ML22340A181
Text
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WASTE DISPOSAL SYSTEM PERFORMANCE DATA
Plant Design Life 40 years
Normal process capacit y, liquids 15 gpm
Evaporator load factor 32%
Annual approximate liquid discharge 1
Volume (2 units) 2,415,000 gal.
Tritium Activity 2 (2 units) 2.0 x 103 curies
Other (2 units) 1.125 curies/ year
Annual gaseous discharge
Activity (2 units) 11,957 curies/ year
Annual drummed solids shipped for burial 3 27,624 ft3/year
1 Estimate based on Table 11.1-4, equilibrium cycle.
2 Volume is an annual average based on actual shipments for two units from 1979 through 1988.
3 Quantity based on approximate actual discharge for 1986.
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WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS1
COMPONENT CODE
Chemical Drain Tank No code Reactor Coolant Drain Tanks A S M E III, 2 Class C Sump Tanks No code Waste Holdup Tank No code Waste Evaporator Condensate Tank No code Laundry and Hot Shower Tank No code Waste Evaporator(s) No code
Waste Filters A S M E III,(2) Class C
Piping and Valves USAS-B31.13, Section 1 ASME III Appendix F 4 Gas Decay Tank A S M E III,(2)Class C
Spent Resin Storage Tank A S M E III,(2) Class C
Waste Evaporator Condensate Demineralizer A S M E III,(2) Class C Waste Evaporator Condensate Filter A S M E III,(2) Class C Waste Evaporator Bottoms Storage Tank No code
1 Repairs and replacements for pressure retaining components within the code boundary, and their supports, are conducted in accordance with ASME Section XI 2 ASME III American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
3 USAS-B31.1 American Standards Association Code for pressure piping and special nuclear cases where applicable.
4 The evaluation criteria of ASME III Appendix F (faulted conditions) is applicable to 1) piping from normally closed PRT drain line isolation valve and the RCDT drain line check valve inside containment to the normally closed isolation valve outside containment (U-1 & U-2); and 2) piping between containment sump pump discharge check valves inside containment and discharge isolation valve outside containment (U-1 only).
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COMPONENT
SUMMARY
DATA TANKS QUANTITY TYPE VOLUME DESIGN DESIGN MATERIAL PRESSURE TEMP.oF (1)
Reactor Coolant Drain 1 Horiz 350 gal 25 psig 267 ss (per unit)
Laundry & Hot Shower 2 (2) Vert 600 gal Atm 180 ss
Chemical Drain 1 (2) Vert 600 gal Atm 180 ss
Clean Sump 1 (2) Vert 600 gal Atm 180 ss
Station Drainage Sump 1 (2) Vert 525 gal Atm 180 ss
Waste Holdup 2 (2) Horiz 24,700 gal Atm 180 ss
Waste Condensate 2 (2) Vert 6,450 gal Atm 180 ss
Gas Decay 8 (2) Vert 600 ft3 150 psig 180 cs Waste Evaporator 1 (2) Vert 4,000 gal Atm 250 ss Bottoms Storage Spent Resin Storage 1 (2) Vert 300 ft3 100 psig 180 ss
FLOW HEAD DESIGN DESIGN MATERIAL PUMPS QUANTITY TYPE gpm ft. PRESSURE TEMP oF (1) psig Reactor Coolant Drain ( A) 1 Horiz canned 50 175 150 300 ss (per unit)
Reactor Coolant Horiz Drain (B) 1 canned 150 175 150 300 ss (per unit)
Chemical Drain 1 (2) Horiz 20 100 150 180 ss Laundry & Hot 1 (2) Horiz (3) 20 100 150 180 ss Sho wer Sump Tank 2 (2) Horiz (3) 20 100 150 180 ss Waste 2 (2) Horiz 20 100 150 180 ss Evaporator Waste 2 (2) Horiz (3) 150 200 150 180 ss Condensate Waste Horiz (3)
Evaporator 1 (2) 20 60 150 180 ss Bottoms
(1) Material contacting fluid (2) Shared by Units 1 and 2 (3) Mechanical seal provided (2) Shared by Units 1 and 2 (3) Mechanical seal provided I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 16.1 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 11.1-3 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS R EE PP OO RR TT Page: 2 of 2 R
Storage Tank
MISCELLANEOUS EQUIPMENT QUANTITY CAPACITY TYPE
Waste Evaporator 1(2) 15 gpm Forced Circulation Flash (Incoloy - 825 tubes)
Boric Acid/Waste Evaporator 1(2) 15 gpm Submerged Tube (Incoloy - 825 tubes)
Waste Gas Compressors 2(2) 40 CFM Liquid piston rotary(3)
DESIGN DESIGN MATERIAL QUANTITY TYPE CAPACITY PRESSURE TEMP. (1) psig oF
Waste Evaporator Condensate Filter 1(2) Disposable cartridge 20 gpm 150 180 ss
Waste Evaporator 30 ft3 Condensate 1(2) Flushable 100 250 ss Demineralizer
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ESTIMATED LIQUID DISCHARGE TO WASTE DISPOSAL SYSTEM
SOURCE TOTAL ANNUAL (Gal)
