1) the last isotopic analysis for radioiodines 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 e-d time of sampling and the radiofodine 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 reactor coolant exceeded 1 microcurie per gram DOSE EQUIVALENT I-131.
SURVEILLANCE REQUIREMENTS - I 4.4.8 The specific activity of the reactor coolant shall be determined to be - within the limits by performance of the sampling and analysis program of Table 4.4-4. l JUL1819k CATAWBA - UNITS 1 AND 2 3/4 4-28 OCT 7 1985
I PROOF & REVIEW COPY
. .;[ .t.. t . . -, r
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- OPERATION o .
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o C ; l w w m o j O O -- 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER 4 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS l PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCi/ gram DOSE EQUIVALENT I-131 i CATAWBA - UNITS 1 AND 2 3/4 4-29 JUL I 81985 0CT 7 685 I t
- _. - . -~ . - _ - - - -
n TABLE 4.4-4 D REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE g AND ANALYSIS PROGRAM e TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE g AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED G 1. Gross Radioactivity At least once per 72 hours 1, 2, 3, 4 [ Determination ** l
> 2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days 1 5 LENT I-131 Concentration m
- 3. Radiochemical for 5 Determination *** 1 per 6 months
- 1 I 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 w EQUIVALENT I-131
, 5 or 100/E pCi/ gram of y gross radioactivity, and 8 b) One sample between 2 1,2,3 and 6 hours following J
a THERMAL POWER change exceeding 15% o of the RATED THERMAL m o POWER within a 1-hour o period.
=
k 8N 5 a M m -< k$ D
l l ' PROOF & REVIEW COPY . TABLE 4.4-4 (Continued) TABLE NOTATIONS .
#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.
**A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radiofodines. The total specific activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities in the sample within 2 hours after the sample is taken and extrapolated back to when the sample was taken. Deter-mination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta emitting radionuclides.
***A radiochemical analysis for [ shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in I the reactor coolant. The specific activities fgr these individual radio-nuclides shall be used in the determination of_E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.
1 i ( i l l l CATAWBA - UNITS 1 AND 2 3/4 4-31
Roof g gNI WCopy REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM 3 LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and i 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 hydrostatic 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 i to within the limit within 30 minutes; perform an engineering evaluation to l 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 be in at least HOT STANDBY within the next 6 hours and reduce the T,yg 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 minutes 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 l Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. 1 i i T l CATAWBA - UNITS 1 AND 2 3/4 4-32 t CT 7 85 I
MATERIAL PROPERTY BASIS PROOF & REVIEW C COPPER CONTENT : CONSERVATIVELY ASSUMED TO BE 0.10 WT% (ACTUAL CONTENT : 0.08 WT%) RTNDT INITIAL : CONSERVATIVELY ASSUMED TO BE 400F 0 (ACTUAL RTNDT = 8 F) RTNDT AFTER 16 EFPY: 1/4T,1100F 3/4T,870F CURVE APPLICABLE FOR HEATUP RATES UP TO 60 F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 l I LE A K -,, I f f TEST y- J / LI M,iT -~f [ j C I r 1 i 1 i3 2000 7 f ;,
- 6. j g g E ! I f I i 8 I ( i m i i !
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1000 l, [ j iE !' ! I I i i l II _L_, H E ATUP- si - / %--CRITICALITY LIMIT BASED ON CURVE h/ t INSERVICE HYDROSTATIC TEST
- U,.k '.( i TEMPERATURE (255 0F) FOR THE i
SERVICE PERIOD UP TO 16 EFPY
< 1
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O 100 200 300 400 500 INDICATED TEMPER ATURE (OF) FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY CATAWBA - UNITS 1 AND 2 3/4 4-33 OCT 7 1985
l M ATERIAL PROPERTY BASIS N YlEW COPY COPPER CONTENT : CONSERVATIVELY ASSUMED TO BE 0.10 WT% (ACTUAL CONTENT = 0.08 WT%) RTNOT INITIAL : CONSERVATIVELY ASSUMED TO BE 400F 0 (ACTUAL RTNDT = -8 F) RT NDT AFTER 16 EFPY: 1/4T.1100F 3/4T,87 F 0 CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100 F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 100F AND 60 PSIG FOR POSSIPLE INSTRUMENT ERRORS 3000 .; 3 _____ 1,_-___
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_ -. 2 COOLDOWN - 4 W RATES o 1I Lm & (CF/HR) 20 - SM I--- 40'CIM
- - 1- I 60 - AT 100 ! ! -
l l l 0 l l f~~~~ 0 100 200 300 400 500 INDICATED TEMPER ATURE (OF) FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY CATAWBA - UNITS 1 AND 2 3/4 4-34 OCT 7 1985
. _ = . .
i TABLE 4.4-5 9
$ REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE 5
' LEAD CAPSULE VESSEL E NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY) 1 U 58.5* 4 First Refueling w
E V 61* 3.69 9 i o j " W 121.5* 4 Standby X 238.5* 4 Standby Y 241* 3.69 6 w Z 301.5 4 Standby 1 1 { t u 5 8n u O r= M
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i PR00Fg ggyggy (9py REACTOR COOLANT SYSTEM PRESSURIZER w LIMITING CONDITION FOR OPERATION ! 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup of 100*F in any 1-hour period, and
- b. A maximum cooldown of 200*F in any 1-hour period.
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 STAND 8Y within the next 6 hours and reduce the pressurizer pressure to less than 500 psig j within the following 30 hours. SURVEILLANCE REQUIREMENTS . 4.4.9.2 The pressurizer temperatures shall be determined to be within the l limits at least once per 30 minutes during system heatup or cooldown. i l jut,18 MS CATAWBA - UNITS 1 AND 2 3/4 4-36 00T 7 85
PROOF & REVIEW COPY 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 power operated relief valves (PORVs) with a lift setting of less than or equal to 450 psig, or
- b. The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.
i APPLICABILITY: MODE 4 when the temperature of any Reactnr Coolant System cold leg is less than or equal to 285*F, MODE 5 and MODE 6 with the reactor vessel head on. ACTION:
- a. With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the Reactor Coolant System through at least a 4.5 square inch vent within the next 8 hours.
- b. With both PORVs inoperable, depressurize and vent the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours,
- c. In the event either the PORVs or the Reactor Coolant System vent (s) are used to mitigate a Reactor Coolant System 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 Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable. ,
1 l 1 i l l jul. 1 g \3BS i l CATAWBA - UNITS 1 AND 2 3/4 4-37 g 11365
PR00F & REVIPN COPY REACTOR COOLANT SYSTEM . SURVEILLANCE REQUIREMENTS I 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. 4.4.9.3.2 The Reactor Coolant System vent (s) shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overpressure protection.
1 "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 i least once per 31 days. I 1 Jyt.181985 CATAWBA - UNITS 1 AND 2 3/4 4-38 00T 7 BBS
. . - _ , - _ _ - _ . _ _ _ - . . - _ . - . - - . _ _ _ - _ = - _ . .
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM l 3/4.4.10 STRUCTURAL INTEGRITY t , t 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. 1 i ACTION: { a. l With the structural integrity of any ASME Code Class I component (s)
. not conforming to the above requirements, restore the structural
! integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations. I b. 3 With the structural integrity of any ASME Code Class 2 component (s)
! not conforming to the above requirements, restore the structural i integrity of the affected component (s) to within its limit or isolate i the affected component (s) prior to increasing the Reactor Coolant 4
System temperature above 200 F. 4
- c. I 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 i
the affected component (s) from service. l i d. The provisions of Specification 3.0.4 are not applicable. 3 l SURVEILLANCE REQUIREMENTS i 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 3 Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. [ t i i l E. I g 18 BD CATAWBA - UNITS 1 AND 2 3/4 4-39 OCT 7 1985
REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS f'R00F & Rf3?E1V C LIMITING CONDITION FOR OPERATION
! 3.4.11 At least one Reactor Coolant System vent path consisting of at least i
two valves in series powered from emergency buses shall be OPERABLE and closed
- at each of the following locations:
- a. Reactor Vessel Head
- b. Pressurizer steam space
- APPLICABILITY
- MODES 1, 2, 3 and 4.
ACTION: i
- a. With one of the above Reactor Coolant System vent paths inoper-able, STAPTUP 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 SHUT-DOWN within the following 30 hours.
- b. With both of the above Reactor Coolant System vent paths inoper-able; maintain the inoperable vent paths closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least one of the vent paths to OPER-ABLE status within 72 hours or be in HOT STANOBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
1 i SURVEILLANCE REQUIREMENTS 4.4.11 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:
- 1. Verifying all manual isolation valves in each vent path are
, locked in the open position.
- 2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN i or REFUELING.
i l' *For the plants using power operated relief valve (PORV) as a vent path, PORV i l block is not required to be closed if the PORV is operable. OCT 7 1985 t CATAWBA - UNITS 1 AND 2 3/4 4-40
M00f REVIEW C opy 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION . 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
- a. The discharge isolation valve open with power removed,
- b. A contained borated water volume of between 7853 and 8171 gallons,
- c. A boron concentration of between 1900 and 2100 ppm,
- d. A nitrogen cover pressure of between 385 and 481 psig, and
- e. A water level and pressure channel OPERABLE.
l APPLICABILITY: MODES 1, 2, and 3*. , ACTION:
- a. With one cold leg injection 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 cold leg injection 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.
s SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE: I 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 cold leg injection accumulator isolation j valve is open and power is removed.
- " Pressurizer pressure above 1000 psig.
r j OCT 71985 CATAWBA - UNITS 1 AND 2 3/4 5-1
EMERGENCY CORE COOLING SYSTEMS N00F& REWy COPP 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 75 gallons by verifying the boron concentration of the accumulator solution;
- c. At least once per 31 days when the Reactor Coolant System pressure is above 1000 psig by verifying that power is removed from the isolation valve operators on Valves NI54A, NI65B, NI76A, and NI88B and that the respective circuit breakers are padlocked; and 4.5.1.1.2 Each cold leg injection accumulator water level and pressure channel shall be demonstrated OPERABLE:
- a. At least once per 31 days by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and
- b. At l' east once per 18 months by the performance of a CHANNEL CALIBRATION.
QM 78 CATAWBA - UNITS 1 AND 2 3/4 5-2
PROOF & REvlEW COPY EMERGENCY CORE COOLING SYSTEMS UPPER HEAD INJECTION LIMITING CONDITION FOR OPERATION 3.5.1.2 Each Upper Head Injection Accumulator System shall be OPERABLE with:
- a. The discharge isolation valves open,
- b. A minimum contained borated water volume of 1807 cubic feet,
- c. A boron concentration of between 1900 and 2100 ppm, and
- d. The nitrogen-bearing accumulator pressurized to between 1185 and 1285 psig.
APPLICABILITY: MODES 1, 2, and 3.* ACTION:
- a. With the Upper Head Injection Accumulator System inoperable, except as a result of closed isolation valve (s), restore the Upper Head Injection Accumulator System 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 the Upper Head Injection Accumulator System inoperable due to the isolation valve (s) being closed, either immediately open the isolation valve (s) or be in HOT STANDBY within 6 hours and be in HOT SHUTDOWN within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.5.1.2 Each Upper Head Injection Accumulator System shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1) Verifying the contained borated water level in the surge tank and nitrogen pressure in the accumulators, and
- 2) Verifying that each accumulator discharge isolation valve is open.
- Pressurizer pressure above 1900 psig.
! g 1B N , l 2 CATAWBA - UNITS 1 AND 2 3/4 5-3 OG l l l _ _ ,__ . _ _ __
1 i l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b
- b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 138.3 gallons by verifying the boron concentration of the solution in the water-filled accumulator;
- c. At least once per 18 months by:
- 1) Verifying that each accumulator discharge isolation valve 3 closes automatically when the water level is 93.2 1 2.7 inches (Unit 1) and 93.1 1 2.7 inches (Unit 2) above the working line on the water-filled accumulator, and
- 2) Verifying that the total dissolved nitrogen and air in the -
water-filled accumulator is less than 80 scf per 1800 cubic feet of water (equivalent to 5 x 10 5 pound of nitrogen per pound of water).
, d. At least once per 5 years and if the requirements of Specifica-tion 4.5.1.2c.2) are not met by replacing the membrane installed
, between the water-filled and nitrogen-bearing accumulators. i i 4 i , l 0 01 71985 e l CATAWBA - UNITS 1 AND 2 3/4 5-4
- ,.,.---n7 w ,,-,,,---,---,,,r,-n-~,--s-.c-m-- . ,~ n-,- -----_nnn mw,,--,--r- ------.------r----,..-~,-n-e,- , -r-- ,er, n .
EMERGENCY CORE COOLING SYSTEMS WCOPy 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:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE Safety Injection pump,
- c. One OPERABLE residual heat removal heat exchanger, j
, d. One OPERABLE residual heat removal pump, and I
- 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 phase of operation.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem
' 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.
l'
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to i
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. l l l l JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 5-5 00T 7 1985
.._ ___ ~ _ , . - - . - - _
PRoop : EMERGENCY CORE COOLING SYSTEMS g REvigy(gpI _ SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: s
- 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 NI-162A Cold Leg Recirc. Open NI-121A Hot Leg Recirc. Closed NI-1528 Hot Leg Recirc. Closed NI-183B Hot Leg Recirc. Closed
, ~
NI-173A Residual Heat Open , Removal Pump Disch. NI-1788 Residual Heat Open Removal Pump Disch. NI-1008 Safety Injection Open Pump Suction from Refueling Water Storage Tank NI-147B Safety Injection Open Pump Mini-flow
- 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 establish-ing 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 residual heat removal 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 i CATAWBA - UNITS 1 AND 2 3/4 5-6 00T 7 1985
i EMERGENCY CORE COOLING SYSTEMS YCOpy SURVEILLANCE REQUIREMENTS (Continued) b) With a simulated or actual Reactor Coolant System pressure ! signal less than or equal to 660 psig the interlocks will l cause the valves to automatically close.
- 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:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on Safety Injection and Containment Sump Recirculation test signals, and
- 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and c) Residual heat removal pump.
- f. By verifying that each of the following pumps develops the indicated differential pressure when tested pursuant to Specification 4.0.5:
- 1) Centrifugal charging pump 3 2380 psid,
- 2) Safety Injection pump 1 1430 psid, and
- 3) Residual heat removal pump 1 165 psid.
- g. By verifying the correct position of each electrical and/or mechanical stop for the following ECCS throttle valves:
- 1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and
- 2) At least once per 18 months.
Centrifugal Charging Pump Injection Throttle Safety Injection Throttle Valve Number Valve Number NI-14 NI-164
, NI-16 NI-166 l NI-18 NI-168 l
NI-20 NI-170 CATAWBA - UNITS 1 AND 2 3/4 5-7 QCT 7 1986 l l
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- h. By performing a flow balar,ce 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 333 gpm, and b) The total pump flow rate is less than or equal to 565 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 462 gpm, i and b) The total pump flow rate is less than or equal to 660 gpm.
- 3) For residual heat removal pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3648 gpm.
JUL 181985 JUL 18 boa CATAWBA - UNITS 1 AND 2 3/4 5-8 GCT 7 1985
EMERGENCY CORE COOLING SYSTEMS PROOF & REVIEW COPY 3/4.5.3 ECCS SUBSYSTEMS - T < 350*F avg 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 residual heat removal heat exchanger,
- c. One OPERABLE residual heat removal pump, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recircula' ion phase of operation.
APPLICABILITY: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours,
- b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T less than 350*F by use of alternate heat removal methods."#9
- 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. l i
# A maximum of one centrifugal charging pump and one Safety Injection pump shall be OPERABLE whenever the temperature of one or more of the Reactor Coolant System cold legs is less than or equal to 285*F.
l JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 5-9 00T 7 1985
EMERGENCY CORE COOLING SYSTEMS PROOF & REVISV COPY fTf, SURVEILLANCE REQUIREMENTS J . 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 charging pumps and Safety Injection pumps, except the above required OPERABLE centrifugal charging pump, shall be demonstrated inoperable by verifying that the motor circuit breakers are secured in the open position or the discharge of each pump has been isolated from the Reactor Coolant System by at least two isolation valves with power removed from the valve motor opera-tors at least once per 12 hours whenever the temperature of one or more of the
, Reactor Coolant System cold legs is less than or equal to 285*F.
t i i i CATAWBA - UNITS 1 aid 2 3/4 5-10 00T 7 1985
EMERGENCY CORE COOLING SYSTEMS YCOP[ 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 4 3.5.4 The refueling water storage tank shall be OPERABLE with:
- a. A minimum contained borated water volume of 363,513 gallons,
- b. A boron concentration of between 2000 and 2100 ppm of boron,
- c. A minimum solution temperature of 70*F, and
- d. A maximum solution temperature of 100*F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the refueling water storage tank inoper~able, restore the tank to OPERABLE i status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. i SURVEILLANCE REQUIREMENTS 4.5.4 The refueling water storage tank shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1) Verifying the contained borated wa'.er level in the tank, and
- 2) Verifying the boron concentration of the water.
- b. At least once per 24 hours by verifying the refueling water storage tank temperature when the outside air temperature is less than 70*F or greater than 100*F.
CATAWBA - UNITS 1 AND 2 3/4 5-11 JUL 181S85 00T 7 1985 l
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: H0 DES 1, 2, 3, and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY 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. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 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 ,14.68 psig, and verifying that when the measured leakage rate for th$se 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 to 0.60 L,. n Except valves, blind flanges, and deactivated automatic valves which are located inside the annulus or the containment and are locked, sealed or , otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 6-1 QCT 7 1995
~- . - . . _- _ . .-- . . --
PR00ra pg : CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of:
- 1) Less than or equal to L,, 0.20% by weight of the containment air per 24 hours at P,,14.68 psig, or
- 2) Less than or equal to L , 0.122% by weight of the containment t
air per 24 hours at a reduced pressure of P , 7.34 psig. t
- 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 P,,
and -
- c. A combined bypass leakage rate of less than 0.07 L, for all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P,.
APPLICABILITY: MODES 1, 2, 3, and 4. l l ACTION: I With: (a) the measured overall integrated containment leakage rate exceeding 0.75 L, or 0.75 L t, as applicable, or (b) the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L,, or (c) the combined bypass leakage rate exceeding 0.07 L,, restore the overall integrated leakage rate to less than 0.75 L, or less than 0.75 Lg , as applicable, and the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than 0.60 L,, and the combined bypass leakage rate to less than 0.07 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 l test schedule and shall be determined in conformance with the criteria speci-l fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972 or the mass plot method: l JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-2 GCT 7 Iga 5
REVIEW COPY CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 1 10 month intervals during shutdown at either P,, 14.68 psig, or at P , 7.34 psig, during each 10 year service t
period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection;
-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 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 L tat 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:
- 1) Confirms the accuracy of the test by verifying that the supple-mental test result, L , cminus the sum of the Type A and the superimposed leak, L , is equal to or less than 0.25 L, or ,
0.25 L ; t
- 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and
- 3) Requires that the rate at which gas is injected into the containment or bled from the containment during the supple-mental test is between 0.75 L, and 1.25 L, or 0.75 Lt ""d 1.25 L t
- d. Type B and C tests shall be conducted with gas at a pressure not less than P,, 14.68 psig, at intervals no greater than 24 months except for tests involving: ,
- 1) Air locks,
- 2) Purge supply and exhaust isolation valves with resilient material seals, and
- 3) Dual-ply bellows assemblies on containment penetrations between the containment building and the annulus.
- e. The combined bypass leakage rate shall be determined to be less than by applicable Type B and C tests at least once per 24 months 0.07 exceptL,for penetrations which are not individually testable; penetra-tions not individually testable shall be determined to have no detect-able leakage when tested with soap bubbles while the containment is pressurized to P,, 14.68 psig, or P g , 7.34 psig, during each Type A test; CATAWBA - UNITS 1 AND 2 3/4 6-3 l JR 18 585 l OUT 7125 1_ _ _ _ _- __ . _ .
PROOF & REVIEW CDP'( CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- f. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.9.3 or 4.6.1.9.4, as applicable;
- g. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3;
- h. The space between each dual ply bellows assembly or; containment penetrations between the containment building and the annulus shall be vented to the annulus during Type A tests. At least once per 24 months, the space between each dual ply bellows assembly shall be subjected to a low pressure test at 3 to 5 psig to verify no detect-able leakage or the dual ply bellows assembly shall be subjected to a leak test with the pressure on the containment side of the dual ply bellows assembly at P,, 14.68 psig, to verify the leakage to be within the limits of Specification 4.6.1.2e. ; and
- i. The provisions of Specification 4.0.2 are not applicable.
I JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-4 OCT 7 1935
TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS h PENETRATION TEST
, NUMBER SERVICE RELEASE LOCATION TYPE
$ M216 Pressurizer Relief Tank Makeup Auxiliary Building Type C M212 Nitrogen to Pressurizer Relief Tank Auxiliary Building Type C
$ M327 Reactor Coolant Pump Motor Drain Tank Auxiliary Buf1 ding Type C g Pump Discharge M259 Reactor Makeup Water Flush Header Auxiliary Building Type C M373 Ice Condenser Glycol Pumps Discharge Auxiliary Building Type C Line y M372 Ice Condenser Glycol Pumps Suction Auxiliary Building Type C
- Line M332 Cont. Hydrogen Purge Inlet Blower Discharge Line Atmosphere Type C :":c, o
M348 Reactor Coolant Drain Tank Gas O Auxiliary Building Type C n 3 Space to Waste Gas System so
- tp M221 Ventilation Unit Condensate Drain Header Auxiliary Building Type C Q
5 E { M356 Equipment Decontamination Line Auxiliary Building Type C O M358 Refueling Water Pump Suction o Auxiliary Building Type C N M377 Refueling Cavity Fill Line Auxiliary Building Type C of M235 Pressurizer Sample Auxiliary Building Type C
--s ?
- M310 Reactor Coolant Hot Leg Sample Auxiliary Building Type C .
m ao 0 u .
TABLE 3.6-1 (Continued) 9
$ SECONDARY CONTAINMENT BYPA3S LEAKAGE PATHS E
- PENETRATION TEST NUMBER SERVICE RELEASE LOCATION TYPE C
h M323 Component Cooling to Component Cooling Drain Sump Auxiliary Building Type C
@ M240 Nuclear Service Water to Reactor o Coolant Pump and Lower Cont.
N Vent. Units Auxiliary Building Type C M230 Nuclear Service Water From Reactor Coolant Pump and Lower Cont. Vent. Units Auxiliary Building lype C w M385 Nuclear Service Water to Upper
} Containment Ventilation Units In Turbine Building Type C M308 Nuclear Service Water to Upper Containment Ventilation Units Out Turbine Building Type C M213 Incore Instrumentation Room Purge $
o In Auxiliary Building Type C 0 3 Ro M140 Incore Instrumentation Room Purge Out Auxiliary Building Type C g rn M456 Upper Compartment Purge Inlet Auxiliary Building Type C 5 M432 Upper Compartment Purge Inlet Auxiliary Building Type C o o M357 Lower Compartment Purge Inlet Auxiliary Building Type C 4 M368 Containment Purge Exhaust Auxiliary Building Type C h M433 Containment Purge Exhaust Auxiliary Building Type C ,D
" M434 Lower Compartment Purge Inlet Auxiliary Building Type C m
y D
J 4 TABLE 3.6-1 (Continued)
; SECONDAR) CONTAINMENT BYPASS LEAKAGE PATHS s
2- PENETRATION TEST I NUMBER T SERVICE RELEASE LOCATION TYPE C 1 M386 Containment Air Release Auxiliary Building Type C l w M204 Containment Air Addition Auxiliary Building Type C l
$ M316 Int. Fire Protection Header - Auxiliary Building Type C
} ro Hose Racks ' l M337 Demineralized Water Auxiliary Building Type C M220 Instrument Air Auxiliary Building Type C 4 t
- M219 Station Air Auxiliary Building Type C 3
l { M215 Breathing Air Auxiliary Building Type C l l T M329 Reactor Coolant Pump Motor Oil Fill Auxiliary Building Type C ' i l M361 Int. Fire Protection Header - Auxiliary Building Type C C 3 i Sprinklers 0 i r, l M119 Containment Purge Exhaust Auxiliary Building Type C _o i t M331 Nitrogen Supply to Cold Leg Auxiliary Building Type C Accumulators k c~> } M322 Safety Injection Test Line Auxiliary Building Type C O o M454 UHI Test Line Auxiliary Building Type C ! O M328* Component Cooling to Reactor Auxiliary Building Type C Vessel Support and RCP Coolers n
*Not applicable for Unit 1 until after the first refueling outage.
4 I
CodTAINMENT SYSTEMS CONTAINMENT AIR LOCKS PROOF & gglEW COPY 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 exit through the containment, then at least one air lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 L*
at P,, 14.68 psig. 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 tne 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 inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTCOWN within the following 30 hours.
l ' l l l l
'NJb 10 BOD CATAWBA - UNITS 1 AND 2 3/4 6-8 f OCT 7 1985
CONTAINHENT SYSTEMS PROOF & REVIEW COPY 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 0.01 L, as determined by i
precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of 14.68 psig; b. By conducting overall air lock leakage tests at not less than P,, 14.68 psig, and verifying the overall air lock leakage rate is within i its limit: '
- 1) At least once per 6 months,# and
- 2) Prior to establisM ng CONTAINMENT INTEGRITY when maintenance has been performed on 1.ne 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.
- d. At least once per 6 months by conducting a pressure test at not less
, than P ,14.68 psig, to verify door seal integrity, with a measured j leak r$te of less than 15 sccm per door seal.
4 4 The provisions of Specification 4.0.2 are not applicable. This represents an exemption to Appendix J of 10 CFR Part 50. l JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-9 OCT 7 1985 l
- - , - . - -.. . , - . , , , - - - - - , , _ - - . - . .,-,v.- . .--- - - _ . ~ . - - - - . - - - - - - ,
d CONTAINMENT SYSTEMS INTERNAL PRESSURE PROOF g OEYfEW COPY f' 1 LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between
-0.1 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. f SURVEILLANCE REQUIREMENTS , 4.6.1.4 The primary containment internal pressure shall be determined to be l within the limits at least once per 12 hours. i JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-10 ocT 7 1985
CONTAINMENT SYSTEMS PROOF & ggIEW COPY AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall be maintained:
- a. Between 75* and 100*F in the containment upper compartment, and
- b. Between 100* and 120*F in the containment lower compartment.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment average air temperature not conforming to the above limits, restore the air temperature to within the limits within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.1.5.1 The primary containment upper compartment average air temperature shall be the arithmetical average of ambient air temperature monitoring stations located in the upper compartment. Temperature readings shall be obtained at least once per 24 hours from the elevation of 653 feet at the inlet of each operating upper containment ventilation unit. 4.6.1.5.2 The primary containment lower compartment average air temperature shall be the arithmetical average of ambient air temperature monitoring stations located in the lower compartment. Temperature readings shall be obtained at least once per 24 hours from the elevation of 570 feet at the inlet of each operating lower compartment ventilation unit. f I l
- Lower limit may be reduced to 60*F in MODE 2, 3 or 4.
1 I ( CATAWBA - UNITS 1 AND 2 3/4 6-11 JUL 10 0 01 7 1985 1 1 _-
CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY PROOF & REVIEW COP'( 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: With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200'F. SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degra-dation of the containment vessel detected during the above required inspec-tions shall be reported to the Commission within 15 days as a Special Report pursuant to Specification 6.9.2. f l CATAWBA - UNITS 1 AND 2 3/4 6-12 jyt i 81985 OCT 7 1985 l.__. .
CONTAINMENT SYSTEMS PRO 0f g g ' REACTOR BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION I 3.6.1.7 The structural integrity of the reactor building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the structural integrity of the reactor building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature aLove 200*F. SURVEILLANCE REQUIREMENTS 4.6.1.7 The structural integrity of the reactor building shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the reactor building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. Any abnormal degradation of the reactor building detected during the above required inspections shall be reported to the Commission within 15 days as a Special Repcrt pursuant to Specification 6.9.2. ' l 1 i CATAWBA - UNITS 1 AND 2 3/4 6-13 JUL 181985 OCT 7 1985 l \ _ _, , - - . _ _ _ - - - - . -. . - - - -- -
CONTAINMENT SYSTEMS ANNULUS VENTILATION SYSTEM PRoopgON/fWCOPy l LIMITING CONDITION FOR OPERATION s 3.6.1.8 Two independent Annulus Ventilation Systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one Annulus 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. , SURVEILLANCE RE0VIREMENTS 4.6.1.8 Each Annulus Ventilation 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 carbon adsorbers and verifying that the system operates for at least 10 continuous hours with the pre-heaters operating;
- b. At least once per 18 months or after any structural maintenance on the HEPA filter or carbon adsorber housings by:
- 1) Verifying that the cleanup system satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Posi-tions C.S.a, C.S.c, and C.5.d* of Regulatory Guide 1.52, Revi-sion 2, March 1978, and the system flow rate is 9000 cfm i 10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria i
of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%; and
- 3) Verifying a system flow rate of 9000 cfm 10% during system operation when tested in accordance with ANSI N510-1980.
- Purging of residual refrigerant is not mandatory. '
l OCT 7 1985 l CATAWBA - UNITS 1 AND 2 3/4 6-14
l l l CONTAINMENT SYSTEMS WCOPy SURVFILLANCE REQUIREMENTS (Continued)
- c. After every 720 hours of carbon adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regtriatory 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, carbon adsorber banks, and moisture separators is less than 8 inches Water Gauge while operating the system at a flow rate of 9000 cfm i 10%;
- 2) Verifying that the system starts automatically on any Phase "A" Isolation test signal,
- 3) Verifying that the filter cooling electric motor-operated bypass valves can be manually opened,
- 4) Verifying that each system produces a negative pressure of greater than or equal to 0.5 inch Water Gauge in the annulus within 1 minute after a start signal, and
- 5) Verifying that the pre-heaters dissipate 45 6.7 kW.
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetra-
, tion and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a 00P test aerosol while operating the system at a flow rate of 9000 cfm i 10%;
- f. After each complete or partial replacement of a carbon adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 9000 cfm i 10%; and
- g. Following painting, fire or chemical release in any ventilation zone communicating with the system, by verifying within 31 days after i
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 cri-teria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1%. I 00T 7# . CATAWBA - UNITS 1 AND 2 3/4 6-15 l
CONTAINMENT SYSTEMS l CONTAINMENT PURGE SYSTEMS PR00f g REVIEW r CDPo LIMITING CONDITION FOR OPERATION
- 3. 6.1. 9 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:
- a. Each containment purge supply and/or exhaust isolation valve for the lower compartment and the upper compartment (24-inch), instrument room (12-inch), and the Hydrogen Purge System (4-inch) shall be sealed closed, and
- b. The Containment Air Release and Addition System (4-inch) isolation valve (s) may be open for up to 2000 hours during a calendar year.
APPLICABILITY: MODES 1, 2, 3, and 4. 1 ACTION:
- a. With any containment purge supply and/or exhaust isolation valve for the lower compartment and the upper compartment, or instrument room, or Hydrogen Purge System open or not sealed closed, close and/or seal closed that valve or isolate the penetrations (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 Containment Air Release and Addition System isolation valve (s) open for more than 2000 hours during a calendar year, close the open 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 folicwing 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 Specifi-cations 4.6.1.9.3 and/or 4.6.1.9.4, 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. .
l SURVEILLANCE REQUIREMENTS 4.6.1.9.1 Each containment purge supply and/or exhaust isolation valves for the lower compartment and the upper containment, or instrument room, or Hydrogen Purge System shall be verified to be sealed closed at least once per 31 days. 4.6.1.9.2 The cumulative time that the Containment Air Release and Addition System has been open during a calendar year shall be determined at least once per 7 days. CATAWBA - UNITS 1 AND 2 3/4 6-16 OCT 7 1985 l
CONTAINMENT SYSTEMS 00Fg RE W SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.9.3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard valves with resilient material in each sealed closed containment purge supply and exhaust penetration for the lower compartment and the upper compartment, or instrument room, or Hydrogen Purge System shall be demonstrated OPERABLE pressurized to by verifying that the measured leakage rate is less than 0.05 L, when P,. 4.6.1.9.4 At least once per 3 months the Containment Air Release and Addition System with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to P,. 1 l P i l l l CATAWBA - UNITS 1 AND 2 3/4 6-17 OCT 7 1935
l l PRoop , CONTAINMENT SYSTEMS EW COPY 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the refueling water storage tank and transferring suction to the containment sump. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2 Each Containment Spray 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 positiun,
- b. By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 185 psid 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 Phase "B" Isolation test signal, '
and
- 2) Verifying that each spray pump starts automatically on a Phase "B" Isolation test signal.
- 3) Verifying that each spray pump is prevented from starting by the Containment Pressure Control System when the containment atmosphere pressure is less than or equal to 0.25 psid, and is allowed to start at greater than or equal to 0.45 psid relative to the outside atmosphere,
( CATAWBA - UNITS 1 AND 2 3/4 6-18 JUL 181985 0CT 7 1985
CONTAINMENT SYSTEMS REVIEW COPY SURVEILLANCE REQUIREMENTS (Continued)
- 4) Verifying that each spray pump discharge valve closes or is prevented from opening by the Containment Pressure Control l System when the containment atmosphere pressure is less than or equal to 0.25 psid and is allowed to open at greater than or equal to 0.45 psid relative to the outside atmosphere, and
- 5) Verifying that each spray pump is automatically deenergized by the Containment Pressure Control System when the containment atmosphere pressure is less than or equal to 0.25 psid relative to the outside atmosphere.
- d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
1 1 ( JUL 181985 CATAM3A - UNITS 1 AND 2 3/4 6-19 00T 7 1985
00F & REVIEW Copy
- CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.6-2 shall be OPERABLE with isolation times as shown in Table 3.6-2.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the isolation valve (s) specified in Table 3.6-2 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 isolation valves specified in Table 3.6-2 shall be demonstrated OPERABLE prior to returning the valve to service af ter maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test and verification of isolation time. . JUL 181985 CATAWBA,- UNITS 1 AND 2 3/4 6-20 007 71985
CONTAINMENT SYSTEMS EWCOPY SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 i months by: l I a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;
! b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position;
- c. Verifying that on a Containment Radioactivity-High test signal, each purge and exhaust valve actuates to its isolation position; and
- d. Verifying that on a High Relative Humidity (>70%) isolation test signal, each upper and lower containment purge supply and exhaust valve actuates to its isolation position.
1 4.6.3.3 The isolation time of each power-operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to
- Specification 4.0.5.
i s l i i i l l CATAWBA - UNITS 1 AND 2 3/4 6-21 UL 18 85 OCT 71985
TABLE 3.6-2 l' 5 32 CONTAINHENT ISOLATION VALVES 6, x HAXIMUM ll
-o VALVE NUMBER FUNCTION ISOLATION TIME (s)
- 1. Phase "A" Isolation s
r 4 3l BB-57B# Steam Generator IA Blowdown Containment Outside Isolation <10 88-218# Steam Generator 1B Blowdown Containment Outside Isolation 210 BB-61B# Steam Generator IC Blowdown Containment Outside Isolation 210 BB-10B# Steam Generator ID Blowdown Containment Outside Isolation 210 BB-56A# Steam Generator IA Blowdown Containment Inside Isolation 210 BB-19A# Steam Generator IB Blowdown Containment Inside Isolation 210
- BB-60A# Steam Generator IC Blowdown Containment Inside Isolation 210 i
BB-8A# Steam Generator ID Blowdown Containment Inside Isolation 210 u, BB-148B# Steam Generator IA Blowdown Containment Isolation Bypass 210 3: BB-150B# Steam Generator 1B Blowdown Containment Isolation Bypass 510 as 88-1498# Steam Generator IC Blowdown Containment Isolation Bypass <10 43 BB-1478# Steam Generator ID Blowdown Containment Isolation Bypass -210
; no CA-149# Steam Generator 1A Main Feedwater to Auxiliary Feedwater Nozzle Isolation 15 CA-150# Steam Generator IB Main Feedwater to Auxiliary Feedwater Nozzle Isolation 15 g CA-151# Steam Generator IC Main Feedwater to Auxiliary Feedwater Nozzle Isolation 15 c)
CA-152# Steam Generator ID Main feedwater to Auxiliary Feedwater Hozzle Isolation <5 CD CA-185# Auxiliary Nozzle Temper SGIA ' 55 , CA-186# Auxiliary Nozzle Temper SGIB <5 ) CA-187# Auxiliary Nozzle Temper SGIC {5 57
! CA-188# Auxiliary Nozzle Temper SGID <5 25 i CF-604 Steam Generator ID Feedwater Containment Isolation 25 CU CF-51# Steam Generator IC Feedwater Containment Isolation '#
25
! CF-42# Steam Generator 18 Feedwater Containment Isolation 25 E)
- CF-33# Steam Generator IA Feedwater Containment Isolation 25 29' cs!!~
CF-90# Steam Generator IA Feedwater Purge Valve 25
, CF-89# Steam Generator IB Feedwater Purge Valve 55 (l"")
f3'll CF-88# CF-87# Steam Steam Generator Generator IC ID Feedwater Feedwater Purge Purge Valve Valve 15
-a __ 15 mW -
8 "' ---f 1
a i TABLE 3.6-2 (Continued) , n i g CONTAINMENT ISOLATION VALVES I $ 1 a MAXIMUM
- VALVE NUMBER FUNCTION ISOLATION TIME (s)
U 1. Phase "A" Isolation (Continued) e
;e KC-3058# Excess Letdown Hx Supply Containment Isolation (Outside) <20 6 KC-3158# Excess Letdown Hx Return Header Containment Isolation (Outside) 520 N KC-320A# NCDT Hx Supply Hdr Containment Isolation (Outside) $20 KC-3328# NCDT Hx Return Hdr Containment Isolation (Inside) <20 KC-333A# NCDT Hx Return Hdr Containment Isolation (Outside) 220 ! KC-4298 RB Drain Header Inside Containment Isolation 210 KC-430A RB Drain Header Outside Containment Isolation 310
) , NB-260B Reactor Makeup Water Tank to Flush Header $10 t x l [ NC-53B Nitrogen to Pressurizer Relief Tank #1 Containment Isolation Outside $10 1 4 NC-54A Nitrogen to Pressurizer Relief Tank #1 Containment Isolation Inside $10 w NC-568 RMW Pump Disch Cont Isolation <10 l NC-1958 NC Pump Motor 011 Containment Isolation Outside 210 I l NC-196A NC Pump Motor Oil Containment Isolation Inside 510 m' o
) NF-228A Unit 1 Air Handling Units Glycol Supply Containment Isolation Outside <10 O i NF-233B Unit 1 Air Handling Units Glycol Return Containment Isolation Inside 210 Ro NF-234A Unit 1 Air Handling Units Glycol Return Containment Isolation Outside 210
~
m Q NI-47A NI-95A Accumulator N2 Supply Outside Containment Isolation Test Hdr Inside Containment Isolation
$10
<10 g
NI-968 Test Hdr Outside Containment Isolation 210 0 j NI-120B NI-122B# Safety Injection Pump to Accumulator Fill Line Isolation Hot leg Injection Check INI124, INI128 Test Isolation 510 @ X
$10 ; NI-1548# Hot Leg Recirculation Check 1NI125, 1NI129 Test Isolation $10 NI-255B UHI Check Valve Test Line Isolation <10 i
- r-
$c C NI-258A MI-264B UHI Check Valve Test Line Isolation UHI Check Valve Test Line Outside Containment Isolation 310 g $10 m
]
. y, g
- l j
TABLE 3.6-2 (Continued)
$ CONTAINMENT ISOLATION VALVES l5 MAXIMUM E VALVE NUMBER FUNCTION ISOLATION TIME (s)
- 1. Phase "A" Isolation (Continued)
[
@ NI-266A UHI Check Valve Test Line Inside Containment Isolation <10 NI-267A UHI Check Valve Test Line Inside Containment Isolation 210 NI-153A# Hot Leg Injection Check NI156, NI159 Test Isolation 210 NM-3A Pressurizer Liquid Sample Line Inside Containment Isolation <10 NM-6A Pressurizer Steam Sample Line Inside Containment Isolation 210 NM-7B Pressurizer Sample Header Outside Containment Isolation 210 NM-22A NC Hot Leg A Sample Line Inside Containment Isolation 210 w NM-25A NC Hot Leg C Sample Line Inside Containment Isolation 210 1 NM-26B NC Hot Leg Sample Hdr Outside Containment Isolation 210 m NM-72B NI Accumulator IA Sample Line Inside Containment Isolation 210 4
NM-758 NI Accumulator IB Sample Line Inside Containment Isolation 210 NM-788 NI Accumulator IC Sample Line Inside Containment Isolation 210 4 NM-81B NM-82A NI Accumulator ID Sample Line Inside Containment Isolation NI Accumulator Sample Hdr Outside Containment Isolation 510
<10 M
NM-187A# SG 1A Upper Shell Sample Containment Isolation Inside 210 h NM-190A# SG 1A Blowdown Line Sample Containment Isolation Inside 210 % NM-1918# NM-1978# SG 1A Sample Hdr Conta'nment Isolation Outside SG 1B Upper Shell Sample Containment Isolation Inside 510
$10 h
{D NM-200B# SG 1B Blowdown Line Sample Containment Isolation Inside <10 'O NM-201A# SG 1B Sample Hdr Containment Isolation Outside 510
<10 Q
73 NH-207A# SG IC Upper Shell Sample Containment Isolation Inside NH-210A# SG IC Blowdown Line Sample Containment Isolation Inside 210 NM-2118# SG IC Sartple Hdr Containment Isolation Outside 210 NM-2178# SG ID Upper Shell Sample Containment Isolation Inside 510 NM-220B# SG ID Blewdown Line Sample Containment Isolation Inside $10 o ~ HM-221A# SG 10 Sample Hdr Containment Isolation Outside $10 Q Letdown Containment Isolation Outside
- c
,_. NV-ISB $10 h M co NV-89A NC Pumps Seal Return Containment Isolation Inside <10 gg m
NV-918 NC Pumps Seal Return Containment Isolation Outside <10
$ NV-3148# Charging Line Containment Isolation Outside $10
1 TABLE 3.6-2 (Continued) S g CONTAINMENT ISOLATION VALVES x MAXIMUM g VALVE NUMBER FUNCTION ISOLATION TIME (s) G m 1. Phase "A" Isolation (Continued) s
& NV-11A 45 gpm Letdown Orifice Outlet - Containment Isolation <10 O NV-13A 75 gpm Letdown Orifice Outlet - Containment Isolation 210 ; ro NV-10A High Pressurizer Letdown Orifice Outlet - Containment Isolation 210 NV-872A Standby Makeup Pump to RCS seals 310 RF-389B Interior Fire Protection Containment Hose Rack Isolation Valve 55 (Outside Containment)
RF-4478 Reactor Building Sprinklers Containment Isolation Valve -<5 m (Outside Containment) 1
; , VB-838 Breathing Air Unit 1 Containment Isolation $10 VY-188** Containment Hz Purge to Annulus Inside Containment Isolation <10 j VY-17A** Containment H2 Purge to Annulus Outside Containment Isolation 210 VY-ISB** Containment H2 Purge Blower Outlet, Containment Isolation (Outside) 510 g o
VI-312A RB Isolation Valve for VI Supply to annulus Vent. 510 Q Qu VP-1B** Upper Containment Purge Supply #1 Outside Isolation 55 m VP-2A** Upper Containment Purge Supply #1 Inside Isolation <5 rri j VP-38** VP-4A** Upper Upper Containment Containment Purge Purge Supply Supply
#2
#2 Outside Isolation Inside Isolation 55 55 k
q VP-68** Lower Containment Purge Supply #1 Outside Isolation <5 o VP-7A** Lower Containment Purge Supply #1 Inside Isolation 55 Q VP-88** Lower Containment Purge Supply #2 Outside Isolation <5 -< VP-9A*" Lower Containment Purge Supply #2 Inside Isolation 25 g VP-10A** Upper Containment Purge Exhaust #1 Inside Isolation 55 Q ' OE D 2::= l NE
~
ww . I
. _ _ _ - - _ _ - - _ _ - _ _ _ _ _ _ - . -- _ _ ___ ~ ._.
