ML22340A197
Text
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 5 LIST OF REACTOR TRIPS AND ACTUATION MEANS OF: ENGINEERED SAFETY FEATURES, CONTAINMENT AND STEAM LINE ISOLATION & AUXILIARY FEEDWATER Initiator Coincidence Circuitry and Interlocks Comments Reactor Trip
- 1. Manual 1/2, no interlocks
- 2. Neutron flux (Power Range):
High and low settings; manual block and automatic reset 2A. High neutron flux 2/4, low setpoint interlocked with P-10 of low setting by P-10, Table 7.2-2 2B. High positive neutron flux rate 2/4, no interlocks
- 3. Overtemperature T 2/4, no interlocks
- 4. Overpower T 2/4, no interlocks
- 5. Low pressurizer pressure 2/4, interlocked with P-7
- 6. High pressurizer pressure 2/4, no interlocks
- 7. High pressurizer water level 2/3, interlocked with P-7 Blocked below P-7. Low flow in 1 loop permitted below
- 8. Low reactor coolant flow 2/3 signals per loop, interlocked with P-7 and P-8 P-8.
Monitored electrical supply to 9.
reactor coolant pumps:
2/4 bus undervoltage signals will actuate a reactor trip (interlocked with P-7) and will also actuate the autostart 9A. Undervoltage 2/4, interlocked with P-7 of the turbine driven auxiliary feedwater pump (not interlocked with P-7).
2/4 underfrequency signals will actuate a reactor trip:
9B. Underfrequency 2/4, interlocked with P-7 (interlocked with P-7).
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 5 LIST OF REACTOR TRIPS AND ACTUATION MEANS OF: ENGINEERED SAFETY FEATURES, CONTAINMENT AND STEAM LINE ISOLATION & AUXILIARY FEEDWATER Initiator Coincidence Circuitry and Interlocks Comments 2/4 breakers open will cause a reactor trip above P-7.
9C. Reactor coolant pump breaker position 2/4, interlocked with P-7 Blocked below P-7.
Trips main feedwater pumps. Closes feedwater control Manual 1/2 panel switches (per train).
valves. Closes feedwater isolation valves. Initiates Phase 2/3 low pressurizer pressure.
A isolation. Actuation by pressurizer pressure may be 2/3 high containment pressure.
- 10. Safety injection signal manually blocked below P11 and is automatically 2/3 differential steam line pressure signals of one unblocked above P-11.
line compared with the other three lines.
2/3 high containment pressure actuates the Containment 2/4 Low steam line pressure.
Air Recirculation Hydrogen Skimmer System Fans.
Interlocked with P-8 OR 4/4 stop valves closed
- 11. Turbine-generator trip 31 percent power (P-8) signal interlocked P-8.
1/2 steam/feedwater flow mismatch in
- 12. Low Feedwater Flow coincidence with 1/2 low steam generator water level, per loop.
- 13. Low-low steam generator water level 2/3, per loop
- 14. Intermediate range neutron flux 1/2, manual block permitted by P-10 Manual block and automatic reset 1/2, manual block permitted by P-6, interlocked
- 15. Source range neutron flux Manual block and automatic reset with P-10 Containment Isolation Actuation Actuates all non-essential process lines containment
- 16. Containment pressure Same as Item 10 or 1/2 manual isolation trip valves. (Isolation Phase A)
Actuates all remaining trip valves (except those required Coincidence of 2/4 containment Hi-Hi pressure for operation of engineered safeguard system).
or 1/2 manual (Isolation Phase B)
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 5 LIST OF REACTOR TRIPS AND ACTUATION MEANS OF: ENGINEERED SAFETY FEATURES, CONTAINMENT AND STEAM LINE ISOLATION & AUXILIARY FEEDWATER Initiator Coincidence Circuitry and Interlocks Comments 1/3 high activity signal, from containment area, Closes containment purge supply, exhaust ducts and all
- 17. High containment activity air particulate, or noble gas monitors or manual other necessary to isolate containment atmosphere.
initiation of Phase A or Phase B isolation (Containment Ventilation Isolation)
Engineering Safeguards Systems Actuation
- 18. Safety injection signal See Item 10 Hi-Hi-containment pressure (2/4) or manual 1/1
- 19. Containment spray signal (per train)
- 20. NaOH addition Containment Spray Actuation Signal Steam Lines Isolation Actuation High Steam flow (any 2 steam lines out of 4)
- 21. Steam flow coincident with low-low Tavg .
- 22. Steamline Pressure 2/4 Low Steamline Pressure
- 23. Containment pressure 2/4 Hi-Hi containment pressure signal
- 24. Manual (per steam line) 1/1 per steam line (per train)
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 5 LIST OF REACTOR TRIPS AND ACTUATION MEANS OF: ENGINEERED SAFETY FEATURES, CONTAINMENT AND STEAM LINE ISOLATION & AUXILIARY FEEDWATER Initiator Coincidence Circuitry and Interlocks Comments Auxiliary Feedwater Actuation Coincidence of 2/3 low-low level in two steam generators or undervoltage on 2/4 RCP busses or 2/3 high level in steam generator trips main feedwater
- 24. Turbine driven pump 3/4 low feedwater flow coincident with power pumps above 40% (AMSAC) or manual (local and remote) 2/3 low-low level in any steam generator: or trip of both main feedwater pumps, or safety injection signal, manual (local and remote); 3/4
- 25. Motor driven pumps low feedwaterflow coincident with power 40%
(AMSAC); or 2/3 per bus (T11A, T11D) to start pump; valve actuation or 2/3 per bus on 2 busses (T11A and T11B or T11C and T11D)
Main Feedwater Isolation Feedwater isolation signal closes main feedwater control valves (fast closure) and feedwater isolation valves. SI
- 26. Safety Injection See No. 10 signal also trips main feedwater pumps which causes subsequent feedwater pump discharge valve closure.
Feedwater isolation signal closes main feedwater control valves (fast closure) and feedwater isolation valves. Hi-27 Hi-Hi Steam Generator level. 2/3 high-high in 1/4 steam generators. Hi steam generator level signal also trips main feedwater pumps which causes subsequent feedwater pump discharge valve closure.
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 5 of 5 LIST OF REACTOR TRIPS AND ACTUATION MEANS OF: ENGINEERED SAFETY FEATURES, CONTAINMENT AND STEAM LINE ISOLATION & AUXILIARY FEEDWATER Initiator Coincidence Circuitry and Interlocks Comments Feedwater isolation signal closes main feedwater control valves (fast closure) and feedwater isolation valves.
Reactor Trip coincident with low Tavg.
28 Reactor Trip Interlock P-4 signal also trips main feedwater pumps (Interlock P-4, see Table 7.2-2) which causes subsequent feedwater pump discharge valve closure
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 4 INTERLOCK CIRCUITS Note: Where applicable setpoints can be found in the Technical Specifications for each unit, or in appropriate engineering documents.