Laundry and Shower 390,000
Equipment drains, leaks, laboratory 1,950,000
Decontamination 75,000
Totals 2,415,000
Load Factor1 32%
1 Based on 15 gpm Radwaste Evaporator Capability.
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ESTIMATED LIQUID RELEASE BY ISOTOPE (Two Units)
ISOTOPE ANNUAL RELEASE µ ISE RELEE µ Sr 89 6.12E2 Cs 134 2.78E4 Sr 90 1.54E1 Cs 136 5.36E3 Y 90 1.49E1 Cs 137 1.69E5 Sr 91 3.19E1 Cs 138 3.24E-12
Y 91 1.13E3 Te 132 3.10E4
Sr 92 5.93E-2 I 132 1.72E1 Y 92 4.52E-1 Te 134 6.36E-10 Zr 95 1.29E2 Ba 140 5.92E2 Nb 95 1.25E2 La 140 1.49E2 Mo 99 3.38E5 Ce 144 7.55E1 I 133 2.11E5 I 134 5.04E-6 I 131 3.27E5 I 135 1.31E4
Notes: Other waste disposal 11.25E5 µC i / yr.
All Isotopes with total activity per year <1.0E -12 were ignored.)
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ESTIMATED ANNUAL GASEOUS RELEASE BY ISOTOPE
ISOTOPE ACTIVITY TO ENVIRONMENT (Ci/Yr)
Kr 85 10,808
Kr 85m, 87, 88 Negligible
Xe 133 1149
Xe 133m, 135, 135m, 138 Negligible
Total 11,957
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PLANT ZONE CLASSIFICATIONS
MAXIMUM EXPOSURE RATE ZONE ACCESS CONDITIONS (1% failed fuel) mrem/hr.
1 Unlimited <0.25
2 Occupational 0.25 - 2.499
3 Periodic 2.5 - 4.999
4 Limited 5.0 - 100
5 Restricted >100
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PRIMARY SHIELD DESIGN PARAMETERS, NEUTRON AND GAMMA FLUXES
DESIGN PARAMETERS Core thermal power 3391 MW Active core height 144 in.
Effective core diameter 132.7 in.
Baffle wall thickness 1.125 in.
Barrel wall thickness 2.25 in.
Thermal shield wall thickness 2.75 in.
Reactor vessel I.D. 173.0 in.
Reactor vessel wall thickness 8.625 in.
Reactor coolant cold leg temperature 536°F Reactor coolant hot leg temperature 600°F Maximum thermal neutron flux exiting primary concrete 8.4 x 103n/cm2sec.
Reactor shutdown dose exiting primary concrete <15 mrem/hr
CALCULATED NEUTRON FLUXES ENERGY GROUP INCIDENT FLUXES LEAKAGE FLUXES (n/cm2 - sec) (n/cm2 - sec)
E 1 Mev 7.7 x 108 2.5 x 101 5.3 Kev < E < 1 Mev 1.3 x 1010 5.6 x 101
.625 ev < E < 5.3 Kev 7.8 x 109 9.5 x 101 E <.625 ev 2.0 x 109 8.4 x 103
CALCULATED GAMMA FLUXES ENERGY GROUP INCIDENT FLUXES LEAKAGE FX
(/ 2 s) (/ 2 s)
E = 7.5 Mev 4.5 x 109 4.4 x 105 E = 4.0 Mev 1.2 x 109 3.1 x 105 E = 2.5 Mev 2.2 x 109 3.4 x 105 E = 0.8 Mev 7.6 x 108 2.8 x 104
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SECONDARY SHIELD DESIGN PARAMETERS
Core power density @ 3391 MWt 103.9 w/cc
Reactor coolant liquid volume 12,600 ft3 1
Reactor coolant transit times:
Core 0.8 sec.