TABLE 3.6-2 (Continued) S
$ CONTAINMENT ISOLATION VALVES 5
MAXIMUM E VALVE NUMBER FUNCTION ISOLATION TIME (s) [ 1. Phase "A" Isolation (Continued) E VP-11B** Upper Containment Purge Exhaust #1 Outside Isolation <5 VP-12A** Upper Containment Purge Exhaust #2 Inside Isolation 25
" VP-138** Upper Containment Purge Exhaust #2 Outside Isolation 25 VP-ISA** Lower Containment Purge Exhaust #1 Inside Isolation 25 VP-16B** Lower Containment Purge Exhaust #1 Outside Isolation 25 VP-17A** Incore Instru. Room Purge Supply Inside Isolation 25 VP-188** Incore Instru. Room Purge Supply Outside Isolation 55 u, VP-19A** Incore Instru. Room Purge Exhaust Inside Isolation <5 1 VP-20B** Incore Instru. Room Purge Exhaust Outside Isolation 55 VQ-2A** Containment Air Release Inside Isolation <5 VQ-38** Containment Air Release Outside Isolation 55 g VQ-15B** Containment Air Addition Uutside Isolation <5 O VQ-16A** Containment Air Addition Inside Isolation 55 k Go VS-54B Unit 1 Containment Header Outside Isolation $15 m
, in WL-8078# NCOT Pumps Discharge Outside Containment Isolation $10 rn WL-805A# NCDT Pumps Discharge Inside Containment Isolation $10 Q WL-450A NCDT Vent Inside Containment Isolation <10 o WL-451B WL-825A#** NCDT Vent Outside Containment Isolation RB Sump Pump Discharge Inside Containment Isolation 510
<10
'C WL-827B# *
- RB Sump Pump Discharge Outside Containment Isolation 510 YM-119B Demin. Water Containment Outside Isolation $10
$ {=
--* r-
- 2. Phase "B" Isolation
" KC-3388# NC Pump Supply Header Pent. Isolation (Outside) <40 h
[]
$ g KC-4248#
KC-425A# NC Pumps Return Hdr. Pent. Inside Isolation NC Pumps Return Hdr. Outside Isolation 540
$40 CJ1 yi .
TABLE 3.6-2 (Continued) 9 32 CONTAINMENT ISOLATION VALVES I5 an MAXIMUM gj VALVE NUMBER FUNCTION ISOLATION TIME (s) u' 2. Phase "B" Isolation (Continued) w g; RN-4378 Supply to NC Pumps and LCVU Supply Outside Containment Isolation $60 o RN-484A Return from NC Pumps and LCVU Return Inside Containment Isolation ~<60 N RN-4878 Return from NC Pumps and LCVU Return Outside Containment Isolation 60 3 RN-4048 Supply to Upper Containment Supply Ventilation Units Containment 210
~
I Isolation (Outside) RN-429A Return from Upper Containment Ventilation Units Containment Isolation
$10 1
(Inside) RN-43?B Return from Upper Containment Ventilation Units Containment Isolation <10
, (Outside) 1 30
, VI-778 Instrument Air Containment Outside Isolation $10 e
D$ SM-1 # Main Steam ID Isolation <5 50 SM-3 # Main Steam IC Isolation 35 Ej SM-5 # Main Steam IB Isolation <5 n
; SM-7 # Main Steam 1A Isolation 35 Ro l SM-9 # Main Steam ID Isolation Bypass Ctrl. <S 30 SM-10 # Main Steam IC Isolation Bypass Ctr1. 55 C2 SM-11 # Main Steam IB Isolation Bypass Ctr1. 55 F5 SM-12 # Main Steam 1A Isolation Bypass Ctrl. $5 25 c3 SV-19 # Main Steam 1A PORV $5 $$
SV-13 # Main Steam IB PORV ~<
$5 SV-7 # Main Steam IC PORV $5 SV-1 # Main Steam ID PORV <5 c- WL-867A*
- Containment Vent Unit Drains Inside Containment Isolation 210 (([
jgEE WL-869B *
- Containment Vent Unit Drains Outside Containment Isolation 510 l=g p
- ' - . h;y t
aEon gg M
TABLE 3.6-2 (Continued) 2>
$ CONTAINMENT ISOLATION VALVES 20 HAXIMUM Ei VALVE NUMBER FUNCTION ISOLATION TIME (s)
Z
- 3. Manual w
$l NC-141 NC Pump 2H Drain Tank Pump Discharge N.A.
NC-142 NC Pump 2H Drain Tank Pump Discharge N.A. NI-3 Boron Injection Tank Line to Cold legs N.A. FW-11 Refueling Water Pump Suction N.A. FW-13 Refueling Water Pump Suction N.A. CF-91# Feedwater 1A N.A. CF-93# Feedwater IB N.A. CF-95# Feedwater IC N.A. u, CF-97# Feedwater 10 N.A. 3: CA-121# Aux. Feedwater IA N.A. as BW-1# Aux. Feedwater IA H.A. 4 CA-120# Aux. Feedwater 18 N.A. BW-26# Aux. Feedwater IB N.A. no CA-119# Aux. Feedwater IC N.A. 3D BW-17# Aux. Feedwater IC N.A. h$ CA-118# Aux. Feedwater ID N.A. 'n BW-10# Aux. Feedwater ID N.A. D* SM-16# Main Steam 1A N.A. 30 SM-73#* Main Steam 1A N.A. C2 SM-105# Main Steam 1A N.A. h3 SM-121# Main Steam 1A N.A. "O SM-143# SM-72#* Main Steam 1A Main Steam IB N. A. N.A. [} 9a SM-104# Main Steam IB N.A. SM-120# Main Steam IB N.A. SM-142# SM-1# Main Steam IB Main Steam IB N.A. N.A. (lll} , g, l gll r SM-17# Main Steam IB N.A. l3:3m c3 {r- SM-18# Main Steam IC N.A. r* SM-71#* Main Steam IC N.A. oo 03 .- EE USw -
TABLE 3.6-2 (Continued) S 32 CONTAINMENT ISOLATION VALVES S:n 8 MAXIMUM gi VALVE NUMBER FUNCTION ISOLATION TIME (s) i :3 v' 3. Manual (Continued) w B SM-103# Main Steam IC N.A. E5 SM-119# Main Steam IC N.A. k) SM-141# Main Steam IC N.A. SA-4# Main Steam IC N.A. SM-19# Main Steam ID N.A. SM-70#* Main Steam 1D N.A. SM-102# Main Steam ID N.A.
. SM-118# Main Steam 10 N.A.
,, SM-140# Main Steam ID N.A.
); WE-20* Cont Bldg Supply Isol N.A.
c, WE-22* Cont Bldg Supply Isol N.A. g, WE-56* Cont Bldg Supply Isol N.A. , u) FW-4* Refueling Water N.A. so NV-862#* Pressurizer Auxiliary Spray ND Outside Containment N.A. [j
! WLA-21#* Steam Generator Drain Pump Discharge Outside Containment Isolation N.A. ,,
WLA-24#* Steam Generator Drain Pump Discharge Outside Containment Isolation H.A. Do 33 C2 TABLE NOTATIONS {3
-c i
- May be opened on an intermittent basis under administrative control. c3 1
c,o)
** Valve also receives a High Radiation (H) and/or a High Relative Humidity isolation signal. ~<
j # Not subject to Type C leakage tests. g , NOTE: Times are for valve operation only, and do not include any sensor response or circuit delay times. EI i See Specification 3/4 3.2 for system actuation response times. IIE3" 4 r-h5 *
.O OD Eb ES "
E$
l
. \
CONTAINMENT SYSTEMS PROOF & REVIE%/ COPY ) ; 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN HONITORS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION:
- a. With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within j the next 6 hours.
1 b. With both hydrogen monitors inoperable, restore at least one monitor 4 ' to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS i 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using hydrogen gas mixtures to obtain calibration points of:
- a. One volume percent hydrogen, and i
! b. Four volume percent hydrogen. 1 J l l , i l t CATAWBA - UNITS 1 AND 2 3/4 6-30 Jul 181985 00T 7 1985
CONTAINMENT SYSTEMS & REVIEW ggpy hh ELECTRIC HYDROGEN RECOMBINERS LIMITING CONDITION FOR OPERATION t 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next d 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 Recombiner System functional test that the minimum heater sheath temperature increases to greater than or equal to 700*F within 90 minutes. Upon reaching 700*F, increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 kW; and I b. At least once per 18 months by:
! 1) Performing a CHANNEL CALIBRATION of all recombiner I instrumentation and control circuits, l 2) Verifying through a visual examination that there is no i evidence of abnormal conditions within the recombiners I 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 abnve ' required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms. l A JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-31 00T 7 1985
CONTAINMENT SYSTEMS REVIEW COPY - HYOROGEN MITIGATION SYSTEM LIMITING CONDITION FOR OPERATION . i 3.6.4.3 Both trains of the Hydrogen Mitigation System shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: i With one train of the Hydrogen Mitigation System inoperable, restore the 1 inoperable train to OPERABLE status within 7 days or increase the surveillance i interval of Specification 4.6.4.3a. from 92 days to 7 days on the OPERABLE train j until the inoperable train is returned to OPERABLE status. I l SURVEILLANCE REQUIREMENTS l 4.6.4.3 Each train of the Hydrogen Mitigation System shall be demonstrated OPERA 8f.E:
- a. At least once per 92 days by energizing the supply breakers and verifying that at least 35 of 36 igniters are energized,* and
- b. At least once per 18 months by verifying the temperature of each igniter is a minimum of 1700*F.
j i
!
- Inoperable igniters must not be on corresponding redundant circuits which l provide coverage for the same region, i
l l ) { i i l l ! l l JUL 18 bn CATAWBA - UNITS 1 AND 2 3/4 6-32 00T 71985 _ . _ . . _ _ _ . . ~ , _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ , .___ _ _ _ _ _ _ __
4 CONTAINMENT SYSTEMS PROOF & REVIEV/ COPh i 3/4.6.5 ICE CONDENSER ICE BED i LIMITING CONDITION FOR OPERATION 3.6.5.1 The ice bed shall be OPERABLE with: i
- a. The stored ice having a boron concentration of at least 1800 ppm boron as sodium tetraborate and a pH of 9.0 to 9.5,
- b. Flow channels through the ice condenser, i
- c. A maximum ice bed temperature of less than or equal to 27'F, a
- d. A total ice weight of at least 2,368,652 pounds at a 95% level of confidence, and j e. 1944 ice baskets.
APPLICABILITY: H0 DES 1, 2, 3, and 4. ACTION: With ti.* ice bed inoperable, restore the ice bed to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT-i DOWN within the following 30 hours. 1 SURVEILLANCE REQUIREMENTS d i ! 4.6.5.1 The ice condenser shall be determined OPERABLE: j a. At least once per 12 hours by using the Ice Bed Temperature Monitor-8 ing System to verify that the maximum ice bed temperature is less than or equal to 27'F, l b. At least once per 9 months by: Chemical analyses which verify that at least nine representative 4
- 1) l l samples of stored ice have a boron concentration of at least '
] 1800 ppm as sodium tetraborate and a pH of 9.0 to 9.5 at 25'C; i 2) Weighing a representative sample of at least 144 ice baskets and verifying that each basket contains at least 1218 lbs of ice. The representative sample shall include six baskets from each of the 24 ice condenser bays and shall be constituted of 1 i I' CATAWBA - UNITS 1 AND 2 3/4 6-33 JUL 181965 00T 7 1985 i i
- . . . , _ . - - - . . - - . . . ~ , . - - _ - - , - , _ _ _ _ _ - . . - . . _ - - . - . - - - - . . . , - - . . _ - - _ , - - _ - _ , . . - - - . - _ _ . _ - - -. . - .
CONTAINMENT SYSTEMS PROOF & REVIEW COPY g[ SURVEILLANCE REQUIREMENTS (Continued) one basket each from Radial Rows 1, 2, 4, 6, 8, and 9 (or from the same row of an adjacent bay if a basket from a designated row cannot be obtained for weighing) within each bay. If any basket is found to contain less than 1218 pounds of ice, a representative sample of 20 additional baskets from the same bay shall be weighed. The minimum average weight of ice from the 20 additional baskets and the discrepant basket shall not be less than 1218 pounds / basket at a 95% level of confidence. The ice condenser shall also be subdivided into 3 groups of baskets, as follows: Group 1 - Bays 1 through 8, Group 2 - Bays 9 through 16, and Group 3 - Bays 17 through 24. The minimum average ice weight of the sample baskets from Radial Rows 1, 2, 4, 6, 8, and 9 in each group shall not be less than 1218 pounds / basket at a 95% level of confidence. The minimum total ice condenser ice weight at a 95% level of confidence shall be calculated using all ice basket weights determined during this weighing program and shall not be less than 2,368,652 pounds; and
- 3) Verifying, by a visual inspection of at least two fiow passages per ice condenser bay, that the accumulation of frost or ice on flow passages between ice baskets, past lattice frames, through the top deck floor grating, or past the lower inlet plenum support structures and turning vanes is restricted to a thickness of less than or equal to 0.38 inch. If one flow passage per bay is found to have an accumulation of frost or ice with a thickness of greater than or equal to 0.38 inch, a representative sample of 20 additional flow passages from the same bay shall be visually inspected. If these additional flow passages are found acceptable, the surveillance program may proceed considering the single deft-ciency as unique and acceptable. More than one restricted flow passage per bay is evidence of abnormal degradation of the ice condenser.
- c. At least once per 40 months by lifting and visually inspecting the accessible portions of at least two ice baskets from each one-third of the ice condenser and verifying that the ice baskets are free of detri-mental structural wear, cracks, corrosion or other damage. The ice baskets shall be raised at least 12 feet for this inspection.
l JUL 181985 l CATAWBA - UNITS 1 AND 2 3/4 6-34 OCT 7 1985 l t -
CONTAINMENT SYSTEMS ICE BED TEMPERATURE MONITORING SYSTEM PROOF & REVl&V COPY h h'
~
LIMITING CONDITION FOR OPERATION i 3.6.5.2 The Ice Bed Temperature Monitoring System shall be OPERABLE with at least two OPERABLE RTD channels in the ice bed at each of three basic elevations (< 11', 30'9" and 55' above the floor of the ice condenser) for each one-third of the ice condenser. APPLICABILITY: MODES 1, 2, 3, and 4. ) ACTION:
- a. With the Ice Bed Temperature Monitoring System inoperable, POWER OPERATION may continue for up to 30 days provided:
- 1. The ice compartment lower inlet doors, intermediate deck doors, and top deck doors are closed;
- 2. The last recorded mean ice bed temperature was less than or equal to 20*F and steady or decreasing ; and
- 3. The ice condenser cooling system is OPERABLE with at least:
a) Twenty-one OPERABLE air handling units, b) Two OPERABLE glycol circulating pumps, and c) Three OPERABLE refrigerant units. Otherwise, be in at least HOT STANDBY within 6 hours and in COLD j SHUTDOWN within the following 30 hours,
- b. With the Ice Bed Temperature Monitoring System inoperable and with i the Ice Condenser Cooling System not satisfying the minimum compo-i nents OPERABILITY requirements of ACTION a.3 above, POWER OPERATION i
may continue for up to 6 days provided the ice compartment lower inlet doors, intermediate deck doors, and top deck doors are closed , f and the last recorded mean ice bed temperature was less than or equal I to 15'F and steady; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.5.2 The Ice Bed Temperature Monitoring System shall be determined OPERABLE j by performance of a CHANNEL CHECK at least once per 12 hours. 1 JUL 18 ISB) CATAWBA - UNITS 1 AND 2 3/4 6-35 00T 7 1935
CONTAINMENT SYSTEMS ICE CONDENSER 000RS PROOF & REVlEW COPY LIMITING CONDITION FOR OPERATION 3.6.5.3 The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be closed and OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one or more ice condenser doors open or otherwise inoperable (but capable of opening automatically), POWER OPERATION may continue for up to 14 days provided the ice bed temperature is monitored at least once per 4 hours and the maximum ice bed temperature is maintained less than or equal to 27*F; otherwise, restore the doors to their closed positions or OPERABLE status (as applicable) within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one or more ice condenser door inoperable (not capable of opening automatically), restore all doors to OPERA 8LE status within 1 hour or be in HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.5.3.1 Inlet Doors - Ice condenser inlet doors shall be:
- a. Continuously monitored and determined closed by the inlet door position monitoring system, and
- b. Demonstrated OPERABLE during shutdown at least once per 9 months by:
- 1) Verifying that the torque required to initially open each door is less than or equal to 675 inch pounds;
- 2) Verifying that each door is capable of opening automatically in that it is not impaired by ice, frost, debris, or other obstruction;
- 3) Testing a sample of at least 50T of the doors and verifying that the torque required to open each door is less than 195 inch pounds when the door is 40 degrees open. This torque is defined as the
! " door opening torque" and is equal to the nominal door torque ( plus a frictional torque component. The doors selected for I determination of the " door opening torque" shall be selected to ensure that all doors are tested at least once during two test intervals; CATAWBA - UNITS 1 AND 2 3/4 6-36 0CT 7 1985 l 1 l _
l CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 4) Testing a sample of at least 50% of the doors and verifying that the torque required to keep each door from closing is greater than 78 inch pounds when the door is 40 degrees open.
This torque is defined as the " door closing torque" and is equal to the nominal door torque minus a frictional torque
- component. The doors selected for determination of the " door closing torque" shall be selected to ensure that all doors are tested at least once during two test intervals; and
- 5) Calculation of the frictional torque of each door tested in l accordance with Specification 4.6.5.3.lb.3) and 4), above.
The calculated frictional torque shall be less than or equal to 40 inch pounds. 4.6.5.3.2 Intermediate Deck Doors - Each ice condenser intermediate deck door shall be:
- a. Verified closed and free of frost accumulation by a visual inspection at least once per 7 days, and
- b. Demonstrated OPERABLE at least once per 3 months during the first year after the ice bed is initially fully-loaded and at least once per 18 months thereafter by visually verifying no structural
' deterioration, by verifying free movement of the vent assembif es, 1 and by ascertaining free movement when lifted with the applicable force shown below: I Door Lifting Force i 1) Adjacent to Crane Wall 5 37.4 lbs,
- 2) Paired w/ Door Adjacent to Crane ' 5 33.8 lbs, i
Wall 1
- 3) Adjacent to Containrent Wall 5 31.8 lbs, and
- 4) Paired w/ Door Adjacent to < 31.0 lbs.
Containment Wall 4.6.5.3.3 Top Deck Doors - Each ice condenser top deck door shall be deter-mined closed and OPERABLE at least once per 92 days by visually verifying:
- a. That the coors are in place, and
- b. That no condensation, frost, or ice has formed on the doors or blankets which would restrict their lifting and opening if required, l
1 ! CATAWBA - UNITS 1 AND 2 3/4 6-37 OCT 7 1985 i
CONTAINMENT SYSTEMS INLET 000R POSITION MONITORING SYSTEM PROOF & REVIEW COPY LIMITING CONDITION FOR OPERATION 3.6.5.4 The Inlet Door Position Monitoring System shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Inlet Door Position Monitoring System inoperable, POWER OPERATION may continue for up to 14 days, provided the Ice Bed Temperature Monitoring System is OPERABLE and the maximum ice bed temperature is less than or equal to 27'F when monitored at least once per 4 hours; otherwise, restore the Inlet Door Position Monitoring System to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS i 1 j 4.6.5.4 The Inlet Door Position Monitoring System shall be determined OPERABLE by: j a. Performing a CHANNEL CHECK at least once per 12 hours,
- b. Performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once
, per 18 months, and l
- c. Verifying that the Monitoring System correctly indicates the status
- of each inlet door as the door is opened and reclosed during its
! testing per Specification 4.6.5.3.1. l i l , l 1 CATAWBA - UNITS 1 AND 2 3/4 6-38 JUL 121:S5 OCT 7 1985
- - . . - . _- - - . _ = - . - . -- --- _ . - - _. .- - -
1 CONTAINMENT SYSTEMS PR0OF & REVIEW CdPY DIVIDER BARRIER PERSONNEL ACCESS DOORS AND EQUIPMENT HATCHES LIMITING CONDITION FOR OPERATION 3.6.5.5 The personnel access doors and equipment hatches between the contain-ment's upper and lower compartments shall be OPERABLE and closed. t APPLICABILITY: MODES 1, 2, 3, and 4. - t i ACTION: I l With a personnel access door or equipment hatch inoperable or open except for l personnel transit entry, restore the door or hatch to OPERABLE status or to ! its closed position (as applicable) within 1 hour or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following
- 30 hours.
! SURVEILLANCE REQUIREMENTS 1 I J 4.6.5.5.1 The personnel access doors and equipment hatches between the con-l tainment's upper and lower compartments shall be determined closed by a visual 1 inspection prior to increasing the Reactor Coolant System T above 200*F and
- after each personnel transit entry when the Reactor Coolant'5kstem T is j above 200*F. **U i
l 4.6.5.5.2 The personnel access doors and equipment hatches between the con-
- tainment's upper and lower compartments shall be detemined OPERABLE by
} visually inspecting the seals and sealing surfaces of these penetrations and l verifying no detrimental misalignments, cracks or defects in the sealing ) surfaces, or apparent deterioration of the seal material: l a. Prior to final closure of the penetration each time it has been l opened, and r l b. At least once per 10 years for penetrations containing seals fabri- ' I cated from resilient materials. l l l i l t t CATAW8A - UNITS 1 AND 2 3/4 6-39 M 18125 QCT 7 1985
._ - a
_. _. - .__ - ~_ __ - - - _- _ l l [' CONTAINMENT SYSTEMS O/ COPY CONTAINMENT AIR RETURN AND HYDROGEN SKIMMER SYSTEMS l l LIMITING CONDITION FOR OPERATION 3.6.5.6 Two independent Containment Air Return and Hydrogen Skimmer Systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4.
; ACTION:
With one Containment Air Return and Hydrogen Skimmer System inoperable, restore the inoperable system 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. J I SURVEILLANCE REQUIREMENTS 1 4.6.5.6.1 Each Containment Air Return and Hyurogen Skimmer System shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by:
- a. Verifying that the air return and hydrogen skimmer fans start '
automatically on a Containment Pressure-High-High test signal
- after a 9 1 1 minute delay and operate for at least 15 minutes;
- b. Verifying that during air return fan operation with the air return fan damper closed and with the bypass dampers open, the fan motor current is less than or equal to 59 amps when the fan speed is i 1187 1 13 rpm; I
a
- c. Verifying that with the hydrogen skimmer fan operating and the motor-operated valve in its suction line closed, the fan motor current is less than or equal to 69 amps when the fan speed is 3580 1 20 rpm; -
- d. Verifying that with the air return fan off, the motor-operated damper
- in the air return fan discharge line to the containment's lower com-
] partment opens automatically with a 10 1 1 second delay after a l Containment Pressure-High-High test signal; ! e. Verifying that with the air return fan operating, the check damper
- in the air return fan discharge line to the containment's lower compartment is open; CATAWBA - UNITS 1 AND 2 3/4 6-40 00T 7 1985
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- f. Verifying that the motor-operated valve in the hydrogen skimrer suction line opens automatically and the hydrogen skimmer ft rs receive a start permissive signal; and
- g. Verifying that with the fan off, the air return fan check damper is closed.
4.6.5.6.2 At least once per 18 months, each Containment Air Return and Hydrogen Skimmer System shall be demonstrated OPERABLE by:
- a. Verifying that each air return fan is deenergized or is prevented from starting by the Containment Pressure Control System when the contain-ment internal pressure is less than or equal to 0.25 psid, relative to the outside atmosphere; and
- b. Verifying that each air return fan isolation damper closes or is prevented from opening by the Containment Pressure Control System when the containment internal pressure is less than or equal to 0.25 psid and is allowed to open at greater than or equal to 0.45 psid, relative to the outside atmosphere.
1 l CATAWBA - UNITS 1 AND 2 3/4 6-41 JUL 181985 0CT 7 1985
l i CONTAINMENT SYSTEMS , FLOOR ORAINS PROOF & REVIEW COPY LIMITING CONDITION FOR OPERATION 3.6.5.7 The ice condenser floor drains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With an ice condenser floor drain inoperable, restore the floor drain to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.5.7 Each ice condenser floor drain shall be demonstrated OPERABLE at
- least once per 18 months during shutdown by
- a. Verifying that the valve gate opening is not impaired by ice, frost, or debris, i
- b. Verifying that the valve seat is not damaged,
- c. Verifying that the valve gate opens when a force of less than or equal to 66 lbs is applied, and
- d. Verifying that the drain line from the ice condenser floor to the containment lower compartment is unrestricted.
i i l i t I , CATAWBA - UNITS 1 AND 2 3/4 6-42 JUL 181985 00T 7 1985 _ ,- - _ _ - , , , .__ .__--_-..,,,m,
, , _ ~ , . - - _ _ _ , . - - _ . . - . . _ _ . _ , .
. 1 CONTAINMENT SYSTEMS F & REVIEW CdPY '
REFUELING CANAL DRAINS LIMITING CONDITION FOR OPERATION 3.6.5.8 The refueling canal drains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With a refueling canal drain inoperable, restore the drain to OPERABLE status within 1 hour or be in at least HOT STANOBY within the next 6 hours and in at least COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS , l 4.6.5.8 Each refueling canal drain shall be demonstrated OPERABLE:
- a. Prior to increasing the Reactor Coolant System temperature above 200*F after each partial or complete filling of the canal with water by verifying that the valves in the drain line are locked open and that the drain is not obstructed by debris, and i
- b. At least once per 92 days by verifying, through a visual inspec-tion, that there is no debris that could obstruct the drain.
1 JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 6-43 QCT 7 1985
I CONTAINMENT SYSTEMS DIVIDER BARRIER SEAL PROOF & REVlEW COPY {' LIMITING CONDITION FOR OPERATION 3.6.5.9 The divider barrier seal shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the divider barrier seal inoperable, restore the seal to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.5.9 The divider barrier seal shall be determined OPERABLE at least once ! per 18 months during shutdown by: l a. Removing two divider barrier seal test coupons and verifying that the physical properties of the test coupons are within the acceptable range of values shown in Table 3.6-3, and
- b. Visually inspecting at least 95% of the seal's entire length and:
- 1) Verifying that the seal and seal mounting bolts are properly installed, and
- 2) Verifying that the seal material shows no visual evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearances.
JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-44 OCT 7 1985
N TA8LE 3.6-3 & REVlEW COPY l l DIVIDER BARRIER SEAL ACCEPTABLE PHYSICAL PROPERTIES + 1 MATERIAL TENSILE STRENGTH Membrane Type Seals I Mk 10 39.7 lbs Mk 11 39.7 lbs i [ ] l i I l l i t JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 6-45 00T 7 1985 l
. - - - - - ~ . - - - -
CONTAINMENT SYSTEMS CONTAINMENT VALVE INJECTION WATER SYSTEM PRO 0f & REVl8V COPY LIMITING CONDITION FOR OPERATION 3.6.6 Both trains of the Containment Valve Injection Water System shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one train of the Containment Valve Injection Water System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.6.1 Each train of the Containment Valve Injection Water System shall be demonstrated OPERABLE at least once per 31 days by verifying that the system is pressurized to greater than or equal to 1.10 P, (16.2 psig) and has ade-quate capacity to maintain system pressure for at least 30 days. 4.6.6.2 Each train of the Containment Valve Injection Water System shall be demonstrated OPERABLE at least oncia per 18 months by verifying that the valve seal injection flow rate is less than 1.7 gpm for Train A and 1.4 gpm for Train B with a tank pressure greater than or equal to 45 psig and each auto-matic valve in the flow path actuotes to its correct position on a Containment Pressure-High or a Containment Pressure-High-High test signal. h l JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 6-46 0CT 7 1985
i 3/4.7 PLANT SYSTEMS PR0OF & REVlEW COPY 3/4.7.1 TURBINE CYCLE SAFETY VALVES 4 LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2. APPLICABILITY: MODES 1, 2, and 3. s 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 hours, 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 HOT SHUTDOWN within the following 6 hours.
i b. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. OCT 7 1985
, CATAWBA - UNITS 1 AND 2 3/4 7-1
PROOF & ggIBVN CDP TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES OURING FOUR LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SEIPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 87 2 65 3 43 1 i i 4 l JUL 181965 CATAWBA - UNITS 1 AND 2 3/4 7-2 OCT 7 1985
n TABLE 3.7-2 - e h STEAM LINE SAFETY VALVES PER LOOP i 5 VALVE NUMBER LIFT SETTING (1 1%)* ORIFICE SIZE
~
E
*1 Loop A Loop B Loop C Loop D vs
- 1. SV-20 SV-14 SV-8 SV-2 1175 psig 14.18 in.2 3E
- 2. SV-21 SV-15 SV-9 SV-3 1190 psig 14.18 in.2
- 3. SV-22 SV-16 SV-10 SV-4 1205 psig 14.18 in.2
- 4. SV-23 SV-17 SV-11 SV-5 1220 psig 14.18 in.2
- 5. SV-24 SV-18 SV-12 SV-6 1230 psig 14.18 in.2 w
. k i
y "The lift setting pressure shall correspond to ambient conditions of the valve at nominal J, operating temperature and pressure. 8 m Qo
- D i a
0 l m
, x 8
f
,.v-....
PLANT SYSTEMS PROOF & REVIEW COPY 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 steara 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 SHUTDOWN within the following 6 hours,
- b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY witin 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 total dynamic head of greater than or equal to 3470 feet at a flow of greater, than or equal to 400 gpm;
- 2) Verifying that the steam turbine-driven pump develops a total dynamic head of greater than or equal to 3550 feet at a flow of greater than or equal to 400 gpm when the secondary steam supply pressure is greater than 600 psig and the auxiliary feedwater pump turbine is operating at 3600 rpm. The provisions !
of Specification 4.0.4 are not applicable for entry into MODE 3; ' CATAWBA - UNITS 1 AND 2 3/4 7-4 E 181985 00T 7 1885
PLANT SYSTEMS PROOF & REVIEW COPY 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;
- 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; and
- 5) Verifying that the isolation valves in the auxiliary feedwater pump suction lines are open and that power is removed from the valve operators on Valves CA-2, CA-7A, CA-98, and CA-11A and that the respective circuit breakers are padlocked.
- b. At least once per 18 months during shutdown by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal,
- 2) Verifying that each motor-driven auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal,
- 3) Verifying that the turbine-driven auxiliary feedwater pump steam supply valves open upon receipt of an Auxiliary Feedwater Actua-tion test signal, and
- 4) Verifying that the valve in the suction line of each auxiliary feedwater pump from the Nuclear Service Water System automatically actuates to its full open position within less than or equal to 15 seconds
- on a Loss-of-Suction test signal.
4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of 30 days or longer prior to entering MODE 2 by verifying normal flow to each steam generator. 1 . 6
- Includes 5 second time delay.
1 CATAWBA - UNITS 1 AND 2 3/4 7-5 M 18 585 007 7 585 _ __
PLANT SYSTEMS PROOFg g g SPECIFIC ACTIVITY LIMITING CGiOITION FOR OPERATION 3.7.1.3 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 / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.1.3 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. CATAWBA - UNITS 1 AND 2 3/4 7-6 M I 81985 00I 7 1985
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 At least once per 72 hours.
Determination
- a) Once per 31 days, when-
- 2. Isotopic Analysis for DOSE ever the gross radio-EQUIVALENT I-131 Concentration activity determination indicates concentrations greater than 10% of the allowable limit for radiciodines.
b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10% of the allowatle limit for radioiodines.
"A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of ti.e secondary coolant except for radio-nuclides with half-lives less than 10 minutes. Determination of the contri-butors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.
I CATAWBA - UNITS 1 AND 2 3/4 7-7 JUl I 81985 l 7 1985 i
f l PLANT SYSTEMS REVIEW COPY MAIN STEAM LINE ISOLATION VALVES l LIMITING CONDITION FOR OPERATION 3.7.1.4 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. The provisions of Specification 3.0.4 are not applicable. Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 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. l l 1 JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 7-8 OCT 7 1985 l
PLANT SYSTEMS l 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 generators 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 thais 70*F. l s a l CATAWBA - UNITS 1 AND 2 JUL I 81985 l 3/4 7-9 007 7 1985
e,* n A-PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM 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; and
- b. At least once per 18 montns during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment actuates to its correct position on a Safety Injection, Phase "A" Isolation, or Phase "B" Isolation test signal, and
- 2) Each Component Cooling Water System pump starts automatically on a Safety Injection test signal.
i I l NL 181985 CATAWBA - UNITS 1 AND 2 3/4 7-10 007 7g
PLANT SYSTEMS 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent nuclear service water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one nuclear service 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 4.7.4 At least two nuclear service 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; and b. At least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment actuates to its correct position on a Safety Injection, or Phase "B" Isolation test signal, and i
- 2) Each Nuclear Service Water System pump starts automatically on a Safety Injection, or Loss-of-Offsite Power test signal.
1 - 1 I f ! E I 1 CATAWBA - UNITS 1 AND 2 JUL 181965 3/4 7-11 OCT 7 1985
PLANT SYSTEMS 3/4.7.5 STANDBY NUCLEAR SERVICE WATER POND LIMITING CONDITION FOR OPERATION 3.7.5 The standby nuclear service water pond (SNSWP) shall be OPERABLE with:
- a. A minimum water level at or above elevation 570 feet Mean Sea Level, !
USGS datum, and
- b. An average water temperature of less than or equal to 86.5'F at elevation 540 feet in the SNSWP intake structure.
APPLICABILITY: MODES 1, 2, 3, and 4. I ACTION: With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. i SURVEILLANCE REQUIREMENTS I 4.7.5 The SNSWP shall be determined GPERABLE:
- a. At least once per 24 hours by verifying the water level to be within its limit,
- b. At least once per 24 hours during the months of July, August, and September by verifying the water temperature to be within its limit, and
- c. At least once per 12 months by visually inspecting the SNSWP dam and .
verifying no abnormal degradation, erosion, or excessive seepage. ,' 1 JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 7-12 00T 71985
, PLANT SYSTEMS 00ra 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM Py LIMITING CONDITION FOR OPERATION '
3.7.6 Two independent Control Room Area Ventilation Systems shall be OPERABLE. APPLICABILITY: ALL MODES ACTION: MODES 1, 2, 3 and 4: With one Control Room Area Ventilt. tion 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 Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Area Ventilation System with flow through the HEPA filters and carbon adsorbers.
- b. With both Control Room Area Ventilation Systems inoperable, or with
' the OPERABLE Control Room Area Ventilation System, required to be operating by ACTION a., not capable of being powered by an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Area 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 90*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 carbon adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating; OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 7-13
PLANT SYSTEMS PROOF &REVsVCopy SURVEILLANCE REQUIREMENTS (Continued)
- c. At least once per 18 months or after any structural maintenance on the HEPA filter or carbon adsorber housings by:
- 1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Position C.5.a, C.5.c, and C.S.d* of Regulatory Guide 1.52, Revi-sions 2, March 1978, and the sytem flow rate is 6000 cfm i 10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance 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 Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1%; and
- 3) Verifying a system flow rate of 6000 cfm 110% during system operation when tested in accordance with ANSI N510-1980.