Designation Derivation Function 1/2 Reactor trip breakers and its corresponding P-4 Actuates main turbine trip.
bypass breaker open Initiates feedwater isolation on Tavg below setpoint.
Actuates main feedwater pump trip.
Blocks re-actuation of safety injection after manual reset of safety injection actuation signal.
Allows manual block of source range P-6 1/2 Intermediate range neutron flux above setpoint reactor trip.
Prevents or defeats the manual block of P-6 Reset 2/2 Intermediate range neutron flux below reset point source range reactor trip.
Enabled by P-10 or P 2/4 Power range neutron Permits reactor trip when any of the P-7 flux channels above setpoint (P-10) or 1/2 turbine following conditions are sensed:
first stage pressure above setpoint (P-13)
- Greater than one loop reactor coolant flow low
- Greater than one reactor coolant pump breaker open
- Reactor coolant pump bus undervoltage or underfrequency
- Pressurizer low pressure
- Pressurizer high level Prevents or defeats the reactor trip when P-7 Reset P-10 reset and P-13 reset any of the following conditions are sensed:
- Greater than one loop reactor coolant flow low
- Greater than one reactor coolant pump breaker open
- Reactor coolant pump bus undervoltage or underfrequency
- Pressurizer low pressure
- Pressurizer high level
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 4 INTERLOCK CIRCUITS Note: Where applicable setpoints can be found in the Technical Specifications for each unit, or in appropriate engineering documents.
Designation Derivation Function Prevents or defeats the automatic block of P-8 2/4 power range channels above setpoint reactor trip on low coolant flow condition in a single loop.
Permits reactor trip when the Turbine Trip condition is sensed.
Permits the automatic block of reactor trip P-8 Reset 3/4 NIS power range channels below reset point on low flow in a single loop.
Prevents or defeats the reactor trip when the Turbine Trip condition is sensed.
2/4 power range neutron flux channels above P-10 Inputs to P-7 permissive.
setpoint Permits the manual block of reactor trip on:
- Intermediate range high neutron flux level.
- Power range channel low setpoint high neutron flux level.
Permits manual block of intermediate range channel rod stop.
Permits automatic block of source range channel trip.
Prevents or defeats the manual block of P-10 Reset 3/4 power range channels below reset point reactor trip on:
- Intermediate range channel high neutron flux level.
- Power range channel low setpoint high neutron flux level.
Prevents or defeats the manual block of intermediate range channel rod stop.
Permits manual block of safety injection P-11 2/3 pressurizer pressure below setpoint actuation on low pressurizer pressure.
Prevents or defeats manual block of safety P-11 Reset 2/3 pressurizer pressure above reset point injection actuation on low pressurizer pressure.
Permits manual block of safety injection P-12 2/4 Tavg channels below setpoint on low steam line pressure.
Permits or causes steam line isolation on high steam line flow.
Blocks condenser steam dump.
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 4 INTERLOCK CIRCUITS Note: Where applicable setpoints can be found in the Technical Specifications for each unit, or in appropriate engineering documents.
Designation Derivation Function Prevents or defeats manual block of safety P-12 Reset 3/4 Tavg channels above reset point injection on low steam line pressure.
Prevents or defeats steam line isolation on high steam flow.
Permits condenser steam dump.
1/2 turbine first stage pressure channel above P-13 Inputs to P-7.
setpoint P-13 Reset 2/2 turbine first stage pressure below reset point Inputs to P-7.
P-14 2/3 hi-hi steam generator level (any steam Permits the initiation of:
generator) greater than or equal to setpoint
- Feedwater isolation
- Main feedwater pump trip.
- Main turbine trip.
P-14 Reset 2/3 hi-hi steam generator level (any steam Prevents or defeats initiation of:
generator) less than or equal to reset point
- Feedwater isolation.
- Main feedwater pump trip.
- Min turbine trip.
Blocks automatic and manual control C-1 1/2 Intermediate range neutron flux above setpoint rod withdrawal.
May be manually blocked above P-10.
Automatically unblocked below P-10.
Blocks automatic and manual control C-2 1/4 Power range neutron flux above setpoint rod withdrawal.
Blocks automatic and manual control C-3 2/4 Overtemperature Delta T above setpoint rod withdrawal.
Actuates turbine runback via load reference.
Blocks automatic and manual rod C-4 2/4 overpower Delta T channels above setpoint withdrawal and initiates turbine runback C-5 Turbine impulse pressure equivalent below setpoint Blocks automatic rod withdrawal
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 4 INTERLOCK CIRCUITS Note: Where applicable setpoints can be found in the Technical Specifications for each unit, or in appropriate engineering documents.
Designation Derivation Function 1/1 Time Derivative (absolute value) of turbine Makes steam dump valves available C-7A, C-B First Stage pressure (decrease only) above setpoint for either tripping or modulation.
2/3 Turbine Emergency Trip Fluid pressure below Blocks steam dump control via load C-8 setpoint or 4/4 stop valves closed rejection T avg controller.
2/3 Turbine emergency trip fluid pressure above Blocks steam dump control via turbine setpoint or 4/4 stop valves not closed trip T avg controller.
Any condenser vacuum above setpoint or All C-9 circulation water pump breakers open or Loss of Blocks steam dump to condenser.
CRID II Power Control Bank D demand signal above withdrawal Blocks automatic control rod C-11 limit setpoint withdrawal.
C-20 2/2 Turbine first stage pressure high Defeats the block of AMSAC.
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 16.1 D. C. COOK NUCLEAR PLANT Table: 7.2-3 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 ROD STOPS Rod Stop Designation Rod Motion Blocked Nuclear Overpower
- a. Intermediate Range C-1 Automatic and manual withdrawal
- b. Power Range C-2 Automatic and manual withdrawal High T:
- a. Over-temperature C-3 Automatic and manual withdrawal
- b. Over-power C-4 Automatic and manual withdrawal Turbine Power:
- a. Low Demand C-5 Automatic withdrawal
- b. High Demand C-11 Automatic withdrawal
See Table 7.2-2 for more detail.
May be manually overridden above interlock P-10.
Automatically re-instated below interlock P-10.