Core exit to steam generator inlet 2.1 sec.
Steam generator inlet channel 0.7 sec.
Steam generator tubes 3.7 sec.
Steam generator tubes to vessel inlet 2.1 sec.
Vessel inlet to core 2.2 sec.
Total Out of Core 10.8 sec.
Total power dose rate outside secondary shield <1 mrem/hr
1 This value has been conservatively chosen for the purpose of shield design. Actual best-estimated reactor coolant systems volumes can be obtained from the current Westinghouse IMP databases.
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ACCIDENT SHIELD DESIGN PARAMETERS
TID-14844 RELEASE
Core thermal power 3391 MW Minimum full power operating time 650 days Equivalent fraction of core melting 1.0 Fission product fractional releases:
Noble gases 1.0 Halogens 0.5 Remaining fission product inventory 0.01 Clean-up rate following accident 0 Maximum integrated dose (infinite exposure) in the control room <1 rem GAP ACTIVITY RELEASE Core Thermal Power, MW 3391 Minimum full power operating time, days 650 Equivalent fraction fuel rod failure 1.0 Fraction of gap activity absorbed by sump water:
Noble Gases 0.0 All Other 1.0 Cleanup rate following accident 0.0 Sump water volume, ft 3:
Reactor Coolant 12,560 Refueling Water 46,800 Accumulators 4,000
Total 63,360 ft 3
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ORIGINAL REFUELING SHIELD DESIGN PARAMETERS 1
Total number of fuel assemblies 193
Minimum full power exposure 1000 days
Minimum time between shutdown and fuel handling 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />
Maximum exposure rate adjacent to spent fuel pit 1.0 mrem/hr
Maximum exposure rate at water surface 2.5 mrem/hr
1 These parameters are kept for historical reasons. The dose rates are no longer applicable since the design of the spent fuel pit has been changed.
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PRINCIPAL AUXILIARY SHIELDING
Design parameters for the auxiliary shielding include:
Core thermal power 3391 MWt Fraction of fuel rods containing small clad defects 0.01 Reactor coolant liquid volume 12600 ft.3 1
Letdown flow (normal purification) 75 gpm Cesium purification flow (intermittent) 75 gpm Cut-in concentration deborating demineralizer 100 ppm Dose rate outside auxiliary building <1 mrem/hr Dose rate in the building outside shield walls <2.5 mrem/hr COMPONENT CONCRETE SHIELD THICKNESS Ft. - In.
Mixed Bed Demineralizers 4 - 0 Charging pumps 2 - 6 Liquid holdup tanks 2 - 8 Volume control tank 3 - 9 Reactor Coolant filter 2 - 6 Boric Acid Evaporator 2 - 4 Gas decay tanks 3 - 3 Waste Gas Compressors 2 - 8 Waste Evaporator 2 - 0 Liquid Waste Holdup Tank 2 - 0 Spent Resin Storage Tank 4 - 0
1 This value has been conservatively chosen for the purpose of shield design. Actual best-estimated reactor coolant systems volumes can be obtained from the current Westinghouse IMP databases.
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CORE AND GAP ACTIVITIES Assumptions: Operation at 3391 MWt for 650 days.