- d. After every 720 hours of carbon adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;
- e. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA filters, carbon adsorber banks, and moisture separators is less than 8 inches Water Gauge while operating the system at a flow rate of 6000 cfm 1 10%;
- 2) Verifying that on a High Radition-Air Intake, or Smoke Density-High test signal, the system automatically isolates the affected intake from outside air with recirculating flow through the HEPA filters and carbon adsorber banks;
- 3) Verifying that the system maintains the control room at a post-tive pressure of greater than or equal to 1/8 inch Water Gauge relative to adjacent areas at less than or equal to pressuriza-tion flow of 4000 cfm to the control room during system operation;
- 4) Verifying that the heaters dissipate 25 1 2.5 kW, and
- Purging of residual refrigerant is not mandatory.
4 OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 7-14
PLANT SYSTEMS PROOF & REVIEW SURVEILLANCE REQUIREMENTS (Continued)
- 5) Verifying that on a High Chlorine / Toxic Gas test signal, the system automatically isolates the affected intake from outside i air with recirculating flow through the HEPA filters and carbon j adsorbers banks within 10 seconds (plus air travel time between l the detectors and the isolation dampers). '
- 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 acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 6000 cfm i 10%;
- g. After each complete or partial replacement of a carbon adsorber bank, by verifying that the cleanup system satisfies the in place -
penetration and bypass leakage testing acceptance criteria of less
, than 1% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 6000 cfm i 10%; and
- h. Following painting, fire, or chemical release in any ventilation zone communicating with the system, 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 cri-teria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1%.
CATAWBA - UNITS 1 AND 2 3/4 7-15 OCT 71985
l PLANT SYSTEMS EN00f & REVIEW Co 3/4.7.7 AUXILIARY BUILDING FILTERED EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 The Auxiliary Building Filtered Exhaust System shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Auxiliary Building Filtered Exhaust System inoperable, restore the inoperable system 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. SURVEILLANCE REQUIREMENTS 4.7.7 The Auxiliary Building Filtered Exhaust System shall be demonstrated OPERABLE:
- a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and carbon adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;
- b. At least once per 18 months by:
- 1) Verifying that the cleanup system satisfies the in place
] penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regula- > tory 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 30,000 cfm i 10% per fan;
- 2) Verifying, within 31 days after removal, that a laboratory i
analysis of a representative carbon sample obtained in accor- ! dance 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, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1%; and
- Purging of residual refrigerant is not mandatory.
l 6 i CATAWBA - UNITS 1 AND 2 3/4 7-16 GCT 7 1985 ! l
- _ - __ - . =
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 3) Verifying a system flow rate of 30,000 cfm t10% per fan during system operation when tested in accordance with ANSI N510-1980. j
- c. After every 720 hours of carbon 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, I 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%; j d. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA filters, carbon adsorber banks, and moisture separators of less than 8 inches Water Gauge while operating the system at a flow rate of 30,000 cfm + 10% per fan,
- 2) Verifying that the system starts on a Safety Injection test signal, and directs its exhaust flow through the HEPA filters and carbon adsorbers,
! 3) Verifying that the system maintains the ECCS pump room at a negative pressure relative to adjacent areas,
- 4) Verifying that the filter cooling bypass valves can be manually opened, and
- 5) Verifying that the heaters dissipate 40 2 4 kW.
F e. After any structural maintenance on the filter housing or, 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% in accordanca with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 30,000 cfm 2 10% per fan; i f.
After any structural maintenance on the filter housing or, after each complete or partial replacement of a carbon adsorber bank, by verify- } ing that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in accord-ance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 30,000 cfm i 10% per fan; and i g. After any structural maintenance on the filter housing or, following ' i painting, fire or chemical release in any ventilation zone communicat-j l ing with the system by verifying, within 31 days after removal, that t I a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, j Revision 2, March 1978, meets the laboratory testing criteria of Regu-t latory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%. CATAWBA - UNITS 1 AND 2 3/4 7-17 00T 7 BBS
.- _ _. _ _ _-__ = __ --
PROOF & REVEW COPY 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 requirements are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system. APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES. ACTION: With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.8g. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.8 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program in lieu of the requirements of Specification 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. 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 mechan-ical snubbers. If less than two snubbers of each type are found in-j operable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months i 25% from
, the date of the first inspection. Otharwise, subsequent visual in-1 spections shall be performed in accordance with the following schedule: No. of Inoperable Snubbers of Each Subsequent Visual 1 Type Found During Inspection Inspection Period *# - 0 18 months i 25% 1 12 months 25% 2 6 months i 25% 3,4 124 days i 25% 5,6,7 62 days 1 25% 8 or more 31 days i 25%
"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 l 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.
l CATAWBA - UNITS 1 AND 2 3/4 7-16 gg7 7
l PROOF & REVIEW PLANT SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) I
- c. Visual Inspection Acceptance Criteria .
Visual inspections shall verify that: (1) there are no visible indi-cations of damage or impaired OPERABILITY, (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 pur-pose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and reme-died for that particular snubber and for other snubbers (regardless of type) that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found coadition and deter-mined OPERABLE per Specification 4.7.8f. When a fluid port of a hy-draulic snubber is found to be uncovered, the snubber shall be declared inoperable and may be determined OPERABLE via functional testing only if the test is started with the piston in the as-found setting extending the piston rod in the tension mede direction. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers.
- d. Refueling Outage Inspections At each refueling, the systems which have the potential for a severe dynamic event, specifically, the Nin Steam System (upstream of the main steam isolation valves) the main steam safety and power-operated relief valves and pi' ping, Auxiliary Feedwater System, main steam supply to the auxiliary feedwater pump turbine, and the letdown and charging portion of the CVCS System shall be inspected to determine if there has been a severe dynamic event. In the case of a severe dynamic event, mechanical snubbers in that system which experienced the event shall be inspected during the refueling outage to assure that the mechanical snubbers have freedom of movement and are not frozen up.
The inspection shall consist of verifying freedom-of-motion using one of the following: (1) manually induced snubber movement; or (2) eval-uation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel. If one or more mechanical snubbers are found to be frozen up during this inspection, , those snubbers shall be replaced or repaired before returning to power. The requirements of Specification 4.7.8b. are independent of the requirements of this specification.
- e. Functional Tests During the first refueling shutdown and at least once per 18 months l thereafter during shutdown, a representative sample of snubbers of each type shall be tested using one of the following sample plans.
The large-bore steam generator hydraulic snubbers shall be treated as i JUL 18 085 CATAWBA - UNITS 1 AND 2 3/4 7-19 , OCT 7 1985
PROOF & REVIEW COPY PLANT SYSTEMS DRAFT SURVEILLANCE REQUIREMENTS (Continued)
- e. Functional Tests (Continued) a separate type (population) for functional test purposes. A 10%
random sample shall be tested at least once per 18 months during refueling with continued testing based on a failure evaluation. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be noti-fied 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 all snubbers shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Spe-cification 4.7.8f., an additional 10% of all snubbers shall be functionally tested until no more failures are found or until all snubbers have been functicnally tested; or
- 2) A representative sample of all snubbers shall be functionally tested in accordance with Figure 4.7-1. "C" is the total number of snubbers of a type found not meeting the acceptance require-ments of Specification 4.7.8f. The cumulative number of snub-bers tested is denoted by "N". At the end of each day's test-ing, the new values of "N" and "C" (previous day's total plus current day's increments) shall be plotted on Figure 4.7-1. If at any time the point plotted falls in the " Reject" region, all snubbers of that type shall be functionally tested. If at any time the point plotted falls in the " Accept" region, testing of
- snubbers of that type may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of
, that type shall be tested until the point falls in the " Accept" region or the " Reject" region, or 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 t
the functional test acceptance criteria, another sample of at j 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-l 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 may be terminated. If the point plotted falls above the " Accept" line, testing must continue until the point falls in the " Accept" region or all the snubbers of that type have been tested. i JUL 181985 I CATAWBA - UNITS 1 AND 2 3/4 7-20 gg7
? 1985 i
PROOF & REVn
, 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 random-ly selected from all snubbers and reviewed before beginning the test-ing. 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. Snubbers placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, addi-tional sampling is required due to failure of only one type of snub-ber, 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 functional 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 co:::pression, except that inertia de-
, pendent, acceleration limiting mechanical snubbers may be tested to verify only that activation takes place in both directions of travel;
- 2) Snubber bleed, or release rate where required, is present in both tension and compression, within the specified raage;
- 3) For mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel; and
- 4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.
Testing methods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.
- g. Functional Test Failure Analysis i An engineering evaluation shall be made of each failure to meet the
- functional test acceptance criteria to determine the cause of the i failure. The results of this evaluation shall be used, if applicable, !
. in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
l JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 7-21 GCT 7 ;gg
PROOF & REVIEW COPY PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l
- g. Functional Test Failure Analysis (Continued) -
For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service. If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen-in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.8e. for snubbers not i meeting the functional test acceptance criteria. i j h. Functional Testing of Repaired and Replaced Snubbers
! Snubbers which fail the visual inspection or the functional test ! acceptance criteria shall be repaired or replaced. Replacement i snubbers and snubbers which have repairs which might affect the l functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall i
have met the acceptance criteria subsequent to their most recent , service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit. , ! i. Snubber Service Life Program i The service performance of all snubbers shall be monitored. If a 1 service lifetime limit is associated (established) with any snubber (or critical part) based on manufacturer's information, qualification tests, or historical service results, then the service life shall be monitored to ensure that the service life is not exceeded between surveillance inspections. Established snubber service life shall be 1 extended or shortened based on monitored test results and failure 1 history. The replacements (snubbers or critical parts) shall be docu-
- 1 mented and the documentation shall be retained in accordance with
! Specification 6.10.2. t l i l l jut 181965 CATAWBA - UNITS 1 AND 2 3/4 7-22 get 7 1985 l
~---e.- . ,,
PROOF & RE 10 O \ i 8 7
/ ,,
REJECT x Gj CONTINUE 3 2 / TESTING y
, [
/
2 [ ACCEPT 1 r J 0 10 20 30 40 50 60 70 80 90 100 : N 1 l FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 7-23 OCT 7 1985
PROOF & REVIEW COP PLANT SYSTEMS . 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 ard repair the sealed source, or
- 2. Dispose of the sealed source in accordance with Commission Regulations.
- b. The provisions of Lpecifications 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. '
\
CATAWBA - UNITS 1 AND 2 3/4 7-24 JUL 181985 j OCT 7 1985 l
- ~
PLANT SYSTEMS PROOF & REVISY COPY , 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 withcut a certificate indicating the last test date shall te tested prior to being placed into use; and i
- 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 or fuel assemblies and following repair or maintenance to the source.
4 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 I . I i , : i l i I l f l JUL 181985 I CATAWBA - UNITS 1 AND 2 3/4 7-25 MT 7M
PLANT SYSTEMS F& ggyIEW Copy 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM , LIMITING CONDITION FOR OPERATION 3.7.10.1 The Fire Suppression Water System shall be OPERABLE with: i
- a. At least two fire suppression purps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header, and
- b. An OPERABLE flow path capable of taking suction from Lake Wylie and transferring the water through distribution piping with OPERABLE sectionalizing control valves and isolation valves for each sprinkler, hose standpipe, or Spray System riser required to be OPERABLE per Specifications 3.7.10.2 and 3.7.10.4.
3 APPLICABILITY: At all times. 1 ACTION:
- a. With one of the above required pumps and/or one Water Supply /Distri-i bution System inoperable, restore the inoperable equipment to OPERABLE status within 7 days or provide an alternate backup pump or supply. 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.
i l JUL 181965 CATAWBA - UNITS 1 ANO 2 3/4 7-26 0CT 7 1985 i _-_ _. _ - _ - - - - _ . - . - _ . - .__ ___ . _ _ - _ _
PRoop g W lEW. Copy PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.10.1 The Fire Suppression Water System shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by starting each electric motor-driven pump and operating it for at least 15 minutes on recirculation flow,
- b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path which is accessible during plant operations is in its correct position,
- c. At least once per 6 months by performance of a system flush of the outside distribution loop to verify no flow blockage by fully opening the hydraulically most remote hydrant,
- d. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- e. At least once per 18 months by verifying that each valve (manual, power-cperated, or automatic) in the flow path which is inaccessible during plant operations is in its correct position,
- 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 automatic valve in the flow path actuates to its correct position,
- 2) Verifying that each pump develops at lea 3t 2500 gpm at a net pressure of 144 psig by testing at three points on the pump performance curve,
- 3) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and
- 4) Verifying that each fire suppression pump starts within 10 psig '
of its intended starting pressure (A pump, primary switch-95 psig; 8 pump, primary switch-90 psig; and C pump, primary switch-85 psig). ]
- g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 8, Section 16 of the Fire Protection Handbook, 15th Edition, published by the National Fire Protection Association.
JUL 181985 CATAWBA - UNITS 1 AND 2 3/4 7-27 0CT 7 1985 l ~ . - . _- .-.
- ^
PLANT SYSTEMS RO'ISVCOPY SPRAY AND/OR SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION ' 3.7.10.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:
- a. Elevation 522 + 0 ft - Auxiliary Building Room No. Equipment 100 RHR & Containment Spray Sump Pump Area 101 Corridor
-. .. 104 RHR Pump 18 -
105 RHR Pump 1A 106 Corridor 109 RHR Pump 2B 110 RHR Pump 2A 111 Corridor - 112 Corridor
- b. Elevation 543 + 0 ft - Auxiliary Building 230 Cent. Chg. Pump 1A
! 231 Cent. Chg. Pump 1B 240 Cent. Chg. Pump 2A 241 Cent. Chg. Pump 2B 250 ' Unit 1 Aux. Feedwater Pump Room 260 Unit 2 Aux. Feedwater Pump Room
- c. Elevation 554 + 0 ft - Auxiliary Building 340 Battery Room Corridor (DD-EE) 350 Battery Room Corridor (DD-EE)
- d. Elevation 560 + 0 ft - Auxiliary Building 300 Component Cooling Pumps 1A1, 1A2, 181 & 182
- e. Elevation 574 + 0 ft - Auxiliary Building 480 Cable Room Corridor (DD-EE) 490 Cable Room Corridor (DD-EE)
- f. Elevation 577 + 0 ft - Auxiliary Building 400 Component Cooling Pumps 2A1, 2A2, 2B1, & 282
! g. Reactor Buildings Annulus Pipe Corridor CATAWBA - UNITS 1 AND 2 3/4 7-28
PLANT SYSTEMS W COPY SURVEILLANCE REQUIREMENTS 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.
4.7.10.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path which is accessible during -
plant operations 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 by verifying that each valve (manual, power-operated, or automatic) in the flow path which is inaccessible during plant operations is in its correct position, and i
- d. At least once per 18 months:
- 1) By performing a system functional test which includes simulated automatic actuation of the system, and 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 each Sprinkler System starting at the system isolation valve to verify the system's integrity; and
- 3) By a visual inspection of each nozzle's spray area to verify the spray pattern is not cbstructed.
I 001 y coa CATAWBA - UNITS 1 AND 2 3/4 7-29 . . .
- ~
PROOF & REVIEW COPY PLANT SYSTEMS CO2 SYSTEMS LIMITING CONDITION FOR OPERATION s 3.7.10.3 The following High Pressure and Low Pressure CO2 Systems shall be OPERABLE:
- a. Low Pressure CO 2 System - Diesel generator rooms, and
- b. High Pressure CO2 System - Auxiliary feedwater pump rooms.
APPLICABILITY: Whenever equipment protected by the CO2 Systems is required to be OPERA 8LE. ACTION:
- a. With one or more of the above required CO2 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 4.7.10.3.1 Each of the above required CO 2 Systems shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path is in its correct position. 4.7.10.3.2 Each of the above required Low Pressure CO 2 Systems shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying the CO 2 storage tank level to be greater than 44% of full capacity, and
- b. At least once per 18 months by verifying
1
- 1) Each system actuates manually and automatically, upon receipt of a simulated actuation signal,
- 2) Damper closure devices receive an actuation signal upon system operation, and ,
- 3) By a visual inspection of discharge nozzles to assure no blockage.
CATAWBA - UNITS 1 AND 2 3/4 7-30 007 7 1985
l PROOF & REVIEW PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.7.10.3.3 Each of the above required High Pressure CO2 Systems shall be demonstrated OPERABLE: ,
- a. At least once per 6 months by verifying the weight of each CO2 storage cylinder to be at least 90% of full charge weight, and
- b. At least once per 18 months by:
- 1) Verifying each system actuates manually and automatically upon receipt of a simulated actuation signal,
- 2) Verifying that damper closure devices receive an actuation ,
signal upon system operation, and
- 3) A visual inspection of the discharge nozzles to assure no blockage.
l 1 l CATAWBA - UNITS 1 AND 2 3/4 7-31 007 7 1985
ROOF & R&l&/ CDPy PLANT SYSTEMS FIRE HOSE STATIONS 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 given in Table 3.7-3 inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the standard length of hose provided for the hose station. The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station. Where it can be demonstrated that the physical routing of the fire hose would result in a recognizable hazard to
' station personnel, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station. Signs shall be mounted above the gated wye (s) to identify the proper hose to use. The above ACTION requirement shall be accomplished within 1 hour if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours,
- b. The provisions of 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: ' i 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 200 psig or l at least 50 psig above maximum fire main operating pressure, l
whichever is greater. CATAWBA - UNITS 1 AND 2 3/4 7-32 OCT 7 1S85
TABLE 3.7-3 PROOF & REWE%g FIRE HOSE STATIONS LOCATION ELEVATION HOSE RACK #
- 1. Auxiliary Building -
59, FF 522+0 1RF235 55, FF 522+0 1RF248 63-64, KK 543+0 1RF210 63, Mi 543+0 1RF211 60, MM 543+0 1RF212 58, PP 543+0 1RF218 59, GG-HH 543+0 1RF236 60-61, FF-GG 543+0 1RF237 63, CC 543+0 1RF238
- 57, JJ 543+0 1RF242 -
54-55, GG 543+0 1RF249 57,FF 543+0 1RF250 52-53, GG 543+0 1RF255 51, CC 543+0 1RF256 50-51, JJ-KK 543+0 1RF262 53, MM 543+0 1RF268 50-51, NN 543+0 1RF271 62, MM-NN 560+0 1RF203 63, JJ-KK 560+0 1RF213 58, PP 560+0 1RF219 56, NN 560+0 1RF220 57, KK 560+0 1PF243 54-55, FF-GG 560+0 1RF251 51, KK 560+0 1RF263 52, MM-NN 560+0 1RF269 58, BB 554+0 1RF484 65, BB-CC 560+0 1RF485 [ 62, AA-BB 560+0 1RF486
- 56, BB 554+0 1RF487 t
52, AA-BB 560+0 1RF488 l 49, BB-CC 560+0 1RF489 l 68-69, BB 560+0 1RF996 45-46, BB 560+0 1RF997 63, NN 577+0 1RF204 61,LL 577+0 1RF214 i 63, KK-LL 577+0 1RF215 58, PP 577+0 1RF221 59, JJ 577+0 1RF230 l 58, GG 577+0 1RF240 i 56, KK 577+0 1RF244 54, GG 577+0 1RF252 52-53, KK 577+0 1RF258 51, KK 577+0 1RF264 51-52, NN 577+0 1RF272 56, PP 577+0 1RF278 68-69, BB 577+0 1RF478 65, BB-CC 577+0 1RF479 CATAWBA - UNITS 1 AND 2 3/4 7-33 007 7tg
i i TABLE 3.7-3 (Continued) IS R$/gYCOPy FIRE HOSE STATIONS i . LOCATION ELEVATION HOSE RACK # 59, DD 574+0 iRF480 60, AA 574+0 1RF481 49, BB-CC 577+0 1RF490 45, BB 577+0 1RF491 55, DD 574+0 1RF492 54, AA 574+0 1RF493 63, AA 577+0 1RF993 51, AA 577+0 1RF998 62, NN 594+0 1RF205 57, M 594+0 1RF222
-- -63, JJ 594+0 1RF231 -
57, HH 594+0 1RF245 57, EE 594+0 1RF253 51, JJ 594+0 1RF259 53, NN 594+0 1RF275 64, BB 594+0 1RF984 50, BB 594+0 1RF985 51, JJ 605+10 1RF265 63, JJ 605+10 1RF233 4 63-64, M 631+6 1RF483 50-51, MM 631+6 1RF495 4
- 2. Fuel Pools 48, TT-UU 605+10 1RF208 48, TT-UU 605+10 1RF276 50-51, MM 605+10 1RF483 50-51, MM 605+10 1RF822 r
- 3. Nuclear Service Water Pump Structure East Section 600+0 1RF939 West Section 600+0 1RF940 l
I i
)
l l l CATAWBA - UNITS 1 AND 2 3/4 7-34
i j PRoopg PLANT SYSTEMS
/EWCOPy 3/4.7.11 FIRE BARRIER PENETRATIONS , LIMITING CONDITION FOR OPERATION
) - 1 3.7.11 All fire barrier penetrations (walls, floor / ceilings, cable tray i enclosures and other fire barriers) separating safety related fire areas or separating portions of redundant systems important to safe shutdown within a j fire area and all sealing devices in fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable, piping, and ventilation duct pene-I tration seals) shall be OPERABLE. j APPLICABILITY: At all times. f , ACTION-t , I a. With one or more of the above required fire barrier penetrations )' and/or sealing devices inoperable, within I hour either establish a continuous fire watch on at least one side of the affected pene-
! tration, or verify the OPERABILITY of fire detectors on at least
] one side of the inoperable penetration and establish an hourly fire watch patrol. i
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i i
SURVEILLANCE REQUIREMENTS i : i 4.7.11.1 At least once per 18 months the above required fire barrier penetrations and sealing devices shall be verified OPERABLE by performing a visual inspection of:
- a. The exposed surfaces of each fire rated assembly; I
- b. At least 10% of all fire dampers. If apparent changes in appearance ,
or abnormal degradation are found, a visual inspection of an j additional 10% of the dampers shall be made. This inspection process shall continue untti a 10% sample with no apparent changes i in appearance or abnormal degradation is found. Samples shall be selected such that each fire damper will be inspected every 15 years; and
- c. At least 10% of each type of sealed penetration. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10% of each type of sealed penetration shall be made, f This inspection process shall continue until a 10% sample with no apparent changes in appearance or abnormal degradation is*found.
Samples shall be selected such that each penetration seal will be ! inspected every 15 years. . l i I. 1 l CATAWBA - UNITS 1 AND 2 3/4 7-35
PROOF g gEVIEW CO PLANT SYSTEMS : i j SURVEILLANCE REQUIREMENTS (Continued) I 4.7.11.2 Each of the above required fire doors shall be verified DPERABLE by 4 inspecting the closing mechanism and latches at least once per 6 months, and , by verifying:
- a. The position of each interior closed fire door at least once per 24 l hours, j
'l b. The OPERABILITY of the fire door supervision system for each i electrically supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days, and
] ,
- c. That each locked closed fire door is closed at least once per ,
i 7 days. i i 1 i i 1 i l l i i i i i e 1 I i i, i 1 l l I 001 i ! CATAWBA - UNITS 1 AND 2 3/4 7-36 l l l i
PLANT SYSTEMS RN/f$ 3/4.7.12 GROUNDWATER LEVEL LIMITING CONDITION FOR OPERATION s 3.7.12 The groundwater level shall be maintained at or below the I.op of the adjacent floor slabs of the Reactor Containment Building and the Auxiliary Building. APPLICABILITY: At all times. ACTION:
- a. With the groundwater level above the top of the adjacent floor slab
.~~ ~ by less than or equal to 5 feet, reduce the groundwater level to or.
below the top of the affected adjacent floor slab within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hears.
- b. With the groundwater level above the top of the adjacent floor slab by greater than 5 feet but less than 15 feet, reduce the groundwater level to less than or equal to 5 feet above the top of the affected adjacent floor slab within 24 hours and to or below the top of the affected adjacent floor slab within 7 days of initially exceeding the above limits or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the folicwing 30 hours.
- c. With the groundwater level above the top of the adjacent floor slab by greater than or equal to 15 feet, be in at least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the next 30 hours.
Perform an engineering evaluation to determine the effe ts of this higher groundwater level on the affected building (s) and submit the , results of this evaluation and any corrective action determined ! necessary to the Commission as a Special Report pursuant to Specifi-cation 6.9.2 prior to increasing T,yg above 200 F.
- d. Determine the rate of rise of groundwater when the level reaches the tcp of the floor slab. If the rate of rise of the groundwater level is greater than or equal to 0.3 foot per hour, determine the rate of rise at least once per 30 minutes. If the rate of rise exceeds 0.5 foot per hour for more than 1 hour, be in at least HOT STANDBY within 1 hour and in COLD SHUTOOWN within the following 30 hours.
If the rate of rise is less than 0.5 feet per hour, comply with the requirements of ACTIONS a. , b. , and c. above. CATAWBA - UNITS 1 AND 2 3/4 7-37
ptANT SYSTEMS PROOFgpgyggy SURVEILLANCE REQUIREMENTS 4.7.12 The groundwater level shall be determined at the following frequencies by monitoring the water level and by verifying the absence of alarlh in the six groundwater monitor wells as shown in FSAR Figure 2.4.13-14 instaljed around the perimeter of the Unit 1 Reactor and Auxiliary Building:
- a. At least once per 7 days when the groundwater level is at or below the top of the adjacent floor slab, and
- b. At least once per 24 hours when the groundwater level is above the top of the adjacent floor slab.
O 9 OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 7-38
PLANT SYSTEMS SYl[W(0Py 3/4.7.13 STANDBY SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION 3.7.13 The Standby Shutdown System (SSS) shall be OPERABLE. . APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the Standby Shutdown System inoperable, restore the inoperable equipment to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
- b. With the total leakage from UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE and reactor coolant pump seal leakage greater than 26 gpm, declare the Standby Makeup Pump inoperable and take ACTION a., above.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.13.1 The Standby Shutdown System diesel generator shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying:
- 1) The fuel level in the fuel storage tank is greater than or equal to 67 inches, and
- 2) The diesel starts from ambient conditions and' operates for at a least 30 minutes at greater than or equal to 700 kW.
- 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-0270-1975, is within the acceptable limits specified in Table 1 of ASTM-0975-1977 i
when checked for viscosity and 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.
4.7.13.2 The Standby Shutdown System 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.
CATAWBA - UNITS 1 AND 2 3/4 7-39 OCT 7 G85
4 R&ig, j PLANT SYSTEMS Py l SURVEILLANCE REQUIREMENTS (Continued)
, b. At least once per 92 days by verifying that the specific' gravity is j 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 j 2) The battery-to-battery and terminal connections are clean, tight, and free of corrosion. 4.7.13.3 The Standby Makeup Pump water supply shall be demonstrated OPERABLE
. by:
) .. , ,
- a. Verifying at least once per 7 days:
- 1) That the requirements of Specification 3.9.10 are met and the boron concentration in the storage pool is greater than or
- equal to 2000 ppm, or
- 2) That a contained borated water volume of at least 112,320 gallons with minimum baron concentration of 2,000 ppm is available and capable of being aligned to the Standby Makeup Pump.
j b. Verifying at least once per 92 days that the Standby Makeup Pump 1 develops a flow of greater than or equal to 26 gpm at a pressure greater than or equal to 2488 psig. 4.7.13.4 The Standby Shutdown System 250/125-Volt Battery Bank and its ) associated charger shall be demonstrated OPERABLE: :
- a. At least once per 31 days by verifying: I
- 1) That the electrolyte level of each battery is above the plates, i and
- 2) The total battery terminal voltage is greater than or equal to i 258/129 volts on float charge.
i
- 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: i
' ?
- 1) The batteries, cell plates, and battery racks show no visual i j indications of physical damage or abnormal deterioration, and l
- 2) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.
OCT 7 1985 CATAWBA - (JNITS 1 AND 2 3/4 7-40 i L- - - _ - - .- - . . - - -= - - - - - - -
__ _ _ _ _ _ _.__.__ __. _ - _ = _ _ _ . . _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ . _ . _ - - _ - . _ _ . -
-J s :
i I i Dfg + PLANT SYSTEMS IEly COPY i I l l SURVEILLANCE REQUIREMENTS (Continued) I i 4.7.13.5 The Steam Turbine Driven Auxiliary Feedwater Pump and associated l components shall be demonstrated OPERA 8LE at least once per 18 months by veri-fying that the system functions as designed from the Standby Shutdown System. 4.7.13.6 Each Standby Shutdown System instrumentation device shall be i demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per i j 31 days and a CHANNEL CALIBRATION at least once per 18 months, i i i m . . l 4 1 1 i i r i l J i i 1 i i I i i 1 i ? 4 l ! 71585 OCT l CATAWBA - UNITS 1 AND 2 3/4 7-41 l
{ l 3/4.8 ELECTRICAL POWER SYSTEMS PROOF & REVIEW COPY - 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 OPERABLE:
! a. Two physically independent circuits between the offsite transmission network and the Onsite Essential Auxiliary Power System, and
- b. Two separate and independent diesel generators, each with:
- 1) A separate day tank containing a minimum volume of 518.5 gallons i
of fuel,
~
- 2) A separate Fuel Storage System containing a minimum volume of 1 82,056 gallons of fuel,
) 3) A separate fuel transfer valve, and
- 4) A separate 125 VOC battery and charger connected to the diesel generator control loads.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: '
- a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstaate the OPERABILITY of the remaining A.C. sources by performing Specifica-
; tions 4.8.1.1.la. and 4.8.1.1.2a.4) within 1 hour and,at least once per 8 hours thereafter; restore at least two offsite circuits and ) two diesel generators 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. } 4
- b. With one offsite circuit and one diesel generator of the above I
required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-1 tions 4.8.1.1.la. and 4.8.1.1.2a.4) within 1 hour and at least once per 8 hours thereafter; restore at least one of the .
, inoperable sources to OPERABLE status within 12 hours or be in at i
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 OPERA 8LE status within 72 hours from 1 the time of initial loss or be in at least HOT STANDBY within the j 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 i
source of emergency power are also OPERABLE, and i i JUL 18 G5 CATAWBA - UNITS 1 AND 2 3/4 8-1 i OCT 7 235 _ _ . _ _ . _ _ _ _ _ _ _ _ . - _ _ . . - _ . _ . _ . ~ . - . . , _ . . _ _ . , . .
I J
\
ELECTRICAL POWER SYSTEMS PROOF & REylEW COPY LIMITING CONDITION FOR OPERATION ACTION (Continued) I 2. When in MODE 1, 2, or 3 with a steam pressure greater than 900
- psig, the steam-driven auxiliary feedwater pump is OPERABLE.
If these conditions are not satisfied within 2 hours be in at least i HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the l following 30 hours,
- d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Specfication 4.8.1.1.2a.4) within 1 hour and at least once per 8 hours thereaf ter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE 1 status within 24 hours or be in at least HOT STANDBY within the next i 6 hours. With only one offsite source restored, restore at least i
! two offsite circuits to OPERABLE status within 72 hours from time of ' i initial loss or be in at least HOT STANDBY within the next 6 hours i
; and in COLD SHUTDOWN within the following 30 hours. i
- e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Specifica-tion 4.8.1.1.la, within 1 hour and at least once per 8 hours there-after; restore at least one of the inoperable diesel generators to
! OPERABLE status within 2 hours or be in at least HOT STANOBY within 7 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 at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- f. With a diesel generator operating at greater than 5750 kW, within 1 hour reduce the diesel generator output to less than or equal to 5750 kW.
- g. With the Cathodic Protection System inoperable, restore the System to OPERABLE status within 10 days or prepare and submit a Special Report pursuant to Specification 6.9.2 outlining the cause of the
.i inoperability and the plans for restoring the System to OPERABLE. SURVEILLANCE REQUIREMENTS j 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Essential Auxiliary Power System shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the
~ normal circuit to the alternate circuit. 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: i a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by: , 1) Verifying the fuel level in the day tank, JUL 181935 ! CATAWBA - UNITS 1 AND 2 3/4 8-2 i OCT 7 1985
- - - - , . - - - , - , - , ~ , ----r-r-r--..-.- -,,,-r,n n,,,,,--w, e,--,n-- - - - n,,n - --,- ,. - , , - - , - - - - , -,--,,,,--,,-e -,,---c.,---- ,-eer- -~ --,-
PROOFg ngy;gy (9py ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2) Verifying the fuel level in the fuel storage tank,
- 3) Verifying the fuel transfer valve can be operated to, allow fuel to be transferred from the storage system to the day tank,
- 4) Verifying the diesel starts frcm ambient condition and accelerates to at least 441 rpm in less than or equal to 11 seconds.
The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 11 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) Simulated loss of offsite power in conjunction with an ESF Actuation test signal, or d) An ESF Actuation test signal by itself.
- 5) Verifying the generator is synchronized, loaded to greater than or equal to 5600 kW but less than or equal to 5750 kW in less than or equal to 60 seconds, and operates 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 tank;
, c. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks;
- d. By verifying that the Cathodic Protection System is OPERABLE
- by verifying: -
- 1) At least once per 60 days that cathodic protection rectifiers are OPERABLE and have been inspected in accordance with the manufacturer's inspection procedures, and
- 2) At least once per 12 months that adequate protection from corrosion is provided in accordance with manufacturer's inspection procedures.
, e. By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to storage tanks and:
*The Cathodic Protection System need not be OPERABLE until after June 1, 1985.
CATAWBA - UNITS 1 AND 2 3/4 8-3 I8135 1 OCT 7 1985
l
! l l
l ELECTRICAL POWER SYSTEMS PROOF & REVIEW COPY i SURVEILLANCE REQUIREMENTS (Continued) l 1) By verifying in accordance with the tests specified in . ASTM-0975-81 prier to addition to the storage tanks that the l sample has: . } j a) An API Gravity of within 0.3 degrees at 60*F, or a specific 1 gravity of within 0.0016 at 60/60*F, when compared to the ' j supplier's certificate, or an absolute specific gravity at
; 60/60*F of greater than or equal to 0.83 but less than or j equal to 0.89, or an API gravity of greater than or equal j to 27 degrees but less than or equal to 39 degrees; j 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 ! (alternatively, Saybolt viscosity, SUS at 100*F of greater j than or equal to 32.6, but less than or equal to 40.1), !
if gravity was not determined by comparison with the I supplier's certification; j c) A flash point equal to or greater than 125'F; and i d) A clear and bright appearance with proper color when ' l 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 4 when tested in accordance with ASTM-0975-81 except that the analysis for sulfur may be performed in accordance with
- ASTM-01552-79 or ASTM-D2622-82. '
\ i f. 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 i , with ASTM-D2276-78, Method A;
.' g. At least once per 18 months, during shutdown, by:
i
- 1) Subjecting the diesel to an inspection in accordance with ,
i procedures prepared in conjunction with its manufacturer's l l recommendations for this class of standby service; I l 2) Verifying the generator capability to reject a load of greater than or equal to 825 kW while maintaining voltage at 4160 1 420 volts and frequency at 60 1 1.2 Hz; f i 3) Verifying the generator capability to reject a load of greater l than or equal to 5600 kW but less than or equal to 5750 kW j without tripping. The generator speed shall not exceed 500 rpm j during and following the load rejection; j 4) Simulating a loss-of-offsite power by itself, and: i a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and i i l l CATAWBA - UNITS 1 AND 2 3/4 8 4 I 8 1985 l OCT 71985 l
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) PROOF & REVIEW CO b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 11 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. Af ter ener-gi:ation, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 + 420 volts and 60 1 1.2 Hz during this test.
- 5) Verifying that on an ESF Actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be at 4160 1 420 volts and 60 1 1.2 Hz within 11 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test;
- 6) Simulating a loss-of-offsite power in conjunction with an ESF Actuation test signal, and a) Verifying deenergization of the emergency busses and load shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 11 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 1 420 volts and 60 + 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engine overspeed, loclow lube oil pressure, .
generator differential, and the 2 out of 3 voltage controlled overcurrent relay scheme, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safety Injection Actuation signal.
- 7) Verifying the diesel generator operates for at least 24 hours.
The diesel generator shall be loaded to greater than or equal to 5600 kW but less than or equal to 5750 kW. The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 11 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits hl 181985 CATAWBA - UNITS 1 AND 2 3/4 8-5 00T y 085
ELECTRICAL POWER SYSTEMS PROOF & N WCOPy SURVEILLANCE REQUIREMENTS (Continued) during this test. Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2g.6)b) *
- 8) Verifying that the auto-connected loads to each diesel generator do not exceed 5750 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) Vorifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;
- 11) Verifying that the fuel transfer valve transfers fuel from each fuel storage tank to the day tank of each diesel via the in-stalled cross-connection lines;
- 12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within the tolerances given in Table 4.8-2;
- 13) Verifying that the voltage and diesel speed tolerances for the accelerated sequencer permissives are 92.511% and 98 + 1%,
respectively, with a minimum time delay of 2 1 0.2 s; and
~
- 14) Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a) Turning gear engaged, or b) Maintenance mode.
*If Specification 4.8.1.1.2g.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated at greater than or equal to 5600 kW but less than or equal to 5750 kW for 1 hour or until operating temperature has stabilized.
EL 181985 CATAWBA - UNITS 1 AND 2 3/4 8-6 W 7138$
ELECTRICAL POWER SYSTEMS PROOF & REVIEW COP SURVEILLANCE REQUIREMENTS (Continued)
- h. At least once per 10 years or af ter 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 441 rpm in less than or equal to 11 seconds; and
- 1. 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 its 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 1107,of the system design pressure.
- 3) Performing tank wall thickness measurements. The resulting data shall be evaluated and any abnormal degradation will be justified or corrected. Any abnormal degradation will be documented in a report to the Commission.
4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, 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. 4.8.1.1.4 Diesel Generator Batteries - Each diesel generator 125-volt battery a 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 at or above the low mark and at or below the high mark, .
- 2) The overall battery voltage is greater than or equal to 125 volts on float charge, and
- 3) The individual cell voltage is greater than or equal to 1.36 volts on float charge.*
- 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) There is no visible corrosion at either terminals or connectors, and
*Two different cells shall be tested each month.
#A maximum of two cells may be jumpered out at a time.
CATAWBA - UNITS 1 AND 2 3/4 8-7
ELECTRICAL POWER SYSTEMS ! SURVEILLANCE REQUIREMENTS (Continued) i i 2) The average electrolyte temperature of six connected cells is j above 60'F. i 1 ! c. At least once per 18 months by verifying that: ;
, 1) The batteries, cell plates and battery racks show no visual j indication of physical damage or abnormal deterioration, !
- 2) The cell-to-cell and terminal connections are clean, tight, ;
l 4 free of corrosion and coated with anticorrosion material in l l accordance with manufacturer's recommendations, i i
- 3) The cell-to-cell pole screws torque setting is 14.5 1 0.5 ft-lbs, ;
} 4) The battery charger will supply at least 75 amperes at a minimum of 125 volts for at least 8 hours, and -
- 5) The battery capacity is adequate to supply and maintain in 7
OPERABLE status its emergency loads when subjected to a ! battery service test. The battery shall supply a current of greater than or equal to 171.6 amps for the first minute j and a current of greater than or equal to 42.5 amps for
; the remaining 119 minutes, while maintaining a terminal j voltage of greater than or equal to 105 volts.
i d. At least once per 60 months, during shutdown, by verifying that the I i battery capacity is at least 80% of the manufacturer's rating when i subjected to a performance discharge test. Once per 60 month } interval, this performance discharge test may be performed in lieu j of the battery service test." l
. 4. 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 l performance tests, or is below 90% of the manufacturer's rating. ! I 1 l I l "First test to be conducted within 60 months following OL issuance date. g 18 Db6 < CATAWBA - UNITS 1 AND 2 3/4 8-8 g gp 0 01
j - i ! l TABLE 4.8-1 i DIESEL GENERATOR TEST SCHEDULE
%00rg R&ty COPY I NUMBER OF FAILURES IN -
LAST 100 VALID TESTS
- TEST FREQUENCY 1
.' i1 At least once per 31 days 2 At least once per 14 days l 3 At least once per 7 days f i
>4 At least once per 3 days l
i i I I ; l i l < 8 . 4 i r 1 i i i
!