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 17 D. C. COOK NUCLEAR PLANT Table: 7.2-4 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 SYMBOLS AND ABBREVIATIONS
- T_ ABBREVIATIONS Transmitter DNBR Departure from nucleate boiling ratio
- S/ Test Switch TH Reactor Coolant Hot Leg temperature TC Reactor Coolant Cold Leg temperature
- P/ Test Point Tavg (TH+TC (per loop)
- D/ Computer Input T TH - TC (per loop)
AUCT Auctioneered (highest)
- Q_ Power Supply LVDT Linear Variable Differential Transformer (position sensor)
I/I Signal Isolator Ch. Protection channel L/L Lead/Lag Unit S. I. Safety Injection MDAFP Motor Driven Auxiliary Feedwater Pump
- I-Indicator TDAFP Turbine Driven Auxiliary Feedwater Pump CIA Containment Isolation Phase "A"
- L Test Lamp CIB Containment Isolation Phase "B" Channel in Test TL CVI Containment Ventilation Isolation Warning Lamp FWI Feedwater Isolation
- B- Signal Monitor (Bistable)
De-energize (trip) on SLI Steam Line Isolation signal set point Signal Monitor (Bistable) CTS Containment Spray
- B-De-energize (trip) on signal set point WR Rec. Wide Range Recorder Signal Monitor (Bistable) NR Rec. Narrow Range Recorder
- B-Energize (trip) on signal set point
- =P=Pressure Channel
- =L=Level Channel
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 16.2 D. C. COOK NUCLEAR PLANT Table: 7.2-5 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 PROCESS CONTROL BLOCK DIAGRAM DRAWING INDEX Item # Function Drawing # Drawing Title 1 Index 2 Reactor Coolant Flow Protection Set I 99010 Reactor Coolant Flow Loops 1,2,3,4 3 Reactor Coolant Flow Protection Set II 99010 Reactor Coolant Flow Loops 1,2,3,4 4 Reactor Coolant Flow Protection Set III 99010 Reactor Coolant Flow Loops 1,2,3,4 Wide-Range Reactor Wide Range Reactor Coolant Temp &
5 Control 99029 Coolant Temp. (Hot Leg) Press Wide-Range Reactor Wide Range Reactor Coolant Temp &
6 Control 99029 Coolant Temp. (Cold Leg) Press Reactor Coolant Delta T Protection 7 Loop 1 T/AVG Protection Set I 99001 Channel 1 Reactor Coolant Delta T Protection 8 Loop 2 T/AVG Protection Set II 99002 Channel 2 Reactor Coolant Delta T Protection 9 Loop 3 T/AVG Protection Set III 99003 Channel 3 Reactor Coolant Delta T Protection 10 Loop 4 T/AVG Protection Set IV 99004 Channel 4 Protection Set I, Pressurizer, RWST, & Condensate 11 Pressurizer Levels 99031 II, III Storage Tank Levels Protection Set I 12 Pressurizer Pressure 99022 Pressurizer Pressure CH 1,2, & 3 Thru IV Stm Gen Hdr Press Ch 3,4 Pressurizer Protection Set I 13 Pressurizer Pressure 99034 Pressure Ch 4& Lower Containment Thru IV Press Channel 4 Steam Generators 1 & 2 Steam Generator 1 & 2 Mismatch 14 Protection Set I 99012 Mismatch Channel 1 Steam Generators 3 & 4 Steam Generators 3 & 4 Mismatch 15 Protection Set I 99014 Mismatch Channel 1 Steam Generators 1 & 3 Steam Generators 1 & 3 Mismatch 16 Protection Set II 99015 Mismatch Channel 2 Steam Generators 2 & 4 Steam Generator 2 & 4 Mismatch 17 Protection Set II 99013 Mismatch Channel 2 Protection Set I, 18 Steam Generator Levels 99032 Steam Generator Level Channel 1 & 2 II Upper & Lower Containment Ch.
Turbine Impulse Chamber Protection Set I, 19 99033 1,2,3 & Turbine Impulse Ch. 1, 2 Pressure II Pressure 20 Steam Generator Levels Protection Set III 99018 Steam Generator level Channel 3 21 Steam Generator Levels Protection Set IV 99019 Steam Generator level Channel 4
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 16.2 D. C. COOK NUCLEAR PLANT Table: 7.2-5 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 PROCESS CONTROL BLOCK DIAGRAM DRAWING INDEX Item # Function Drawing # Drawing Title Stm Gen Hdr Press Ch. 3,4 Pressurizer Protection Set 22 Steam Pressures 99034 Pressure Ch. 4 & Lower Containment III, IV Press Channel 4 23 None 24 Rod Control Control 99007 Rod Control 25 Rod Control Control 99007 Rod Control Steam Dump Feed Steam Dump, Turbine & Feed Pump 26 Control 99008 Pump Speed Control Control Pressurizer Pressure 27 Control 99023 Prz pressure Control Control 28 Pressurizer Level Control Control 99024 Pressurizer Level Control System Volume Control Tank 29 Control 99027 RWST, CST and VCT Level Level Control 30 Boric Acid Blend Control Control 99061 Boric Acid and Primary Water 31 Rod Insertion Limit Control 99006 Rod Insertion Limits T/Auctioneered T Tavg & T/Auctioneered Tavg & T 32 Control 99005 Deviation Alarms Deviation Alarms TAVG/Auctioneered Tavg & T/Auctioneered Tavg & T 33 Control 99005 TAVG Deviation Alarms Deviation Alarms Steam Generator Level Steam Generators 1,2 & 3 Level 34 Control 99016 Control Control System Steam Generator Level 35 Control 99017 S. G. #4 Level Control Control Steam Generator Level Steam Pressure & Wide Range S.G.
36 Control 99020 (Wide Range) Level Reactor Coolant Delta T Protection 37 T & T - S.P. Recording Control 99001 Channel 1
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.2-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES1 Functional Unit Response Time Power Range, Neutron Flux 1 Less than or equal to 0.5 seconds 2 (High and Low setpoint)
For Unit 1, see Table 14.1-2 (2) 2 Overtemperature T -----------------------------------------------------
For Unit 2, see Table 14.1.0-4(2)
For Unit 1, see Table 14.1-2 3 Overpower T -----------------------------------------------------
For Unit 2, see Table 14.1.0-4 4 Pressurizer Pressure - Low Less than or equal to 2.0 seconds 5 Pressurizer Pressure - High Less than or equal to 2.0 seconds 6 Pressurizer Water Level - High Less than or equal to 2.0 seconds 7 Loss of Flow - Single Loop (Above P-8) Less than or equal to 1.0 seconds Loss of Flow - Two Loops 8 Less than or equal to 1.0 seconds (Above P-7 and below P-8)
- 9. Steam Generator Water Level -Low-Low Less than or equal to 2.0 seconds 10 Undervoltage - Reactor Coolant Pumps Less than or equal to 1.5 seconds 11 Underfrequency - Reactor Coolant Pumps Less than or equal to 0.6 seconds 1
Response times previously in Technical Specifications 4.3.1.1.3 as documented in Reference 19.
2 Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in the channel.