Temperature Distribution Specified in Table 11.2-9
Isotope Curies in the Core Percent of Core Curies in the Gap (x 10 7) Activity in the Gap (x 10 5)
I-131 8.26 2.3 19.0 I-132 12.65 0.26 3.29 I-133 18.76 0.79 14.82 I-134 21.92 0.16 3.51 I-135 17.02 0.43 7.32 Xe-133 18.00 1.85 33.30 Xe-133m 0.45 1.27 0.57 Xe-135 5.31 0.54 2.87 Xe-135m 5.22 0.086 0.45 Kr-85 0.095 21.57 2.05 Kr-85m 4.30 0.29 1.25 Kr-87 7.79 0.20 1.56 Kr-88 10.60 0.29 3.07
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INSTANTANEOUS RADIATION SOURCES RELEASEDTO THE CONTAINMENT
FOLLOWING TID-14844 ACCIDENT RELEASE - MEV/SEC
GAMMA ENERGY (MEV/PHOTON)
TIME AFTER RELEASE 0.4 0.8 1.3 1.7 2.2 2.5 3.5
0 HR 2.94x1018 1.42x1019 3.29x1018 1.51x1019 1.24x1019 6.24x1018 6.31x1018
0.5 HR 2.82x1018 1.17x1019 2.51x1018 1.57x1018 8.10x1018 5.09x1018 2.34x1017
1 HR 2.74x1018 9.97x1018 2.18x1018 1.32x1018 6.48x1018 4.24x1018 1.20x1017
2 HR 2.61x1018 7.46x1018 1.68x1018 1.01x1018 5.15x1018 3.01x1018 3.56x1016
8 HR 2.04x1018 2.76x1018 5.70x1017 3.16x1017 2.21x1018 5.53x1017 1.19x1015
1 DY 1.15x1018 1.28x1018 1.00x1017 1.30x1017 3.63x1017 3.08x1016 4.27x1014
1 WK 4.41x1017 2.15x1017 6.07x1015 8.04x1016 1.66x1015 7.39x1015 3.29x1014
1 MO 2.76x1017 1.41x1017 2.25x1015 2.63x1016 1.58x1015 2.41x1015 1.13x1014
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CORE TEMPERATURE DISTRIBUTION
% of Core Fuel Volume Local Temperature, oF Above the Given Temperature 0.0 4100 0.2 3700 1.8 3300 7.0 2900 14.5 2500
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CONCENTRATION OF IODINE ISOTOPES IN THE RECIRCULATION LOOP
ISOTOPES RECIRCULATION LOOP CONCENTRATION(c/cc)
I-131 1.06X103 I-132 1.83X102 I-133 8.26X102 I-134 1.96X102 I-135 4.08X102
The radiation sources circulating in the residual heat removal loop are shown in Table 11.2-11 and are used for whole body radiation doses in the auxiliary building.
The radioactivity in the containment also would be additional source of radiation to the auxiliary building following a loss -of-coolant accident.
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GAP ACTIVITY CIRCULATING IN RESIDUAL HEAT REMOVAL LOOP, MEV/CC-SEC
GAMMA ENERGY (MEV/PHOTON)
TIME AFTER RELEASE 0.4 0.8 1.3 1.7 2.2 2.5 3.5
0 HR 1.63x107 1.31x108 8.54x106 4.90x106 4.61x106 1.70x106 4.50x105
0.5 HR 1.51x107 1.23x108 7.56x106 4.16x106 4.16x106 1.61x106 3.78x105
1 HR 1.39x107 1.14x108 6.18x106 3.46x106 3.67x106 1.20x106 2.78x105
2 HR 1.28x107 1.03x108 4.59x106 2.53x106 3.01x106 8.24x105 2.00x105
8 HR 1.11x107 7.75x107 7.16x105 4.16x105 5.61x105 1.30x105 2.51x104
1 DY 1.03x107 6.99x107 4.84x104 1.82x104 1.75x105 7.07x103 9.96x101
1 WK 9.54x106 4.88x107 1.16x102 2.93x102
1 MO 1.21x106 4.69x107
6 MO 4.16x104 1.56x107
1 YR 1.22x103 1.31x107
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DOSE RATE (REM/HR) RHR OR CONTAINMENT SPRAY
Time Pu mp Roo m Heat Exchanger Room Safety Injection Pump Room
0 2.8 22.7 37.6 0 - 5 hr 2.3 18.6 32.2 1 hr. 2.0 16.3 27.4 2 hr. 1.6 13.4 22.1 8 hr. 0.83 6.6 10.5 1 day 0.l8 l.5 2.6 1 week 0.02 0.2 0.41 1 month 0.008 0.08 0.18
Under the assumptions of: (1) increased sump dilution by melted ice, (2) core and halogen releases in accordance with Safety Guide No. 4 (in effect on September 1971), and (3) washdown of 50% of the core halogens to the sump occurs as a result of the action of the containment sprays.
INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
U1 Containment-Air Particulate ERS-1301, 1401 -4 Cs137, Radioactive U2 Containment-Air Particulate ERS-2301, 2401 Air 1x10 to 10 Ci Particulates
U1 Containment-Air Iodines ERS-1303, 1403 Air 2x10-4 to 3 Ci I131, Radioiodine U2 Containment-Air Iodines ERS-2303, 2403
U1 Containment Normal Range Noble-Gas ERS-1305, 1405 Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gases U2 Containment Normal Range Noble-Gas ERS-2305, 2405
U1 Steam Jet Air Ejector Normal Range Gas SRA-1905-A, 1905-B Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gases U2 Steam Jet Air Ejector Normal Range Gas SRA-2905-A, 2905-B U1 Steam Jet Air Ejector Accident Range Gas SRA-1909 Air 1x10-2 to 9x104 Ci/cc Xe133, Noble Gases U2 Steam Jet Air Ejector Accident Range Gas SRA-2909 1x10-2 to 9x104 Ci/cc
-5 -2 Co60, Mixed Fission U1 Component Cooling Loop Liquid 1-CRA-415 & 1-CRA-425 Water 1x10 to 1x10 Ci/cc Products
2-CRA-415 & 2-CRA-425 -5 -2 Co60, Mixed Fission U2 Component Cooling Loop Liquid Water 1x10 to 1x10 Ci/cc Products INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
RRS-1001-A Co60, Mixed Fission Waste Disposal System Liquid Effluent RRS-1001-B Water 1x10-7 to 4.43x10-2 Ci/cc Products
Cs137,Mixed Fission U1 Steam Generator Blowdown Liquid 1-DRA-300 Products Water 2x10-6 to 2x100 Ci/cc Co60, Mixed Fission U2 Steam Generator Blowdown Liquid 2-DRA-300 Products
Cs137,Mixed Fission U1 Essential Service Water Liquid 1-WRA-713 Water 1x10-5 to 4x10-1 Ci/cc Products Co60, Mixed Fission Products Cs137,Mixed Fission U2 Essential Service Water Liquid 2-WRA-714 Water 1x10-5 to 4x10-1 Ci/cc Products Co60, Mixed Fission Products Turbine Room Sump Compositor Water Not Applicable Not Applicable INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
U1 Steam Generator Blowdown Treatment 1-DRA-353 System Liquid Co60, Mixed Fission Water 1x10-6 to 2x10-1 Ci/cc Products U2 Steam Generator Blowdown Treatment System Liquid 2-DRA-353
U1 Unit Vent Air Particulate VRA-1501 Air -4 Cs137, Radioactive U2 Unit Vent Air Particulate VRA-2501 1x10 to 10 Ci Particulates
U1 Unit Vent Radioiodine VRA-1503 Air 2x10-4 to 3 Ci I131, Radioiodine U2 Unit Vent Radioiodine VRA-2503 Unit Vent Normal Noble Gas VRS-1505-A, 1505-B, Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gas 2505-A, 2505-B Unit Vent Accident Noble Gas VRS-1509 Air 1x10-4 to 9x104 Ci/cc Xe133, Noble Gas
Unit Vent Accident Noble Gas VRS-2509 Air 1x10-4 to 9x104 Ci/cc Xe133, Noble Gas
Gland Seal Condenser Exhaust Monitor SRA-1805, 2805 Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gas INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
SRA-1809 Air 1x10-2 to 9x104 Ci/cc Xe133, Noble Gas
SRA-2809 Air 1x10-2 to 9x104 Ci/cc Xe133, Noble Gas
Co60, Mixed Fission U1 Essential Service Water Liquid 1-WRA-717 Water 1x10-5 to 1x10-2 Ci/cc Products
-5 -2 Co60, Mixed Fission U2 Essential Service Water Liquid 2-WRA-718 Water 1x10 to 1x10 Ci/cc Products
Containment Area at Personnel Lock VRS-1101, 2101 Air 1x10-1 to 1x104 mR/hr
Upper Containment Area Monitor VRS-1201, 2201 Air 1x10-1 to 1x104 mR/hr
Steam Generator Power Operated Relief Valve MRA-1600, 2600 Vapor 1x10-1 to 1x10+2 Ci/cc Xe133, Noble Gas Monitor 1700, 2700 Sampling Room Iodine ERA-7003 Air 2x10-4 to 3 Ci I131, Radioiodine INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 5 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
Sampling Room Low Range Noble Gas ERA-7005 Air 1x10-7 to 1x10-1 Ci/cc Xe133, Noble Gas
Sampling Room Area ERA-7006 Air 1x10-2 to 1x107 mR/hr
-2 7 Spent Fuel Area 12-RRC-330 Air 1x10 to 1x10 mR/hr
ERA-7402 (Unit 1) Air 1x10-4 to 1x104 R/hr 1 In-Core Instrumentation Room Area ERA-8402 (Unit 2) Air 1x10-4 to 1x104 R/hr 1