- 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 detemined on a per j nuclear unit basis. For purposes of this schedule, only valid tests conducted i after the completion of the preoperational test requirements of Regulatory Guide r ! 1.108. Revision 1, August 1977, shall be included in the computation of the ' ! "last 100 valid tests." JUL 18 ud5 l , i OCT 7 ras ' ! CATAwsA - UNITS 1 AND 2 3/4 8-9 _ _ ~ . ~ . . - - _ _ _ _ , _ _ _ , . _ _ _ _ . . .- _ __ _ . ....._
i . i i TABLE 4.8-2 f LOAD 3EQUENCING TIMES PROOF & WgeN l ! LOAD GROUP SEQUENCE TIME . + NUMBER (Seconds) \
- Initiate Timer (T,) 9.7 2 0.3 1 1 (T3) T,+ 0.9 1 0.1 2 (T2 ) T, + 1.9 1 0.1 l
3 (T3 ) T,+ 4.7 1 0.3 4 (T4) T, + 9.4 1 0.6 l ! 5 (TS ) T,+ 14.1 1 0.9 6 (T6 ) T, + 18.8 1 1.2 3 7 (T7 ) T, + 23.5 2 1.5 8 (T8) T,+ 28.2 1 1.8 l 9 (T9 ) T, + 37.6 1 2.4 i j 10 (Tig' T,+ 47.0 1 3.0 l I 11 (T33) T, + 555.0 1 35.0 , l
] 12 (T12) T33 + 56.4 1 3.6 '
l 13 (T13) T33 + 112.8 i 7.2 I i I i 1 I i' l i i l l i i t i i I f
@L1 6 ue5 l
l OCT 7 un , CATAW8A - UNITS 1 AND 2 3/4 8-10 i
- i l
.._. _ - - - - ~= - - - - -
I i ELECTRICAL POWER SYSTEMS REWEW COPY f i A.C. SOURCES e SHUTDOWN - l ! LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be ] OPERABLE: 1 i a. One circuit between the offsite transmissior: network and the Onsite l Essential Auxiliary Power System, and
- b. One diesel generator with:
l 1) A day tank containing a minimum volume of 518.5 gallons of fuel, j 2) A fuel storage system containing a minimum volume of 82,056 , j gallons of fuel, I 3) A fuel transfer valve, and ]
; 4) A 125 VOC battery and charger connected to the diesel generator control loads.
- APPLICABILITY
- MODES 5 and 6.
l ACTION: l l With less than the above minimum required A.C. electrical power sources , ) OPERA 8LE, immediately suspend all operations involving CORE ALTERATIONS, ! i positive reactivity changes, movement of irradiated fuel, or crane operation 1 with loads over the fuel storage pool, and within 8 hours, depressurize and t vent the Reactor Coolant System through at least a 4.5 square inch vent. In ! addition, when in MODE 5 with the Reactor Coolant loops not filled, or in ' 4 MODE 6 with the water level less than 23 feet above the reactor vessel flange, i immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible. f SURVEILLANCE REQUIREMENTS i \ t 1 i 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated i 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), 4.8.1.1.3, and , i 4.8.1.1.4. l
?
I Jul.181965 ' CATAWBA - UNITS 1 ANO 2 3/4 8-11 g g nos
i i i ELECTRICAL POWER SYSTEMS PROOF & ggylEW CCPy i 3/4.8.2 0.C. SOURCES OPERATING 4 LIMIT'NG CONDITION FOR OPERATION ! ! I { 3.8.2.1 The following D.C. channels and trains shall be OPERABLE and energized: t { a. Channel I consisting of 125-Volt D.C. Bus No. EDA, 125-Volt D.C. 1 4 Battery Bank No. E8A and a full-capacity charger,*
- b. Channel 2 consisting of 125-Volt D.C. Bus No. EDB,125-Volt D.C. i j Battery Bank No. EBB and a full-capacity charger,*
- s. . c. Channel 3 cansisting of 125-Volt D.C. Bus No. EDC, 125-Volt D.C. -
- Battery Bank No. E8C and a full-capacity charger,*
- d. Channel 4 consisting of 125-Volt D.C. Bus No. E00, 125-Volt D.C.
Battery Bank No. E80 and a full-capacity charger," . e. Train A consisting of 125-Volt D.C. Bus No. EDE, and
- f. Train 8 consisting of 125-Volt D.C. Bus No. EDF.
- APPLICABILITY
- MODES 1, 2, 3, and 4.
j ACTION: i l
- a. With 125 VOC Bus EDE or EDF inoperable, restore the inoperable bus to OPERA 8LE status within 2 hours or be in at least HOT STAN08Y within '
the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one 125 VOC Bus EDA, EDB, EDC or E00 inoperable, restore the i inoperable bus to OPERABLE status within 8 hours or be in at least I
HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the i l following 30 hours. l c. With either 125 VOC Battery Bank No. EBB or E8C and/or its full- { capacity charger inoperable, restore the inoperable battery and/or
- full-capacity charger to OPERABLE status within 10 days or be in at
! least HOT STAN08Y within the next 6 hours and in COLD SHUTOOWN within i the following 30 hours. , I
*A vital bus may be disconnected from its D.C. source for up to 24 , hours for the purpose of performing an equalizing charge on its associated battery bank j provided that the vital busses associated with the other battery banks are i OPERABLE and energized. Also when the spare charger is being used as a l l replacement for the nonsal battery charger verify that the A.C. input to the i
charger is from the same A.C. division as the normal charger which it is j replacing, t . CATAWBA - UNITS 1 AND 2 3/4 8-12 W
ELECTRICAL POWER SYSTEMS PROOF & REVIEW COPY LIMITING CONDITION FOR OPERATION ACTION (Continued)
- d. With either 125 VDC Battery Bank No. EBA or E80 and/or its full-capacity charger inoperable and 125 VDC diesel generator Batteries OG8A and DGBB and their full-capacity chargers in service powering Busses EDE and EDF during this period of time, restore the inoperable battery and/or full-capacity charger to OPERABLE status within 10.
da3: or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- e. With two 125 VDC batteries and/or their full-capacity chargers in-operable and 125 VDC Batteries EBA and EBC and/or their full-capacity chargers in service, or 125 VDC Batteries EBB and E80 and/or their full-capacity chargers in service duaing this period of time, restore at least one battery and/or its full-capacity charger to OPERABLE a status within 72 hours or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the f ollowing 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.2.1.1 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 4.8-3 meet the Category A limits,
- 2) The total battery terminal voltage is greater than or equal to 125 volts on float charge, and
- 3) There is no visible indication of electrolyte leakage.
~
- 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 4.8-3 meet the category B limits,
- 2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 8 ohm, and
- 3) The average electrolyte temperature of six connected cells is above 60*F.
- c. At least once per 18 months by verifying that:
- 1) The cells, cell plates (if visible), and battery racks show no visual indication of physical damage or abnormal deterioration, 00T 71985 CATAWBA - UNITS 1 AND 2 3/4 8-13
(
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2) The cell-to-cell and terminal connections are clean tight, and coated with anticorrosion material,
- 3) The resistance of each cell-to-cell an1 terminal connection is less than or equal to 150 x 10 8 ohm, and
- 4) The battery charger will supply at least 200 amperes at a minimum of 125 volts for at least 8 hours.
- d. At least once per 18 months during shutdown, by veritying that the battery capacity is adequate to either:
- 1) Supply and maintain in OPERABLE status all of the actual erergency i loads for 1 hour when the battery is subjected to a ratterj service test; or i
- 2) Supply a dummy load from Batteries EBA and EB0 and fror., Bat- '
J 1 teries EBB and EBC of greater than or equal to 373 amperes for the first minute of the first hour, greater than or equal to , t 213 amperes for the next 59 minutes of the first hour and a c'unty load only from Batteries EBA and E80 of greater than or equal tt- 1 210 amperes for the second hour while maintainirn the bat.tery terminal voltage greater than or equal to 105 volts. , e. At least once per 60 months, during shutdown, by verifying that tht battery capacity is at least 80% of the manufacturer's ratirg when subjected to a performance discharge test. Once per G] month interval this performance discharge test may be pe* formed in lieu of the battery service test required by Specification 4.8.2.1.1d. ; and
- f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacit of degradation or has reached 85%ofy the to any batterylife service that shows signs 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 manufact.srer's rating.
4.8.2.1.2 Each D.C. channel shall be determined OPERABLE and energized with tie breakers open between redundant busses at least once per 7 days by veri-fying correct breaker alignment, indicated power availability from the charger and battery, and voltage on the bus of greater than or equal to 125 volts. e 00T 7 1985 CATAWBA - UNITS 1 AND 2 3/4 8-14 i
PROOF & REVIEW COP \ TABLE 4.8-3 h y { l BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY B(2) PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH
- CELL CONNECTED CELL ,
s Ele:trolyte > Minimum level >Hinimum level Above top of Level indication mark, indication mark, plates, i and < k" above and < %" above and not maxiium level maxiEum level overflowing indication mark indication mark Float Voltage 1 2.13 volts 1 2.13 volts (6) > 2.07 volts Not more than 0.020 below the average of all Specific > 1.200(5) > 1.195
~
connected cells Gravity (4) or 1 1.195 Average of all Average of all connected cells connected cells
> 1.205 1.195(5)
TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. . (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERA 8LE provided that the Category B parameters are within their allowable values and provided the Category B 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. (5) Or battery charging current is less than 2 amps when on charge. (6) Corrected for average electrolyte temperature. jul.181985 CATAWBA - UNITS 1 AND 2 3/4 8-15 gg 7 WB5 1
PR0OF & REVIEW COPY ELECTRICAL POWER SYSTEMS D.C. SOURCES SHUTDOWN 4 LIMITING CONDITION FOR OPERATION 3.3.2.2 As a minimum, the following D.C. electrical equipment and busses shall be OPERABLE and energized:
- a. One - 125-volt D.C. train related bus (EDE or EDF),
- b. Two - 125-volt D.C. channel related busses (one of which must be EDA or EDD), and
- s. .. .
- c. Two - 125-volt battery banks and full-capacity chargers associated with the above D.C. busses.
3 APPLICABILITY: MODES 5 and 6. ACTION: With the required battery banks and/or full-capacity chargers or 125-volt D.C. busses inoperable, immediately suspend all operations involvine CORE ALTERATIONS, positive reactivity changes or movement of irradiated fuel; iritiate corrective action to restore the required battery bank and full-capacit:: charger to OPERABLE status as soon as possible, and within 8 hours, degressurize and vent the Reactor Coolant System through at least a 4.5 square inch vent. i l SURVEILLANCE REQUIREMENTS 4.8.2.2.1 The above required 125-volt D.C. busses shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and voltage on the bus. 4.8.2.2.2 The above required 125-volt battery banks and chargers shall be demonstrated OPERABLE per Specifications 4.8.2.1.1 and 4.8.2.1.2. CATAWBA - UNITS 1 AND 2 3/4 8-16 00
l ELECTRICAL POWER SYSTEMS PROOF & REVIEW COPY 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The fc11owing A.C. electrical busses and inverters shall be OPERABLE and energized with tie breakers open between redundant busses:
- a. 4160-Volt Essential Bus # ETA,
- b. 4160-Volt Essential Bus #ETB,
- c. 600-Volt Essential Dus #ELXA,
- d. 600-Volt Essential Bus #ELXB, ,
- e. 600-Volt Essential Bus #ELXC,
- f. 600-Volt Essential Bus #ELXD,
- g. 120-Volt A.C. Vital Bus #ERPA energized from Inverter #EIA connected to D.C. Channel 1,*
- h. 120-Volt A.C. Vital Bus # ERPB energized from Inverter #EIB connected to D.C Channel 2,* -
- 1. 120-Volt A.C. Vital Bus #ERPC energized from Inverter #EIC connected
., to D.C. Channel 3,*
- j. 120-Volt A.C. Vital Bus #ERPD energized from Inverter #EID connected to D.C. Channel 4.*
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With less than the above complement of A.C. busses OPERABLE and energized, restore the inoperable UJsses to OPERABLE and energized status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one inverter inoperable, energize the associated A.C. vital bus within 2 hours; restore the inoperable inverter to OPERABLE and energized 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.
~
*An inverter may be disconnected from its D.C. source for up to 24. hours for the purpose of performing an equalizing charge on its associated battery bank provided: (1) its vital bus is OPERABLE and energized, and (2) the vital busses associated with the other battery banks are OPERABLE and energized from their associated inverters an( ';onnected to their associated D.C. bus.
{ eg $B5 CATAWBA - UNITS 1 AND 2 3/4 8-17 i I
ELECTRICAL POWER SYSTEMS PROOF & REVIEW C ONSITE POWER DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION
. . - > ~
SURVEILLANCE REQUIREMENTS
~ - n.
4.8.3.1 The specified A.C. busses and inverters shall be s tfr M " U P jized in the required manner at least once per 7 days by verifyin 'w it t n.4er alignment and indicated voltage on the busses. 3.8. E 2 As a minimum, the following A.C. electrical busses and inverters shall be OPERABLE and energized:
- a. One - 4160-volt essential bus,
- b. Two - 600-volt essential busses in a single train, and
- c. Two - 120-volt A.C. vital busses energized from their respective inverters connected to their respective D.C. channels.
APPLICABILITY MODES S and 6. ACTION: With any of the above required A.C. busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required A.C. busses in the specified manner as soon as possible, and within 8 hours depressurize and vent the Reactor Coolant System through at least a 4.5 square inch vent. SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified A.C. busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker align-ment and indicated voltage on the busses. , CATAWBA - UNITS 1 AND 2 3/4 8-18 jy( l 8 Nb 7 1985 l
ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES PRO 0Fg CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES WCOPy LIMITING CONDITION FOR OPERATION 3.8.4 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 containment penetration conductor overcurrent protective device (s) given in Table 3.8-1 inoperable:
- a. Restore the protective device (s) to OPERABLE status or de-energize 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 inoper-able 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 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 medium voltage (4-15 kV) circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of '
the circuit breakers of each vo'.tage level, and performing the following: a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated protective system functional test which includes simulated automatic actuation of the system and t verifying that each relay and associated circuit breakers function as designed, and BBS CATAWBA - UNITS 1 AND 2 3/4 8-19 M 10 g i BBS
1 PROOF EVIEW COPY ELECTRICAL POWER SYSTEMS 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 nr all circuit breakers of that type have been functionally t.ested.
- 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 func-tional 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 a least 10% of all the circuit breakers of the inoper-able type 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 semple 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 por 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. t l i CATAWBA - UNITS 1 AND 2 3/4 8-20 M1 ! OCT 7 1985
TABLE 3.8-la NEWEWCOPy UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 1. 6900 VAC Swgr Primary Bkr RCP1A Reactor Coolant Pump 1A Backup Bkr 1TA-3 Primary Bkr RCP1B Reactor Coolant Pump 1B Backup Bkr ITB-3
,~'
Primary BKR RCPIC Reactor Coolant Pump IC '
' Backup Bkr ITC-3 Primary BKR RCP10 Reactor Coolant Pump 1D Backup Bkr ITD-3
- 2. 600 VAC MCC 1EMXC-F01B Primary Bkr Accumulator 1C Discharge Backup Fuse Isol Viv 1NI76A 1EMXC-F01C Primary Bkr Check Valve Test Header Backup Fuse Cont Isol V1v INI95A 1EMXC-F02A Primary Bkr Train A Alternate Power Backup Fuse To ND LTDN V1v 1N018 1EMXC-F028
- Primary Bkr Hot Leg Inj. Check Viv Backup Fuse Test Isol V1v 1NI153A 1EMXC-F02C Primary Bkr Cont Isol at 134 Deg Backup Fuse Annulus Area Viv IVI312A j 1EMXC-F03A Primary Bkr NC Pump IC Thermal Barrier Outlet i Backup Fuse Isol V1v 1KC345A 1EMXC-F03B Primary Bkr N2 to Prt Cont Isol Inside Backup Fure Viv 1NC54A 1EMXC-F03C Primary Bkr Pressurizer Power-Operated l Backup Fuse Relief Isol Vlv INC33A
' 1 OCT CATAWBA - UNITS 1 AND 2 3/4 8-21 l l
l __ - .- - - _ -, - -- . - -..-l
-.m .
TABLE 3.8-la (Continued) NNIW/ COPY UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXC-F05A Primary Bkr NCDT Vent Inside Cont Isol Backup Fuse Vlv 1WL450A 1EMXC-F058 Primary Bkr Cont Sump Pumps Discharge Inside Backup Fuse Cont Isol Vlv IWL825A IEMXC-T05C Primary Bkr Vent Unit Cond Drn Tank Backup Fuse Outside Cont Isol Vlv 1WL867A 1
1EMXC-F06A Primary Bkr NCDT Pumps Disch Inside Cont Isol Backup Fuse Viv 1WL805A 1EMXC-F07B Primary Bkr Cont H2 Purge Outlet Cont Isol Backup Fuse Viv IVY 17A 1EMXD-F01A Primary Bkr ND Pump 1A Suction From NC Backup Fuse Loop B Viv 1ND1B IEMXD-F018 Primary Bkr Accumulator 18 Discharge Backup Fuse Isol Viv 1NI65B IEMXD-F01C Primary Bkr NI Pump A to Hot leg Check Backup Fuse Vlv Test Isol Viv 1NI1228 1EMXD-F02A Primary Bkr ND Purnp 1B Suction from NC Backup Fuse Loop C Viv 1ND36B IEMXD-F028 Primary Bkr ND to Hot Legs Chk 1NI125, 1NI129 Backup Fuse Test Isol V1v 1NI1548, I CATAWBA - UNITS 1 AND 2 3/4 B-22 0 01
PROOF & REVIEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXD-F02C Primary Bkr Pressurizer Power-Operated Relief Backup Fuse Isol Viv 1NC318 1EMXD-F05A Primary Bkr Pressurizer Power-Operated Relief Backup Fuse Isol V1v 1NC358 IEMXD-F05B Primary Bkr Rx Bldg Drain Hdr Inside Cont Backup Fuse Isol Viv 1KC429B IEMXD-F05C 4
Primary Bkr NCDT Hx Cing Water Return Inside j Backup Fuse Isol Viv 1KC3328 1EMXD-F06A Primary Bkr NC Pump 1B Thermal Barrier Outlet Backup Fuse Isol V1v 1KC3648 IEMXD-F06B Primary Bkr NC Pumps Rtn Hdr Inside Cont Backup Fuse Isol Viv IKC4248 IEMXK-F01A o Primary Bkr UHI Check Viv Test Line Inside Backup Fuse Cont Isol Viv 1NI266A 1EMXK-F01B Primary Bkr Upper Cont Vent Units Return Backup Fuse Cont Isol Viv 1RN429A 1EMXK-F01C Primary Bkr Backup N2 to PORV INC34A From Backup Fuse Accum Tnk 1A Viv 1NI438A 1EMXK-F02A Primary Bkr NC Pump 1A Thermal Barrier Backup Fuse Outlet Isol Viv 1KC394A 1EMXK-F02B ' Primary Bkr Lower Cont Vent Units Return Backup Fuse Cont Isol Viv 1RN484A l l CATAWBA - UNITS 1 AND 2 3/4 8-23 0 l
PROOF & REVIEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXK-F02C Primary Bkr NV Supply to Pressurizer Viv Backup Fuse INV037A 1EMXK-F03A Primary Bkr S/G C Blowdown Line Sample Backup Fuse Inside Cont Isol Vlv 1NM210A ,
IEMXK-F04A Primary Bkr S/G A Upper Shell Sample Inside Backup Fuse Cont Isol Vlv 1NM187A 1EMXK-F048 Primary Bkr S/G A Blowdown Line Sample Backup Fuse Inside Cont Isol Vlv INM190A 1EMXK-F04C Primary Bkr S/G C Upper Shell Sample Backup Fuse Inside Cont Isol Viv INM207A 1EMXK-F06A Primary Bkr Hydrogen Skimmer Fan 1A Backup Fuse Inlet Viv IVX1A 1EMXK-F07C Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 1A 1EMXK-F09A Primary Bkr Accumulator 1A Discharge Isol Backup Fuse Viv 1NI54A 1EMXK-F09B Primary Bkr UHI Check Vlv Test Line Inside Backup Fuse Cont Isol Viv 1NI267A 1EMXK-F09C Primary Bkr NC Pump Oil Fill Header Backup Fuse Cont Isol Vlv INC196A, CATAWBA - UNITS 1 AND 2 3/4 8-24 l
i PROOI& REVIEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXK-F10A Primary Bkr Containment Air Return Damper Backup Fuse 1 ARF-D-2 1EMXK-F108 Primary Bkr VQ Fans Suction From Containment Backup Fuse Isol Viv IVQ2A ,
IEMXK-F10C Primary Bkr Cont Air Addition Containment Backup Fuse Isol Viv IVQ16A 1EMXK-F11A Primary Bkr Containment Air Return Fan Backup Fuse Motor IA 1EMXK-F118 Primary Bkr Hydrogen Skimmer Fan Motor 1A Backup Fuse IEMXL-F018 Primary Bkr Trn 8 Alternate Power to ND Backup Fuse Letdn Viv IND37A 1EMXL-F01C Primary Bkr NI Accum D Sample Line Inside Backup Fuse Cont Isol Viv INM818 1EMXL-F02A Primary Bkr NC Pump 10 Thermal Barrier Backup Fuse Outlet Isol Viv 1KC4138 1EMXL-F028 Primary Bkr Air Handling units Glycol Return Backup Fuse Cont Isol Viv INF2338 1EMXL-F02C Primary Bkr NI Accum C Sample Line Inside Backup Fuse Cont Isol Viv INM78B , IEMXL-F03A - Primary Bkr S/G D Blowdown Sample Line Inside Backup Fuse Cont Isol Viv INM220B CATAWBA - UNITS 1 AND 2 3/4 8-25
PR0bf & REVlEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT P10TECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXL-F03B Primary Bkr NI Accum A Sample Line Inside Backup Fuse Cont Isol Vlv 1NM72B IEMXL-F03C Primary Bkr NI Accum B Sample Line Inside Backup Fuse Cont Isol Viv INM75B ,
1EMXL-F04A Primary Bkr S/G B Upper Shell Sample Inside Backup Fuse Cont Isol Viv 1NM197B 1EMXL-F048 Primary Bkr S/G B Blowdown Sample Line Inside Backup Fuse Cont Isol Viv 1NM2008 1EMXL-F04C Primary Bkr S/G D Upper Shell Sample Inside Backup Fuse Cont Isol Viv 1NM217B 1EMXL-F06A Primary Bkr Hydrogen Skimmer Fan 1B Backup Fuse Inlet Viv IVX2B IEMXL-F06B Primary Bkr Backup N2 to PORV INC328 Backup Fuse from Accum Tnk 1B Viv 1NI439B IEMXL-F07C Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel IB 1EMXL-F09A Primary Bkr Accumulator ID Discharge Backup Fuse Isol Viv 1NI8CB 1EMXL-F10A Primary Bkr Containment Air Return Damper Backup Fuse 1 ARF-0-4 , CATAWBA - UNITS 1 AND 2 3/4 B-26
PROOF & REVIEW COPY
, TABLE 3.8-la (Continued) J UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXL-F10B Primary Bkr Reactor Vessel Head Vent Backup Fuse V1v 1NC2518 IEMXL-F10C Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse INC2528 ,
IEMXL-Fila Primary Bkr Containment Air Return Backup Fuse Fan Motor 18 IEMXL-F118 Primary Bkr Hydrogen Skimmer Fan Motor 1B Backup Fuse IEMXS-F01B Primary Bkr NC Pumps Seal Rtn Backup Fuse Inside Cont Isol Viv 1NV89A 1EMXS-F02A Primary Bkr NO Pump IB Suction from NC Backup Fuse Loop C V1v 1NO37A 1EMXS-F028 Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse INC250A 1EMXS-F03C Primary Bkr NO Pump 1A Suction from NC Backup Fuse Loop B V1v 1ND2A 1EMXS-F03D Primary Bkr Reactor Vessel Head Vent V1v Backup Fuse 1NC253A 1EMXS-F048 Primary Bkr S/G ID Blowdown Inside Cont Backup Fuse Isol V1v 1BB8A ,
~
IEMXS-F04C Primary Bkr S/G 18 Blowdown Inside Cont Backup Fuse Isol Viv 1BB19A QCT CATAWBA - UNITS 1 AND 2 3/4 8-27
, TABLE 3.8-la (Continued)
UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1EMXS-F05A Primary Bkr S/G 1A Blowdown Inside Cont Backup Fuse Isol Viv 1BB56A 1EMXS-F05B Primary Bkr S/G IC Blowdown Inside Cont Backup Fuse Isol Viv 1BB60A ,
IEMXS-F05C Primary Bkr Pzr Liquid Sample Line Inside Backup Fuse Cont Isol Viv 1NM3A 1EMXS-F06A Primary Bkr Pzr Steam Sample Line Inside Backup Fuse Cont Isol Viv INM6A 1EMXS-F06B Primary Bkr NC Hot Leg A Sample Line Backup Fuse Inside Cont Isol Viv 1NM22A 1EMXS-F06C Primary Bkr NC Hot Leg C Sample Line Backup Fuse Inside Cont Isol Viv 1NM25A IMXM-F01A Primary Bkr Reactor Coolant Pump Motor Backup Fuse Drain Tank Pump Motor 1MXM-F02A Primary Bkr NC Pump 1B Oil Lift Backup Fuse Pump Motor 1 1MXM-F02B Primary Bkr NC Pump IC Oil Lift Backup Fuse Pump Motor 1 1MXM-F03A Primary Bkr Ice Condenser Power Backup Fuse Transformer ICTIA , 1MXM-F038 - Primary Bkr Ice Condenser Air Handling Unit Backup Fuse 186 Fan Motor A & B g 'l CATAWBA - UNITS 1 AND 2 3/4 8-28
,-- . - - - _ = _ _ _ _ _
TABLE 3.8-la (Continued) PROOF & REVIEW COPY UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued)
IMXM-F03C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Holst Motor 1A 1MXM-F04D Primary Bkr Lighting Transformer ILR10 Backup Fuse 1MXM-F04E Primary Skr Lighting Transformer ILR13 Backup Fuse 1MXM-F05A Primary Bkr 175 Ton Polar Crane and 25 Backup Fuse Ton Aux Crane No. R013 and R015 1MXM-F05C Primary Bkr Upper Containment Weldi- Feeder Backup Fuse 1MXM-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A7 Fan Motor A & B 1MXM-F068 i o Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 188 Fan Motor A & B 1MXM-F06C 1 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A9 Fan Motor A & B IMXM-F06D l Primary Bkr Ice Condenser Air Handling l Backup Fuse Unit 1810 Fan Motor A & B 1MXM-F078 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A13 Fan Motor A,& B IMXM-F07C - Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1814 Fan Motor A & B CATAWBA - UNITS 1 AND 2 3/4 8-29
. _. . _ . - = . . - . - _. . - . - . .- - . - - . _ . - - .- .__ __ -
I t PROOF & REVIEW COPY TABLE 3.8-la (Continued) . UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES i DEVICE NUMBER & LOCATION SYSTEM POWERED
; 2. 600 VAC MCC (Continued)
IMXM-F08D Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Defrost Heater 1A IMXM-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A1 Fan Motor A & B , IMXM-F098 Primary Bkr Ice Condenser Air Handling 4 Backup Fuse Unit 182 Fan Motor A & B IMXM-F09C
) Primary Bkr Backup Fuse Ice Condenser Air Handling l Unit 1A3 Fan Motor A & B .i 1MXM-F03D j Primary Bkr Ice Condenser Air Handling i Backup Fuse Unit 184 Fan Motor A & B 1 IMXM-F10A ! Primary Bkr Containment Floor and Equipment j Backup Fuse Sump Pump' Motor 1A1 1
1MXM-F108 i Primary Skr Containment Floor and Equipment Backup Fuse Sump Pump Motor 181
- IMXN-F01F Primary Bkr Stud Tensioner Backup Fuse Hoist 18 1MXN-F02A j Primary Bkr NC Pump 1B Oil Lift Pump Motor 2 4 Backup Fuse
! IMXN-F02B Primary Bkr NC Pump IC Oil Lift Pump Motor 2 ' Backup Fuse , i
~
i 1MXN-F02E Primary Bkr Stud Tensioner Hoist 1C . Backup Fuse I g1 N CATAWBA - UNITS 1 AND 2 3/4 8-30 4
- . . , _ . ~ . - . . , , . . . + , , , . . _ _ . - , , - . - - , . - - , , , , - - - - .. - - - - . , , - - - , - - . .. ,_--,n- , . - . , , . , - ..n. .
w, - - . . -
PROOF & REVIEW COPY TABLE 3.8-la (Continued) . UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1MXN-F03A Primary Bkr Ice Condenser Power Transformer Backup Fuse ICT1B 1MXN-F03B Primary Bkr Ice Condenser Bridge Crane 1 Backup Fuse Crane No. R0n ,
INXN-F03E Primary Bkr Stud Tensioner Hoist 1A Backup Fuse 1MXN-F04D Primary Bkr Lighting Transformer ILR5 Backup Fuse IMXN-F04E Primary Bkr Lighting Transformer ILR6 Backup Fuse IMXN-F05A Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Defrost Heater 1B 1MXN-F05B Primary Bkr Ice Condenser Refrigeration Floor Backup Fuse Cool Pump Motor 1B 1MXN-F05C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 1B 1MXN-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 181 Fan Motor A & B ' 1MXN-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A2 Fan Motor A &,B IMXN-F06C ' Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 183 Fan Motor A & B q Q$6 l ocT CATAWBA - UNITS 1 AND 2 3/4 8-31 t 1
PROOF &~REYlEW CO TABLE 3 8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1MXN-F06D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A4 Fan Motor A & B 1MXN-F078 Primary Bkr Ice Condenser Air Handling j Backup Fuse Unit 185 Fan Motor A & B ,
1MXN-F07C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A6 Fan Motor A & B 1MXN-F08A - Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 187 Fan Motor A & B INXN-F088 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A8 Fan Motor A & B 1MXN-F08C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 189 Fan Motor A & B IMXN-F08D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A10 Fan Motor A & B IMXN-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1811 Fan Motor A & B 1MXN-F098 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A12 Fan Motor A & 2 1MXN-F09C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1B13 Fan Motor A,& B 1MXN-F090 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A14 Fan Motor A & B CATAWBA - UNITS 1 AND 2 3/4 8-32 0 01
TABLE 3.8-la (Continued) PROOF & REVIOV UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued)
IMXN-F10A Primary Skr Containment Floor and Equipment Backup Fuse Sump Pump Motor 1A2 1MXN-F108 Primary Skr Containment Floor and Equipment Backup Fuse Sump Pump Motor 182 IMXN-F10C Primary Bkr Incore Instrumentation Backup Fuse Sump Pump Motor 1 IMXN-F100 - Primary Bkr Ice Condenser Air Handling
- Backup Fuse Unit 1815 Fan Motor A & B 1MX0-F01A Primary Bkr Upper Containment Air Return Backup Fuse Fan Motor 1C 1MX0-F01B Primary Bkr Incore Instrument Tunnel Booster Backup Fuse Fan Motor 1A 1MX0-F028 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1A 1MX0-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit IC Fan Motor IMX0-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 1C Fan Motor 1MX0-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 1A ,
1MXP-F01A - Primary Bkr Upper Containment Return Backup Fuse Air Fan 18 gz 'l N CATAWBA - UNITS 1 AND 2 3/4 8-33
l l TABLE 3.8-la (Continued) REVlEW COPY UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1MXP-F018 Primary Bkr Incore Instrument Tunnel Backup Fuse Booster Fan Motor 18 IMXP-F028 Primary Bkr Control Rod Drive Vent
.., ,, Backup Fuse Fan Motor IB .
1MXP-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 18 Fan Motor IMXP-F04C Primary Bkr Upper Containment Ventilation i Backup Fuse Unit 18 Fan Motor 1MXP-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 1B IMXQ-FOIA Primary Bkr Upper Containment Return Backup Fuse Air Fan Motor IA 1MXQ-F01B Primary Bkr Incore Instrument Room Venti-Backup Fuse lation Unit 1A Fan Motor IMXQ-F02B Primary Skr Control Rod Drive Vent Fan Backup Fuse Motor 1C 2 ' 1MXQ-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 1A Fan Motor 1MXQ-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 1A Fan Motor . 1MXR-F01A Primary Bkr Upper Containment Return Air Backup Fuse Fan Motor 10 CATAWBA - UNITS 1 AND 2 3/4 8-34 N
i l l TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 1MXR-F01B Primary Bkr Incore Instrument Room Ventila-Backup Fuse tion Unit IB Fan Motor IMXR-F028 Primary Bkr Control Rod Drive Vent Backup Fuse Fan Motor 10 IMXR-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit ID Fan Motor IMXR-F04C ~
Primary Bkr Upper Containment Ventilation Backup Fuse Unit ID Fan Motor 1MXY-F02A Primary Bkr NC Pump 1A Oil Lift Pump Motor 1 Backup Fuse IMXY-F028 Primary Bkr - NC Pump 10 Oil Lift Pump Motor 1 Backup Fuse 1MXY-F03A
. Primary Bkr Reactor Coolant Drain Tank Pump Backup Fuse Motor 1A 1MXY-F03D Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Pump Motor 1A 1MXY-F054 Primary Bkr Lighting Transforme.r Backup Fuse 1LR8 IMXY-F05B Primary Bkr Lighting Transformer i
Backup Fuse ILR11 . 9 l OCT CATAWBA - UNITS 1 AND 2 3/4 8-35
PROOF & REVIEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued)
IMXY-F05C Primary Bkr Lighting Transformer Backup Fuse ILR14 1MXY-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A5 Fan Motor A & B , IMXY-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1All Fan Motor A & B 1MXY-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1812 Fan Motor A & B IMXY-F06D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A15 Fan Motor A & B 1MXY-F08A Primary Bkr Incore Drive Assembly Backup Fuse Motor 1A 1MXY-F088 Primary Bkr Incore Drive Assembly Backup Fuse Motor 1C 1MXY-F08C Primary Bkr Incore Drive Assembly Backup Fuse Motor IE IMXY-F08D ! Primary Bkr Lower Containment Auxiliary Backup Fuse Charcoal Filter Unit Fan Motor 1A 1MXZ-F02A Primary Bkr NC Pump 1A Oil Lift Pump Backup Fuse Motor 2 , 1MXZ-F028 - , Primary Bkr NC Pump 10 Oil Lift Pump Backup Fuse Motor 2 CATAWBA - UNITS 1 AND 2 3/4 8-36
4 PROOF & REVIEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued)
JMXZ-F03A Primary Bkr Reactor Coolant Drain Tank Backup Fuse Puen Motor 18 1MXZ-F04B Primary Bkr Lighting Transformer ILR1 Backup Fuse IMXZ-F04C Primary Bkr Lighting Transformer ILR2 Backup Fuse 1MXZ-F04D Primary Bkr Lighting Transformer ILR3 l Backup Fuse IMXZ-FOSA Primary Bkr Reactor Coolant Pump Jib a Backup Fuse Holst No. R019 TH R022 1MXZ-F05C Primary Bkr Lower Containment Auxiliary - Backup Fuse Charcoal Filter Unit Fan Motor IB ] IMXZ-F06A Primary Bkr Incore Drive Assembly Motor 18
- Backup Fuse i
1MXZ-F06B Primary Bkr Incore Drive Assembly Motor ID Backup Fuse IMXZ-F06C Primary Bkr Incore Drive Assembly Motor IF Backup Fuse 1MXZ-F078 Primary Bkr Lighting Transformer ILR4 Backup Fuse 1MXZ-F07C
- Primary Bkr 5 Ton Jib Crane in Containment Backup Fuse Crane No. R005 i
CATAWBA - UNITS 1 AND 2 3/4 8-37 l_ ___ . __ - _ _ _
TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES l DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued)
IMXZ-F070 Primary Bkr Reactor Cavity Manipulator Backup Fuse Crane No. R007 & R027 1MXZ-F08A Primary Bkr Steam Generator Drain Pump Backup Fuse Motor 1 , IMXZ-F08C Primary Bkr 15 Ton Equipment Access Hatch Backup Fuse Hoist Crane No. R009 1MXZ-F08D Primary Bkr Control Rod Drive 2 Ton Jib Backup Fuse Holst Crane No. R017 1MXZ-F08E Primary Bkr Reactor Side Fuel Handling Backup Fuse Control Console SMXG-F01C Primary Bkr Standby Makeup Pump Drain Isol Backup Fuse Viv INV876 SMXG-F05C Primary Bkr Pressurizer Heaters 28, 55 & 56 Backup Fuse SMXG-F06A Primary Bkr Standby Makeup Pump to Seal Backup Fuse Water Line Isol Viv 1NV877
- 3. 600 VAC Pressurizer Heater Power Panels PHP1A-F01A Primary Bkr Pressurizer Heaters Backup Fuse 1, 2, & 22 PHP1A-F018
~
Primary Bkr Pressurizer Heatars Backup Fuse 5, 6, & 27 ' CATAWBA - UNITS 1 AND 2 3/4 8-38
, TABLE 3.8-la (Continued)
UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 3. 600 VAC Pressurizer Heater Power Panels (Continued)
PHP1A-F01C Primary Bkr Pressurizer Heaters Backup Fuse 9, 10 & 32 PHP1A-F02C Primary Skr Pressurizer Heaters Backup Fuse 11, 12 & 35 , PHP1A-F02D Primary Bkr Pressurizer Heaters Backup Fuse 13, 14 & 37 PHP1A-F02E Primary Bkr Pressurizer Heaters Backup Fuse 17, 1B & 42 PHPIB-F01A Primary Bkr Pressurizer Heaters Backup Fuse 21, 47 & 48 PHP18-F01B Primary Bkr Pressurizer Heaters Backup Fuse 26, 53 & 54 PHPIB-F01C Primary Bkr Pressurizer Heaters Backup Fuse 31, 59 & 60 PHP18-F02C Primary Bkr Pressurizer Heaters Backup Fuse 36, 65 & 66 PHP18-F02D Primary Bkr Pressurizer Heaters Backup Fuse 41, 71 & 72 l
- PHP18-F02E Primary Bkr Pressurizer Heaters Backup Fuse 46, 77 & 78 .
~
PHP1C-F01A l Primary Bkr Pressurizer Heaters l Backup Fuse 7, 8 & 30 1 00T l CATAWBA - UNITS 1 AND 2 3/4 8-39
i PROOF & R5 VIEW COPY TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED 4 l 3. 600 VAC Pressurizer Heater Power Panels (Continued) PHPIC-F01B
- Primary Bkr Pressurizer Heaters '
Backup Fuse 19, 20 & 45 PHPIC-F01C Primary Bkr Pressurizer Heaters
... . Backup Fuse 24, 51 & 52 .