UFSAR Revision 31.0 I NDI A NA MI C HI G A N POW E R Revision: 27.0 D. C . C OOK NU C L E A R PL A NT Table: 7.2-7 U PDA TE D F I NA L SA F E TY A NA L Y SI S RE PORT Page: 1 of 3 E ngineered Saf ety F eature A ctuation Sy stem Response Tim es1 I nitiating Signal and F unction Response Tim e I n Seconds
- 1. C ontainm ent Pressure - High
- a. Safety Injection (ECCS) Less than or equal to 27.02 / 27.03
- b. Reactor Trip (from SI) Less than or equal to 3.0
- c. Essential Service Water System Less than or equal to 47.04 Greater than or equal to 270.0 and less than or equal to 300.0 [ Unit 1] ;
- d. Containment Air Recirculation Fan Greater than or equal to 108.0 and less than or equal to 132.0 [ Unit 2]
- e. Feedline Isolation Less than or equal to 44
- 2. Pressuriz er Pressure - L ow
- a. Safety Injection (ECCS) Less than or equal to 27.02 / 27.03
- b. Reactor Trip (from SI) Less than or equal to 3.0
- c. Feedwater Isolation Less than or equal to 8.0
- d. Essential Service Water Less than or equal to 47. 04 1
Response times previously in Technical Specifications 4.3.2.1.3 as documented in Reference 19.
2 Diesel generator starting and sequence loading delays NOT included. Offsite power available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the V CT to the RWST (RWST valves open, then V CT valves close) is included.
3 Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging, SI and RHR pumps. Sequential transfer of charging pump suction from the V CT to the RWST (RWST valves open, then V CT valves close) is NOT included. For Unit 1 only, an additional allowance of 20 sec. is provided from the LOCA analysis (47.0 seconds total).
4 Essential Service Water System is implicitly assumed available for safety injection and containment spray pump operability in addition to heat exchangers ultimate heat sink.
UFSAR Revision 31.0 I NDI A NA MI C HI G A N POW E R Revision: 27.0 D. C . C OOK NU C L E A R PL A NT Table: 7.2-7 U PDA TE D F I NA L SA F E TY A NA L Y SI S RE PORT Page: 2 of 3 E ngineered Saf ety F eature A ctuation Sy stem Response Tim es1 I nitiating Signal and F unction Response Tim e I n Seconds
- 3. Steam L ine Pressure - L ow
- a. Safety Injection (ECCS) Less than or equal to 27.02 / 37.05
- b. Reactor Trip (from SI) Less than or equal to 3.0
- c. Feedwater Isolation Less than or equal to 8.0
- d. Steam Line Isolation Less than or equal to 11.0
- 4. C ontainm ent Pressure - High- High Greater than or equal to 244.0 and
- a. Containment Spray less than or equal to 300.0 [ Unit 1] ;
Less than or equal to 45.0 [ Unit 2]
- b. Steam Line Isolation Less than or equal to 11.0
- 5. Steam G enerator W ater L evel - High - High
- a. Turbine Trip Less than or equal to 2.5
- b. Feedwater Isolation Less than or equal to 11.0
- 6. Steam G enerator W ater L evel- L ow - L ow
- a. Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0
- b. Turbine Driven Auxiliary Feedwater Pumps Less than or equal to 60.0 5
Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the V CT to the RWST (RWST valves open, then V CT valves close) is included.
UFSAR Revision 31.0 I NDI A NA MI C HI G A N POW E R Revision: 27.0 D. C . C OOK NU C L E A R PL A NT Table: 7.2-7 U PDA TE D F I NA L SA F E TY A NA L Y SI S RE PORT Page: 3 of 3 E ngineered Saf ety F eature A ctuation Sy stem Response Tim es1 I nitiating Signal and F unction Response Tim e I n Seconds
- 7. 4160 v olt E m ergency B us L oss of V oltage
- a. Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0
- 8. L oss of Main F eedwater Pum ps
- a. Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0
- 9. Reactor C oolant Bus U ndervoltage
- a. Turbine Driven Auxiliary Feedwater Pumps Less than or equal to 60.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 21.2 D. C. COOK NUCLEAR PLANT Table: 7.5-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 PROCESS INSTRUMENTATION FOR RPS & ESF ACTUATION Parameter Quantity Sensor Type Protect./Safeguards Use Taps Reactor Coolant Temperature 16 RTD T trips, Tavg permissives Installed in thermowells Pressurizer Pressure 4 Transmitter Hi/Lo pressure trips, Safety Injection 3 (top level), one shared Pressurizer Level 3 P Transmitter Reactor Trip 3 (top level), 3 (bottom level)
Steam Flow 8 P Transmitter Mismatch Trip, Safety Injection 1 pair each Feedwater Flow 8 P Transmitter Mismatch Trip 1 pair each Steam Pressure 12 Transmitter Safety Injection 1 each Steam Generator Level 12 P Transmitter Mismatch Trip, Low Level Trip 1 pair each 1 High press shared/loop, Reactor Coolant Flow 12 P Transmitter Low Flow Trip 1 Low press each Containment Pressure 4 Transmitter Safety Injection (3), Containment Spray (4) 4 Turbine 1st Stage Pressure 2 Transmitter Set Point Programs, Turbine Power permissives 1 each
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 24.0 D. C. COOK NUCLEAR PLANT Table: 7.5-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 E NG I NE E RE D SA F E TY F E A TU RE S E Q U I PME NT E X POSE D TO HA RSH E NV I RONME NT I nside Outside Operating Mode Design Operating E q uipm ent Q uantity C ontainm ent C ontainm ent L OC A or MSL B Duration Miscellaneous Containment Air Recirculation/ Approximately 2 minutes after 2/3 hi 2 X 1 Year Hydrogen Skimmer Fan containment pressure.
Safeguards Equipment Power, As required by X X As required by equipment serviced Control and Instrument Cable equipment serviced V alves Containment Isolation 34 14 20 2 open, 8 remain as is, 2 switch & 22 close V arious ECCS Process 40 8 32 14 remain as is, 6 close & 20 switch V arious Other ESF Process 38 8 30 16 open, 9 remain as is, 4 close & 9 switch V arious
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 24.0 D. C. COOK NUCLEAR PLANT Table: 7.5-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 E NG I NE E RE D SA F E TY F E A TU RE S E Q U I PME NT E X POSE D TO HA RSH E NV I RONME NT I nside Outside I nitiate Design Operating E q uipm ent Q uantity Monitor C ontainm ent C ontainm ent L OC A or MSL B Duration I nstrum entation Auxillary Feedwater Flow 4 X X 4 Months Containment Pressure 4 X X X X 4 Months Main Feedwater Flow 2/Stm. Gen. X X 1 Minute Main Steam Flow 2/Stm. Gen. X X 1 Minute Main Steam Pressure 3/Stm. Gen. X X X 4 Months Pressurizer Level 3 X X 4 Months 3 X X X X 4 Months Pressurizer Pressure 1 X X X 1 Minute Wide Range Reactor Coolant Pressure 2 X X 4 Months Narrow Range Reactor Coolant 2/Loop X X X 1 Minute Temperature Wide Range Reactor Coolant 2/Loop X X 4 Months Temperature Steam Generator Narrow Range Level 3/Stm. Gen. X X X 4 Months Containment Water Level 8 X X 4 Months
Includes a redundant pair of overlapping level monitors and a redundant pair of level switches.