Drumming Station Area 12-RRA-322 Air 1x10-2 to 1x107 mR/hr 12-ERA-7505
1 These monitors are calibrated to the appropriate range for the expected radiation levels in a particular area.
INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 6 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
VRA-1310, High Range Containment Area Monitor VRA-2310, Air 1 to 1x107 R/HR VRA-1410, VRA-2410, Vestibule Elevation 591' ERA-1306, -2306 Air 1x10-3 to 1x102 mR/hr
Outside Containment Spray Pump Rooms ERA-1406, -2406 Air 1x10-3 to 1x102 mR/hr Elevation 573' West of Equipment Hatch Elevation 650' VRA-1506, -2506 Air 1x10-3 to 1x102 mR/hr
Turbine Building, Elevation 609' SRA-1906 Air 1x10-3 to 1x102 mR/hr
Turbine Building, Elevation 591' SRA-2906 Air 1x10-3 to 1x102 mR/hr
North of Boric Acid Tanks Elevation 587' RRA-1003 Air 1x10-1 to 1x104 mR/hr
Unit 1 E CCP Room ERA-7303 Air 1x10-4 to 1x104 R/hr 1 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 7 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
Unit 1 W CCP Room ERA-7304 Air 1x10-4 to 1x104 R/hr 1
Unit 1 E RHR Pump Room ERA-7305 Air 1x10-4 to 1x104 R/hr 1
Unit 1 W RHR Pump Room ERA-7306 Air 1x10-4 to 1x104 R/hr 1
Unit 1 N SIS Pump Room ERA-7307 Air 1x10-4 to 1x104 R/hr 1
Unit 1 S SIS Pump Room ERA-7308 Air 1x10-4 to 1x104 R/hr 1
Unit 1 Reactor Coolant Filter Cubicle ERA-7309 Air 1x10-4 to 1x104 R/hr 1
Unit 2 E CCP Room ERA-8303 Air 1x10-4 to 1x104 R/hr 1
Unit 2 W CCP Room ERA-8304 Air 1x10-4 to 1x104 R/hr 1
Unit 2 E RHR Pump Room ERA-8305 Air 1x10-4 to 1x104 R/hr 1
Unit 2 W RHR Pump Room ERA-8306 Air 1x10-4 to 1x104 R/hr 1 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 8 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
Unit 2 N SIS Pump Room ERA-8307 Air 1x10-4 to 1x104 R/hr 1
Unit 2 S SIS Pump Room ERA-8308 Air 1x10-4 to 1x104 R/hr 1
Unit 2 Reactor Coolant Filter Cubicle ERA-8309 Air 1x10-2 to 1x104 R/hr 1
Unit 1 Control Room ERS-7401 Air 1x10-1 to 1x104 mR/hr
Access Control Facility ERA-7403 Air 1x10-1 to 1x104 mR/hr
Radio Chemistry Lab ERA-7404 Air 1x10-1 to 1x104 mR/hr
Unit 1 N Seal Water Injection Filter Cubicle ERA-7407 Air 1x10-4 to 1x104 R/hr 1
Unit 1 S Seal Water Injection Filter Cubicle ERA-7408 Air 1x10-4 to 1x104 R/hr 1
Unit 1 Seal Water Filter Cubicle ERA-7409 Air 1x10-4 to 1x104 R/hr 1
Unit 2 Control Room ERS-8401 Air 1x10-1 to 1x104 mR/hr INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 9 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
609' Elevation Passageway ERA-8403 Air 1x10-4 to 1x104 R/hr 1
Unit 2 N Seal Water Injection Filter Cubicle ERA-8407 Air 1x10-4 to 1x104 R/hr 1
Unit 2 S Seal Water Injection Filter Cubicle ERA-8408 Air 1x10-4 to 1x104 R/hr 1
Unit 2 Seal Water Injection Filter, Filter ERA-8409 Air 1x10-4 to 1x104 R/hr 1 Cubicle 587' Elevation Passageway ERA-7504 Air 1x10-4 to 1x104 R/hr 1
Emergency Sampling Location ERA-7507 Air 1x10-1 to 1x104 mR/hr
573' Elevation Passageway ERA-7508 Air 1x10-4 to 1x104 R/hr 1
Refueling Water Purification Filter Cubicle ERA-7509 Air 1x10-4 to 1x104 R/hr 1
Unit 1 Vent Sampling Area ERA-7601 Air 1x10-1 to 1x104 mR/hr INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 11.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 10 of 10
Radiation Monitoring System Channel Sensitivities and Detecting Medium
Monitor Name Channel Number Medium Typical Range Detected Isotopes
Unit 1 Vent Sampling Flow Adjacent Area ERA-7602 Air 1x10-4 to 1x104 R/hr 1
Unit 2 Vent Sampling Area ERA-7603 Air 1x10-1 to 1x104 mR/hr
Unit 2 Vent Sampling Flow Adjacent Area ERA-7604 Air 1x10-4 to 1x104 R/hr 1
633' Elevation Passageway ERA-7605 Air 1x10-1 to 1x104 mR/hr
I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 16.1 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 11.3-2 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS RR EE PP OO RR TT Page: 1 of 1