PHPIC-F010 Primary Bkr Pressurizer Heaters
- Backup Fuse 29, 57 & 58 l PHPIC-F02C
; Primary Bkr Pressurizer Heaters Backup Fuse 34, 63 & 64 PHPIC-F02D Primary Skr Pressurizer Heaters Backup Fuse 39, 69 & 70
- PHP1C-F02E Primary Bkr Pressurizer Heaters Backup Fuse 44, 75 & 76 PHP1D-F01A Primary Bkr Pressurizer Heaters j Backup Fuse 3, 4 & 25 PHP10-F01B Primary Bkr Pressurizer Heaters i Backup Fuse 15, 16 & 40 i
! PHP10-F01C l Primary Bkr Pressurizer Heaters Backup Fuse 23, 49 & 50 PHP10-F02C Primary Bkr Pressurizer Heaters Backup Fuse 33, 61 & 62 - PHP1D-F020 Primary Bkr Pressurizer Heaters Backup Fuse 38, 67 & 68 i l 00T 1 66 CATAWBA - UNITS 1 AND 2 3/4 8-40
sa PROOF & REVIEW COPY l TABLE 3.8-la (Continued) UNIT 1 CONTAINMENT PEllETRATION C0flDVCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 3. 600 VAC Pressurizer Heater Power Panels (Continued)
PHPlD-F02E Primary Bkr Pressurizer Heaters Backup Fuse 43, 73 & 74
- 4. 250 VDC Reactur Building Deadlight Panelboard
-' IDLD-2 ,
Primary Bkr Lighting Panelboard No. 1LR1, Backup Fuse ILR2, ILR3, ILR4 1DLD-3 Primary Bkr Lighting Panelboard No. 1LR13, Backup Fuse ILR14 1DLD-4 Prim?.ry Bkr Lighting Panelboard No. ILRS, Backup Fuse ILR6 1DLD-5 Primary Bkr Lighting Panelboard No. ILR10, Backup Fuse ILR11 1DLD-10 Primary Bkr Lighting Panelboard No. ILR8 Backup Fuse
- 5. 120 VAC Panelboards 1ELB-5 Primary Bkr Emergency A.C. Lighting Backup Fuse IELB-7 Primary Bkr Emergency A.C. Lighting Backup Fuse IELB-13 Primary Bkr Emergency A.C. Lighting Backup Fuse .
1ELB-15 Primary Bkr Emergency A.C. Lighting Backup Fuse g i DN CATAWBA - UNITS 1 AND 2 3/4 8-41
TABLE 3.8-la (Continued) REVIEW COPY UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PPOTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 5. 120 VAC Pane 1 boards (Continued) 1ELB-17 Primary Bkr Emergency A.C. Lighting Backup Fuse 1KPM-1 Primary Bkr NC Pump Motor IA Space Heater Backup Fuse .
IKPM-2 Primary Bkr NC Pump Motor 1C Space Heater. Backup Fuse 1KPM-7-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 1A Fan Motor Space Heater 1KPM-8-1 Primary Bkr Lower Containment Vent Unit Backup Fuse IC Fan Motor Space Heater 1XPM-24-1 Primary Skr Control Rod Drive Vent Fan Backup Fuse Motor 1A Space Heater 1XPM-24-2 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1B Space Heeter IKPM-24-3 l Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1C Space Heater 1KPM-24-4 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 10 Space Heater IKPM-33-3, 4, 5, 6, 7 Primary Bkr Safety Injection' System Backup Fuse Temperature Transtaitters 1KPN-1 Primary Bkr NC Pump Motor 1B Space lieater i Backup Fuse OCT 7N CATAWBA - UNITS 1 AND 2 3/4 8-42
3 TABLE 3.8-la (Continued) PROOF & REVIEW COPY UNIT 1 CONTAINMENT PEfjETRATION CONDUCTOR OVERCilRRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED '
- 5. 120 FAC Pane 1 boards (Continued) 1KPN-2 Primary Bkr NC Pump Motor ID Space Heater Backup Fuse 1KPN-7-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 18 Fan Motor Space Heater 1XPN-8-1 Primary Bkr Lower Containment Vent Unit Backup Fuse ID Fan Motor Space Heater 1KPN-11 Primary Bkr Misc Control Power Backup Fuse for 1ATC 24 O
CATAWBA - UNITS 1 AND 2 3/4 8-43
TABLE 3.8-1b PROOF & RWG m UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 1. 6900 VAC Swgr Primary Bkr RCP2A Reactor Coolant Pump 2A Backup Bkr 2TA-3 Primary Bkr RCP2B Reactor Coolant Pump 28 Backup Bkr 2TB-3 Primary BKR RCP2C Reactor Coolant Pump 2C Backup Bkr 2TC-3 Primary BKR RCP2D Reactor Coolant Pump 2D Backup Bkr 2TD-3
- 2. 600 VAC MCC 2EMXC-F01B '
Primary Bkr Accumulator 2C Discharge Backup Fuse Isol Viv 2NI76A I 2EMXC-F01C Primary Bkr Check Valve Test Header Backup Fuse Cont Isol Vlv 2NI95A 2EMXC-F02A Primary Bkr Train A Alternate Power Backup Fuse To ND LTDN Viv 2ND1B 2EMXC-F02B Primary Bkr Hot Leg Inj. Check Viv Backup Fuse Test Isol Vlv 2NI153A 2EMXC-F02C Primary Bkr Cont Isol at 134 Deg Backup Fuse Annulus Area V1v 2VI312A 2EMXC-F03A Primary Bkr a Pump 2C Thermal Barrier Outlet Backup Fuse 'aol Viv 2KC345A 2EMXC-F03B Primary Bkr N2 to Prt Cont Isol Inside
. Backup Fuse Viv 2NC54A 2EMXC-F03C Primary Bkr Pressurizer Power-Operated Backup Fuse Relief Isol Viv 2NC33A CATAWBA - UNITS 1 AND 2 3/4 8-44
REVIEW COPY TABLE 3.8-Ib (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2EMXC-F05A Primary Bkr NCDT Vent Inside Cont Isol Backup Fuse Viv 2WL450A 2EMXC-F05B Primary Bkr Cont Sump Pumps Discharge Inside Backup Fuse Cont Isol Viv 2WL825A ,
2EMXC-F05C Primary Bkr Vent Unit Cond Drn Tank Backup Fuse Outside Cont Isol Viv 2WL867A 2EMXC-F06A Primary Bkr NCDT Pumps Disch Inside Cont Isol Backup Fuse Viv 2WL805A 2EMXC-F078 Primary Bkr Cont H2 Purge Outlet Cont Isol Backup Fuse Vlv 2VY17A 2EMXD-F01A Primary Bkr ND Pump 2A Suction From NC Backup Fuse Loop B Viv 2ND1B 2EMXD-F01B Primary Bkr Accumulator 28 Discharge Backup Fuse Isol Viv 2NI65B 2EMXD-F01C Primary Bkr N1 Pump A to Hot leg Check Backup Fuse Viv Test Isol Viv 2 nil 228 2EMXD-F02A Primary Bkr ND Pump 28 Suction from NC Backup Fuse Loop C Viv 2ND36B 2EMXD-F028 Primary Bkr ND to Hot Legs Chk 2NI125, 2NI129 Backup Fuse Test Isol Viv aNIl548, 1 OCT CATAWBA - UNITS 1 AND 2 3/4 8-45
TABLE 3.8-1b (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES l DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2EMXD-F02C Primary Bkr Pressurizer Power-Operated Relief
- . Backup Fuse Isol Viv 2NC31B 2EMXD-F05A Primary Bkr Pressurizer Power-Operated Relief Backup Fuse Isol V1v 2NC35B 2EMXD-F058 Primary Bkr Rx Bldg Drain Hdr Inside Cont Backup Fuse Isol Viv 2KC4298 2EMXD-F05C Primary Bkr NCDT Hx Cing Water Return Inside Backup Fuse Isol Viv 2KC332B 2EMXD-F06A Primary Bkr NC Pump 28 Thermal Barrier Outlet Backup Fuse Isol Viv 2KC364B 2EMXD-F06B Primary Bkr NC Pumps Rtn Hdr Inside Cont Backup Fuse Isol Viv 2KC4248 2EMXK-F01A Primary Bkr UHI Check Viv Test Line Inside l Backup Fuse Cont Isol Viv 2NI266A 2EMXK-F01B Primary Bkr Upper Cont Vent Units Return Backup Fuse Cont Isol Viv 2RN429A 2EMXK-F01C
} Primary Bkr Backup N2to PORV 2NC34A From Backup Fuse Accum Tnk 2A Viv 2NI438A t ]' 2EMXK-F02A Primary Bkr NC Pump 2A Thermal Barrier Backup Fuse Outlet Isol Vlv 2KC394A 2EMXK-F028 4 Primary Bkr Lower Cont. Vent Units Return Backup Fuse Cont Isol Viv 2RN484A f CATAWBA - UNITS 1 AND 2 3/4 8-46 l 1
- , - - - , - - , ---c , ,-,, --,.,,,,.-r n.,-.- .,,..-n-. - - , .
,.,,,.n---------------..,.--.--.,-.-m , , - - -
-.m,n,------ a
PROOF g gEVIEW COPY TABLE 3.8-lb (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2EMXX-F02C Primary Bkr NV Supply to Pressurizer Vlv Backup Fuse 2NV037A 2EMXK-F03A Primary Bkr S/G C Blowdown Line Sample Backup Fuse Inside Cont Isol Viv 2NM210A .
2EMXK-F04A Primary Bkr S/G A Upper Shell Sample Inside Backup Fuse Cont Isol Viv 2NM187A 2EMXK-F04B Primary Bkr S/G A Blowdown Line Sample Backup Fuse Inside Cont Isol Viv 2NM190A 2EMXK-F04C Primary Bkr S/G C Upper Shell Sample Backup Fuse Inside Cont Isol Viv 2NM207A 2EMXK-F06A Primary Bkr Hydrogen Skimmer Fan 2A Backup Fuse Inlet Viv 2VX1A 2EMXK-F07C . Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 2A 2EMXK-F09A Primary Bkr Accumulator 2A Discharge Isol Backup Fuse Viv 2NI54A 2EMXK-F098 Primary Bkr UHI Check Viv Test Line Inside Backup Fuse Cont Isol Viv 2NI267A 2EMXK-F09C Primary Bkr NC Pump 011 Fill Header Backup Fuse Cont Isol Viv 2NC196A. CATAWBA - UNITS 1 AND 2 3/4 8-47
_ . _..__ _ ~ . TABLE 3.8-lb (Continued) PROOF & REVIEW UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2EMXK-F10A Primary Bkr Containment Air Return Damper Backup Fuse 2 ARF-D-2 2EMXK-F108 Pri: nary Bkr VQ Fans Suction From Containment Backup Fuse Isol Viv 2VQ2A ,
2EMXK-F10C Primary Bkr Cont Air Addition Containment Backup Fuse Isol V1v 2VQ16A 2EMXK-F11A Primary Bkr Containment Air Return Fan Backup Fuse Motor 2A 2EMXK-F118 Primary Bkr Hydrogen Skimmer Fan Motor 2A Backup Fuse 2EMXL-F018 Primary Bkr Trn 8 Alternate Power to ND Backup Fuse Letdn Viv 2NO37A 2EMXL-F01C Primary Bkr NI Accum D Sample Line Inside Backup Fuse Cont Isol Viv 2NM818 2EMXL-F02A Primary Bkr NC Pump 20 Thermal Barrier Backup Fuse Outlet Isol Vlv 2KC413B 2EMXL-F028 Primary Bkr Air Handling units Glycol Return Backup Fuse Cont Isol Vlv 2NF2338 2EMXL-F02C Primary Bkr NI Accum C Sample Line Inside Backup Fuse Cont Isol Viv 2NM788 , 2EMXL-F03A
- Primary Bkr S/G 0 Blowdown Sample Line Inside Backup Fuse Cont Isol Viv 2NM2208 g 7 %%5 CATAWBA - UNITS 1 AND 2 3/4 8-48
TABLE 3.8-1b (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES l DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2EMXL-F03B Primary Bkr NI Accum A Sample Line Inside Backup Fuse Cont Isol Viv 2NM72B 2EMXL-F03C Primary Bkr NI Accum B Sample Line Inside Backup Fuse Cont Isol Viv 2NM75B 2EMXL-F04A Primary Bkr S/G B Upper Shell Sample Inside Backup Fuse Cont Isol Viv 2NM1978 2EMXL-F04B Primary Bkr S/G B Blowdown Sample Line Inside Backup Fuse Cont Isol Vlv 2NM200B 2EMXL-F04C Primary Bkr S/G D Upper Shell Sample Inside Backup Fuse Cont Isol Viv 2NM217B 2EMXL-F06A Primary Bkr Hydrogen Skimmer Fan 28 Backup Fuse Inlet Viv 2VX2B 2EMXL-F068 Primary Bkr Backup N2to PORV 2NC32B Backup Fuse from Accum Tnk 2B Viv 2NI439B 2EMXL-F07C 1 Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 28 2EMXL-F09A Primary Bkr Accumulator 2D Discharge Backup Fuse Isol Viv 2NI88B 2EMXL-F10A Primary Bkr Containment Air Return Damper Backup Fuse 2 ARF-D-4 0 01 CATAWBA - UNITS 1 AND 2 3/4 8-49
f i TABLE 3.8-Ib (Continued) PRO 0f&RnIEWCO UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED i a 2. 600 VAC MCC (Continued) 2EMXL-F10B Primary Bkr Reactor Vessel Head Vent Backup Fuse V1v 2NC251B
, 2EMXL-F10C l Primary Bkr Reactor Vessel Head Vent V1v
- Backup Fuse 2NC2528 2EMXL-F11A i Primary Bkr Containment Air Return Backup Fuse Fan Motor 28 2EMXL-F118 Primary Bkr Hydrogen Skimmer Fan Motor 2B l Backup Fuse 2EMXS-F01B 3 Primary Bkr NC Pumps Seal Rtn
. Backup Fuse Inside Cont Isol V1v 2NV89A
!l 2EMXS-F02A l Primary Bkr NO Pump 28 Suction from NC Backup Fuse Loop C Viv 2ND37A l 2EMXS-F02B Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 2NC250A
, 2EMXS-F03C Primary Bkr ND Pump 2A Suction from NC i Backup Fuse Loop B V1v 2N02A
! 2EMXS-F030 { Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 2NC253A 2EMXS-F048 Primary Bkr S/G 2D Blowdown Inside Cont
; Backup Fuse Isol Viv 2BB8A , 2EMXS-F04C
!' Primary Bkr S/G 2B Blowdown Inside Cont Backup Fuse Isol Viv 2BB19A OCT 7# l CATAWBA - UNITS 1 AND 2 3/4 8-50 l l t
COPY TABLE 3.8-lb (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES
~
DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2EMXS-F05A Primary Bkr S/G 2A Blowdown Inside Cont Backup Fuse Isol Vlv 2BB56A 2EMXS-F05B Primary Ekr S/G 2C Blowdown Inside Cont Backup Fuse Isol Viv 2BB60A .
2EMXS-F05C Primary Bkr Pzr Liquid Sample Line Inside Backup Fuse Cont Isol Vlv 2NM3A 2EMXS-F06A Primary Bkr Pzr Steam Sample Line Inside Backup Fuse Cont Isol Viv 2NM6A 2EMXS-F068 Primary Bkr NC Hot Leg A Sample Line Backup Fuse Inside Cont Isol Viv 2NM22A 2EMXS-F06C Primary Bkr NC Hot Leg C Sample Line Backup Fuse Inside Cont Isol Viv 2NM25A 2MXM-F01A Primary Bkr Reactor Coolant Pump Motor Backup Fuse Drain Tank Pump Motor 2MXM-F02A Primary Bkr NC Pump 1B Oil Lift Backup Fuse Pump Motor 1 2MXM-F028 Primary Bkr NC Pump 2C Oil Lift Backup Fuse Pump Motor 1 2MXM-F03A Primary Bkr Ice Condenser Power Backup Fuse Transformer ICT2A .
~
2MXM-F03B Primary Bkr Ice Condenser Air Hanuling Unit Backup Fuse 286 Fan Motor A & B I 0 01 CATAWBA - UNITS 1 AND 2 3/4 8-51
PROOF k REVIEW C TABLE 3.8-1b (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXM-F03C Primary Skr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 2A 2MXM-F04D Primary Bkr Lighting Transformer 2LR10 Backup Fuse .
2MXM-F04E Primary Bkr Lighting Transformer 2LR13 Backup Fuse 2MXM-F05A Primary Bkr 175 Ton Polar Crane and 25 Backup Fuse Ton Aux Crane No. R014 and R016 2MXM-F05C Primary Bkr Upper Containment Welding Feeder Backup Fuse 2MXM-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A7 Fan Motor A & B 2MXM-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 288 Fan Motor A & B 2MXM-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A9 Fan Motor A & B
- 2MXM-F060 i
Primary Bkr Ice Condenser Air Handling i Backup Fuse Unit 2810 Fan Motor A & B 2MXM-F078 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A13 Fan Motor A.& B l 2MXM-F07C , Primary Bkr Ice Condenser Air Handling l Backup Fuse Unit 2B14 Fan Motor A & B lb CATAWBA - UNITS 1 AND 2 3/4 8-52 l
TABLE 3.8-1b (Continued) F k REVIEW C ' UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXM-F080 Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Defrost Heater 2A 2MXM-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A1 Fan Motor A & B .
2MXM-F098 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 282 Fan Motor A & B 2MXM-F09C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A3 Fan Motor A & B 2MXM-F090 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 284 Fan Motor A & B 2MXM-F10A Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 2A1 2MXM-F10B
. Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 2B1 2MXN-F01F Primary Bkr Stud Tensioner Backup Fuse Hoist 28 2MXN-F02A Primary Bkr NC Pump 28 Oil Lift Pump Motor 2 Backup Fuse 2MXN-F028 Primary Bkr NC Pump 2C Oil Lift Pump Motor 2 Backup Fuse .
2MXN-F02E Primary Bkr Stud Tensioner Hoist 2C Backup Fuse g 885 g
- CATAWBA - UNITS 1 AND 2 3/4 B-53
_ . . ~ . s 00F& R&ty(gpy TABLE 3.8-Ib (Continued) I UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXN-F03A Primary Bkr Ice Condenser Power Transformer Backup Fuse ICT2B 2MXN-F03B Primary Bkr Ice Condenser Bridge Crane 2 j .. , ,
Backup Fuse Crane No. R012 . 1 i 2MXN-F03E Primary Bkr Stud Tensioner Hoist 2A Backup Fuse 2MXN-F04D Primary Skr Lighting Transformer 2LR5 Backup Fuse 2MXN-F04E Primary Bkr Lighting Transformer 2LR6 Backup Fuse 2MXN-F05A Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Defrost Heater 28 2MXN-F05B Primary Bkr Ice Condenser Refrigeration Floor Backup Fuse Cool Pump Motor 28 2MXN-F05C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 28 2MXN-F06A Primary Bkr Ice Condenser Air Handling a Backup Fuse Unit 281 Fan Motor A & B 2MXN-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A2 Fan Motor A & B 2MXN-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 283 Fan Motor A & B
)
i , ocT 7 8B5 i CATAWBA - UNITS 1 AND 2 3/4 8-54
----n -,- . - - . . . - , - -
. ,,.- ,.,-.- - _ , - - _..n.-- . - - - - . . - . . . _ -.. ~ ~- --- . - -._.--- - ,.. .- ., -.,--- - -.- ,- - ,.--.
PRO 0i& REVIEW COPY TABLE 3.8-Ib (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) l 2MXN-F060 i
Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A4 Fan Motor A & B 2MXN-F078 Primary Bkr Ice Condenser Air Handling m , Backup Fuse Unit 285 Fan Motor A & B , 2MXN-F07C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A6 Fan Motor A & B 2MXN-F08A Primary Skr Ice Condenser Air Handling Backup Fuse Unit 287 Fan Motor A & B 2MXN-F08B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A8 Fan Motor A & B 2MXN-F08C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 289 Fan Motor A & B 2MXN-F08D ' Primary Bkr Ice Condenser Air Handling
- Backup Fuse Unit 2A10 Fan Motor A & B 2MXN-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B11 Fan Motor A & B '
2MXN-F09B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A12 Fan Motor A & B 2MXN-F09C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2813 Fan Motor A.& B 2MXN-F090 Primary Bkr Ice Condenser Air Handling i Backup Fuse Unit 2A14 Fan Motor A & B 001 7 1985 CATAWBA - UNITS 1 AND 2 3/4 8-55
TABLE 3.8-Ib (Continued) PROOF & REVIEW C UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXN-F10A Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 2A2 2MXN-F10B Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 282 .
2MXN-F10C Primary Bkr Incore Instrumentation Backup Fuse Sump Pump Motor 2 2MXN-F100 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2815 Fan Motor A & B 2MX0-F01A Primary Bkr Upper Containment Air Return Backup Fuse Fan Motor 2C 2MX0-F028 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2A 2MX0-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 2C Fan Motor 2MX0-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 2C Fan Motor 2MX0-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 2A 2MXP-F01A Primary Bkr Upper Containment Return Backup Fuse Air Fan 2B . OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 8-56
1 l l . PROOF & REVIEW COPY TABLE 3.8-lb (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXP-F028 Primary Bkr Control Rod Drive Vent Backup Fuse Fan Motor 28 2MXP-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 28 Fan Motor .
2MXP-F04C Primary Bkr Upper Containment Ventilation
; Backup Fuse Unit 28 Fan Motor 2MXP-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 2B 2MXQ-F01A Primary Bkr Upper Containment Return Backup Fuse Air Fan Motor 2A 2MXQ-F018 Primary Bkr Incore Instrument Room Venti-Backup Fuse lation Unit 2A Fan Motor 2MXQ-F02B Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2C 2MXQ-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 2A Fan Motor 2MXQ-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 2A Fan Motor 2MXR-F01A Primary Bkr Upper Containment Return Air Backup Fuse Fan Motor 2D .
1 66 CATAWBA - UNITS 1 AND 2 3/4 8-57 1
TABLE 3.8-Ib (Continued) PROOF 5 REVIEW CO UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXR-F018 Primary Bkr Incore Instrument Room Ventila-Backup Fuse tion Unit 28 Fan Motor 2MXR-F02B Primary Bkr Control Rod Drive Vent Backup Fuse Fan Motor 2D 2MXR-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 2D Fan Motor 2MXR-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 20 Fan Motor 2MXY-F02A Primary Bkr NC Pump 2A Oil Lift Pump Motor 1 Backup Fuse 2MXY-F02B Primary Bkr NC Pump 20 Oil Lift Pump Motor 1 Backup Fuse 2MXY-F02C Primary Bkr Reactor Building Lower Containment Backup Fuse Welding Machine Receptacle 2RCPL0185
~
2MXY-F03A Primary Bkr Reactor Coolant Drain Tank Pump Backup Fuse Motor 2A 2MXY-F030 Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Pump Motor 2A 2MXY-F05A Primary Bkr Lighting Transformer Backup Fuse 2LR8 2MXY-F058 Primary Bkr Lighting Transformer Backup Fuse 2LR11 0 01 CATAW8A - UNITS 1 AND 2 3/4 8-58
, - - . , - - . . - - - - ._- - , - , . . . , , - - - - , . . , - - ,_ , - - - , , - - . ., ,.n. - _ , , , , - - - .
PROOF & REVIEW COP TABLE 3.8-lb (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXY-F05C Primary Bkr Lighting Transformer Backup Fuse 2LR14 2MXY-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A5 Fan Motor A & B ,
2MXY-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2All Fan Motor A & B 2MXY-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B12 Fan Motor A & B 2MXY-F060 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A15 Fan Motor A & B 2MXY-F07C Primary Bkr EXH Reactor Building Equipment Backup Fuse Hatch Jib Cranes R035 and R036 2MXY-F08A Primary Bkr Incore Drive Assembly Backup Fuse Motor 2A 2MXY-F088 Primary Bkr Incore Drive Assembly Backup Fuse Motor 2C i 2MXY-F08C i Primary Bkr Incore Drive Assembly Backup Fuse Motor 2E 2MXY-F080 , Primary Bkr Lower Containment Auxiliary Backup Fuse Charcoal Filter Unit Fan Motor 2A 2MXZ-F02A
- Primary Bkr NC fump 2A Oil Lift Pump Backup Fuse Motor 2 g 1N CATAWBA - UNITS 1 AND 2 3/4 8-59
l TABLE 3.8-2 (Continued) REVI W COPY UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXZ-F02B Primary Bkr NC Pump 20 Oil Lift Pump Backup Fuse Motor 2 2MXZ-F03A Primary Bkr Reactor Coolant Drain Tank Backup Fuse Pump Motor 28 2MXZ-F04B Primary Bkr Lighting Transformer 2LR1 Backup Fuse 2MXZ-F04C Primary Bkr Lighting Transformer 2LR2 Backup Fuse 2MXZ-F040 Primary Bkr Lighting Transformer 2LR3 Backup Fuse 2MXZ-F05A Primary Bkr Reactor Coolant Pump Jib Backup Fuse Hoist No. R023 TH R026 2MXZ-F05C Primary Bkr Lower Containment Auxiliary Backup Fuse Charcoal Filter Unit Fan Motor 2B
~
2MXZ-F06A Primary Bkr Incore Drive Assembly Motor 28 Backup Fuse 2MXZ-F06B Primary Bkr Incore Drive Assembly Motor 20 Backup Fuse 2MXZ-F06C Primary Bkr Incore Drive Assembly Motor 2F Backup Fuse 2MXZ-F07B Primary Bkr Lighting Transformer 2LR4 Backup Fuse 1 bu3 CATAWBA - UNITS 1 AND 2 3/4 8-60 l l
EW COPY TABLE 3.8-1b (Continued) EIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEV'.E NUMBER & LOCATION SYSTEM POWERED
- 2. 600 VAC MCC (Continued) 2MXZ-F07C Primary akr 5 Ton Jib Crane in Containment Backup Fuse Crane No. R006 2MXZ-F070 Primary Bkr Reactor Cavity Manipulator
. _ , ,, Backup Fuse Crane No. R008 & R028 ,
2MXZ-F08A Primary Bkr Steam Generator Drain Pump Backup Fuse Motor 2 2MXZ-F08C Primary Bkr 15 Ton Equipment Access Hatch Backup Fuse Hoist Crane No. R010 2MXZ-F08D Prima y Bkr Control Rod Drive 2 Ton Jib Backup Fuse Hoist Crane No. R018 2MXZ-F08E Primary Bkr Reactor Side Fuel Handling Backup Fuse Control Console SMXG-F06B Primary Bkr Standby Makeup Pump Drain Isol Backup Fuse Viv 2NV876 SMXG-R05B Primary Bkr Pressurizer Heaters 28, 55 & 56 Backup Fuse SMXG-F06C Primary Bkr Standby Makeup Pump to Seal Backup Fuse Water Line Isol V1v 2NV877 1
- 3. 600 VAC Pressurizer Heater Power Panels
~
PHP2A-F01A
~
Primary Bkr Pressurizer Heaters
- Backup Fuse 1, 2, & 22
- PHP2A-F01B Primary Bkr Pressurizer Heaters Backup Fuse 5, 6, & 27 g
- CATAWBA - UNITS 1 AND 2 3/4 8-61
TABLE 3.8-1b (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 3. 600 VAC Pressurizer Heater Power Panels (Continued)
PHP2A-F01C Primary Bkr Pressurizer Heaters Backup Fuse 9, 10 & 32 PHP2A-F02C Primary Bkr Pressurizer Heaters
... ,, Backup Fuse 11, 12 & 35 .
PHP2A-F02D Primary Bkr Pressurizer Heaters . Backup Fuse 13, 14 & 37 i PHP2A-F02E Primary Bkr Pressurizer Heaters Backup Fuse 17, 18 & 42 i PHP28-F01A Primary Bkr Pressurizer Heaters Backup Fuse 21, 47 & 48 PHP28-F01B Primary Bkr Pressurizer Heaters j Backup Fuse 26, 53 & 54
, PHP28-F01C l o Primary Bkr Pressurizer Heaters Backup Fuse 31, 59 & 60 PHP28-F02C . Primary Bkr Pressurizer Heaters i
Backup Fuse 36, 65 & 66 , PHP28-F020 Primary Bkr Pressurizer Heaters Backup Fuse 41, 71 & 72 PHP2B-F02E Primary Bkr Pressurizer Heaters Backup Fuse 46, 77 & 78 . PHP2C-F01A l Primary Bkr Pressurizer Heaters Backup Fuse 7, 8 & 30 00s 1 $6 CATAWBA - UNITS 1 AND 2 3/4 8-62
4 PROOF & REVIEW CO i TABLE 3.8-lb (Continued)
- UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 3. 600 VAC Pressurizer Heater Power Panels (Continued) f PHP2C-F01B
; Primary Bkr Pressurizer Heaters Backup Fuse 19, 20 & 45
, PHP2C-F01C Primary Bkr Pressurizer Heaters Backup Fuse 24, 51 & 52 , PHP2C-F010
) Primary Bkr Pressurizer Heaters Backup Fuse 29, 57 & 58 .
PHP2C-F02C i Primary Bkr Pressurizer Heaters ), Backup Fuse 34, 63 & 64 PHP2C-F020 j Primary Bkr Pressurizer Heaters
- Backup Fuse 39, 69 & 70 ,
) , PHP2C-F02E Primary Bkr Pressurizer Heaters Backup Fuse 44, 75 & 76 i
- PHP2D-F01A Primary Bkr Pressurizer Heaters Backup Fuse 3, 4 & 25 PHP20-F01B Primary Bkr Pressurizer Heaters Backup Fuse 15, 16 & 40 PHP2D-F01C
! Primary Bkr Pressurizer Heaters Backup Fuse 23, 49 & 50 PHP20-F02C Primary Bkr Pressurizer Heaters , Backup Fuse 33, 61 & 62 -
1 PHP20-F020 - Primary Bkr Pressurizer Heaters Backup Fuse 38, 67 & 68 OCT 7# l CATAWBA - UNITS 1 AND 2 3/4 8-63 \
, - ~ . - - - - . . , _ . . _ . _ _ . . . . . . - , _ . _ - - , - . . . . _ . . _ _ _ _ _ . . . . - _ . - . . _ . - - . _ , _ - . . , . - _ . - - - . - . - _ . _ . - . _ . _ - . - -
TABLE 3.8-lb (Continued) i UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 3. 600 VAC Pressurizer Heater Power Panels (Continued)
PHP20-F02E i Primary Bkr Pressurizer Heaters Backup Fuse 43, 73 & 74 i
- 4. 250 VOC Reactor Building Deadlight Panelboard
. ' 20LO-2 Primary Bkr Lighting Panelboard No. 2LR1, Backup Fuse 2LR2, 2LR3, 2LR4 20LD-3 ; Primary Bkr Lighting Panelboard No. 2LR13, Backup Fuse 2LR14 20LD-4 Primary Bkr Lighting Panelboard No. 2LRS, Backup Fuse 2LR6 20LD-5 Primary Bkr Lighting Panelboard No. 2LR10, Backup Fuse 2LR11 20LD-10 Primary Bkr Lighting Panelboard No. 2LR8 Backup Fuse e
- 5. 120 VAC Panelboards 2ELB-5 Primary Bkr Emergency A.C. Lighting Backup Fuse 2ELB-7 Primary Bkr Emergency A.C. Lighting Backup Fuse 2ELB-13 Primary Bkr Emergency A.C. Lighting Backup Fuse 2ELB-15 "
Primary Bkr Emargency A.C. Lighting Backup Fuse gqi 7 1985 CATAWBA - UNITS 1 AND 2 3/4 8-64 i
--. . - ~ , . , - , - , , . . - . . - . - - - . - , , . - , _ - - - . . _ , - - , - . - - - - , . - . . - . . - . - ~ - - . . , , - - - - -
PROOF & REVIEW COPY TABLE 3.8-lb (Continued) UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED
- 5. 120 VAC Pane 1 boards (Continued) 2ELB-17 Primary Bkr Emergency A.C. Lighting Backup Fuse 2KPM-1 Primary Bkr NC Pump Motor 2A Space Heater Backup Fuse .
2KPM-2 Primary Bkr NC Pump Motor 2C Space Heater Backup Fuse 2KPM-7-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 2A Fan Motor Space Heater 2KPM-8-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 2C Fan Motor Space Heater 2KPM-24-1 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2A Space Heater 2KPM-24-2 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 28 Space Heater 2KPM-24-3 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2C Space Heater 2KPM-24-4 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2D Space Heater 2KPM-33-3, 4, 5, 6, 7 Primary Bkr Safety Injection System Backup Fuse Temperature Transmitters
~
2KPN-1 Primary Bkr NC Pump Motor 28 Space Heater Backup Fuse i OCT 70 CATAWBA - UNITS 1 AND 2 3/4 8-65
,- -- - , , , - , ~ - - . - , - - + , , - - , r--,,,.y-., . , . _ - .-,-u -. . - - - - - , . , _ _ - - - - .-- .-w--.-ru
-- - - + , - - - - - , . - -
PROOF & REVIEW COPY l TABLE 3.8-Ib (Continued) l i UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES. j DEVICE NUMBER & LOCATION SYSTEM POWERED I 5. 120 VAC Panelboards (Continued) 2KPN-2 Primary Bkr NC Pump Motor 2D Space Heater Backup Fuse
- 2KPN-7-1 Primary Bkr Lower Containment Vent Unit l Backup Fuse 28 Fan Motor Space Heater 1 2KPN-8-1 ,
Primary Bkr Lower Containment Vent Unit Backup Fuse 20 Fan Motor Space Heater 2KPN-11 Primary Bkr Misc Control Power
- Backup Fuse for 2ATC 24
- 6. DC Welding Circuits 2EQC80001
; Primary Bkr - AA Iower Containment Backup Bkr - AB DC Welding Circuit 2EQCB0002 ,
Primary Bkr - AA Upper Containment i Backup Skr - AB DC Welding Circuit i 1 d i D l , oss CATAWBA - UNITS 1 AND 2 3/4 8-66
PROOF & REVIEW COPY 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION F01 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 either:
- 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 requirenents 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 of a solution containing greater than or equal to 7000 ppm boron or its equivalent until K is reduced to less than or equal to 0.95 or the boron concentrationisrINoredtogreaterthanorequalto2000 ppm,whicheveris 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.
4.9.1.0 tie boron corcentration of the Reactor Coolant System and the refueling canal 6 il be determined by chemical analysis at least once per 72 hours.
- 4. 9.1. 3 Valves NV-231, NV-237, NV-240, NV-241, and NV-244 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.
- The reactor shall be maintained in HODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
CATAWBA - UNITS 1 AND 2 3/4 9-1 g5 M g 7 95
REFUELING OPERATIONS PROOF & RNU # 3/4.9.2 INSTRUMENTATION j LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE i and operating with Alarm Setpoints at 0.5 decade above steady-state count rate, 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.
. 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 within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
- g 18 CATAWBA - UNITS 1 AND 2 3/4 9-2 001 72
PROOF & REVIE'.7 COPY REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least 72 hours. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION: With the reactor subcritical for less than 72 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 72 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. g 18 CATAWBA - UNITS 1 AND 2 3/4 9-3 007 71885 1
REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS T LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following sta+.us:
- a. The equipment hatch 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) Exhausting through an OPERABLE Rea.: tor Building Containment Purge System HEPA filters and charcoal adsorbers.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within 1 the containment. ACTION: i With the requirements of the above specification not satisfied, immediately i suspend all operations involving CORE ALTERATIONS or movement of irradiated l fuel in the containment building. SURVEILLANCE REQUIREMENTS j 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or exhausting through
- an OPERABLE Reactor Building Containment Purge System with the capability of
< being automatically isolated upon heater failure within 72 hnurs prior to the I start of and at least once per 7 days during CORE ALTERATIONS or movement of . l irradiated fuel in the containment building by: l
- a. Verifying the penetrations are in their closed / isolated condition, or l
- b. Verifying the upper and lower containment purge supply and exhaust valves close upon a High Relative Humidity test signa?.
i l g 18 M 7 3/4 9-4 OCT 7 1985 CATAWBA - UNITS 1 AND 2
~_ - _ - _ . - . - . . ._ ._ _ ._
~ _ . - _ .
l 1 REFUELINS OPERATIONS PROOF & REVIE'.V COPY SURVEILLANCE REQUIREMENTS (Continued) 4.9.4.2 The Reactor Building Containment Purge System shall be demonstrated OPERABLE: )
- a. At least once per 31 days by initiating flow through the HEPA filters and carbon adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating; s b. At least once per 18 months or after any structural maintenance on the HEPA filter or carbon adsorber housings, by:
- 1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of 1 '
i less than 1% and uses the test procedures guidance in Regula-tory Positions C.S.a. C.S.c, and C.S.d* of Regulatory Guide 1.52 Revision 2,'&rch 1978, and the system flow rate is 25,000 cfm i 10% (both exhaust fans operating);
, 2) Verifying within 31 days after removal, that a laboratory analysis of a presentative carbon sample obtained i.i accordance l with Regulatory Position C.6.b of Regulatory Guide 1.52, Revi-sion 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 6%;
and
- 3) Verifying a system flow rate of 25,000 cfm i 10% (both exhaust fans operating) during system operation when tested in accordance with ANSI N510-1980.
i
- c. After every 720 hourt, of carbon adsorber operation, by verifying, t
within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory i' Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, l meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl i iodide penetration of less than 6%;
- d. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA fil-ters, carbon adsorber banks, and preffiters is less than 8 inches Water Gauge while operating the system at a flow rate
, of 25,000 cfm 10% (both exhaust fans operating); and
- Purging of residual refrigerant is not mandatory.
l i I 1 i i l ] 0 01 CATAWBA - UNITS 1 AND 2 3/4 9-5 1
PROOF & REVIEW COPY. REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) .
- 2) Verifying that the filter train duct heater dissipates 120 i 12 kW. .
- 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% in accordance with ANSI N510-1980 fer a D0P test aerosol while operating the system at a flow rate of 25,000 cfm i 10% (both exhaust fans operating);
- f. After each complete or partial replacement of a carbon adsorber bank, by verifying that the cleanup system bank satisifies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 25,000 cfm i 10% (both exhaust fans operating); and
- g. Following painting, fire, or chemical release in any ventilation zone communicating with the system, 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 cri-teria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 6%.
e l CATAWBA - UNITS 1 AND 2 3/4 9-6 f
h REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Directatcommunications personnel shall be maintained between the control room and 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. SURVEILLANCE REQUIREMENTS 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. s OCT 7 1985 CATAWBA - UNITS 1 AND 2 3/4 9-7 jg( 18 D0b
PROOF & REVlEW COPY REFUELING OPERATIONS 3/4.9.6 MANIPULATOR CRANE LIMITING CONDITION FOR OPERATION 3.9.6 The reactor building manipulator crane and auxiliary hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:
- a. The manipulator crane used for movement of fuel assemblies having:
- 1) A minimum capacity of 3250 pounds, and
- 2) An overload cutoff limit less than or equal to 2900 pounds.
- b. The auxiliary hoist used for latching and unlatching drive rods having:
- 1) A minimum capacity of 610 pounds, and
- 2) A load indicator which shall be used to prevent lifting loads in excess of 600 pounds.
APPLICABILITY: During movement of drive rods or fuel assemblies within ( the reactor vessel. ACTION: With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of drive rods and fuel assembTies within the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.6.1 Each manipulator crane used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 72 hours prior to the start of such operations by performing a load test of at least 3250 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2850 pounds. 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 72 hours prior to the start of such operations by performing a load test of at least 610 pounds. gz 7 GB5 CATAWBA - UNITS 1 AND 2 3/4 9-8 )g( 18 Eb
Proof & R&lEW COPY REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 3000 pounds
- shall be prohibited from travel over fuel assemblies in the storage pool.
APPLICABILITY: With fuel assemblies in the storage pool. 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 The weight of each load, other than a fuel assembly and control rod, shall be verified to be less than or equal to 3000 pounds prior to moving it over fuel assemblies.*
- Weir gates of the spent fuel pool may be moved by crane over the stored fuel provided the spent fuel has decayed for at least 17.5 days since last being part of a core at power.
q BBS gg] CATAWBA - UNITS 1 AND 2 3/4 9-9 jQL 10
REFUELING OPERATIONS PROOF & REVlEW COPY 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL . LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be OPERABLE and in operation.* APPLICABILITY: HODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no residual heat removal loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required residual heat removal loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmos-phere within 4 hours. SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one residual heat removal loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours.
- The residual heat removal loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
CATAWBA - UNITS 1 AND 2 3/4 9-10 jgg,18 Tb q N6Tb QS l 1 l
~
REFUELING OPERATIONS PROOF & REVIEW COPY LOW WATER LEVEL LIMITING CONDITICN FOR OPERATION 3.9.8.2 Two independent residual heat removal loops shall be OPERABLE, and at least one residual heat removal 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 residual heat removal loops OPERABLE, immediately initiate corrective action to return the required resi-dual heat removal loops to OPERABLE status, or establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
- b. With no residual heat removal 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 residual heat removal 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 residual heat removal loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours.