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Verify ECCS Flow 0 to 110% design flow Flow in HPI System IFI-51 IFI-52 Prior to Manual Control Room A-1 (Centrifugal Charging ----------------------------
Stop of Reactor Panel SIS Pump Flow) IFI-53 IFI-54 0-200 GPM Coolant Pumps Control Room 0-3000 psig Panel SIS Manual Trip of RC NPS-110 NPS 111 RCS Pressure A-2 pumps based on ----------------------------
(wide range) Control Room RCS pressure MR-13 0-3000 psig Panel RHR for MR-13 A-3 Not used A-4 Not used A-5 Not used A-6 Not used Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value MPP-210 MPP-211 From atmospheric Determination of MPP-212 MPP-220 pressure to 20% above required core exit MPP-221 MPP-222 the lowest safety valve Control Room A-7 S/G Pressure setting temperature by SG MPP-230 MPP-231 Panel SG pressure MPP-232 MPP-240 ----------------------------
MPP-241 MPP-242 0-1200 psig NLI-320 NLI-321 599-3 to 614 Determination of elevation Containment Water Control Room A-8 Adverse Level (Containment floor to Panel RHR Containment NLI-330 NLI-331 NLI-340 NLI-341 max flood level) 602'-2 3/4" to 613'-0" Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value From below 1st stage BLP-110 BLP-111 separator to 2nd stage Manual reduction BLP-112 BLP-120 separator of ECCS Flow S/G Level BLP-121 BLP-122 Control Room A-9 ----------------------------
(secondary heat Narrow Range BLP-130 BLP-131 Panel SG sink capability) BLP-132 BLP-140 From below 1st stage BLP-141 BLP-142 separator to 2nd stage separator Top to Bottom NLP-151 Manual Reduction ---------------------------- Control Room A-10 Pressurizer Level NLP-152 of ECCS Flow 0-100% (96% of Panel PRZ NLP-153 indicated volume)
A-11 Not used Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 R/Hr to VRA-1310 VRA-1410 Containment Area 1 x 107 R/Hr Determination of for U1 Radiation Monitor Control Room A-12 adverse ----------------------------
High Range Panel RMS-CT containment VRA-2310 VRA-2410 1 R/Hr to for U2 1 x 107 R/Hr Manually establish PPP-300 PPP-301 -5 to design pressure Containment Pressure Control Room A-13 or trip containment ----------------------------
(narrow range) Panel SPY spray PPP-302 PPP-303 -5 to +12 psig A-14 Not used Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 5 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Manual reduction 0-110 % design flow FFI-210 FFI-220 of ECCS Flow Auxiliary Feedwater Control Room A-15 ----------------------------
(Secondary heat Flow Panel SG FFI-230 FFI-240 0 to 250 x 103 PPH sink capability)
Essentially Top (bottom of overflow) to Bottom (bottom of Control Room ILS-950 safety injection pipe) Panel Spy for Manual transfer to (100% of total volume) MR-36 cold leg A-16 RWST Level ILS-951 ----------------------------
recirculation in low Essentially Top Control Room level in RWST MR-36 for ILS-950 (bottom of overflow) to Panel BA for Bottom (bottom of ILS-951 safety injection pipe)
(100% of total volume)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 6 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Inputs are received from 200F subcooling to variables Manual trip or 35F superheat Control Room reduction of A-17 Degrees subcooling A-2, A-25, A-38, and A-39 ---------------------------- Panel BA Pressurizer Spray and ECCS flow 425F subcooling to for Hot and Cold Leg Variables 75F superheat A-18 Not used A-19 Not used A-20 Not used A-21 Not used A-22 Not used A-23 Not used A-24 Not used Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 7 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Control Room 200 - 2300F Panel FI SG-30 and SG-31 Manual reduction TC Readings A-25 Core Exit Temp for TC 1-65 1 ----------------------------
of ECCS Flow Monitored By 200 - 2300F PPC and Recorders A-26 Not used A-27 Not used Unit 1 & 2 101-TD7 (Pump PP-50E) Open/Close Manual Trip of by BKR T11D7 Control Room A-28 CCP Breaker status ----------------------------
RCPs Panel BA 101-TA8 (Pump PP-50W) Open/Close by BKR T11A8 1
Technical Specifications (Unit 1-3.3.3.8 and Unit 2-3.3.3.6) require two TCs per channel per quadrant minimum.
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 8 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Unit 1 & 2 101-TD5 (Pump PP-26N) Open/Close Manual Trip of by BKR T11D5 Control Room A-29 SI pump breaker status ----------------------------
RCPs Panel SIS 101-TA1 (Pump PP-26S) Open/Close by BKR T11A1 A-30 Not used A-31 Not used A-32 Not used A-33 Not used 0-500 gpm Manual trip of Safety Injection Pump Control Room A-34 IFI-260 IFI-266 ----------------------------
RCPs Flow Panel SIS 0-500 gpm A-35 Not used A-36 Not used Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 9 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value A-37 Not used MR-9 and MR-11 for 50-400F RCS Cold Leg RCS Cold Leg Control Room A-38 ----------------------------
water temperature water temperature Panel DTU NTR-210 NTR-230 0-700F MR-9 and MR-11 for 50-700F RCS Hot Leg water Control Room A-39 Core Cooling ----------------------------
temperature Panel DTU NTR-110 NTR-130 0-700F Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 10 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value PAS-H2-A-CRI and 0 to 30 volume % for PAS-H2-B-CRI ice-condenser type Post-Accident Containment Hydrogen for containment Control Room A-40 Sampling concentration ESR-1, ESR-2, ESR-3, Panel IV Containment ----------------------------
ESR-4, ESR-5, ESR-6, ESR-7, ESR-8, ESR-9 0-30 volume %
(101-TD3) PP-10E by BKR T11D3 Open/Close CCW flow to ESF CCW Pump Breaker Control Room A-41 ----------------------------
system Status Panel CCW (101-TA7) PP-10W Open/Close by BKR T11A7 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 11 of 51 TYPE "A" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Determination of excessive fouling or Containment NLI-300
Control Room A-42 blockage of Recirculation Sump Above/Below Vortex Panel RHR recirculation sump Water Level NLI-301 strainers Limit Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 12 of 51 TYPE "B" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value NRI-21 10-6 to 100% power Control Room B-1 Reactivity Control Neutron Flux ---------------------------- Panel NIS NRI-23 10-8 to 200% power CA1-8 (U1), CB1-4 (U1)
CA1-4 (U2), CB1-8 (U2)
Full In or Not Full In Reactivity Control Control Room B-2 Control Rod Position CC1-8 CD1-9 ----------------------------
(continued) Panel RC Full In or Not Full In SA1-8 SB1-8 SC1-4 SD1-4 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 13 of 51 TYPE "B" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 0-6000 ppm NSX-101 Diluted = 375-10000 Reactivity Control RCS soluble Boron B-3 ppm NA (continued) concentration NSX-103 Undiluted = 0.375-10000 ppm Reactivity Control 50-400F (continued) RCS Cold Leg water B-4 ----------------------------