REACTOR COOLANT FISSION AND CORROSION PRODUCT ACTIVITIES DURING STEADY STATE OPERATION AND PLANT SHUTDOWN OPERATION
OPERATING PWR PLANT DONALD C. COOK PLANT - 1% FUEL DEFECTS
MEASURED ACTIVITY MEURED µE CAUTEDCTI µTK ISOTOPE BEFORE SHUTDOWN SHWN ACTI BEFOHWN SHWN ACTITY
µCim µCip µCim µCim
I-131 0.83 14.9 2.4 43.0 Xe-133 127.0 65.0 1 254.0 130.0 (1)
Cs-134 1.29 1.7 0.19 0.25 Cs-137 1.67 2.14 1.1 1.4 Cs-144 0.00068 0.0058 0.00051 0.0044 Sr-89 0.0033 0.40 0.0042 0.51 Sr-90 0.00057 0.013 0.0001 0.0023 Co-58 --- 0.95 0.025 1.0
1 Activity reduced from steady state level by approximately one day of s ystem degassification prior to plant shutdown.
INNDDIIAANNAAI MIICCHHIIGGAANN PPOOWWEERRM Revision: 30.0 D.. CC.. CCOOOOKK NNDUCCLLEEAARR PPLLAANNTTU Table: 11.3-3 UPPDDAATTEEDD FFIINNAALL SSAAFFEETTYYU ANNAALLYYSSIISS RREEPPOORRTTA Page: 1 of 2
RADIATION MONITORING SYSTEM CHANNELS
CHANNEL PURPOSE ASSOCIATED TRIP FUNCTION OVERVIEW ERS-1301, 1401, 2301, 2401 Containment Airborne Particulates - Detection Containment ventilation isolation, prevent further release ERS-1303, 1403, 2303, 2403 Containment Radioiodine - Detection Containment ventilation isolation, prevent further release ERS-1305, 1405, 2305, 2405 Containment Normal Range Noble Gas - Detection Containment ventilation isolation, prevent further release ERS-7401, 8401 Control Room Area Monitor Isolate Control Room Ventilation CRA-415, CRA-425 Component Cooling Water Loop Liquid Monitor - Detect Isolate CCW surge tank vent leaks from RCS or RHR into the CCW system DRA-300 Steam Generator Blowdown Liquid Monitor - detect primary Isolate steam generator blowdown system.
to secondary leakage via common blowdown header WRA-713, WRA-714, WRA-Essential Service Water Liquid Monitor - Detect leakage in None 717, WRA-718 the containment spray heat exchangers, (post LOCA)
Steam Generator Blowdown Treatment System Liquid DRA-353 Monitor - measure activity in the blowdown liquid after it Isolate steam generator blowdown system passes the treatment demineralizer 12-RRC-330 SFP Area Monitor Place SFP ventilation into service RRS-1001-A, RRS-1001-B Waste Disposal System Liquid Effluent Monitor Automatic valve closure to prevent further release SRA-1805 Gland Seal Condenser Exhaust - Normal Range Detection None SRA-1809 Gland Seal Condenser Exhaust - Accident Range Detection None SRA-1905-A, 1905-B Steam Jet Air Ejector Normal Range Noble Gas - Detect None primary and secondary leakage SRA-1909 Steam Jet Air Ejector Accident Range Noble Gas - Detect None primary and secondary leakage SRA-2805 Gland Seal Condenser Exhaust - Normal Range Detection None INNDDIIAANNAAI MIICCHHIIGGAANN PPOOWWEERRM Revision: 30.0 D.. CC.. CCOOOOKK NNUCCLLEEAARR PPLLAANNTT D U Table: 11.3-3 UPPDDAATTEEDD FFIINNAALL SSAAFFEETTYYU ANNAALLYYSSIISS RREEPPOORRTTA Page: 2 of 2
RADIATION MONITORING SYSTEM CHANNELS
CHANNEL PURPOSE ASSOCIATED TRIP FUNCTION OVERVIEW SRA-2809 Gland Seal Condenser Exhaust -Accident Range Detection None SRA-2905-A, 2905-B Steam Jet Air Ejector Normal Range Noble Gas - Detect None primary and secondary leakage SRA-2909 Steam Jet Air Ejector Accident Range Noble Gas - Detect None primary and secondary leakage Unit Vent Continuous air flow Tritium sampling None sampler VRA-1501, 2501 Unit Vent Airborne Particulates - Detection None VRA-1503, 2503 Unit Vent Radioiodines - Detection None VRS-1505-A, 1505-B Unit Vent Normal Range Noble Gas - Detection Gas decay tank isolation valves 1 VRS-1509 Unit Vent Accident Range Noble Gas - Detection Sample pathway bypass of channels 1, 3, 5 to sample panel VRS-2505-A, 2505-B Unit Vent Normal Range Noble Gas - Detection Gas Decay Tank isolation valves 1 VRS-2509 Unit Vent Accident Range Noble Gas - Detection Sample pathway bypass of channels 1, 3, 5 to sample pallet