- Prior to initial criticality, the residual heat removal loop may be removed from ope. cation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
g $6 CATAWBA - UNITS 1 AND 2 3/4 9-11 g 1 $N
REFUELING OPERATIONS 3/4.9.9 WATER LEVEL - REALTOR VESSEL PROOF & REVIEW COPY h LIMITING CONDITION FOR OPERATION 3.9.9 At least 23 feet of water shall be maintained over the top of the reactor vessel flange. APPLICABILITY: During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.9 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 essemblies or control rods. l CATAWBA - UNITS 1 AND 2 3/4 9-12
$L 1B \%%h g $35 Q@
PROOF & REVIEW COPY REFUELING OPERATIONS 3/4.9.10 WATER LEVEL-STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.10 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 storage pool. ACTION:
- a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.10 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. CATAWBA - UNITS 1 AND 2 3/4 9-13 18 196b 1 \3B6 0 01 I
REFUELING OPERATIONS PROOF & REVIEW COPY 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION 3.9.11 At least one train of the Fuel Handling Ventilation Exhaust System shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is in the storage pool. ACTION: a. With both trains of the Fuel Handling Ventilation Exhaust System inoperable, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until the Fuel Handling Ventilation Exhaust System is restored to OPERABLE status. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.11.1 One train of the Fuel Handling Ventilation Exhaust System shall be determined to be operating and discharging through the HEPA filter and carbon adsorbers at least once per 12 hours whenever irradiated fuel is being moved in the storage pool and during crane operation with loads over the storage pool. 4.9.11.2 Both trains of the Fuel Handling Ventilation Exhaust System shall be demonstrated OPEP.ABLE: a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and carbon adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;
- b. At least once per 18 months by:
- 1) Verifying that the cleanup system satisfies the in place perietration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Positions C.5.a, C.S.c, and C.5.d* of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 16,565 cfm 110% per fan;
- Purging of residual refrigerant is not mandatory.
1 \N CATAWBA - UNITS 1 AND 2 3/4 9-14 ocT
- -. -. ~
REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) i
?.) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor- !
dance with Regulatory Positions C.6.b of Regulatory Guide 1.52, , Revision 2, March 1978, meets the laboratory testing criteria l of Regulatory Position C.6.a of Regulatory Guide 1.52. Revision 2, March 1978, for a methyl iodide penetration of less than 1%; and
- 3) Verifying a system flow rate of 16,565 cfm 110% per fan during system operation when tested in accordance with ANSI N510-1980.
- c. After every 720 hours of carbon adsorber operation in any train by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regula-tory 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 for each train by:
1
- 1) Verifying that the pressure drop across the combined HEPA filters, carbon adsorber banks, and moisture separators is less than 8 inches Water Gauge while operating the system at a flow rate of 16,565 cfm i 10% per fan.
- 2) Verifying that the system maintains the spent fuel storage poo'l area at a negative pressure of greater than or equal to % inch Water Gauge relative to the outside amosphere during system j operation,
- 3) Verifying that the filter cooling bypass valves can be manually opened, and
- 4) Verifying that the heaters dissipate 80 i 8 kW.
- e. After any structural maintenance on the filter housing, or after each complete or partial replacement of a HEPA filter bank in any train, by verifying that the cleanup system satisfies the in place penetra-tion and bypass leakage testing acceptance criteria of less than 1%
in accordance with ANSI N510-1980 for a 00P test aerosol while operat-ing the system at a flow rate of 16,565 cfm i 10% per fan
- f. After any structural maintenance on the filter housing, or after each complete or partial replacement of a carbon adsorber bank in any train, by verifying that the cleanup system satisfies the in place penetra-tion and bypass leakage testing acceptance criteria of less than 1%
in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrig-erant test gas while operating the system at a flow rate of 16,565 cfm i 10% per fan; and CATAWBA - UNITS 1 AND 2 3/4 9-15 gg 1N i i
.. - = -
REFUELING OPERATIONS COPY SURVEILLANCE REQUIREMENTS (Continued)
- g. After any structural maintenance on the filter housing dr, following '
painting, fire, or chemical release in any ventilation zone communicat-ing with the system by verifying, within 31 days after rimoval, 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%. I t 4 CATAWBA - UNITS 1 AND 2 3/4 E 16
l 1 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN , PROOF & REVIEW COPY 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). APPLICA5ILITY: 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, immedi-ately initiate and continue beration 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 MAF]IN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS
~
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 and being relied orf for the SHUTDOWN MARGIN requirement shall be demonstrated capable of full inser-tion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. l CATAWBA - UNITS 1 AND 2 3/4 10-1 QU l
- - , , - - - - , -. , - - - , ,m.-,,
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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 equal 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.
I SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. t 4.10.2.2 The requirements of the below listed specifications shall be performed l 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.
l Y80 Jst 1B CATAWBA - UNITS 1 AND 2 3/4 10-2 q g65 Q@ 1
SPECIAL TEST EXCEPTIONS PROOF & REVIEW COPY 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 is greater thanCoolant System or equal lowest operating loop temperature (T**9) to 541*F. APPLICABILITY: 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, g to within its limit within 15 minutes or be in at least HOT STANOBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS l 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. l l l CATAWBA - UNITS 1 AND 2 3/4 10-3 M 1g Mg OGT
SPECIAL TEST EXCEPTIONS DRAFT PROOF & REVlEW COPY 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
APPLICABILITY: During operation below the P-7 Interlock Setpoint. ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers. 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 Intermediate and Power Range channel, and P-7 Interlock shall be
- subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to i initiating startup and PHYSICS TESTS.
CATAWBA - UNITS 1 AND 2 3/4 10-4 jul 1g $5qU> DCT
SPECIAL TEST EXCEPTIONS PROOF & REVIEW CO { 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;
- a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and
- b. The rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements. ACTION: With the Position Indication System inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers. SURVEILLANCE REQUIREMENTS 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 Oemand Position Indication System and the Digital Rod Position Indication System agree:
- a. Within 12 steps when the rods are stationary, and
- b. Within 24 steps during rod motion.
I l l
*This requirement is not applicable during the initial calibration of the Position Indication System provided: (1) K is maintained less than or equalto0.95,and(2)onlyoneshutdownor'bntrolrodbankiswithdrawn from the fully inserted position at one time.
I i CATAWBA - UNITS 1 AND 2 3/4 10-5 N g \0J05 1 $65 0 01
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS PROOF & REVIEW COPY h CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 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. 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 analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. i l CATAWBA - UNITS 1 AND 2 3/4 11-1 y)\.1 g W
,, ji QGT L __ .. -- _. _. -
PROOF & REVlBV CO[y. . TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(1) TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml) , s
- 1. Batch Waste P P Release Each Batch Each Batch Principal Gamma 5x10 7 Tanks (2) Emitters (3)
I-131 1x10 6 Any tank which discharges P M Dissolved and 1x10
- liquid wastes One Batch /M Entrained Gases by the liquid (Gamma Emitters) effluent moni-tor, EMF-49 P M H-3 1x10 6 Each Batch Composite (4)
Gross Alpha 1x10 7 P Q Sr-89, Sr-90 5x10 8 Each Batch Composite (4) Fe-55 1x10 8
- 2. Continuous W Principal Gamma 5x10 7 Releases (5) Continuous (6) Composite (6- Emitters (3)
I-131 1x10 8 Conventional Waste Water M M Dissolved and 1x10 b Treatment Grab Sample Entrained Gases Line (Gamma Emitters) M H-3 1x10 5 Continuous (6) Composite (6) Gross Alpha 1x10 7 Q Sr-89, Sr-90 5x10 8 Continuous (6) Composite (6) Fe-55 1x10 8 CATAWBA - UNITS 1 AND 2 3/4 11-2 g 18 Mb g5 gG s
PROOF & REVIEW COPI 5 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: 4.66 s b E V 2.22 x 108 " Y exp (-AAt) Where: LLD = the "a priori" lower limit of detection (microcurie 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 samp'.e 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 l 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.
- l 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 then thoroughly mixed to assure representative sampling. CATAWBA - UNITS 1 AND 2 3/4 11-3 M 1g YA5q \@S 0 01
TABLE 4.11-1 (Cont ued) 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, In-65, Mo-99,
s-134, Cs-137 and Ce-141. The LLD for Ce-144 is 5 x 10 8 pCi/ml. This sist does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuc-lides, shc11 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 released. (5)A cantinuous 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)To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for t.he composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release. CATAWBA - UNITS 1 AND 2 3/4 11-4 g 'l N
RADI0 ACTIVE EFFLUENTS PROOF & REVIEW COPY g DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose coruitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to 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: a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to 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. This Special Report shall also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Safe Drinking Water Act.* b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid affluents for the current calendar quarter and the current calendar year shail be determined in accordance . with the methodology and parameters in the ODCM at least once per 31 days. ( l "The requirements of ACTION a.(1) and (2) are applicable only if drinking water ! supply is taken from the receiving water body within 3 miles downstream of ' the plant discharge. 8 l g\M CATAWBA - UNITS 1 AND 2 3/4 11-5 1
RADI0 ACTIVE EFFLUENTS F E REVISV COPY ff LIQUID RADWASTE TREATMENT SYSTEM i l LIMITING CONDITION FOR OPERATION I 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 5.1-3) 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.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid re'ieases 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 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. l g N q $5 CATAWBA - UNITS 1 AND 2 3/4 11-6 gq1
RADI0 ACTIVE EFFLUENTS LIQUID HOLDUP TANKS PROOF & mg 5 kk LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each temporary un:: tr-+ed outdoor tank shall be limited to less than or equal to 10 Curies, exed' y tritium and dissolved or entrained noble gases. APPLICAdILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, 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 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. i O l 1g 6N 3% CATAWBA - UNITS 1 AND 2 3/4 11-7 gg 1 $6
I RADI0 ACTIVE EFFLUENTS CHEMICAL TREATMENT PONOS proof & Rag g j{ i LIMITING CONDITION FOR OPERATION 3.11.1.5 The quantity of radioactive material contained in each chemical treatment pond shall be limited by the following expression: l A ' 264 -
< 1.0
- J -
- excluding tritium and dissolved or entrained noble gases, Where
A) = pond inventory limit for single radionuclide "j", in Curies; i C.J
= 10 CFR Part 20, Appendix B, Table II, Column 2, concentration for single radionuclide "j", microCuries/ml; V = design volume of liquid and slurry in the pond, in gallons; and '
- 264 = conversion unit, microcuries/ Curie per milliliter / gallon.
i l APPLICABILITY: At all times.
- ACTION
- a. With the quantity of radioactive material in any of the above listed i ponds exceeding the above limit, immediately suspend all additions of radioactive material to the pond and initiate corrective action j to reduce the contents to within the limit.
- b. The. provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
4
- SURVEILLANCE REQUIREMENTS i 4.11.1.5 The quantity of radioactive material contained in each batch of j resin / water slurry to be transferred to the chemical treatment ponds shall be determined to be within the above limit by analyzing a representative sample
} of the batch to be transferred to the chemical treatment ponds shall be limited i by the expression: 1 I
< 0.006 j C) -
Where: ' c) = radioactive resin / water s durry concentration for radionuclide "j" i entering the UNRESTRICTE'1 AREA chemical treatment ponds, in microcuries/ milliliter; and ' Cy = 10 CFR Part 20, Appendix B, Table II, Column 2, concentration for single radionuclide "j", in microcuries/ milliliter. I h i CATAWBA - UNITS 1 AND 2 3/4 11-8
$U\@
l l t
RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS PROOF & REVIEW C0P h 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 5.1-4) 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, for Iodine-133, for tritium, and for all radio-nuclides 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 and to be parameters within in the the above limits in accordance with the methodology ODCM. 4.11.2.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing specified analyses in Table 4.11-2.in accordance with the sampling and analysis program e 6 I
$6
\ #O$ q ffs 1 i CATAWBA - UNITS 1 AND 2 3/4 11-9 ' i
TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM h MINIMUM LOWER LIMIT OF SAMPLING ANALYSIS TYPE OF DETECTION (LLD)(y) g GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml) M
- 1. Waste Gas Storage P P Tank Each Tank Each Tank Principal Gamma Emitters (2) 1x10 4 Grab Sample R 2. Containment Purge P P Each PURGE (3) Each PURGE (3) Principal Gamma Emitters (2) 1x10 4 Grab Sample M H-3 (oxide) 1x10 8
- 3. Unit Vent W(3),(4) Principal Gamma Emitters (2) 1x10 4 Grab Sample W I3) H-3 (oxide) 1x10 8 w 4. Containment Air D(3)(5) D(3)(5) Principal Gamma Emitters (2) 1x10 4
) Release and g Addition System Grab Sample M H-3 (oxide) 1x10 6 T
1 5. All Release Types Continuous (6) D II) I-131 1x10 11 o as listed in 3. Charcoal above. Sample I-133 1x10 8 8 m Continuous (6) D(7) Principal Gamma Emitters (2) 1x10 10 i Particulate Sample g Continuous (6) M Gross Alpha (0) 1x10 11 l Composite Par-ticulate Sample Qo Continuous (6) Q Sr-89, Sr-90 1x10 11 Composite Par-ticulate Sample s, gc - V - cp A N' o g@ . i i
1 N00F & REVIEN COPY
, TABLE 4.11-2 (Continued)
TABLE NOTATIONS (1) The LLD is defined, for purposes of these specifications, as the smalle'st 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 (microcurie per unit mass or volume), s b = 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 collecticn 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. T365 l 3.1B gW CATAWBA - UNITS 1 AND 2 3/4 11-11 'l N h
' ~ ' - + - -
PROOFgggyggycopy TABLE 4.11-2 (Continued) 1 TABLE NOTATIONS (Continued) (2) The principal gamma emitters for which the LLD specification applies - include the following radionuclides: Kr-87, Kr-88, X -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, and Ce-141 in Iodine and particulate releases. The LLD for Ce-144 is 5x10 9 pCi/ml. 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 anal-yzed 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 stabilization (power level constant at desired power level) after a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period, for at least one of the three gaseous release types with this notation. (4) Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. (5) Required sampling and analysis frequency during effluent release via this pathway. (6) The ratio of the sample flow volume to the sampled stream flow voluma shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2, aad 3.11.2.3. (7) Samples shall be changed at least once per 24 hours and analyses shall be completed within 48 hours after changing, or after removal from sampler. (8) The composite filter (s) will be analyzed for alpha activity by analyzing one filter per week to ensure that at least four filters are analyzed per collection period. CATAWBA - UNITS 1 AND 2 3/4 11-12 g i g 1985 gg
RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The ait dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-4) 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 1
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 00CM at least once per 31 days. 1 CATAWBA - UNITS 1 AND 2 3/4 11-13 1 0@
RADI0 ACTIVE EFFLUENTS , DOSE - IODINE-131, IODINE-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, Iodine-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 BOUNDARY (see Figure 5.1-4) 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, Iodine-133, tritium, and radionuclides in particu? ate 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) fer exceeding the limit and defines the corrective actions that have been taken to reduce the releases and tf:e 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 4.11.2.3 Cuaulative dose contributions for the current calendar quarter and current calendar year for Iodine-131, Iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. s g\,,1 B CATAWBA - UNITS 1 AND 2 3/4 11-14 g 1
i RADIOACTIVE EFFLUENTS PROOF & REVIEW COPY DRAFT GASEOUS RADWASTE TREATMENT SYSTEM L'MITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLOUP 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 5.1-4) would exceed either:
- 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.
SURVEILLANCF REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and . beyond the SITE BOUNDARY 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 sha?1 be considered OPERABLE by meeting Specification 3.11.2.1 and 3.11.2.2 or 3.11.2.3. 1365 SL18 9 CATAWBA - UNITS 1 AND 2 3/4 11-15 g 1
RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE 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 2% by volume whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION:
- a. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 2% 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 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 HOLOUP 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. CATAWBA - UNITS 1 AND 2 3/4 11-16 JUL 181985 1@
1 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 97,000 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 radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank. ( i f a l l l CATAWBA - UNITS 1 AND 2 JUL 18.1985 i 3/4 11-17 qM 1 003
l RADI0 ACTIVE EFFLUENTS F&kN/[WCOPY 3/4.11.3 SOLID RADIOACTIVE 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 Radwaste System as necessary to pre vent 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.
SURVEILLANCF 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 sludges, 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 the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM;
- b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste; and -
- c. With the installed equipment incapable of meeting Specification 3.11.3 or declared inoperable, restore the equipment to OPERABLE status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.
CATAWBA - UNITS 1 AND 2 3/4 11-18 'l g
PR007 g pgy;gy(9p, RADI0 ACTIVE EFFLUENTS RAFI, 3/4.11.4 TOTAL DOS _E_ _ LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THEsources cycle PUBLIC shaldue to releases of radioactivity and to radiation from uranfun f body or any organ,l be limited to less than or equal to 25 mrems to the whole equal to 75 mrems. except the thyroid, which shall be limited to less than or APPLICABILITY: At all times. ACTION: a. With the calculated doses from the release of radioactive ma in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or 3.11.2.3b., calculations shall be made including direct radiation contributions from the units and from outside storage tanks to ; exceeded. whether the above limits of Specification 3.11.4 have been determine If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report releases to prevent recurrence of exceeding the abo includes limits. the schedule for achieving conformance with the above This Special Report, as defined in 10 CFR 20.405c shall include an analysis that estimates the radiation exposure (, dose) to all effluent pathways and direct radiation, for the ca that includes the release (s) covered by.this report. It shall also describe involved, and levels of radiation the cause and levels of the exposure concentrations of radioactive or concentrations. If the estimated dose (s) exceeds the above limits, and if the release been corrected, the Special Report shall include a req 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 shall be determined in accordance with SpecificationsCumulative do " 4.11.2.3, and in accordance with the methodology and parameters in the 00CM4. 4.11.4.2 . and from radwaste storage tanks shall be determined in acc methodology and parameters in the 00CM. under conditions set forth in ACTION a. of Specification 3.11.4.This requirement E CATAWBA - UNITS 1 AND 2 3/4 11-19 EL 131985 g@ g g
1 PROOF & REVIEW COPY 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM l LIMITING CONDITION FOR OPERATION 3.12.1 The Radiological Environmental Monitoring Program shall be conducted 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 S.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 dose
- to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) concentration (2) + ...> 1.0 reporting level (1) + reporting level (2) When radionuclides other than those in Teble 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 Specification 3.11.1. 2, 3.11. 2. 2, o r 3.11. 2. 3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be '
reported and described in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6.
- The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
CATAWBA - UNITS 1 AND 2 3/4 12-1 Jyt i 81985 gW6 0D1
PROOF & REVIEW COPY RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION ACTION (Continued) '
- c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCH. The specific locations from which samples 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 loca-tion (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 requireo. by Table 4.12-1. CATAWBA - UNITS 1 AND 2 3/4 12-2 jut.18 1985 OCT 7 1985
i TABLE 3.12-1
$ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM E
NUMBER OF E REPRESENTATIVE Z EXPOSURE PATHWAY SAMPLES AND SAMPLING AND AND/OR SAMPLE SAMPLE. LOCATIONS (3) TYPE AND FREQUENCY COLLECTION FREQUENCY OF ANALYSIS O
- 1. Direct RadiationI) Forty routine monitoring stations Quarterly.
Ganuna dose quarterly. N either with two or more dostmeters or with one instrument for measuring and recording dose rate continuously, placed as follows: An inner ring of stations, one in w each meteorological sector in the
) general area of the SITE BOUNDARY h
w An outer ring of stations, one in each meteorological sector in the 6- to 8-km range from the site; and The balance of the stations to be placed in special interest areas such as population centers. nearby residences, schools, and n in one or two areas to serve as y control stations. o Ro h O
" ::o r- .X3 .Q o - , 3::,, -
a- Q n@ E -< i
i TABLE 3.12-1 (Continued) S
$ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 5
3* NUMBER OF REPRESENTATIVE E EXPOSURE PATHWAY SAMPLES AND gy) SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE SAMPLE LOCATIONS
] COLLECTION FREQUENCY OF ANALYSIS e 2. Airborne 2 Radiotodine and Samples from five locations. Continuous sampler oper- Radiofodine Cannister:
$ Particulates ation with sample collec- I-131 analysis weekly.
m tion weekly, or more Three samples from close frequently if required by to the three SITE BOUNDARY dust loading. Particulate Sampler: locations, in different Gross beta radioactivity sectors, of the highest analysis following calculated annual average ground-level D/Q: filter change;(3) and gamma isotopic analysis 2 of composite (by g One sample from the location) quarterly.
'?
vicinity of a community , having the highest calcu- $ lated annual average ground-level D/Q; and b 9* O One sample from a control location, as for example 15 to N W@ "P5 30 km distant and in the least prevalent wind direction.
- 3. Waterborne o
- a. Surface (5) One sample upstream. Composite sample over Gamma isotopic analysis )
One sample downstream. 1-month pariod.IN
- c: an y a erly.
c( 4
- b. Ground Samples from one or two sources Quarterly. Gamma isoto,,1c ) and
" # "" *"* hs s quar erly.
only if likely to be affected . s
~ ,
ns w 4.7s
n TABLE 3.12-1 (Continued)
$ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ai
' NUMBER OF REPRESENTATIVE E EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY
] AND/0R SAMPLE SAMPLE LOCATIONS II) COLLECTION FREQUENCY OF ANALYSIS s 3. Waterborne (Continued) 5 c. Drinking One sample of each of one to Composite sample over low level I-131 analysis m three of the nearest 2-week period (6) when on each composite when the '
water supplies that could be dose calculated for the affected by its discharge. low level I-131 ana- consumption of the water lysis is performed; monthly composite is greater than 1 mrem One sample from a control location. otherwise. per year (8) . Composite for gross beta and gamma t'
- isotopic analyses (4) monthly. Composite for y tritium analysis quarterly.
- d. Sediment One sample from downstream area Semiannually. Gamma isotopic analysis (4) from with er.isting or potential semiannually.
Shoreline recreational value.
- 4. Ingestion
- a. Milk Samples from milking animals Semimonthly when Gamma isotopicI4) and in three locations within 5 km animals are on pasture; low level I-131 analysis distance having the highest monthly at other times. semi-monthly when animals dose potential. If there are are on pasture; monthly none, then one sample from at other times.
. , milking animals in each of . c o
three areas between 5 to 8 km distant where doses are calcu-
.g o
S latedtobegrgerthan Q 1 mrem per yr One sample h a from milking animals at a is control location 15 to 30 km ' 8? distant and in the least preva-lent wind direction. k e
~
N
TABLE 3.12-1 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 5 3" NUMBER OF REPRESENTATIVE E EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE SAMPLE LOCATIONS ( ) COLLECTION FREQUENCY OF ANALYSIS
]
- 4. Ingestion (Continued) 2 E b. Fish and One sample each of a predatory Sample in season, or Gamma isotopic analysis (
N Inverte- species, a bottom feeder and a semiannually if they on edible portions. brates forage species in vicinity of are not seasonal. plant discharge area. One sample each of a predatory species, a bottom feeder and a forage species in areas not influenced by plant discharge. M c. Food One sample of each principal At time of harvest I9) . Gamma isotopic analyses I4) a Products class of food products from on edible portion. any area that is irrigated by water in which liquid plant wastes have been discharged. Samples of three different Monthly, when Gamma isotopicI ) and I-131 kinds of broad leaf vegeta- available. analysis. tion grown nearest each of @ two different offsite loca- % tions of highest predicted y annual average ground level D/Q if milk sampling is not , performed. Gamma isotopic ) and I-131 h""* One sample of each of the similar broad leaf vegeta-Monthly, when available, analysis. ' y tion grown 15 to 30 km dis- , m*
@ tant in the least prevalent
- wind direction if milk sam- %
pling is not performed. p
DRAFT TABLE 3.12-1 (Continued) TABLE NOTATIONS M00F& mm m (1) Specific parameters of distance and direction sector from the centerline of the station, and additional description where pertinent, shall be pro-vided for each and every sample location in Table 3.12-1 in a table and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of auto-matic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, 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 Environ-mental Operating Report pursuant to Specification 6.9.1.6. It is recognized that, at times, it may not be possible or practicable to con-tinue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appro-priate substitutions made within 30 days in the Radiological Environmental Monitoring Program. In lieu of any Licensee Event Report required by Specification 6.9.1 and pursuant to Specification 6.9.1.7, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s). (2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addi-tion 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., at an ocean site, 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 speci-fic system used and should be selected to obtain cptimum dose information within minimal fading.) (3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 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. CATAWBA - UNITS 1 AND 2 3/4 12-7 S n. 1 8 N 7g I
DRAFT TABLE 3.12-1 (Continued) _ TABLE NOTATIONS (Continued) PR0OF & REVIEW COP (4) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility. (5) 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. " Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for recreational activities. (6) A composite sample is one in which the rate at which the liquid sampled is uniform and in which the method of sampling employed results in a specimen that is representative of the time averaged concentration at the location being sampled. 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. (7) 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. (8) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCH. (9) 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 rcot food products. CATAWBA - UNITS 1 AND 2 3/4 12-8 g 18 W 7 1985
1 , TABLE 3.12-2 h
> REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES h REPORTING LEVELS E
Q WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS ANALYSIS (pCi/1) OR GASES (pC1/m3 ) (pCi/kg, wet) (pCi/1) (pci/kg, wet) h H-3 20,000 I1) Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 4
- Zn-65 300 20,000 M
l a Zr-Nb-95 400 I-131 20(2) 1 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 o 3 II)For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, value M . of 30,000 pCi/t say be used. ' 9e l o (2) Reporting level for gamma isotopic analysis. If the calculated dose for drinking water exceeds 1 arem/yr, o u low-level I-131 analysis shall be performed and a reporting level of 2 pCi/1 shall be used. ~ f w a 8 8 i
- i
! TABLE 4.12-1 f DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (1) (2) LOWER LIMIT OF DETECTION (LLD)I3) . E d WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS SEDIMENT g ANALYSIS (pCi/1) OR GASES (pci/m3 ) (pCi/kg, wet) (pCi/t) (pCi/kg, wet) (pCi/kg, dry) E o Gross Beta 4 to H-3 2000* Mn-54 15 130 Fe-59 30 260
$ Co-58,60 15 130 y Zn-65 30 260 i o Z r-Nb-95 15 -
4 I-131 15(5) 0.07 1 60 i Cs-134 15 0.05 130 15 60 150 i Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 "If no drinking water pathway exists, a value of 3000 pCi/t may be used.
. , e e .
O O a a h . w- e D l 4 ,
. & REVIEW COPY DRAFT TABLE 4.12-1 (Continued)
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 Radiolog! cal Environmental Operating Report pursuant to Specification 6.9.1.6. (2) Required detection capabilities for thermoluminescent dosimeters used for environmental tions of Regulatorymeasurements Guide 4.13. shall be in accordance with the recommenda-(3) The LLD is defined, for purposes of these specifications, as the smallest concentrations 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: 4.66 s b ~ E - V - 2.22 - Y - exp(- Aat) Where: LLD
= the "a priori" lower limit of detection (picoCuries per unit mass or volume),
s b = 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 = the number of disintegrations per minute per picocurie, Y
= the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (s 1), and at = the elapsed time between environmental 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. CATAWBA - UNITS 1 AND 2 3/4 12-11 El I 81985 OCT 7 1985
1 PROOF & REVIEW COPY 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 a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. AneTyses shall be performed in such a manner that the stated LLDs will be acnieved 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) ' Reporting level for gamma isotopic analysis. If the calculated dose for' drinking water exceeds 1 mrem /yr, low level I-131 analysis shall be per-i formed and a reporting level of 2 pCi/1 shall be used. (5) LLD for gamma isotopic analysis. If low level I-131 analysis is performed, an LLD of 1 pCi/1 shall apply. 1 1 i CATAWBA - UNITS 1 AND 2 3/4 12-12
PROOF & REVIEW COPY RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than 50 m2 (500 f tz) 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 loca-tion (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 calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new 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 conducted. 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 (s) for the ODCM reflecting the new location (s), with information supporting the change in the sampling locations.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- Broad leaf vegetation sampling of at least three different kinds of 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. Speci-fications for broad leaf vegetation sampling in Table 3.12-1.4.c. shall be followed, including analysis of control samples.
CATAWBA - UNITS 1 AND 2 3/4 12-13 jyl 181985 001 7 1985
PROOF & REVIEW COPY I O l RADIOLOGICAL ENVIRONMENTAL MONITORING U <
}
SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at* 1 east 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. J l CATAWBA - UNITS 1 AND 2 3/4 12-14 Mll 1 g 1985 i 00T 7 1985
PROOF & REVIEW COPY RADIOLOGICAL ENVIRONMENTAL MONITORING [ 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM l LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on all 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. A summary of the results obtained as part of the above required Interlaboratory Ccmparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. i l CATAWBA - UNITS 1 AND 2 3/4 12-15 JUL 1 g BBS 0 01 7 ud5
PROOF & REVIEW COPY g ILgu i 1 BASES FOR SECTIONS 3.0 AND 4.0 4 LIMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE REQUIREMENTS
- l 1
t i j i j i i i i I i t i \ ' I l l i jut 181985 l l 001 7 1985 r -
i I i PROOF & REVIEW COPY - j d NOTE The BASES contained in succeeding pages sumarize the reasons for the Specifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are
- not part of these Technical Specifications.
t 4 l l l [ I l l' I g( 18 t OCT 7 1985 i
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S ' ' 3/4.0 APPLICABILITY l&/ COPY 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 those 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 This specification delineates the measures to be taken for those circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specifi-cation 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 Specification 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 6 hours, and in at least HOT SHUTDOWN within the following 6 hours. As a further example, Specifica-tion 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one Spray System is inoperable. Under the requirements of Specification 3.0.3 f f both the required Containment Spray Systems are inoperable, within 1 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 provided for operation in subsequent lower OPERATION HODE(S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability is discovered but the times provided for achieving a mode reduction are not applicable if the inoperability is discovered in a mode l 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. l 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. However, 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 l 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 CATAWBA - UNITS 1 AND 2 B 3/4 0-1 M g5
a,- w w-APPLICABILITY EW COPY BASES ' met without regard for allowable deviations and out-of-service pro' visions contained in the ACTION statements. " The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other speci-fled limits being exceeded. Exceptions to this provision have been provided for a limited number of specifications when STARTUP with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications.
.~ 3.0.5 This specification delineates the applicability of each specifica.
~ tion'to Unit 1 and Unit 2 operation.
4.0.1 This specification provides that surveillance activities necessary to ensure 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 r 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 this 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 tnis j provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable althqugh still l meeting the Surveillance Requirements. Items may be determined inoperable during use, during surveillance tests, or in accordance with this specification. , Therefore, ACTION statements are entered when the Surveillance Requirements
- should have been performed rather than at the time it is discovered that the tests were not performed.
CATAWBA - UNITS 1 AND 2 8 3/4 0-2 l l
I . , - . APPLICABILITY PROOF & REVIEW COPY ' BASES 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 applicable condition. 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 per-forming the inservice inspection and testing activities required by Section XI of the ASME Boffer and Pressure Vessel Code and applicable Adder.da. This clarification 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 l 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, i
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.
&%.1g $N g 1 BBS CATAWBA - UNITS 1 AND 2 B 3/4 0-3 l
_ - ._ -_l-_- . _ _ _
l
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3/4.1 REACTIVITY CONTROL SYSTEMS BASES { 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating cot.ditions, (2) the reactivity transients asso-ciated with postulited accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, boron concentration, and T,yg. The most restrictive condition occurs at EOL, with T, g at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled Reactor Coolant System cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% ak/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T less than 200*F, the reactivity transients resulting from a postulated ste$r59 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. The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC
- used in the FSAR analyses to nominal operating conditions. These corrections 1
a l g 1B ! CATAWBA - UNITS 1 AND 2 B 3/4 1-1 g T3B5 l
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. transformedintothglimitingMTCvalue-3.7x10ThfsvalueoftheMDCwasthen ak/k/'F. The MTC value of -2.8 x 10 ak/k/*F represents a conservative value (with corrections for burnup and soluble baron) at a core condition of 300 ppm equilibriumboronconcentrationand,fsobtainedbymakingthesecorrectionsto the limiting MTC value of -3.7 x 10 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 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 P-12 interlock is above its setpoint, (4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (5) the reactor vessel is above its minimum RT temperature. NDT 3/4.1.2 B0 RATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators. With the coolant average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN
# g1 $N 1
CATAWBA - UNITS 1 AND 2 B 3/4 1-2 g 1N
PROOF & REVIEW COPY REACTIVITY CONTROL SYSTEMS BASES i BORATION SYSTEMS (Continued) ' MARGIN from expected operating conditions of 1.3% ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppm borated water from the boric acid storage tanks l t or 75,000 gallons of 2000 ppm borated water from the refueling water storage tank. l I With the coolant temperature below 200*F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable ; reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPERABLE aad the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 285'F provides assurance that a mass addition pressure transient can be re11aved by the operation of a single PORV. The boron capability required below 200'F is sufficient to provide a
; SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldcwn from 200*F to 143* F.
This condition requires either 906 gallons of 7000 ppm borated water from the boric acid storage tanks or 3170 gallons of 2000 ppm Dorated water from the refueling water storage tank. 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 boron concentration of the refueling water storage tank also ensure a pH value of between 8.5 and 10.5 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 corresion on mechanical systems and components. The OPERABILITY of one Boron Injection 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-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120 and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges 1 are picked for verification of agreement with demanded position. CATAWBA - UNITS 1 AND 2 B 3/4 1-3 g 9b J g gg5
l PROOF & REVIEW COPY REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) The ACTION statements which permit limited variations frem 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 pro-vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses effected 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,yggreater 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 fre-quent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied. Ub CATAWBA - UNITS 1 AND 2 B 3/4 1-4 g 1B 7 BBS
PROOF & REVIEW COPY 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: (1) maintaining the minimum DNBR in the core greater than or equal to design limit DNBR during normal operation and in short-term transients, and (2) 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: Fq (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F" 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 F,y(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. 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound envelope of 2.32 times the normalized axial peaking factor 9is 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 the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
$66 CATAWBA - UNITS 1 AND 2 8 3/4 2-1 $.18 ggg
PROOF & REVIEW COPY POWER DISTRIBUTION LIMITS j BASES AXIAL FLUX DIFFERENCE (Continued) 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 suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the tier.e duration of the deviation is limited. Accordingly, a 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 deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least two of four or two of three OPERABLE excore channels are outside the target bsnd 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 accu-mulates 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 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, coolant flow rate, 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 i 12 steps, indicated, from the group demand position; i
i b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; 1 l l CATAWBA - UNITS 1 AND 2 B 3/4 2-2 Nt 18 BB6 i 0 01 7D
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INDICATED AXIAL FLUX DIFFERENCE FIGUP.E C 3/4 2-1 TYPICAL INDICATED AXI AL FLUX DIFFERENCE VERSUS THERMAL POWER CATAWBA - UNIT 3 1 AND 2 B 3/4 2-3 Dbb f g iB q 885 0 01
PROOF & REVIEW COPY POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
- c. The control rod insertion limits of Specifications 3.1.3.5 and I 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
- d. above are maintained. As noted on Figure 3.2-3, Reactor Coolant System flow rate and F" may be " traded off" against one another (i.e. , a low measured Reactor Coolant System flow rate is acceptable if the measured F" is also low) to ensure that the calculated DNBR will not be below the design DNBR value. The relaxation of F N as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts for F" less than or equal to 1.49. This value is used in the various accident analyses where F" influences parameters other than DNBR, e.g., peak clad temperature,andthusisthemaximum"asmeasuredyvalueallowed. The rod bow penalty as a function of burnup applied for Fg is calculated with the methods described in WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation," July 1979, and the maximum rod bow penalty is 2.7% DNBR. Since the safety analysis is performed with plant-specific safety DNBR limits of 1.49 and 1.47 compared to the design DNBR limits of 1.34 and 1.32, respectively, for the typical and thimble cells, there is a 10% thermal margin available to offset the rod bow penalty of 2.7% DNBR. When an F q measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full- ore map taken with the Incore Detector Flux Mapping System, and a 3% allowan;e is appropriate for manufacturing tolerance. ( The Radial Peaking Factor, Fxy(Z), is measured periodically to provide ,, assurance that the Hot Channel Factor, F (Z), remains within its limit. The 9 F xy limit for RATED THERMAL POWER (F ) as provided in the Radial Peaking Factor Limit Report per Specification 6.9.1.9 was determined from expected 6 power control manuevers over the full range of burnup conditions in the core. ' 4 E y,18 1985 CATAWBA - UNITS 1 AND 2 B 3/4. 2-4 eg isB5 l
PR0OF & REVIEW COPY POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RAT ENTHALPY RISE HOT CHANNEL FACTOR (Continued) When Reactor Coolant System flow rate and F are measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-3. Measurement errors of 2.1% for Reactor Coolant System total flow rate and 4% for F have been allowed for in determination of the design DN8R value. The measurement error for Reactor Coolant System total flow rate is based Reactor Coolant System flow rate indicators.upon performing a precision h Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the fendwater venturi is included in Figure 3.2-3. Any fouling which might bias the Resctor Coolant System flow rate measurement greater performance thanparameters. 0.1% can be detected by monitoring and trending various plant If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e. , either the effect of the fouling shall be quantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling . The 12-hour periodic surveillance of indicated Reactor Coolant System flow is sufficient to detect only flow degradation which could lead to opera-tion outside the acceptable region of operation shown on Figure 3.2-3. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis. Radial power periodically distribution during measurements are made during STARTUP testing and power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts. A limit the with of 1.02 was selected indicatad to provide an allowance for the uncertainty associated power tilt. The 2-hour time allowance for operation with a tilt condition greater than of a dropped 1.02 butorless than 1.09 misaligned is provided control rod. to allow identification and correction In the event such action does not correct the tilt, the margin for uncertainty on Fqis reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1. 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 CATAWBA - UNITS 1 AND 2 8 3/4 2-S E gg I
PR0OF & REYlEW COPY POWER DISTRIBUTION LIMITS BASES 4 QUADRANT POWER TILT RATIO (Continued) flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. The normal locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. Alternate locations are available if any of the normal locations are unavailable. 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 design limit DNBR throughout each analyzed transient.