temperature See Item A-38 0-700F Core Cooling 50-700F RCS Hot Leg water B-5 ----------------------------
temperature See Item A-39 0-700F Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 14 of 51 TYPE "B" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Core Cooling (continued) 50-700F RCS Cold Leg water B-6 ----------------------------
temperature See Items 0-700F A-38 and B-4 Core Cooling 0-3000 psig (continued)
B-7 RCS Pressure ----------------------------
See Item A-2 0-3000 psig Core Cooling 200-2300F (continued) SG-30 and SG-31 B-8 Core exit temperature ----------------------------
for TC 1-65 See Item A-25 200-2300F Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 15 of 51 TYPE "B" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Bottom of hot leg to top of vessel NLI-110 NLI-111 ----------------------------
Core Cooling Bottom of hot leg to Coolant Inventory Control Room B-9 (continued) NLI-120 NLI-121 top of vessel RVLIS Panel SIS Top of head vent NLI-130 NLI-131 piping to bottom of vessel (100% of volume)
Core Cooling (continued)
B-10 Degrees Subcooling See item A-17 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 16 of 51 TYPE "B" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Maintaining RCS 0-3000 psig Integrity B-11 RCS Pressure See Item A-2 0-3000 psig Maintaining RCS Integrity NLA-310 Containment sump 5895 to 5998 Control Room B-12 (continued) water level (bottom of sump to Panel RHR NLI-311 See Item A-8 containment floor) 0 to design pressure Maintaining RCS (psig)
PPA-310 Integrity Containment pressure Control Room B-13 (continued) (wide range) Panel SPY PPA-312 ----------------------------
-5 to +36 psig Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 17 of 51 TYPE "B" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Closed-not closed Control Room Containment Isolation Maintaining Panel IV, valve position B-14 Containment See Table 5.4-1 Panel BA, (excluding check ----------------------------
Integrity Panel SIS, valves) Closed-not closed Panel SPY Maintaining -5 psig to design Containment pressure Integrity B-15 (continued) Containment pressure ----------------------------
-5 to +12 psig (A-13)
See Items or A-13 & B-13 -5 to +36 psig (B-13)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 18 of 51 TYPE "C" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Fuel Cladding 200-2300F SG-30 and SG-31 C-1 Core Exit Temperature ----------------------------
for TC 1-65 See Item A-25 200-2300F 1/2 Tech Spec limit to Radioactive 100 times Tech Spec Fuel Cladding Concentration or NSX-101 limit C-2 (continued) Radiation Level in ---------------------------- NA Circulating Primary NSX-103 Equipment capable of Coolant measuring 1 Ci/ml to 10 Ci/ml Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 19 of 51 TYPE "C" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 10 Ci/ml to 10Ci/ml or TID-14844 source Fuel Cladding term in coolant volume Analysis of Primary (continued)
C-3 Coolant ----------------------------
(gamma spectrum) Equipment capable of See Item C-2 measuring 1 Ci/ml to 10 Ci/ml Reactor Coolant 0-3000 psig pressure boundary C-4 RCS Pressure ----------------------------
See Item A-2 0-3000 psig Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 20 of 51 TYPE "C" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Reactor Coolant -5 psig to design pressure boundary pressure (continued) ----------------------------
C-5 Containment Pressure
-5 to +12 psig (A-13)
See items or A-13 & B-13 -5 to +36 psig (B-13)
Reactor Coolant See Items pressure boundary A-8 & B-12 (continued) Containment Sump C-6 ----------------------------
Water Level See Items See Items A-8 & B-12 A-8 & B-12 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 21 of 51 TYPE "C" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Reactor Coolant pressure boundary 1 R/Hr to 1 x 104 R/Hr Containment Area C-7 (continued) ----------------------------
Radiation 1 R/Hr to 1 x 107 R/Hr See Item A-12 Effluent Radioactivity- 1 x 10-6 to 1 x 10-2 Reactor Coolant µCi/cc Noble Gas effluent SRA-1900 for Unit 1 pressure boundary Control Room C-8 from condenser air ----------------------------
(continued) FFC Panel removal system SRA-2900 for Unit 2 9x10-7 to 9x104 µCi/cc exhaust (Minimum Range)
Containment 0-3000 psig C-9 RCS Pressure ----------------------------
See Item A-2 0-3000 psig Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 22 of 51 TYPE "C" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 0 to 30 volume % for Containment ice-condenser type PAS-H2-A-CRI and (continued) Containment Hydrogen containment C-10 PAS-H2-B-CRI for Concentration ESR 1 through 9 ----------------------------
See Item A-40 0-30 volume %
-5 psig pressure to 3 times design pressure Containment for concrete (continued) containment C-11 Containment Pressure See Items A-13 & B-13 -5 to +12 psig (A-13) or
-5 to +36 psig (B-13)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 23 of 51 TYPE "C" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 x 10-6 to 1 x 10-2 Containment Effluent Ci/cc Containment VRS-1500 for Unit 1 Radioactivity Noble Control Room C-12 (continued) ----------------------------
Gases from identified FFC Panel VRS-2500 for Unit 2 9x10-7 to 9x104 Ci/cc release points (Minimum Range)
Effluent Radioactivity Noble Gases (from buildings or areas Containment where penetrations and (continued) hatches are located, C-13 eg,) secondary See Item C-12 containment and AUX buildings that are in direct contact with primary containment Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 24 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value IFI-310 IFI-320 0 to 110% design flow (0-1500 GPM)
Control Room D-1 RHR System RHR System Flow 0-1500 GPM Panel RHR IFI- 311 IFI-321 (1500-5000 GPM) 1500-5000 GPM 40-350F Control Room ITI-310 RHR System RHR Heat Exchanger Plant Process D-2 ----------------------------
(continued) Outlet Temperature Computer ITI-320 0-400F (PPC)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 25 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 10 to 90% volume ILA-110 ILA-120 ILA-130 ILA-140 Wide Range 4.148 to (Wide Range) 120.8 or 300 to 1000 Accumulator Tank cubic ft Control Room D-3a SI System (52% of total volume)
Level Panel SIS ILA-111 ILA-121 ILA-131 ILA-141 Narrow Range 104.15 (Narrow Range) to 129.15 or 900 to 1050 cubic ft (7.5% of total volume)
IPA-110 IPA-111 0 to 750 psig Control Room SI System Accumulator Tank IPA-120 IPA-121 D-3b ---------------------------- Panel SIS (continued) Pressure IPA-130 IPA-131 IPA-140 IPA-141 0-700 psig Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 26 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Accumulator Tank IMO-110 IMO-120 Closed or Open SI System Control Room D-4 Isolation Valve ----------------------------