1 Available setpoint is used to accommodate: 1) normal operation, and 2) gas decay tank release.
I NIN DD II AA NN AA MM II CC HH II GG AA NN PP OO WW EE RR Revision: 16.1 D.D. CC.. CC OO OO KK NN UU CC LL EE AA RR PP LL AA NN TT Table: 11.5-1 U PUP DD AA TT EE DD FF II NN AA LL SS AA FF EE TT YY AA NN AA LL YY SS II SS RR EE PP OO RR TT Page: 1 of 1
DESIGN AND MEASURED EQUILIBRIUM REACTOR COOLANT FISSION PRODUCT
ACTIVITIES FOR OPERATING PWR'S AND CALCULATED VALUES FOR THED. C. COOK STATIONS
GINNA BEZNAU COOK STATION 1 STATION (1) STATION DESIGN MEASURED RAT IO DESIGN MEASURED RAT IO DESIGN VALUE 2 VALUE MEASURED VALUE (2) VALUE MEASURED VALUE (2)
µc/cc µc/cc (Design) µc/cc µc/cc (DESIGN) µc/cc
Total Activity 216 71 0.33 299 168 0.73 207 Isotopic Activity (Key Isotopes)
I-131 1.53 0.56 0.37 0.96 0.75 0.78 1.7
I-133 2.55 1.7 0.67 1.74 2.0 1.16 2.6
Xe-133 184 45 0.24 200 119 0.60 178
Cs-134 0.19 0.06 0.32 0.22 0.075 0.35 0.13
Cs-137 0.94 0.37 0.40 1.53 0.22 0.15 0.8
1 Based on an assumed 1% defect level.
2 Amendment 20 to original FSAR (Mar, 1972).
INDIANA MICHIGAN POWER Revision: 28.0 D. C. COOK NUCLEAR PLANT Table: 11.5-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2
Blowdown Treatment System Components
Pu mp
Number 1 per unit Fluid Steam generator blowdown Pressure, Suction Atmospheric Temperature 200°F Head 125 ft.
Flow 60 gpm T ype Horizontal centrifugal Material, Casing Stainless Steel Impeller Stainless Steel NPSH, minimum Ft. H2O 2.5 Heat Exchanger1
Number 1 per unit Shell Side (blowdown liquid)
Inlet Temperature 200°F Outlet Temperature 120°F Max. pressure 70 psi Operating pressure 50 psi Flow 60 gpm Material 304 Stainless Steel Pressure drop, normal 4 psi maximum allowable 15 psi Tube Side (non-essential service water)
Inlet Temperature 76°F
1 The system has been evaluated for a NESW pump discharge temperature of 88.9°F.
INDIANA MICHIGAN POWER Revision: 28.0 D. C. COOK NUCLEAR PLANT Table: 11.5-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2
Blowdown Treatment System Components Outlet Temperature 106°F Max. pressure 150 psig Operating pressure 75 psig Flow 160 gpm Material 304 Stainless Steel Pressure drop, normal 5 psi maximum allowable 9 psi Mixed Bed Demineralizers
Number 3 per unit T ype Flushable Vessel, design pressure, psig 200 Operating pressure, psig 50 Vessel, design temperature, °F 250 Operating temperature, ° F 120 Resin volume, ft3 56 (Nos. 1 & 2) 20 (No. 3)
Design flow rate, gpm 50 Resin type Cation and anion Material of construction Austenitic stainless steel