The indicated T,yg value of 592.75 F and the indicated pressurizer pressure value of 2225.8 psig corre-spond to analytical limits of 594.8 F and 2204.6 psig respectively, with allow-ance for measurement uncertainty. 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. Measure-ment uncertainties cust be accounted for during the periodic surveillance. l 1 CATAWBA - UNITS 1 AND 2 8 3/4 2-6 @ l I _. - . ,, _ _ __ _. --
. ~
PROOF & REVIEW COPY 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 SYSTEM INSTRUMENTATIONREACTOR TRIP SYSTEM and ENGINEERED SAF The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures 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 parameters. system functional capability is available from diverse The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility dasign for the protection and mitigation of accident and transient conditions
; The integrated operation of each of these systems is consistent with the .
assumptions used in the safety analyses. The Surveillance Requirements speci-fied for these systems ensure that the overall system functional capability is
- maintained comparable to the original design standards. The periodic surveil- i lancecapability.
this tests performed at the minimum frequencies are sufficient to demonstrate 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 aessured" 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 Setpoints have been specified in Table 3.3-4. Opera-tion with Setpoints less conservative than the Trip Setpoint but within the Allowable analysis toValue is acceptable accommodate since an allowance has been made in the safety this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exr.eed the Allowable Value. The methodology of this option utilizes the -
- "as measured" deviation from the specified calibration point for rack and ,
i sensor corponents 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 + 5 < 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 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. ( is the difference, in percent span, R or Rack Error is the "as measured" TA or Total Allowance deviation, Trip Setpoint. in the percent span, for the affected channel from the specified S or Seasor Error is either the "as measured" deviation of CATAWBA - UNITS 1 AND 2 8 3/4 3-1 001 7 BBS
PR0OF & REVIEW COPY DMFT INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
- INSTRUMENTATION (Continued) the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows j
for a sensor drift factor, an increased rack drift factor, and provides a 4 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. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess met of the Allowable Value exhibits the behavior that the rack has not its allowance. Being that there is a small statistical chance that this I 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 investigatiol 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 demonstrated 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 time. The Engineered Safety Features Actuation System senses selected plant parameters 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-nf-coolant accident: (1) Safety , Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators. (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps ) start and automatic valves position, (10) nuclear service water pumps start and automatic valves position, and (11) component cooling pumps start and automatic valves position. CATAWBA - UNITS 1 AND 2 B 3/4 3-2 JUL 10 g1 7 1985 l
_ . _ . ~ - PROOF & REVIEW COPY INSTRUMENTATION udln BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i INSTRUMENTATION (Continued) 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 Defeats the manual block of Safety Injection actuation on low pres-surizer pressure and low steam line pressure and defeats steam line isolation on negative steam lire pressure rate. Defeats the manual block of the motor-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level. P-12 On increasing reactor coolant loop temperature, P-12 automatically blocks steam dump and allows manual bypass of steam dump block for the cooldown valves only. On decreasing reactor coolant loop temper-ature, P-12 automatically defeats the manual bypass of the steam dump block. P-14 On increasing steam generator level, P-14 automatically trips all feedwater isolation valves, pumps and turbine and inhibits feedwater control valve modulation. 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The CPERABILITY 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 actuation of Emergency Exhaust or Ventilation Systems. CATAWBA - UNITS 1 AND 2 B 3/4 3-3 jyl l g a 5 gg5 7 885 t ..
PROOF & REVlEW COPY INSTRUMENTATION BASES 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. For the purpose of measuring F (Z) or FN a full incore flux map is used. q Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. 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 capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdowr. 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 Earth-quakes," April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are 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. 3/4.3.3.5 REMOTE SHUTDOWN SYSTEM The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50. The OPERABILITY of the Remote Shutdown System ensures that a fire will not preclude achieving safe shutdown. The Remote Shutdown System instrumentation, CATAWBA - UNITS 1 AND 2 8 3/4 3-4 y B 6 g
PROOF & REVIEW COPY INSTRUMENTATION na BASES REMOTE SHUTDOWN SYSTEM (Continued) control and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shutdown the reactor. This capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR Part 50. 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 consis-tent with the recommendations of Regulatory Guide 1.97, Revision 3. "Instrumenta-tion for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions Dur-ing and Following an Accident," May 1983 and NUREG 0737, " Clarification of TMI Action Plan Requirements," November 1980. 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event af an accidental chlorine release. This capability is required to protect control. room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revision 1, " Protection of Nuclear Power Plant Centrol Room Operators Against an Accidental Chlorine Release," January 1977. 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for 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 l t fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility Fire Protection Program. 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 notifica-tion. Consequently, the minimum number of OPERABLE fire detectors must be greater. I The loss of detection capability for Fire Suppression Systems, actuated ' by fire detectors, represents a significant degradation of fire protection for CATAWBA - UNITS 1 AND 2 B 3/4 3-5 jul 18 ESS gz 71985
PROOF s REVIEW COPY INSTRUMENTATION . BASES b FIRE DETECTION INSTRUMENTATION (Continued) - 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 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, 3/4.3.3.9 LOOSE-PART DETECTION SYSTEM
'The OPERABAILITY of the ' loose part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the Reactor System and avoid or mitigate damage to Reactor 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.
3/4.3.3.10 RADI0 ACTIVE 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 / Trip Setpoints for these instruments shall be calculated and adjusted 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. 3/4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous efflux *.5. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted 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. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitor.used to show compliance with the gaseous effluent release requirements of Specifica-tion 3.11.2.2 shall be such that concentrations as low as 1 x 10-6 mci /cc are measurable. g 1 $85 CATAWBA - UNITS 1 AND 2 B 3/4 3-6 l
PROOF & REVIEW COPY INSTRUMENTATION BASES 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment, or structures. g 18 85 CATAWBA - UNITS 1 AND 2 B 3/4 3-7 OCT 7 1985
PROOF & REVIEW COPY 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 calculated DNBR above the design DNBR value during Condition I and II events. In MODES 1 and 2 with one reactor coolant loop rot in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours. In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat; however, single failure considerations require that three loops be OPERABLE. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or residual heat removal loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either residual heat removal or Reactor Coolant System) be OPERABLE. In MODE 5 with reactor c60lant loops not filled, a single residual heat removal loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two residual heat removal loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one residual heat removal 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 baron reduction will, therefore, be within the capability of operator recogni-tion and control. The restrictions on starting a Reactor Coolant Pump below P-7 with one or more cold legs less than or equal to 285 F are provided to prevent 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 Reactor Coolant System will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50 F above each of the cold leg temperatures. . 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the Reactor Coolant System from being pressurized above its Safety Limit of 2735 psig. Each l safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating residual heat 1 P removal loop, connected to the Reactor Coolant System, provides overpressure l l CATAWBA - UNITS 1 AND 2 B 3/4 4-1 JUL 181985 OCT 7 1985
J REACTOR COOLANT SYSTEM PROOF & REVIEW COPY BASES SAFETY VALVES (Continued) relief capability and will prevent overpressurization. in addition, the Overpressure Protection System provides a diverse means of protection against overpressurization at lov temperatures. During operation, all pressurizer Code safety valves must be OPERABLE to prevent tne Reactor Coolant System from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accardance with the provisions of Section XI of the ASME Boiler and Pressure Code. - 3/4.4.3 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumpcions. 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 Reactor' Coolant System is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve Reactor Coolant System 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. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. Testing of the PORV's includes the emergency N 2 supply from the Cold Leg Accumulators. This testing demonstrates that the valves in the supply line operate satisfactorily and that the non-safety portion of the instrument air system is not necessary for proper iORV operation. 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 Reactor Coolant ! System will be maintained. The program for inservice inspection of steam CATAWBA - UNITS 1 AND 2 B 3/4 4-2 g 1385
. . 59%
REACTOR COOLANT SYSTEM PROOF & REVIEW COPY BASES STEAM GENERATORS (Continued) 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 main- ; tain surveillance of the conditions of the tubes in the event that there is ' evidence of mechanical damage or progressive degradation due to design, manu-facturing 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. 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 plant opera-tion would be limited by the limitation of steam generator tube leakage between the Reactcr Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this ifmit during operation will have an adeqi. ate 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 sht.tdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However .even if a defect should develop in service, it will be found during scheduled inservice steam generator tube axaminations. 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. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be repurted 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 result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The 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 i Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," l May 1973. CATAWBA - UNITS 1 AND 2 B 3/4 4-3 g $5 l ! )
. . . -. --. _ _ - A
PR0OF & REV REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAK,*,SE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. Industry experience has shown that while a limited amount of leakage is expected from the Reactor Coolant System, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This thres-hold value is sufficiently low to ensure early detection of additional leakage. The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the Reactor Coolant System ensures that the dosage contribution from the tube laakage 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 breik. 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 conditions. 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. The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal Reactor Coolant System pres-sure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the safety analyses. The 1 gpm leakage from any Reactor Coolant System 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 substantial length of time, verification of valve integrity is required. Since these vahes are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses . containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillcnce Requirements for Reactor Coolant System pressure isolation valves provide added assurance of valve integrity thereby reducing the prob-ability of gross valve failure and consequent intersystem LOCA. Leakage from the pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. l CATAWBA - UNITS 1 AND 2 B 3/4 4-4 Att 181985 00T 7 1985 i
REACTOR COOLANT SYSTEM PROOF & REVIEW COPY BASES 3/4.4.7 CHEMISTRY
~
The limitations on Reactor Coolant System chemistry minimize the potential for corrosion of the Reactor Coclant System and reduce the potential for Reactor Coolant System leakage or failures 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 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 concen-trations to within the Steady-State Limits. The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. 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 BOUNDARY will not exceed an appro-priately 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 Catawba site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation. The ACTION statement oermitting POWER OPERATION to continue for limited time 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 (approxima:ely 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 BOUNDAFY by a factor of up to 20 following a postulated steam generator i tube rupture. The reporting of cumulative operating time over 500. hours in any 6-month consecutive period with greater than 1 microcurie / gram DOSE l EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the i circumstances prior to reaching the 800-hour limit. 007 7 1985 I CATAWBA - UNITS 1 AND 2 B 3/4 4-5 l
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radioiodine level is to be determined every 4 hours. If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1L In a release of reactor coolant with a typical mixture of radioactivity, the actual radio-iodine contribGtion would probably be about 20L The exclusion of radio-nuclides with half-lives less than 10 minutes from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition. Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour, about 2 hours, about 1 day, about 1 week, and about 1 month. Reducing T,yg 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. 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. l g 181985 CATAWBA - UNITS 1 AND 2 B 3/4 4-6 00T 7 1985 !
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM BASES E 3/4.4.9 PRESSURE / TEMPERATURE LIMITS - The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G and 10 CFR 50 Appendix G:
- 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:
. ..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
- b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure onl For normal operation, other inherent plant characteristics, e.g., y.pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2. These limit lines shall be calculated periodically using methods provided be'ow,
- 3. The secondary side of the steam generator must not be pres:urized above 200 psig if the temperature of the steam generator is below 70 F,
- 4. The pressurizer heatup and cooldown rates shall not exceed 100 F/h and 200*F/h, respectively, and
- 5. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI.
Assurance of adequate fracture toughness of the ferritic materials in the Unit 1 reactor vessel is provided by compliance with Section III of tne 1971 ASME Boiler and Pressure Vessel Code, plus Addenda to Winter 1971. The Unit 2 reac-tor vessel complies with Section III of the 1971 ASME B&PV Code, plus Addenda to Winter 1972. The reactor vessel materials meet the fracture toughness requirements of 10CFR50, Appendix G, to the extent possible. The pressure-temperature limitations on reactor operation, as well as leak and hydrostatic test conditions are determined in accordance with Appendix G to Section III of the ASME B&PV Code and Appendix G, 10CFR50. Since the fracture toughness test-ing performed on vessel material from Units 1 and 2 did not include all of the tests necessary to determine RT in the manner prescribed in NB-2300 of NDT ASME III, Summer 1972 Addenda, the necessary properties were estimated using the procedures provided in Branch Technical Position MTEB 5-2, " Fracture Tough-ness Requirements for Older Plants." CATAWBA - UNITS 1 AND 2 8 3/4 4-7
PR0OF & REVIEW COPY REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) Heatup and cooldown limit curves are eclculated 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 RTNDT at the 1/4T location in the core region is greater than the RT of the limiting unirradiated material. NDT The selection of such a limiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor o; era-tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, 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 NDT computed by either Regulatory Guide 1.99, Revision 1, " Effects of 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 adjust-ments for this shift in RT NDT at the end of 16 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments. 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 with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the pressure 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 aetermined from the surveillance capsule exceeds the calculated NOT f r the equivalent capsule radiation exposure. ART Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion 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 the following paragraphs. CATAWBA - UNITS 1 AND 2 B 3/4 4-8 g ig BBS OCT 7u
i TABLE B 3/4.4-la g REACTOR VESSEL TOUGHNESS (UNIT 1)
-e j 50-FT-LB" SHELF 2
T 35 MIL ENERGY MAT'L CU P NDT TEMP RT NPWD* g COMPONENT HEAT NO. SPEC. NO. % % % *F 'F NDT FT-LB h Closure Head Dome 55888-1 A5338, CL.1 -
.011 -4 70 10 86 s Closure Head Ring 007055 A508, CL.2 -
.006 16 23 16 101 2 Closure Head Flange 527038 A508, CL.2 .05 .013 -4 -2 -4 104 5 Vessel Flange 411212 A508, CL.2 -
.004 -31 -47 -31 153 m Inlet Nozzle ~,26827 A508, CL.2 .05 .010 -13 7 -13 87 Inlet Nozzle 526829 A508, CL.2 .07 .010 -4 43 -4 86 Inlet Nozzle 526859 A508, CL.2 .04 .013 -4 23 -4 81 Inlet Nozzle 526857 A508, CL.2 .05 .012 -13 23 -4 77 Outlet Nozzle 526827 A508, CL.2 .05 .011 -22 27 -22 84 butlet Nozzle 526829 A508, CL.2 .07 .010 -4 2 -4 87 Outlet Nozzle 526859 A508, CL.2 .04 .011 -13 38 -13 81 Outlet Nozzle 526857 A508, CL.2 .05 .013 -4 38 -4 60 t'
Nozzle Shell 411077 A508, CL.2 -
.007 -40 34 -26 101 Inter. Shell 411343 A508, CL.2 .08 .004 -40 52 -8 100 Lower Shell 527708 A508, CL.2 .04 .008 -13 16 -13 101 Bottom Head Ring 527428 A508, CL.2 .06 .013 -4 74 14 68 Bottom Head Segment 55292-1 A533B, CL.1 -
.006 -22 5 -22 79 Bottom Head Segment 55292-1 A5338, CL.1 -
.006 -13 34 -13 79 Bottom Head Segment 55163-2 A533B, CL.1 -
.011 -4 38 -4 80 Bottom Head Segment 55163-2 A533B, CL.1 -
.011 -13 74 14 70 i
Bottom Head Dome 55178-1 A533B, CL.1 -
.010 -31 84 24 64 4
- Nozzle Shell to Weld (P710) u
, Inter Shell to Lower Shell Weld Root (P710) .03 .009 0** - 0** -
- o Lower Shell to Bot. Head Ring Weld (P710) o Inter Shell to Lower Shell Weld (R747) .05 .010 -76 -9 . -51 134
- o o l 6 Q
- Estimated per NRC Standard Review Plan Section 5.3.2 from data obtained in the principal direction.
g o
- ** Estimated per NRC Standard Review Plan Section 5.3.2 from charpy tests performed at 10' F. -<
e TABLE B 3/4.4-1b f 9 REACTOR VESSEL TOUGHNESS (UNIT 2) 5 5 3" 50 Ft-Lb Shelf
' T 35 Mil Energy Pat'l Cu P NDT Temp RT NPWD*
g Component Heat no. Spec. No. % % *F *F F NDT Ft-Lb Z* Closure Head Dome B8607-1 A5338,CL.1 .13 .007 -40 50 -10 106 Closure Head Torus B8608-1 "
.07 .007 -20 57 -3 118 R Closure Head Flange 88601-1 A508 CL.2 -
.010 10 <10 10 152 o Vessel Flange B8602-1 " -
.010 10 <10 10 175 N
Inlet Nozzle B8609-1 " -
.010 -20 <10 -20 119 B8609-2 -
.010 -20 <10 -20 124 B8609-3 -
.008 -20 <40 -20 109 88609-4 -
.006 -20 97 37 97 Outlet Nozzle 88610-1 " -
.008 -10 <10 -10 141
!" B8610-2 " -
.006 -10 <50 -10 m " " 144 88610-3 -
.004 -20 <40 -20 140 R
B8610-4 -
.006 -10 <50 -10 150 Nozzle Shell 88604-1 A533B,CL.1 .11 .007 -10 84 24 96
? B8604-2 "
.11 .007 -10 86 26 89 g " "
88604-3 "
.07 .009 -20 110 50 70 Inter. Shell** 88605-1 .09 .011 -10 75 15 89 B8605-2 .07 .009 -20 93 33 82 B8616-1 .05 .010 0 72 12 92 Lower Shell B8806-1 "
.05 .009 -60 66 6 83 B8806-2 .05 .007 -40 50 -10 102 88806-3 .05 .006 -40 68 8 105 Bottom Head Torus B8613-1 .14 .010 -40 52 -8 113 Bottom Head Dome B8612-1 .14 .010 -40 65 5 124 Nozzle Shell Vert. Weld Seams (G1.36) .15 .012 0* <10 0* >112 e Nozzle Shell to Inter. Shell Weld Seam .13 .016 -40 <20 -40 >102 O
(G1.50) O Inter. & tower Shell Vert. Weld Seams (G1.45) .04 .005 -80 <-20 ' '
-80 >130 4 Inter. to Lower Shell Weld Seam (G1.45) g O
'rt
$ 98
- Estimated per NRC Standard Review Plan Section 5.3.2. ,
**Used for Surveillance Program Weldment.
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e r._ , 1018 2 4 6 8 1019 2 4 6 81020 FAST NEUTRON FLUENCE (N/CM2 , E > 1 MeV) ' FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT NDT FOR REACTOR VESSELS EXPOSED TO 550*F g 55 CATAWBA - UNITS 1 AND 2 B 3/4 4-12 g 1985
PROOF & REVIEW COPY REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) - The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semielliptical sarface defect with a depth of one quarter of the wall thickness, T, and a vength of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities 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 i 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, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, i K7 , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K yg, for the metal temperature at that time. K is obtained from the reference IR fracture toughness curve, defined in Appendix G to the ASME Code. The K curve is given by the equation: IR C KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) 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: CKIM + kit <KIR (2) Where: KIM = the stress intensity factor caused by membrane (pressure) stress, K It = the stress intensity factor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RT NDT f the material, C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations. g 18 M CATAWBA - UNITS 1 AND 2 B 3/4 4-13 g 7 BBS
PROOF & REVIEW CO REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) ~ At any time during the heatup or coaldown transient, K is determined by IR 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 gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, kit, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. 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 thermal 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. The use of the composite curve in the cooldewn analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adiacent to the vessel ID. This condition, of course, is not true for the steady-state situa-tion. It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T location IR for finite cooldown rates than for steady-state operation. Furthermore, if condit. ions 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 contro' 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 as well as finite heatup rate conditions assuming the presence of a 1/4T CATAWBA - UNITS 1 AND 2 8 3/4 4-14 JM.18 1985 00T 7 1985 i
PROOF & REVIEW CO REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) 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 f r the 1/4T crack IR during heatup is lower than the K for 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. Rather, each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit nust 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.
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PROOF & REVIEW COP REACTOR COOLANT SYSTEM BASES ) PRESSURE / TEMPERATURE LIMITS (Continued) 1 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. LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or a Reactor Coolant System vent opening of at least 4.5 square inches ensures that the Reactor Coolant System will be protected from pressure transients which could exceed the limits of Appendix G to 285*F. 10 CFR Part 50 when one or more of the cold legs are less than or equal to Either PORV has adequate relieving capability to protect the Reactor , t Coolant System from overpressurization when the transient is limited to either: (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50*F above the cold leg temperatures, or (2) the start of a Safety Injection pump and its injection into a water , solid Reactor Coolant System. The Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance of the LTOPS assuming various mass input and heat input transients. Opera-tion with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum , pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. those assumed cannot occurTo ensure that mass and heat input transients more severe tha Technical Specifications require lockout of all but one Safety Injection pump a,nd all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50*F above primary temperature. The Maximum Allowed PORV setpoint for the LTOPS 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, 1 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 l ASME Boiler and Pressure Vessel Code and applicable Addenda as required by ! 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). l Components of the Reactor Coolant System were designed to provide access l to permit inservice inspections in accordance with Section XI of the ASME l Boiler and Pressure Vessel Code, and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the g Commission pursuant to 10 CFR 50.55a(g)(6)(i). CATAWBA - UNITS 1 AND 2 8 3/4 4-16 0CT 71985
PROOFg pgyggy(9p, RFACTOR COOLANT SYSTEM BASES E 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cool-ing. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head, and the pressurizer steam space ensures the capability exists to perform this function. The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent ' isolation of the vent path. - The function, capabilities, and testing requirements of the Reactor Coolant System vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. l l 001 7 665 CATAWBA - UNITS 1 AND 2 B 3/4 4-17
PROOF & REVIEW COPY 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs from the cold leg injection accumulators and directly into the reactor vessel from the upper head injection accumulators in the event the Reactor Coolant System pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large 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 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 and 3/4.5.3 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 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 coolant 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. CATAWBA - UNITS 1 AND 2 B 2/4 5-1 JUL 18 boa OCT 7 1985
EMERGENCY CORE COOLING SYSTEMS PROOF & REVIEW COPY BASES t ECCS SUBSYSTEMS (Continued) - The limitation for a maximum of one centrifugal charging pump and one Safety Injection pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety Injection pumps except the required OPERABLE centrifugal charging pump to be inoperable below 285 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. 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 prcper 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. 3/4.5.4 REFUELING WATER STORAGE TAN _K The OPERABILITY of the refueling water storage tank as part of the ECCS ensures that a sufficient supply of barated water is available for injection by the ECCS in the event of a LOCA. The limits on 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 refueling water storage tank and the Reactor Coolant System 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 refueling water storage tank also ensure a pH value of between 8.0 and 9.0 for the solution recirculated within containment after a LOCA. This pH band mini-mizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. CATAWBA - UNITS 1 AND 2 B 3/4 5-2
PROOF & REVIEW COP 3/4.6 CCNTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric-tion, 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 contain-ment 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, or 0.75 Lt; as applicable, during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests. 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 1 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 provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. 3/4.6.'.4 INTERNAL PRESSURE . 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 1.5 psig, and (2) the containment peak pressure does not exceed the design pressure of 15 psig during LOCA conditions. l l l CATAWBA - UNITS 1 AND 2 B 3/4 6-1 IO OCT 7 005
PROOF & REVIEW C CONTAINMENT SYSTEMS BASES INTERNAL PRESSURE (Continued) The maximum peak pressure expected to be obtained from a LOCA event is 14.68 psig. The limit of 0.3 psig for initial positive containment pressure is consistent with the safety analyses. 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that: (1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions, and (2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment. Measurements shall be made at all operating ventilation unit locations, the average air whether temperature. by fixed or portable instruments, prior to determining The containment pressure transient is sensitive to the initially contained air mass during a LOCA. temperature. The contained air mass increases with decreasing The lower temperature limit of 100'F for the lower compartment and 75'F (60'F whers in MODE 2, 3 or 4) for the upper compartment will limit the peak sure of 15pressure psig. Theto 14.7 psig which is less than the containment design pres-upper temperature limit influences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the safety analyses. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that , the vessel will withstand the maximum pressure of 15 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability. 3/4.6.1.7 REACTOR BUILDING STRUCTURAL INTEGRITY ' This limitation ensures that the structural integrity of the containment reactor building will be maintained comparabic to the original design standards for the life of the facility. Structural integrity is required to provide: (1) protection for the steel vessel from external missiles, (2) radiation shielding in the event of a LOCA, and (3) an annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions, A visual inspection is sufficient to demonstrate this capability. i 1 CATAWBA - UNITS 1 AND 2 B 3/4 6-2 jut 181985 00T 7 1985
PROOF & REVIEW l CONTAINMENT SYSTEMS BASES 3/4.6.1.8 ANNULUS VENTILATION SYSTEM The OPERABILITY of the Annulus Ventilation System ensures that durin conditions, containment vessel leakage into the annulus will be filtered thr the HEPA filters and charcoal adsorber trains prior to discharge to the atm Operation of the system with the heaters operating to maintain low ere. 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 adsorbers and HEPA filters. This requirement is necessary to moet the assumptions used in the safety analyses and limit the SITE BOUNDARY radiation doses to within the dose gu line values of 10 CFR Part 100 during LOCA conditions. used as a procedural guide for surveillance testing. ANSI N510-1980 will be 3/4.6.1.9 CONTAINMENT PURGE SYSTEMS The containment purge supply and exhaust isolation valves for the lower compartment and the upper compartment (24-inch), and instrument room (12-inc and the Hydrogen Purge System (4-inch) are required to be sealed closed du , plantaoperation during LOCA. since these valves have not been demonstrated capable o Maintaining these valves sealed closed during plant operations via the Containment Purge System. ensures that excessive quantities of radio valves cannot be inadvertently opened, the valves are sealed closed i ance or lock with operator. Standard the valve closed, orReview preventsPlan power6.2.4 from which includes being supplied mechanical to the valve d The use of the containment purge lines is restricted to the 4-inch Con-tainment Air Release and Addition System valves since, unlike the lower comp ment and the upper compartment, instrument room, and the Hydrogen Purge Syste valves, these 4-inch valves are capable of closing during a LOCA. Therefore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not be excee in the event of an accident during containment purging operation. Operation with 4-inchthe Ifne open will be limited to 2000 hours during a calendar year for the valves. The total time the containment purge (vent) system isolation valves may be open during MODES 1, 2, 3, and 4 in a calendar year is a function of anticipated need and operating experience. , Only safety-related reasons; e.g., containment pressure control or the reduction of airborne radioactivity to facilitate personnel access for surveillance and maintenance activities , may be used to support the additional time requests. Leakage integrity tests with a maximum allowable leakage rate for contain-ment purge supply and exhaust valves will provide early indication of resilient material seal degradation leakage failures and will allow opportunity for repair before gross could develop. The 0.60 L, leakage limit of Specification 3.6.1.2b. shall not be exceeded when the leakage rates determined by the j*tkage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. CATAWBA - UNITS 1 AND 2 8 3/4 6-3 JUL 18 E65 0CT 7 1985
CONTAINMENT SYSTEMS F & R VIETY CO BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS The OPERABILITY of the Containment Spray System ensures that containment depressurization LOCA. and cooling capability will be available in the event of a The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses. 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 those assigned other inoperable ESF equipment. 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 environaent in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close auto-matically ensures that the release of radioactive material to the environ-ment will be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the 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 controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These Hydrogen Control Systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations Following a LOCA," March 1971. The OPERABILITY of at least 35 to 36 igniters per train (70 of 72 for both trains) ensures that the Distributed Ignition System will maintain an effective ' I coverage throughout the containment provided the two inoperable ignitors are not l on corresponding redundant circuits which provide coverage for the same region. I This system of igniters will initiate combustion of any significant amount of hydrogen released af ter a degraded core accident. This system is to ensure l burning in a controlled manner as the hydrogen is released instead of allowing it to be ignited at high concentrations by a random ignition source. CATAWBA - UNITS 1 AND 2 B 3/4 6-4 JUL 18 EOS OCT 7 1985
PROOF & REVlEW C CONTAINMENT SYSTEMS BASES
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3/4.6.5 ICE CONDENSER i The requirements associated with each of the components of the ice con-denser ensure that the overall system will be available to provide sufficient pressure suppression capability to limit the containment peak pressure tran-sient to less than 14.7 psig during LOCA conditions. 3/4.6.5.1 ICE BED The OPERABILITY of the ice bed ensures that the required ice inventory - will: (1) be distributed evenly through the containment bays, (2) contain sufficient boron to preclude dilution of the containment sump following the LOCA, and (3) contain sufficient heat removal capability to condense the Reactor Coolant System volume released during a LOCA. These conditions are consistent with the assumptions used in the safety analyses. The minimum weight figure of 1218 pounds of ice per basket contains a 10% conservative allowance for ice loss through sublimation which is a factor of 10 higher tnan assumed for the ice condenser design. The minimum total weight . of 2,368,652 pounds of ice also contains an additional 1% conseivative allowance to account for systematic error in the weighing instruments. In the event that observed sublimation rates are equal to or lower than design predictions after 3 years of operation, the minimum ice baskets weight may be adjusted downward. In addition, the number of ice baskets required to be weighed each 9 months may be reduced after 3 years of operation if such a reduction is supported by observed sublimation data. 3/4.6.5.2 ICE BED TEMPERATURE MONITORING SYSTEM The OPERABILITY of the Ice Bed Temperature Monitoring System ensures that the capability is available for monitoring the ice temperature. In the event the system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits. 3/4.6.5.3 ICE CONDENSER 000RS The OPERABILITY of the ice condenser doors and the requirement that they , be maintained closed ensures that the Reactor Coolant System fluid released j during a LOCA will be diverted through the ice condenser bays for heat removal i and that excessive sublimation of the ice bed will not occur because of warm ! air intrusion. 1 If an ice condenser door is not capable of opening automatically, then system function is seriously degraded and immediate action must be taken to restore the opening capability of the door. Not capable of opening automatically is defined as those conditions in which a door is physically blocked from opening by installation of a blocking device or by obstruction from temporary or permanent installed equipment or is otherwise inhibited from opening such as may result from ice, frost, debris or increased door opening torque. CATAWBA - UNITS 1 AND 2 B 3/4 6-5 'l N 00T
PROOF & REVIEW COPY , CONTAINMENT SYSTEMS BASES 3/4.6.5.4 INLET 000R POSITION MONITORING SYSTEM The OPERABILITY of the Inlet Door Position Monitoring System ensures that the capability is available for monitoring the individual inlet door position. In the event the system is inoperable, the ACTION requirements provide assur-ance that the ice bed heat removal capacity will be retained within the speci-fied time limits. 3/4.6.5.5 DIVIDER BARRIER PERSONNEL ACCESS DOORS AND EQUIPMENT HATCHES The requirements for the divider barrier pe'sonnel access doors and equipment hatches being closed and OPERABLE ensure that a minimum bypass steam flow will occur from the lower to the upper containment compartments during a LOCA. This condition ensures a diversion of the steam through the ice condenser bays that is consistent with the LOCA analyses. 3/4.6.5.6 CONTAINMENT AIR RETURN AND HYDROGEN SKIMMER SYSTEMS The OPERABILITY of the Containment Air Return and Hydrogen Skimmer Systems ensures that following a LOCA: (1) the containment atmosphere is circulated for cooling by the spray system, and (2) the accumulation of hydrogen in localized portions of the containment structure is minimized. j Since these systems are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment 1 will be made to the test data before it is compared to the Technical Specifica-tion Surveillance Requirements. 3/4.6.5.7 and 3/4.6.5.8 FLOOR AND REFUELING CANAL DRAINS 1 l The OPERABILITY of the ice condenser floor and refueling canal drains ensures that following a LOCA, the water from the melted ice and Containment Spray System has access for drainage back to the containment lower compartment and subsequently to the sump. This condition ensures the availability of the water for long-term cooling of the reactor during the post-accident phase. 1 3/4.6.5.9 DIVIDER BARRIER SEAL ! The requirement for the divider barrier seal to be OPERABLE ensures that a minimum bypass steam flow will occur from the lower to the upper containment l compartments during a LOCA. This condition ensuras a diversion of steam through the ice condenser bays that is consistent with the LOCA analyses. i 1 001 7U CATAWBA - UNITS 1 AND 2 8 3/4 6-6
i 1 l 1 3/4.7 PLANT SYSTEMS PROOF & REVIEW C0P BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES i The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1304 psig) of its design pressure of 1185 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from valve wide-open condition coincident with an assumed loss of condenser heat sink (i.e. , no steam bypass to the condenser). The specified valve lift settings and relieving capacities are in accordance Code, with the requirements of Section III of the ASME Boiler and Pressure 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 16.85 x IOS lbs/h which is 105% of the total secondary steam flow of 16.05 x IOS lbs/h at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steem generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable 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: For four loop operation SP = (U ~ x (109) X Where: 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 1bs/ hour, and ' Y = Maximum relieving capacity of any one safety valve in lbs/ hour CATAWBA - UNITS 1 AND 2 8 3/4 7-1 JUL 181985 OCT 71965
PLANT SYSTEMS PROOF & REVIEW COPY l BASES i L 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 single activein the event of a feedwater line break accident with a worst case failure. The Auxiliary Feedwater System is capable of delivering a total feedwater flow of at least 492 gpm at a pressure of 1210 psig to the entrance of at least two of the steam generators. This capacity is sufficient to ensure that adequate .feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation. 3/4.7.1.3 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. 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. 3/4.7.1.4 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity 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 Require-ments are consistent with the assumptions used in the safety analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator RT f 60 F and are suffitient to prevent brittle fracture. NDT CATAWBA - UNITS 1 AND 2 8 3/4 7-2
PLANT SYSTEMS BASES 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suf-ficient 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. 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM The OPERABILITY of the Nuclear Service Water System ensures that sufficient cooling ment capacity during normalis and available forconditions. accident continued operation of safety-related equip-The redundant cooling capacity of this system, assuming used in the safety analysis.a single failure, is consistent with the assumptions 3/4.7.5 STANDBY NUCLEAR SERVICE WATER POND The limitations on the standby nuclear service water pond level and temperature ensure that sufficient cooling capacity is available to either: (1) provide normal cooldown of the facility, or (2) mitigate the effects of accident conditions within acceptable Ifmits. The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling watnr supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommend-ations1974. March of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants," 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM The OPERABILITY of the Control Room Area Ventilation System ensures that: (1) the ambient air temperatura 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 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 con-sistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50. ANSI N510-1980 will be used as a procedural guide for surveil-lance testing. CATAWBA - UNITS 1 AND 2 8 3/4 7-3 g i a 1985 00T 7 1985
l 2 ) PLANT SYSTEMS
%0f&R g BASES 3/4.7.7 AUXILIARY BUILDING FILTERED EXHAUST SYSTEM '
l The OPERA 8ILITY of the Auxiliary Building FG cered Exhaust System ensures ! that radioactive materials leaking from the D.CS equipment within the auxiliary building following a LOCA are filtered 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 t sufficient to reduce the butidup of moisture on the adsorbers and HEPA filters. l The operation of this system and the resultant effect on offsite dosage calcu-1ations was assumed in the safety analyses. ANSI N510-1980 will be used as a { procedural guide for surveillance testing. ! i 3/4.7.S SNUB 8ERS ' ) 2 i All snubbers are required OPERA 8LE 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.
; size. Snubbers are classified and grouped by design and manufacturer but not by For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip, same type. and 100-kip capacity manufactured by Company "A" are of the 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.
~ 1 i A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance j 8 with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall l be, determined and approved by the Catawba Safety Review Group. The determination shall be based upon the existing radiation levels and the expected time to per-form a visual inspection in each snubber location as well as other factors asso- ! ciated with accessibility during plant operations (e.g. , temperature, atmosphere, i location etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The i addition or deletions of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50. I 1 M 10 i CATAWBA - UNITS 1 AND 2 8 3/4 7-4 QCT I i
l
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PLANT SYSTEMS PRO 0f & RW gg BASES SNUB 8ERS (Continued) 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 tion.is determined by the number of inoperable snubbers found during an inspec-In order to establish the inspection frequency for each type of snubber, it was ssumed that the frequency of snubber failures and initiating events are constant with time and that the failure of any snubber on that system could cause initiating theevent. system to be unprotected and to result in failure during an assumed Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose rasults require a shorter inspection interval will override the previous schedule. The acceptance criteria are to be used in the visual inspection to determine OPERABILITY of the snubbers. l To provide assurance of snubber functional reliability, one of three functional testing methods are used with the stated acceptance criteria:
- 1. Functionally test 10% of a type of snubber with an additional 10% tested 'or each functional testing failure, or
- 2. Functionally test a sample size and determine sample acceptance or rejection usfag Figure 4.7-1, or
'l .
Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation. 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. 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 testing was performed to qualify the snubber for the applicable design conditions. Snubbers so exempted shall be listed in the list of individual snobbers indicating the extent of the exemptions. CATAWBA - UNI 15 1 AND 2 0 3/4 7-5 jut.18 E OCT 7 1985 l
. 1 l
PLANT SYSTEMS i gM . BASES DRUT SNUBBERS (Continued) 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 snubbers, 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. 3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including plutonium. alpha emitters, is Dased on 10 CFR 70.39(a)(3) limits for This limitation will ensure that leakaga from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with sourceSurveillance Requirements commensurate with the probability of damage to a 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. 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 located. in any portion of the facility where safety-related equipment is The Fire Suppression System consists of the water system, sprry, and/or sprinklers, CO2 , and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related tion Program. equipment and is a major element in the fa-ility Fire Protec-In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. 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. i ! l 1 1 1 CATAWBA - UNITS 1 AND 2 8 3/4 7-6 JUL 18 EM I i l OCT 7 BB5 i
-_._ _ , _ __._. - ~_._ __..__-. -._ - __-- - _ , . . _ _ _ . __ _ . _ _ . - _ . _ - . _ . . _
PLANT SYSTEMS BASES FIRE SUPPRESSION SYSTEMS (Continued) 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. 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. These design features minimize the possibility of a single fire rapidly involving guishing of the fire. several areas of the facility prior to detection and extin-The fire barrier penetrations are a passive element in the facility fire protection program and are subject to periodic inspections. Fire barrier penetrations, including cable penetration barriers, fire doors, fire dampers, and other fire barriers 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 evalua-tion shall be performed to show that the modification has not degraded the fire rating of the fire barrier penetration. 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 or at least one side of the affected barrier must be verified OPEPABLE and an hourly fire watch patrol established, until the barrier is restored to functional status. 3/4.7.12 GROUNOWATER LEVEL This specification is provided ta ensure that groundwater levels will be monitored and prevented from rising to unacceptable levels. High ground-water levels could result in unacceptable structural stresses in the Contain-ment designand/or Auxiliary Building due to uplift and hydrostatic forces during basis events. Although these buildings have been statically analyzed to withstand soil pressure along with the uplift and hydrostatic forces resulting from Croundwater rebound to yard elevation (593'6"), this analysis , did not include any other loadings and was not a design condition for these buildings. jut.18 1985 CATAWBA - UNIT 5 1 AND 2 B 3/4 7-7 OCT 7 1985
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PLANT SYSTEMS MOOF& W N @f{ BASES 3/4.7.13 STANDBY SHUTDOWN SYSTEM The Standby Shutdown System (SSS) is designed to mitigate the consequences of certain postulated fire incidents by providing capability to maintain HOT STANDBY conditions and by controlling and monitoring vital systems from loca-tions external to the main control room. This capability is consistent with the requirements of 10 CFR Part 50, Appendix R. The Surveillance Requirements ensure that the SSS systems and components are capable of performing their intended functions. The required level in the SSS diesel generator fuel storage tank ensures sufficient fuel for 72 hours uninterrupted operation. It is assumed that, within 72 hours, either offsite power can be restored or additional fuel can be added to the storage tank. Although the Standby Makeup Pump is not nuclear safety-related and was not designed according to ASME code requirements, it is tested quarterly to ensure its OPERABILITY. The Surveillance Requirement concerning the Standby Makeup Pump water supply ensures that an adequate water volume is available to supply the pump continuously for 72 hours. JUL 1 8 BS$ CATAWBA - UNITS 1 AND 2 B 3/4 7-8 00T 7 1985 l i 4
1 3/4.8 ELECTRICAL POWER SYSTEMS BASES PROOF & REVIEW COPY ' bbb 3/4.8.1. 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES. and ONSITE POWER
- DISTRIBUTION The OPERABILITY of the A.C. and D.C power sources and associated distribu-tion 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. i power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50. l The ACTION requirements specified for the levels of degradation of the power with the sources level ofprovide restriction upon continued facility operation commensurate degradation. t The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and i are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERA 8LE 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 D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability 4 Electrical Power Sources," December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems , subsystems, trains, components and devices, that depend on the remaining OPERA 8LE diesel generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE. This require-ment is intended to provide assurance that a loss-of-offsite power event will not re, ult in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term, verify, as i 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 4
' maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERA 8ILITY of the component.
i The OPERABILITY of the minimum specified A.C. and 0.C. power sources and associated distribution systems during shutdown and refueling ensures that: i ! (1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient instrumentation and control capa-bility is available for monitoring and maintaining the unit status. The Surveillance Requirements for demonstrating the OPERASILITY of the - l diesel generators are in accordance with the recommendations of Regulatory ! Guide 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, August 1977, Regulatory Guide 1.137, " Fuel-011 Systems for Standby Diesel Gener-ators," Revision 1, October 1979, and the NRC Staff Evaluation Report concerning the Reliability of Diesel Generators at Catawba, August 14, 1984. If any other metallic structures (butiding, new or modified piping systems, conduits) are placed in the ground near the Fuel Oil Storage System or if the original system is modified, the adequacy and frequency of inspections for the Cathodic Protec-tion System shall be reevaluated and adjusted in accordance with the manufac-l turer's recommendations, j
CATAWBA - UNITS 1 AND 2 JUL 18 E85 8 3/4 8-1 gg i g .I i
- . , . _ .. ., -__. , _ __.__-----m.,__m,._.________.__.,________.m _ - __ _.__ ,_-. ._ _ ,, , _ ,. m , - _ ,_ . - - m __ _ _ _ _ _ ,
PROOF & REVIEW COPY A ' ELECTRICAL POWER SYSTEMS
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BASES i A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued) The Surveillance Requirements for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129,
" Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testirg, and Replacement of Large Lead Storage
, Batteries for Generating Stations and Substations." 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 ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity. . Table 4.8-3 specifies the normal limits for each designated pilot cell and each gravity. connected cell for electrolyte level, float voltage and specific The limits for the designated pilot cells float voltage and specific
-gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.