(continued) Panel SIS Position IMO-130 IMO-140 Closed or Open SI System 0 to 110% design flow (continued) Boric Acid D-5 QFI-200 ----------------------------
Charging Flow See Item D-24 0-200 GPM SI System 0 to 110% design flow IFI-51 IFI-52 (continued)
D-6 Flow in HPI System ----------------------------
IFI-53 IFI-54 0-200 GPM See Item A-1 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 27 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value SI System 0 to 110% design flow (continued) ----------------------------
D-7 Flow in LPI System 0-1500 GPM See Item D-1 1500-5000 GPM Top to Bottom SI System Essentially Top (continued) (bottom of overflow) to D-8 RWST Level Bottom (bottom of See Item A-16 safety injection pipe)
(100% of total volume)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 28 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value RCP-1 for PP-45-1 Motor Current Primary Coolant RCP-2 for PP-45-2 Control Room D-9 RCP Status ----------------------------
System RCP-3 for PP-45-3 Panel RCP RCP-4 for PP-45-4 0-1200 A Primary Coolant QR-107A QR-107B NA Primary System Safety Control Room D-10 System ----------------------------
Relief Valve Flow Panel RC (continued) QR-107 C QR-107D NA Primary Coolant Top to Bottom System ----------------------------
D-11 (continued) Pressurizer Level 0-100%
(96% of indicated See Item A-10 volume)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 29 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Primary Coolant Group On/Off Pressurizer Heater Control Room D-12 System A1, A2, A3 ----------------------------
Status Panel PRZ (continued) C1, C2, C3 On/Off Top to Bottom Primary Coolant
Control Room D-13 System Quench Tank Level NLA-351 7 above tank bottom Panel PRZ (continued) to 7 below tank top Primary Coolant 50-750F Quench Tank Control Room D-14 System NTA-351 ----------------------------
Temperature Panel PRZ (continued) 50-750F Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 30 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Primary Coolant 0 to design pressure Quench Tank Control Room D-14a System NPA-351 ----------------------------
Pressure Panel PRZ (continued) -10 to 100 psig From tube sheet to BLI-110 BLI-120 separators Secondary System S/G Level Control Room D-15a ----------------------------
(Steam Generator) (wide range) Panel SG BLI-130 BLI-140 From 12 above tube sheet to separators Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 31 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value From atmospheric Secondary System pressure to 20% above (Steam Generator) the lowest safety valve D-15b (continued) S/G Pressure setting See Item A-7 0-1200 psig Secondary System (Steam Generator)
Safety/Relief Valve D-16a (continued)
Positions See Item D-16b Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 32 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value MFC-110 MFC-111 NA Secondary System MFC-120 MFC-121 Control Room D-16b (Steam Generator) Main Steam Flow ----------------------------
MFC-130 MFC-131 Panel SG (continued) 0-4 x 106 PPH MFC-140 MFC-141 Secondary System FFC-210 FFC-211 0 to 110% design flow (Steam Generator) FFC-220 FFC-221 Control Room D-17 Main Feedwater Flow ----------------------------
(continued) FFC-230 FFC-231 Panel SG FFC-240 FFC-241 0-4 x 106 PPH Auxiliary 0 to 110% design flow Feedwater System D-18 Aux Feedwater Flow ----------------------------
See Item A-15 0 to 250 x 103 PPH Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 33 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value CLI-113 CLI-114 Plant specific Auxiliary CLR-110 CLR-111 ----------------------------
Feedwater System Condensate Storage Control Room D-19 Essentially top to (continued) Tank Level Panel CP MR-49 for bottom CLR-110 CLR-111 (95% total volume)
IFI-330 IFI-331 0 to 110% design flow (upper containment)
Containment Containment Spray (0-2500 GPM) Control Room D-20 0-2500 GPM Cooling System Flow Panel SPY PP-9E PP-9W (On/Off) On/Off Containment Heat Removal by D.C. Cook does not have this D-21 Cooling System Containment Heat type of system and therefore it (continued) Removal System does not apply Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 34 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value SG-18 for ETR-11 ETR-12 ETR-13 ETR-14 ETR-15 ETR-16 40-400F Containment Containment Control Room ETR-17 ETR-18 D-22 Cooling System Atmosphere ---------------------------- Aux Relay ETR-19 ETR-20 (continued) Temperature 0-400F Panel A-15 ETR-21 ETR-22 ETR-23 ETR-24 ETR-25 ETR-26 and ETR-27 Containment MR-14 and MR-15 50 to 250F Containment Sump Control Room D-23 Cooling System ----------------------------
Water Temperature Panel RHR (continued) for ITR-311 and ITR-321 50 to 400F Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 35 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Chemical and 0 to 110% design flow Control Room D-24 Volume Control Makeup Flow - In QFI-200 ----------------------------
Panel BA System 0-200 GPM Chemical and 0 to 110% design flow Control Room Volume Control D-25 Letdown Flow - Out QFI-301 ---------------------------- Panel BA and System 0-200 GPM Panel HSD (continued)
Top to Bottom Chemical and QLC-451 ----------------------------
Volume Control Volume Control Tank Control Room D-26 Essentially top to System Level Panel BA QLC-452 bottom (continued)
(65% of total volume)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 36 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value CCW Water SG-10 for 40-200F Cooling Water Control Room D-27 Temperature to ESF CTR-410 CTR-415 ----------------------------
System Panel ESW System CTR-420 CTR-425 0-200F CFI-410 CFI-420 0 to 110% design flow (0-10000 GPM)
Cooling Water CCW Flow to ESF ---------------------------- Control Room D-28 System System 0-10000 GPM Panel CCW CFI-419 CFI-429 (continued) 0-6000 GPM (0-6000 GPM)
Top to Bottom 12-RLS-255 ----------------------------
High Level Radioactive D-29 Rad Waste System Essentially top to Panel WDS Liquid Tank Level 12-RLS-256 bottom (84% of total volume)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 37 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 12-MR-58 and 12-MR-59 for 0 to 150% design Rad Waste System Radioactive Gas 12-RPC-310 12-RPC-320 pressure D-30 Panel WDS (continued) Holdup Tank Pressure 12-RPC-330 12-RPC-340 ----------------------------
12-RPC-350 12-RPC-360 0-250 psig 12-RPC-370 12-RPC-380 Control Room VCR-201 VCR-202 Open/Closed Panel IV VCR-203 VCR-204 Emergency Ventilation D-31 Ventilation System VCR-205 VCR-206 ----------------------------