Operation with a battery cell's parameter outside the normal limit but within 7 days. theDuring allowable value specified in Table 4.8-3 is permitted for up to 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 t5e manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than 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, to perform its designgreater function.than 2.07 volts, ensures the battery's capability CATAWBA - UNITS 1 AND 2 8 3/4 8-2 UU M 7 ggg
ELECTRICAL POWER SYSTEMS BASES PROOF & REVIEW COPY [h(( 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are pro-tected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protec-tion circuit breakers during periodic surveillance. The Surveillance Requirements appifcable to lower voltage circuit breakers and fuses provide assurance of breaker and fuse reliability by testing a', . least one representative sample of each manufacturer's brand of circuit breaker and/or fuse. Each manufacturer's molded 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 neces-sary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuse for surveillance purposes. l l CATAWBA - UNITS 1 AND 2 8 3/4 8-3 , OCT 7 1985 l l 1
3/4.9 REFUELING OPERATIONS PROOF & REVIEW COPY BASES e 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The value of 0.95 or less for K,ff includes a 1% ak/k conservative allowance for uncertainties. Similarly, the boron allowance of 50value concentration ppm boron. of 2000 ppm or greater includes a conservative uncertainty The locking closed of the required valves during refueling operation precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. This action prevents flow to the Reactor Coolant System of unborated water by closing flow paths from sources of unborated water. 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundantofmonitoring condition the core. capability is available to detect changes in the reactivity 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY of the Reactor Building Containment Purge System ensure that a release of radioactive material within containment will be restricted from leakage to the environment or filtered through the HEPA filters and charcoal adsorbers prior to release to the atmosphere. . The OPERABILITY and closure restrictions are rufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurizatior, potential while in the REFUELING MODE. Operation of the Reactor Building Containment Purge System and of thesafety the resulting iodine removal capacity are consistent with the assumption analysis. j Operation of the system with the heaters operating to maintain low humidity using automatic control for at least 10 continuous hours i in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. ANSI N510-1980 will be used as a procedural guide for carveillance testing. l CATAWBA - UNITS 1 AND 2 B 3/4 9-1 jyl.16 00T 7 1%,
REFUELING OPERATIONS BASES 3/4.9.5 COPNUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that: (1) manipulator cranes will be used for movement of control rods and fuel assemblies, (2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and (3:sthe core internals and reactor vessel are protected from excessive lifting force in the event they ar8' inadvertently engaged during lifting operations. 3/4.9.7 CRANE TRAVEL - SPENT FHEL ST3 RAGE POOL BUILDING 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 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 snalyses. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal 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 effec'.s of a boron dilution incident and prevent boron stratification. The requirement to have two residual heat removal loops OPERABLE when ' there is less than 23 feet of water above the reactor vessel flange ensures 1 that a single failure of the operating residual heat removal loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating residual heat removal loop, adequate time is provided to initiate emergency procedures to cool the core. ! CATAWBA - UNITS 1 AND 2 B 3/4 9-2
. aut.1B M 00T 7 1985
REFUELING OPERATIONS BASES 3/4.9.9 and 3/4.9.10 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. 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM The limitations on the Fuel Handling Ventilation Exhaust System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. Operation of the system with the heaters operating to main-tain 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 adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consist? . with the assumptions of the safety analyses. ANSI N510-1980 will be used 6. a procedural guide for surveillance testing. 1 l JUL 181985 l CATAWBA - UNITS 1 AND 2 8 3/4 9-3 GCT 7 1985 l
i 3/4.10 SPECIAL TEST EXCEPTIONS pgoop & REHEW COPY 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 positiened 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 l than or equal to 5% of_ RATED THERMAL POWER with the T,yg slightly lower than normally allowed so that the fundamental nuclear characteristics of the core and related instrumentation can be verified. In order for various charac-teristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, i to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 and the Reactor Coolant System T,yg may 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. 4
- 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. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE. .
i i p 7 885 l CATAWBA - UNITS 1 AND 2 0 3/4 10-1 l
PR00f & RM 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radio-active materials released in liquid wasta effluents to UNRESTRICTED AREAS will be less than the concentration level; 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 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 Inter-national Commission on Radiological Protection (ICRP) Publication 2. This specification applies to the release of radioactive materials in liquid effluents from all units at the site. 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 Radio-chemistry," Annal. 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 Sec-tions 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 calcula-tion methodology and par.sneters in the ODCM implement the requirements in Sec-tion 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 equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid e#fluents are consistent with the methodology provided in Regulatory l Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of CATAWBA - UNITS 1 AND 2 B 3/4 11-1 JUL 181985 OCT 7 1985
RADIOACTIVE EFFLUENTS PROOF & RW 'g BASES DOSE (Continued) Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Apcendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimattag Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicsble, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharirg the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the liquid Radwaste Treatment System ensures that tMs system will be available for use whenever liquid effluents require treatment prior to release to the envieotwent. 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 reasonab7y achievable". This t.pecification 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. Inis specification applies to the release of radioactive materials in l liquid effluents from each unit at the site. When shared Radwaste Treatment l Systems are used by more than one unit or a site, the wastes from all units ' are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and i radioactivity concentrations, or, if not practicable, the treated effluent j releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance l to LCOs, these allocations from shared Radwaste Treatment Systems are to be l added to the releases specifically attributed to each unit to obtain the total I releases per unit. CATAWBA - UNITS 1 AND 2 B 3/4 11-2 R 11985 CUT 7 1985
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks included in this specification are 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 tank's 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. 3/4.11.1.5 CHEMICAL TREATMENT PONDS The inventory limits of the chemical treatment ponds (CTP) are based on limiting the consequences of an uncontrolled release of the pond inventory. The expression in Specification 3.11.1.5 assumes the pond inventory is uniformly mixed, that the pond is located in an uncontrolled area as defined in 10 CFR Part 20, and that the concentration limit in Note 1 to Appendix B of 10 CFR Part 20 applies. The batch limits of the resin / water slurry transferred to the CTP assure that radioactive material transferred to the CTP are "as low is reasonably achievable" in accordance with 10 CFR 50.36a. The expression in Specifica-tion 4.11.1.5 assures no batch will be transferred to the CTP unless the sum of the ratios of the' activity of the radionuclides to their respective concentra-tion limitation is less than the ratio of the 10 CFR Part 50, Appendix I, Section II.A, total body dose level to the 10 CFR 20.105(a), whole body dose limitation, or that: I b 3 mrem /yr
= 0.006
) C) < 500 mrem /yr Where:
c) = radioactive resin / water slurry concentration for radionuclide "j" entering the UNRESTRICTED AREA CTP, in microCuries/ milliliter; and C) = 10 CFR Part 20, Appendix B, Table II, Column 2, concentration for single radionuclide "j", in microcuries/ milliliter. l CATAWBA - UNITS 1 AND 2 B 3/4 11-3 JUL 181985 7g l _ _ - _ . - = - - _ .
. 1 RADI0 ACTIVE EFFLUENTS BASES CHEMICAL TREATMENT PONDS (Continued)
The filter /demineralizers using powdered resin and the blowdown demin-eralizer are backwashed or sluiced to a holding tank. The tank will be agitated to obtain a representative sample of the resin inventory in the tank. A known weight of the wet, drained resin (moisture content approximately 55 to 60%, bulk density of about 58 pounds per cubic foot) will then be counted. The concentra-tion of the resin slurry to be pumped to the chemical treatment ponds will then be determined by the formula: A i R c) = y . Where: Q.3
= concentration of radioactive materials in wet, drained resin for radionuclide "j", excluding tritium, dissolved or entrained noble gases, and radionuclides with less than an 8-day half-life.
The analysis shall include at least Ce-144, Cs-134, Cs-137, Co-58 and Co-60, in microcuries/ gram. Estimates of the Sr-89 and Sr-90 batch concentration shall be included based on the most recent monthly composite analysis (within 3 months); WR = total weight of resin in the storage tank in grams (determined from chemistry logs procedures); and V total volume of resin water mixture in storage tank to be T = transferred to the chemical treatment ponds in milliliters. The batch limits provide assurance that activity input to the CTP will be minimized, and a means of identifying radioactive material in the inventory limitation of Specification 3.11.1.5. 3/4.11.2 GASEOUS 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. i The annual dose limits are the doses associated with the concentrations of i 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE B0UNDARY, 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 l the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be ! sufficiently low to compensate for any increase in the atmospheric diffusion l CATAWBA - UNITS 1 AND 2 B 3/4 11-4 OCT 7 1983
RADIOACTIVE EFFLUENTS PROOF & REVIEW COPY k BASES DOSE RATE (Continued) factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY 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. This specification applies to the release of radioactive materials in gaseous effluents from all units at the site. The required detection capabilities for radioactive material 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 Radio-chemistry," 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 Sections II.B. 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 calcula-tional 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 noble gases 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 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. The ODCM equations provided for determining tne air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions. CATAWBA - UNITS 1 AND.2 B 3/4 11-5 JUL 181985 00T 7E
l l RADI0 ACTIVE EFFLUENTS $00F & WO l BASES DOSE - NOBLE GASES (Continued) This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treat-ment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based.on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactives waste producing units sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases epecifically attributed to each unit to obtain the total releases per unit. 3/4.11.2.3
~
DOSE - 10 DINE-131, 10 DINE-133. TRITIUM, AND RADIOACTIVE 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 de g9 ides 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 Pa 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 equa-tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 Iodine-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 BOUNDARY. The pathways that were examined in the development of the calculations were: (1) individual inhala-tion 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. l CATAWBA - UNITS 1 AND 2 B 3/4 11-6' JUL 181985 l QCT 7 585
RADIOACTIVE EFFLUENTS PROOF & REVIEW COPY I BASES i DOSE - IODINE-131. IODINE-133 TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM (Continued) This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformanco to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. 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 these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR 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 Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents. This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releares cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions,, e.g. , flow rates and , radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units ' sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Syste.ns are to be added to the releases specifically attributed to each unit to obtadn the total releases per unit. l CATAWBA - UNITS 1 AND 2 B 3/4 11-7 jy( 181985 00T 7 moa _ = _ - -. - .- _ _ - _ _ - . - _ _ - - . - - - - - - . - . _ _ . -
DRAFT RADI0 ACTIVE EFFLUENTS BASES M00f &8# p 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is main-tained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concen-trations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration 4 below the flammability limits.) 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 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 wnole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch lechnical 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 l 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 Ifmitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report when-ever the calculated doses due to releases of radioactivity and to 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. CATAWBA - UNITS 1 AND 2 B 3/4 11-8 JUL 181985 l QCT 1 MO
RADIOACTIVE EFFLUENTS g 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 units and from outside storage tanks are kept small. The Special Report will describe a course of action that should result in the Ifmitation 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 corrected), 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 apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 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 out any operation that is part of the nuclear fuel cycle. i l l CATAWBA - UNITS 1 AND 2 B 3/4 11-9 g y g )ggg OCT 71965
l PROOF & REVIEW COPY 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this specification provides representative measurements of radiation and of radio-l active 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 plant 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 affluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is - provided by the Radiological Assessment Branch Technical Position on Environ-i mental Monitoring. 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 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 measure-ment 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 ~ Determination - Application to l Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection
- Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford f
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 i required by the results of this census. The best information from the door- .
to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Sec-tion IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways 1 via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy J vegetables assumed in Regulatory Guide 1.109 for consumption by a child. 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/m 2, CATAWBA - UNITS 1 AND 2 B 3/4 12-1 JUL 18i EqE OCT
PROOF &RN N RADIOLOGICAL ENVIRONMENTAL MONITORING BASES - 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. I CATAWBA - UNITS 1 AND 2 B 3/4 12-2 JUL 181985 0CT 7 1985
PR00f & RNU 0 h5 SECTION 5.0 DESIGN FEATURES I l 1 JUL 181985 0CT 7 1953
5.0 DESIGN FEATURES PROOF & REVIEW COPY 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. I MADS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS 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 defini-tion 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 these 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, established at or beyond the SITE BOUNDARY, is utilized in the Limiting Conditions 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 containment structure is comprised of a steel containment vessel surrounded by a concrete containment having the following design features:
- a. Containment Vessel
- 1) Nominal 1.-ide diameter = 115 feet.
- 2) Nominal inside height = 171 feet.
- 3) Nominal thickness of vessel walls = 0.75 inch.
- 4) Nominal thickness of vessel dome = 0.6875 inch.
- 5) Nominal thickness of vessel bottom = 0.25 inch.
- 6) Net free volume = 1.2 x 108 cubic feet.
- b. Reactor Buildino -
- 1) Nominal Annular space = 6 feet.
- 2) Annulus nominal volume = 484,090 cubic feet.
- 3) Nominal outside height (top of foundation base to top of i dome) = 177 feet.
- 4) Nominal inside diameter = 127 feet.
l 5) Minimum cylinder wall thickness = 3 feet.
- 6) Minimum dome thickness = 2.25 feet.
j 7) Dome inside radius = 87 feet. l CATAWBA - UNITS 1 AND 2 5-1 $5 l JUL 18 l OCT 7 G22
Y PROOF & REVIEW SITE BOUNDARY / PERIMETER FENCE f gy .
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PRO 0F & REVIEW COPY DRAFT SITE BOUNDARY / PERIMETER FENCE NSW
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PROOF & REVIEW COPY DRAFT DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment vessel is designed and shall be maintained for a maximum internal pressure of 15 psig and a temperature of 328*F. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The 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 1619 grams uranium. The initial core loading shall have a maximum enrich-ment of 3.5 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading ar.i shall have a maximum enrichment of 3.5 weight percent U-235. CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length control rod assemblies. The full-length control rod assemblies shall 4:ontain a nominal 142 inches of absorber material of which 102 inches shall be 100% boron carbide and remaining 40-inch tip shall be 80% silver, 15% indium, end 5% cadmium. All control rods shall be 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 normai degradation pursuant to the applicable Surveillance Requi,rements,
- b. For a pressure a 2485 psig, and
- c. For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,040 1 100 cubic feet at a nominal T,yg of 525*F. 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1. l l I CATAWBA <JNITS 1 AND 2 5-6 JUL 181985 00T 7 1985
. . ~ . . . . . .
PROOF & REVIEW COPY OESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY
- 5.6.1 The new and spent fuel storage racks are designed and shall be maintained with:
- a. A k,ff equivalent to less than or equal to 0.95 when flooded with un'aorated water, which includes a conservative allowance for uncer-tainties as described in Section 9.1.2.3.1 of the FSAR, and
- b. A nomin21 21-inch fcr new fuel and a nominal 13.5-inch for spent fuel center-to-center distance between fuel assemblies placed in the storage racks.
DRAINAGE 5.6.2 The spert fuel storage pool is designed and shall be maintained to prevent inadve'.* tent draining of the pool below elevation 596 feet. CAPACITY 5.6.3 Tb2 spent fuel storage pool is designed and shall be maint41ned with a storags capacity limited to no more than 1418 fuel assemblies. 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 CATAWBA - UNITS 1 AND 2 5-7 Al, N
TABLE 5.7-1 9 g COMPONENT CYCLIC OR TRANSIENT LIMITS 5 s CYCLIC OR DESIGN CYCLE g COMPONENT TRANSIENT LI!ilT OR TRANSIENT Reactor Coolant System 200 heatup cycles at < 100*F/h Heatup cycle - T from < 200*F w and 200 cooldown cyclis at **9 ~ to > 550*F. g < 100*F/h. Cooldo.sn cycle - T from o
" > 550*F to 1 200*F**9 200 pressurizer cooldown cycles Pressurizer cooldown cycle at i 200*F/h. temperatures from > 650 F to 5 200*F.
80 less of load cycles, without > 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 Spray water temperature differential actuation cycles. > 320*F. 50 leak tests. Pressurized to > 2485 psig. 5 hydrostatic pressure tests. Pressurized to > 3106 psig. 3 c) Secondary Coolant System E I steam line break. Break in a > 6-inch steam line. '@ r- C 90 5 hydrostatic pressure tests. Pressurized to > 1350 psig. _ ym O
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PROOF & REVIEW COPY DRAFT SECTION 6.0 ADMINISTRATIVE CONTROLS I I 1 i l JUL 181985 00T 7 1985
i l PROOP & REVIEW COPY ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Manager shall be responsible for overall unit opera-tion and shall delegate in writing the succession to this responsibility during his absence.
~
6.1. 2 The Shift Supervisor (or during his absence from the contro1 room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President of Nuclear Production, shall be reissued to all Nuclear Production Department station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and te& .al 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:
- a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1;
- b. At least one licensed Operator for each unit shall be in the control room when fuel is in the reactor. In addition, while either unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;
- c. A Health Physics Technician
- shall be on site when fuel is in either 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 five members
- shall be maintained on site at all times. The Fire Brigade shall not include the three 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 "The Health Physics Technician and Fire Brigade composition 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 is tdken to fill the required positions.
I 00T 7N CATAWBA - UNITS 1 AND 2 6-1 l
PROOF & REVIEW COPY ADMINISTRATIVE CONTROL 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. , licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel).
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40-hour week while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for re-fueling, major maintenance, or major plant modification, on a tempo-rary basis the following guidelines shall be followed:
- 1) An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
- 2) An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period, all excluding shift turnover time.
- 3) A break of at least 8 hours should be allowed between work periods, including shift turnover time.
- 4) Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the Station Manager or his designee, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Station Manager or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized. 1 CATAWBA - UNITS 1 AND 2 6-2 dd t985
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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION Both Units in One Unit in Mode 1, 2, 3 or 4 Both Units in Mode 5 or 6 and Mode 1, 2, 3 or 4 or Defueled One Unit in Mode 5 or 6 Defueled SS 1 1 1 SRO 1 None## 1 R0 3# 2# 3# NEO 3# 3# 3# STA 1 None 1 SS - Shift Supervisor with a Senior Operator license on Unit 1 SRO - Individual with a Senior Operator license on Unit 1 RO - Individual with an Operator license on Unit l' NEO - Nuclear Equipment Operator STA - Shift Technical Advisor The Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. 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 l from the control room while the unit is in MODE 5 or 6, an individual with a l valid Senior Operator license or Operator license shall be designated to assume the control room command function. m
*0n occasion when there is a need for both the Shift Supervisor and the SR0 to be absent from the control room, the STA shall be allowed to assume the centrol room command function and serve as the SRO in the control room provided that: (1) the Shift Supervisor is available to return to the l control room within 10 minutes, (2) the assumption of SRO duties by the STA l be limited to periods not in excess oT 15 minutes duration and a total time not to exceed I hour during any 8-hour shift, and (3) the STA has a Senior Operator license on the unit. #At least one of the required individuals must be assigned to the designated position for each unit. ##At least one licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling must be present during COR'JALTERATIONS on either unit, who has no other concurrent responsibilities.
CATAWBA - UNITS 1 AND 2 6-5 ggi "I N l
ADMINISTRATIVE CONTROLS PROOF & REymw mov 6.2.3 CATAWBA SAFETY REVIEW GROUP
, FUNCTION 4
6.2.3.1 The Catawba Safety Review Group (CSRG) shall function to' examine I i plant operating characteristics, NRC issuances, industry advisories, REPORTABLE EVENTS, and other sources which may indicate areas for improving 4 plant safety. The CSRG shall make detailed recommendations for revised pro- - cedures, equipment modifications, or other means of improving plant safety to the Director, Nuclear Safety Review Board. , COMPOSITION 6.2.3.2 The CSRG shall be composed of a chairman and at least four dedicated, full-time qualified individuals located onsite. i RESPONSIBILITIES -
- 6.2.3.3 The CSRG 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.
l RECORDS 5 6.2.3.4 Records of activities performed by the CSRG shall bg prepared, maintained, Review Board. and forwarded each calendar month to the Dira1stn? Nuclear Safety i i i 6.2.4 SHIFT TECHNICAL ADVISOR i 6.2.4.1 The Shift Technical Advisor shall serve in an advisory capacity to the Shift Supervisor.
- 6. 3 UNIT STAFF QUALIFICATIONS 6.3.1. Each member of the unit staff shall meet or exceed the minimum qualifica-i
= tions of ANSI N18.1-1971 for comparable positions, except for the Radiation Protection Manager, who shall meet or exceed the qualifications outlined in i the Catawba FSAR, Section 13.1.3. The licensed Operators and Senior Operator j
- shall also meet or exceed the minimum qualifications of the supplemental requ,sire-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Station Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified i in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the CSRG. ,
*Not responsible for sign-off function.
CATAWBA - UNITS 1 AND 2 6-6
ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT 6.5.1 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.1.1 Each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, shall be prepared by a qualified individual / organization. Each such procedure, and changes thereto, shall be reviewed by an individual / group other than the individual / group which prepared the procedure, or changes thereto, but who may be from the same organization as the individual / group which prepared the procedure, or changes thereto. 6.5.1.2 Proposed changes tn the Appendix A Technical Specifications shall be prepared by a qualified individual / organization. The preparation of each proposed Technical Specification change shall be reviewed by an individual / group other than the individual / group which prepared the proposed change, but who may be from the same organization as the individual / group which prepared the proposed change. Proposed changes to the Technical Specifications shall be approved by the Station Manager. 6.5.1.3 Proposed modifications to unit nuclear safety-related structt.res, systems, and components shall be designed by a qualified individual / organization. Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modification. Proposed modifications to nuclear safety related structures, systems, and components shall be approved prior to implementation by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent, or the Main-tenance Superintendent, as previously designated by the Station Manager. 6.5.1.4 Individuals responsible for reviews performed in accordance with Specifications 6.5.1.1, 6.5.1.2, and 6.5.1.3 shall be members of the station supervisory staff, previously designated by the Station Manager to perform such reviews. Review of environmental radiological analysis procedures shall l be performed by the Corporate System Health Physicist or his designee. Each [ such review shall include a determination of whether or not additional, cross-j disciplinary, review is necessary. If deemed necessary, such review shall be
- performed by the appropriate designated station review personnel.
l 6.5.1.5 Proposed tests and experiments which affect station nuclear safety ' and are not addressed in the FSAR or Technical Specifications shall be reviewed l by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent or the Maintenance Superintendent, as previously designated by the Station Manager. l l JUL 181985 CATAWBA - UNITS 1 AND 2 6-7 7 1985
ADMINISTRATIVE CONTROLS TECHNICAL REVIEW AND CONTROL ACTIVITIES (Continued) 6.5.1.6 All REPORTABLE EVENTS and all violations of Technical Specifications shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence. Sucn reports shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board. 6.5.1.7 The Station Manager shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President, Nuclear Production. 6.5.1.8 The station security program, and implementing procedures shall be reviewed at least once per 12 months. Recommended changes shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board. 6.5.1.9 The station emergency plan, and implementing procedures, shall be reviewed at least once per 12 months. Recommended changes shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board. 6.5.1.10 The Station Manager shall assure the performance of a review by a qualified individual / organization of every unplanned onsite release of radio-active material to the environs including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective ACTION to prevent recurrence to the Vice President, Nuclear Production and to the Nuclear Safety Review Board. 6.5.1.11 The Station Manager shall assure the performance of a review by a qualifed individual / organization of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and Radwaste Treatment Systems. 6.5.1.12 Reports documenting each of the activities performed under Specifications 6.5.1.1 through 6.5.1.11 shall be maintained. Copies shall be provided to the Vice President, Nuclear Production, and the Nuclear Safety Review Board. 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) FUNCTION 6.5.2.1 The NSRB shall function to provide independent review and audit of designated activities in the areas of:
- a. Nuclear power plant operations, l b. Nuclear engineering,
! c. Chemistry and radiochemistry, CATAWBA - UNITS 1 AND 2 6-8 JUL 8 g5g l l
Proof g p 1 COPY ADMINISTRATIVE CONTROLS FUNCTION (Continued)
- d. Metallurgy,
- e. Instrumentation and control,
- f. Radiological safety,
- g. Mechanical and electrical engineering, and
- h. Administrative control and quality assurance practices.
The NSRB shall report to and advise the Vice President, Nuclear Production, on those areas of responsibility specified in Specifications 6.5.2.8 and 6.5.2.9. ORGANIZATION 6.5.2.2 The Director, members, and alternate members of the NSRB shall be appointed in writing by the Vice President, Nuclear Production, and shall have an academic degree in an engineering or physical science field; and in addition, shall have a minimum of 5 years technical experience, of which a minimum of 3 years shall be in one or more areas given in Specification 6.5.2.1. No more than two alternates shall participate as voting members in NSRB activities at any one time. 6.5.2.3 The NSRB shall be composed of at least five members, including the Director. Members of the NSRB may be from the Nuclear Production Department, from other departments within the Company, or from external to the Company. A maximum of one member of the NSRB may be from the Catawba Nuclear Station staff. 6.5.2.4 Consultants shall be utilized as determined by the NSRB Director to provide expert advice to the NSRB. 6.5.2.5 Staff assistance may be provided to the NSRB in order to promote the proper, timely, and expeditious performance of its functions. 6.5.2.6 The NSRB 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. l 6.5.2.7 The quorum of the NSI.B necessary for the performance of the NSRB review and audit functions of these Technical Specifications shall consist of the Director, or his designated alternate, and at least four other NSRB members including alternates. No more than a minority of the quorum shall have line responsibility for operation of Catawba Nuclear Station. l t 1 i CATAWBA - UNITS 1 AND 2 6-9 00T E 985
, , ~ - - -
PR00fg COPY ADMINISTRATIVE CONTROLS VC REVIEW 6.5.2.8 The NSRB shall be responsible for the review of:
- a. The safety evaluation for: (1) changes to procedures, equipment, or systems, and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR to verify that such actions did not consti-tute an unreviewed safety question.
- b. Proposed changes to procedures, equipment, or systems 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 in 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 affec.t nuclear safety;
- g. All REPORTABLE EVENTS;
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; .
- 1. Quality Assurance Department audits relating to station operations and actions taken in response to these audits; and
- j. Reports of activities performed under the provisions of Specifications 6.5.1.1 through 6.5.1.11.
AUDITS ! 6.5.2.9 Audits of unit activities shall be performed under the cognizance of ' l the NSRB. 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 once per 12 months; CATAWBA - UNITS 1 AND 2 6-10 985 T
_ _ - - . . a. W00f &WIEwcofy ADMINISTRATIVE CONTROLS sob AUDITS (Continued)
- c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or method of operation that affect nuclear safety at least once per 6 months;
- 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 Emergency Plan and implementing procedures at least once per 12 months;
- f. The Security Plan and implementing procedures at least once per 12 months;
- g. The Facility Fire Protection programmatic controls including the implementing procedures at least once per 24 month; by qualified licensee QA personnel;
- h. 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;
- i. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 month;;
- j. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;
- k. The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months;
- 1. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months; and ,
- m. Any other area of unit operation considered appropriate by the NSRB or the Vice President, Nuclear Production.
RECORDS 6.5.2.10 Records of NSRB activites shall be prepared, approved, and distributed as indicated below: l
- a. Minutes of each NSRB meeting shall be prepared, approved, and j forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Power Operations, within 14 days following each meeting; JUL 181985 CATAWBA - UNITS 1 AND 2 6-11 gi 7 $85
PROOF & REVIEW COPY l ADMINISTRATIVE CONTROLS RECORDS (Continued)
- b. Reports of reviews encompassed by Specification 6.5.2.8 above, shall be prepared, approved, and forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Power Operations, within 14 days following completion of the review; and,
- c. Audit reports encompassed by Specification 6.5.2.9 above, shall be forwarded to the Vice President, Nuclear Production, and to the Execu-tive Vice President, Power Operations, and to the management post-tions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.
6.6 REPORTABLE EVENT ACTION 1 l 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: 4
- a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 tn 10 CFR Part 50 and
- b. Each REPORTABLE EVENT shall be reviewed by the Station Manager; or by (1) Operating Superintendent; (2) Technical Services Superintendent; (3) Maintenance Superintendent; or (4) Superintendent of Integrated Scheduling, as previously designated by the Station Manager, and the results of this review shall be submitted to the NSRB and the Vice President-Nuclear Production.
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 possible and in all cases within 1 hour. The Vice President-Nuclear Production and the NSRB shall be notified within 24 hours,
- b. A Safety Limit Violatien Reoort shall be prepared. The report shall be reviewed by the Operating Superintendent and Station Manager.
i 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; I
- c. The Safety Limit Violation Report shall be submitted to the Commission, the NSRB and the Vice President-Nuclear Production within 14 days of the violation; and t
i d. Critical operation of the unit shall not be resumed until authorized by the Commission. 00T 7 BBS l CATAWBA - UNITS 1 AND 2 6-12 1 l- - - . . . . . - , . - - , - . . - - - - - - - . , . , . . - . - - . . . . . - . _ -
- . . . - ~
~-
PROOF & REVIEW COPY ADMINISTRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and nPaintained covering the activities referenced below: ,
- a. The applicable procedures rccommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;
- b. The emergency operating procedures required to ireplement the require-ments of NUREG-0737 and Supplement No. 1 to NUREG-0737 as stated in Generic Letter No. 82-33;
- c. Security Plan implementation;
! d. Emergency Plan implementation; 1
- e. PROCESS CONTROL PROGRAM implementation;
- f. OFFSITE DOSE CALCULATION MANUAL implamentation; and
- g. Quality Assurance Program for effluent monitoring and implemen-tation of the environmental radiological monitoring program.
6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed and approved by the Station Manager; or by: (1) Operating Superin-tenden , (2) Technical Services Superintendent, or (3) Maintenance Superinten-dent, as previously designated by the Station Manager; prior to implementation and shall be reviewed periodically as set forth in administrative procedures. 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made pro-vided:
- a. The intent of the original procedure is not altered;
- b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and
- c. The change is documented, reviewed, and approved by the Station Manager; or by: (1) Operating Superintendent, (2) Technical Services Superintendent, or (3) Maintenance Superintendent, as previously designated by the Station Manager; within 14 days of implementation.
6.8.4 The following programs shall be established, implemented, and maintained: I a. Primarv Coolant Sources Outside Containment . A program to reduce leakage f ; ; .ose portions of systerfis outside containment that could contain ...gnly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the containment spray, Safety Injection, chemical and volume dentrol, and nuclear sa.npling. The program shall include the following: 001 7N CATAWBA - UNITS 1 AND 2 6-13 _ __ __ . _ _ - ~ - - . _ . _ - -
_ _ . . ._ .____) ADMINISTRATIVE CONTROLS ' PROCEDURES AND PROGRAMS (Continued)
- 1) Preventive maintenance and periodic visual inspection requireinents, 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 actions 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. Backup Method for Determining Subcoolina Margin i A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:
- 1) Training of personnel, and
- 2) Procedures for monitoring.
i l JUL 181985 l i CATAWBA - UNITS 1 AND 2 6-14 OCT 1@
- - - ~~~-
ADMINISTRATIVE CONTROLS PROOF & REVIEW COPY PROCEDURES AND PROGRAMS (Continued) .
- e. 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.
- 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.
STARTUP REPORT 6.9.1.1 A sumary 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 unit. 6.9.1.2 The Startup Report shall address each of the tests identified in the l Final Safety Analysis Report 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 specifica-tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in - license conditions based on other commitments shall be included in this report. ,
- 6. 9.1. 3 Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or ,
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. I jl)L 18 N OCT 7N CATAWBA - UNITS 1 AND 2 6-15
3 ADMINISTRATIVE CONTROLS yf & kNbN bb 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 V station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /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 [ describe maintenance], waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermo-luminescent 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 functions. 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 prior to May 1 of each year. The initial report shall be submitted prior to 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 radiologicai environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports snall also include the results of the Land Use Census required by Specification 3.12.2. l l l *A single sut,mittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.
**This tabulation supplements the requirements of 9 20.407 of 10 CFR Part 20.
A single submittal may be made for a multiple unit station. JUL 181985 0g 1 to) l CATAWBA - UNITS 1 AND 2 6-16
l l goot REVIEW COPY l ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) The Annual Radiological Environmental Operating Reports shalf 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 00CM, 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 radiblogical environmental monitoring program; at least two legible maps
- covering all sampling locations keyed to a table giving distances and directions from the centerline of one 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 Specifica-tion 3.12.3; discussion of all deviations from the sampling schedule of Table 3.12-1; reasons for not conducting the Radiological Environmental Monitor-ing Program as required by Specification 3.12.1 and discussion of environmental sample measurements that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to ACTION b. of Specification 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
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, Evaluating, ar.d 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. The Radioactive Effluent l
*0ne map shall cover stations near the SITE BOUNDARY; a second shall include the mare distant stations. -
**A si .gle submittal may be made for a multiple unit station. The submittal shouio combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
OCT 7# CATAWBA - UNITS 1 AND 2 6-17 l [
J ADMINISTRATIVE CONTROLS PROOF & REVIEW COPY SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) Release Reports shall include the following information for each type of solid waste (as defined by 10CFR Part 61) shipped offsite during the report period:
- a. Total Container volume, in meters,
- b. Total Curie quantity (determined by measurement or estimate),
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- c. Principal radionuclides (determined by measurement or estimate),
l d. Type of waste (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),
- e. Number of shipments, and
- f. Solidification agent or absorbent [e.g., cement or other approved agents (media)].
! The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form on an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, 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 i
from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive i liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period. All assump-tions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological 1 conditions concurrent with the time of release of radioactive materials in
~ gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be perforned in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).
The Semiannual Radioactive Effluent Release Report to be submitted within 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 i pathways and direct radiation, for the previous calendar year to show conformance j with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear j 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. l
*In lieu of submission with the Semiannual Radioactive Effluent Release Report, i
the licensee has the option of retaining this summary of required meteoro1c,gical data on site in a fille that shall be provided to the NRC upon request. OCT 1W CATAWBA - UNITS 1 AND 2 6-18 I ( ,
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. = . . _.- . _-. _.. . . .-.
_ . _... _ ._ j PROOFgggyggy(gpy ADMINISTRATIVE CON 1ROLS SEMIANNUAL RADI0ACTI'l- EFFLUENT RELEASE REPORT (Continued) The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period. 1 The Semiannual Radioactive Effluent Release Reports shall include any changes t made during the report.ing period to the PCP and to the ODCM, pursuant to Specifi-l cations 6.13 and 6.14, respectively, as well as any major changes to Liquid, Gaseous or Solid Radwaste Treatment Systems, pursuant to Specification 6.15. j It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2. ! The Semiannual Radioactive Effluent Release Reports shall also include the following: 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 REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, l 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 4 copy to the Regional Administrator of the Regional Office of the NRC, no later ' than the 15th of each month following the calendar month covered by the report. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.9 The F,y limit for RATED THERMAL POWER (FRTP) shall be provided to x the Regional Administrator of the Regional Office of the NRC with a copy to the Director, Nuclear Reactor Regulation, Attention: Chief Core Performance i 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 .PRel) vs Axial Core Height with the limit envelope l at least 60 days prior to each cycle initial criticality unless othentise I 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 Radial ! Peaking Factor Limit Report, it will 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 FxRTP will be by request from the i NRC and need not be included in this report. i i
*In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological
, data on site in a file that shall be provided to the NRC upon request. oup CATAWBA - UNITS 1 AND 2 6-19 OG i i
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PROOF i REVIEW COPY ADMINISTRATIVE CONTROLS ECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report. 6.10 RECORD RETENTION 6.10.1 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. 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 required by these Technical Specifications;
- 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, fuel transfers, and )
assembly burnup histories; ;
- c. Records of radiation exposure for all individuals entering radiation I control areas; CATAWBA - UNITS 1 AND 2 6-20 N
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I ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) j 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 tests and expeH::ent;; ,
- g. Records of training and qualification for current members of the unit staff;
- h. Records of inservice inspections performed pursuant to these Technical Specifications;
; i. Records of quality assurance activities required by the Operational Quality Assurance 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 NSRB and reports required by Specification 6.5.1.10;
- 1. Records of the service lives of all hydraulic and mechanical snubbers
, required by Sptcification 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 analyses 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
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effective at specified times and QA records showing that these procedures were followed. , 6.11 RADIATION PROTECTION PROGRAM 6.11 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. 4 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR Part 20, 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 radiation source or from any curface which j the radiation penetrates shall be barricaded and conspicuously positoi as a high radiation area and entrance thereto shall be controlled by requiring 6-21 00I 7 Nb ; CATAWBA - UNITS 1 AND 2
- - -- = - - . - _ - --- _- - - .. -. - -. - - . - - _ - - - . - _ = - - - - .
COPY ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continu-ously escorted by such individuals may be exempt from the RWP issuance require-ment 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 other-wise following plant radiation protection 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 estab-lished and personnel have been made knowledgeable of them; or
- c. An individual qualified in radiation protection procedures with a radiation cose rate monitoring device, who is responsible for pro-viding positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Station Health Physicist in the RWP.
6.12.2 In addition to the requirements of Specification 6.12.1, areas accessi-ble 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 on duty and/or health physics supervision. Doors shall remain locked except during l periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous sur-veillance may be made by personnel qualified in radiation protection procedures l to provide positive exposure control over the, activities being performed 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 l PWR containment, 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. l l CATAWBA - UNITS 1 AND 2 6-22
PROOFg psyggy (gpy ADMINISTRATIVE CONTROLS i
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6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to impi mentation. 6.13.2 Licensee-initiated changes to the PCP: i
- 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 totally support the i rationale for the change without benefit of additional or j 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 J 3) Documentation of the fact that the change has been reviewed and i
found acceptable by the Station Manager.
- b. Shall become effective upon review and acceptance by a qualified individual / organization.
6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) 6.14.1 The 00CM shall be' approved by the Commission prior to implementation. 6.14.2 Licensee-initiated changes to the ODCM: i l
- a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the 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 l
supplemental information. Information submitted should consist
' of a package of those pages of the ODCM to be changed with each j page numbered, dated, and containing the revision number, ' together with appropriate analyses or evaluations justifying the change (s);
1
- 2) A determination that the change will not reduce the accuracy or reliability of dose calculations of Setpoint determinations; and \
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- 3) Documentation of the fact that the change has been reviewed and r
' found acceptable by the Station Manager.
- b. Shall become effective upon review and acceptance by a qualified individual / organc zation.
g 'l W% CATAWBA - UNITS 1 AND 2 6-23 f
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l PROOF &' REVIEW COPY ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO LIQUID, GASE0US, AND SOLID RADWASTE TREATMENT SYSTEMS
- 6.15 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):
- 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 Station Manager. 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) Sufficie..t detailed information to totally support the reason for the change without benefit of additional or supplemental informations
- 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 chang?, 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;
- 5) 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;
- 7) An estimate of the exposure to plant operating personnel as a result of the change; and.
- 8) Documentation of the fact that,the change was reviewed and found acceptable by the Station Manager.
- b. Shall become effective upon review and acceptance by a qualified individual / organization.
- Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.
001 7O CATAWBA - UNITS 1 AND 2 6-24
PROOFgpgy;gy(9py b [ DELETED] 0 CATAWBA - UNITS 1 AND 2 7-1}}