Damper Position VCR-207 on Open/Closed Control Room VCR-207 Panel SPY Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 38 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value DGAB-AM-1-REM Status of Standby DGCD-AM-1-REM Plant Specific Control Room Power and Other Diesel Generator DGAB-AM-2-REM D-32a ---------------------------- Panel SA Energy Sources Status DGCD-AM-2-REM Important to Safety DGAB-AM-3-REM 0-800 A DGCD-AM-3-REM Status of Standby Bus T11A, T11B, T11C, T11D Power and Other for Unit 1 Plant Specific 4Kv Safety Related Control Room D-32b Energy Sources ----------------------------
Power Systems Status Panel SA Important to Safety Bus T21A, T21B, T21C, T21D 0-150 V (continued) for Unit 2 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 39 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Status of Standby Power and Other Plant Specific 250 Vdc Battery Power BATT-AB-REM-VM Control Room D-32c Energy Sources ----------------------------
System Status BATT-CD-REM-VM Panel SA Important to Safety 0-300 V (continued)
Channel I, II, III, IV Status of Standby Power and Other Plant Specific 120 Vac Safety Related CRID-1-VM Control Room D-32d Energy Sources ----------------------------
Power System Status CRID-2-VM Panel SA Important to Safety 0-150 V CRID-3-VM (continued)
CRID-4-VM Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 40 of 51 TYPE "D" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value XPI-100 (0-150 psig)
Status of Standby Plant Specific Power and Other XPI-50 (0-100 psig)
Control Room D-32e Energy Sources Instrument Air Status Panel SV Important to Safety XPI-20 (0-60 psig)
(continued)
XPI-85 (0-160 psig)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 41 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 R/Hr to 1 x 107 R/Hr Containment Containment Area Radiation E-1 Radiation Monitor ----------------------------
High Range See Item A-12 1 R/Hr to 1 x 107 R/Hr Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 42 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value ERA-7303 ERA-7304 ERA-7305 ERA-7306 ERA-7307 ERA-7308 1 x 10-1 R/Hr ERA-7403 ERA-7404 ERA-7504 ERA-7507 to Radiation Exposure ERA-7508 Rate 1 x 104 R/Hr ERA-7601 ERA-7602 (inside buildings or ERA-7603 ERA-7604 Control Room E-2 Area Radiation ERA-7605 where areas of access FFC Panel are required to service ERA-8303 ERA-8304 equipment important to ERA-8305 ERA-8306 safety) ERA-8307 ERA-8308 Equipment installed is ERA-8403 capable of monitoring within the range ERS-7401 ERS-8401 Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 43 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Noble Gases and Vent Flow Rate Containment or Purge E-3a Effluent See Item E-3e Noble Gases and Vent Flow Rate Reactor Shield E-3b (continued)
Building Annulus See Item E-3e Noble Gases and Vent Flow Rate E-3c (continued) Aux Building See Item E-3e Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 44 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 x 10-6 to SRA-1900 for Unit 1 1 x 105 Ci/cc Noble Gases and Condenser Air SRA-2900 for Unit 2 0 to 110% vent design flow Control Room Vent Flow Rate E-3d Removal System ---------------------------- FFC Panel (continued)
Exhaust SFR-401 (0-250 scfm only) 9 x 10-7 to 9 x 104 Ci/cc (Minimum Range) 0-250 scfm Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 45 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 x 10-6 to VRS-1500 for Unit 1 1 x 103 Ci/cc Noble Gases and 0 to 110% vent design flow VRS-2500 for Unit 2 Vent Flow Rate Control Room E-3e Common Plant Vent (continued) ---------------------------- FFC Panel VFR-315 (0-200k scfm only) 9 x 10-7 to 9 x 104 Ci/cc (Minimum Range) 0-200k scfm Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 46 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value MRA-1600, MRA-1700 (Unit 1) 1 x 10-1 to Noble Gases and 1 x 103 Ci/cc Vent from S/G Safety Control Room E-3f Vent Flow Rate MRA-2600,MRA-2700 ----------------------------
Relief Valves FFC Panel (continued) (Unit 2) 0.1 to 100 Ci/cc2 2
This is the minimum sensitivity of the instrument for normal operation, to follow the course of an accident, and/or take protective actions. Values of the instrument above or below this minimum sensitivity range are acceptable.
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 47 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 x 10-6 to 1 x 102 Ci/cc SRA-1800 (Unit 1) 0 to 110% vent design flow SRA-2800 (Unit 2) ----------------------------
Noble Gases and 1x10-7 to 1x105 Ci/cc (SRA-1800)
Other Identified Control Room E-3g Vent Flow Rate 1x10-7 to 1x105 Ci/cc (SRA-2800)
Release Points FFC Panel (continued) SFR-201 (Minimum Range)
(0-4500 scfm only)
Unit 1 = 0-1500 scfm Unit 2 = 0-4500 scfm Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 48 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value All Identified Release Points 1 x 10-3 to Particulates and 1 x 102 Ci/cc (except S/G safety Halogens E-4 relief valves and 0 to 110% vent design flow condenser air removal See Item E-3e ----------------------------
system exhaust) sampling and onsite analysis 1 x 10-9 to 1 x 103 Ci/cc Airborne Radioactivity Environmental and Particulates E-5a Radiation and NA ----------------------------
Sampling and Analysis Radioactivity 1 x 10-9 to (portable) 1 x 10-3 Ci/cc (minimum)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 49 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value 1 x 10-3 to 1 x 104 R/hr photons 1 x 10-3 to 1 x 104 Rads/hr Beta radiations and low-Environmental Plant and energy photons Radiation and Environmental E-5b NA Radioactivity Radiation (continued) (portable)
Gamma 1 x 10-3 to 1 x 104 R/hr Beta / low energy gamma 1 x 10-3 to 1 x 104 Rad/hr Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 50 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value Environmental Plant and Isotopic analysis Radiation and Environmental E-5c NA Radioactivity Radioactivity ----------------------------
(continued) (portable) Isotopic analysis 0-360 degrees Control Room EFR-410 EFR-412 Plant Process E-6 Meteorology Wind Direction
Computer EFR-413 EFR-414 0-360 degrees (PPC)
EFR-400 EFR-404 0-100 MPH Control Room Meteorology Plant Process E-7 Wind Speed (continued) EFR-402 EFR-403 ---------------------------- Computer 0-125 MPH (PPC)
Last Revised: 30.0
UFSAR Revision 31.0 INDIANA MICHIGAN POWER Revision: 30.0 D. C. COOK NUCLEAR PLANT Table: 7.8-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 51 of 51 TYPE "E" VARIABLES PROVIDED THE OPERATOR FOR MANUAL FUNCTIONS DURING AND FOLLOWING AN ACCIDENT Range Item Reg Guide 1.97 Value Display No. Purpose Variable Tag No. ---------------------------- Location Installed Value
-5C to +10C ETR-400 ETR-402 per 50 meter intervals Control Room Meteorology Estimation of ETR-403 Plant Process E-8 (continued) Atmospheric Stability Computer ETQ-401 (PPC)
-30 to 50 C E-9a Line Item Deleted E-9b Line Item Deleted E-9c Line Item Deleted E-9d Line Item Deleted E-9f Line Item Deleted E-9g Line Item Deleted E-9h Line Item Deleted E-10a Line Item Deleted E-10b Line Item Deleted E-10c Line Item Deleted Last Revised: 30.0