ML22340A181
Text
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.1-1 Page:
1 of 1 WASTE DISPOSAL SYSTEM PERFORMANCE DATA Plant Design Life 40 years Normal process capacity, 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 20.2 Table:
11.1-2 Page:
1 of 1 WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS1 COMPONENT CODE Chemical Drain Tank No code Reactor Coolant Drain Tanks ASME 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 ASME III,(2) Class C Piping and Valves USAS-B31.13, Section 1 ASME III Appendix F 4 Gas Decay Tank ASME III,(2) Class C Spent Resin Storage Tank ASME III,(2) Class C Waste Evaporator Condensate Demineralizer ASME III,(2) Class C Waste Evaporator Condensate Filter ASME 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).
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.1-3 Page:
1 of 2 COMPONENT
SUMMARY
DATA TANKS QUANTITY TYPE VOLUME DESIGN PRESSURE DESIGN TEMP.oF MATERIAL (1)
Reactor Coolant Drain (per unit) 1 Horiz 350 gal 25 psig 267 ss 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 Bottoms Storage 1 (2)
Vert 4,000 gal Atm 250 ss Spent Resin Storage 1 (2)
Vert 300 ft3 100 psig 180 ss PUMPS QUANTITY TYPE FLOW gpm HEAD ft.
DESIGN PRESSURE psig DESIGN TEMP oF MATERIAL (1)
Reactor Coolant Drain (A)
(per unit) 1 Horiz canned 50 175 150 300 ss Reactor Coolant Drain (B)
(per unit) 1 Horiz canned 150 175 150 300 ss Chemical Drain 1 (2)
Horiz 20 100 150 180 ss Laundry & Hot Shower 1 (2)
Horiz (3) 20 100 150 180 ss Sump Tank 2 (2)
Horiz (3) 20 100 150 180 ss Waste Evaporator 2 (2)
Horiz 20 100 150 180 ss Waste Condensate 2 (2)
Horiz (3) 150 200 150 180 ss Waste Evaporator Bottoms 1 (2)
Horiz (3) 20 60 150 180 ss (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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.1-3 Page:
2 of 2 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)
QUANTITY TYPE CAPACITY DESIGN PRESSURE psig DESIGN TEMP.
oF MATERIAL (1)
Waste Evaporator Condensate Filter 1(2)
Disposable cartridge 20 gpm 150 180 ss Waste Evaporator Condensate Demineralizer 1(2)
Flushable 30 ft3 100 250 ss UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.1-4 Page:
1 of 1 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.1-5 Page:
1 of 1 ESTIMATED LIQUID RELEASE BY ISOTOPE (Two Units)
ISOTOPE ANNUAL RELEASE µCi ISOTOPE ANNUAL RELEASE µCi 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 µCi/yr.
All Isotopes with total activity per year <1.0E-12 were ignored.)
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.1-6 Page:
1 of 1 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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.2-1 Page:
1 of 1 PLANT ZONE CLASSIFICATIONS ZONE ACCESS CONDITIONS MAXIMUM EXPOSURE RATE (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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.2-2 Page:
1 of 1 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 (n/cm2 - sec)
LEAKAGE FLUXES (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
(/cm2 - sec)
LEAKAGE FLUXES
(/cm2 - sec)
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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.6 Table:
11.2-3 Page:
1 of 1 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.2-4 Page:
1 of 1 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, ft3:
Reactor Coolant 12,560 Refueling Water 46,800 Accumulators 4,000 Total 63,360 ft 3 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.2-5 Page:
1 of 1 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.6 Table:
11.2-6 Page:
1 of 1 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
11.2-7 Page:
1 of 1 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 (x 107)
Percent of Core Activity in the Gap Curies in the Gap (x 105)
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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
11.2-8 Page:
1 of 1 INSTANTANEOUS RADIATION SOURCES RELEASED TO 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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table:
11.2-9 Page:
1 of 1 CORE TEMPERATURE DISTRIBUTION
% of Core Fuel Volume Above the Given Temperature Local Temperature, oF 0.0 4100 0.2 3700 1.8 3300 7.0 2900 14.5 2500 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table: 11.2-10 Page:
1 of 1 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table: 11.2-11 Page:
1 of 1 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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
23 Table: 11.2-12 Page:
1 of 1 DOSE RATE (REM/HR) RHR OR CONTAINMENT SPRAY Time Pump Room 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.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 U2 Containment-Air Particulate ERS-1301, 1401 ERS-2301, 2401 Air 1x10-4 to 10 Ci Cs137, Radioactive Particulates U1 Containment-Air Iodines U2 Containment-Air Iodines ERS-1303, 1403 ERS-2303, 2403 Air 2x10-4 to 3 Ci I131, Radioiodine U1 Containment Normal Range Noble-Gas U2 Containment Normal Range Noble-Gas ERS-1305, 1405 ERS-2305, 2405 Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gases U1 Steam Jet Air Ejector Normal Range Gas U2 Steam Jet Air Ejector Normal Range Gas SRA-1905-A, 1905-B SRA-2905-A, 2905-B Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gases U1 Steam Jet Air Ejector Accident Range Gas U2 Steam Jet Air Ejector Accident Range Gas SRA-1909 SRA-2909 Air 1x10-2 to 9x104 Ci/cc 1x10-2 to 9x104 Ci/cc Xe133, Noble Gases U1 Component Cooling Loop Liquid U2 Component Cooling Loop Liquid 1-CRA-415 & 1-CRA-425 Water 1x10-5 to 1x10-2 Ci/cc Co60, Mixed Fission Products 2-CRA-415 & 2-CRA-425 Water 1x10-5 to 1x10-2 Ci/cc Co60, Mixed Fission Products UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 Page:
2 of 10 Radiation Monitoring System Channel Sensitivities and Detecting Medium Monitor Name Channel Number Medium Typical Range Detected Isotopes Waste Disposal System Liquid Effluent RRS-1001-A RRS-1001-B Water 1x10-7 to 4.43x10-2 Ci/cc Co60, Mixed Fission Products U1 Steam Generator Blowdown Liquid U2 Steam Generator Blowdown Liquid 1-DRA-300 2-DRA-300 Water 2x10-6 to 2x100 Ci/cc Cs137,Mixed Fission Products Co60, Mixed Fission Products U1 Essential Service Water Liquid 1-WRA-713 Water 1x10-5 to 4x10-1 Ci/cc Cs137,Mixed Fission Products Co60, Mixed Fission Products U2 Essential Service Water Liquid 2-WRA-714 Water 1x10-5 to 4x10-1 Ci/cc Cs137,Mixed Fission Products Co60, Mixed Fission Products Turbine Room Sump Compositor Water Not Applicable Not Applicable UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 System Liquid U2 Steam Generator Blowdown Treatment System Liquid 1-DRA-353 2-DRA-353 Water 1x10-6 to 2x10-1 Ci/cc Co60, Mixed Fission Products U1 Unit Vent Air Particulate U2 Unit Vent Air Particulate VRA-1501 VRA-2501 Air 1x10-4 to 10 Ci Cs137, Radioactive Particulates U1 Unit Vent Radioiodine U2 Unit Vent Radioiodine VRA-1503 VRA-2503 Air 2x10-4 to 3 Ci I131, Radioiodine Unit Vent Normal Noble Gas VRS-1505-A, 1505-B, 2505-A, 2505-B Air 9x10-7 to 5x10-2 Ci/cc Xe133, Noble Gas 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 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 U1 Essential Service Water Liquid U2 Essential Service Water Liquid 1-WRA-717 Water 1x10-5 to 1x10-2 Ci/cc Co60, Mixed Fission Products 2-WRA-718 Water 1x10-5 to 1x10-2 Ci/cc Co60, Mixed Fission 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 Monitor MRA-1600, 2600 1700, 2700 Vapor 1x10-1 to 1x10+2 Ci/cc Xe133, Noble Gas Sampling Room Iodine ERA-7003 Air 2x10-4 to 3 Ci I131, Radioiodine UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 Spent Fuel Area 12-RRC-330 Air 1x10-2 to 1x107 mR/hr In-Core Instrumentation Room Area ERA-7402 (Unit 1)
Air 1x10-4 to 1x104 R/hr 1 ERA-8402 (Unit 2)
Air 1x10-4 to 1x104 R/hr 1 Drumming Station Area 12-RRA-322 12-ERA-7505 Air 1x10-2 to 1x107 mR/hr 1 These monitors are calibrated to the appropriate range for the expected radiation levels in a particular area.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 Page:
6 of 10 Radiation Monitoring System Channel Sensitivities and Detecting Medium Monitor Name Channel Number Medium Typical Range Detected Isotopes High Range Containment Area Monitor VRA-1310, VRA-2310, VRA-1410, VRA-2410, Air 1 to 1x107 R/HR Vestibule Elevation 591' ERA-1306, -2306 Air 1x10-3 to 1x102 mR/hr Outside Containment Spray Pump Rooms Elevation 573' ERA-1406, -2406 Air 1x10-3 to 1x102 mR/hr 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 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 Cubicle ERA-8409 Air 1x10-4 to 1x104 R/hr 1 587' 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 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
30.0 Table:
11.3-1 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 UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.3-2 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 ISOTOPE MEASURED ACTIVITY BEFORE SHUTDOWN
µCi/gm MEASURED PEAK SHUTDOWN ACTIVITY
µCi/gp CALCULATED ACTIVITY BEFORE SHUTDOWN
µCi/gm EXPECTED PEAK SHUTDOWN ACTIVITY
µCi/gm 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 system degassification prior to plant shutdown.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 30.0 Table:
11.3-3 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 leaks from RCS or RHR into the CCW system Isolate CCW surge tank vent DRA-300 Steam Generator Blowdown Liquid Monitor - detect primary to secondary leakage via common blowdown header Isolate steam generator blowdown system.
WRA-713, WRA-714, WRA-717, WRA-718 Essential Service Water Liquid Monitor - Detect leakage in the containment spray heat exchangers, (post LOCA)
None DRA-353 Steam Generator Blowdown Treatment System Liquid Monitor - measure activity in the blowdown liquid after it passes the treatment demineralizer Isolate steam generator blowdown system 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 primary and secondary leakage None SRA-1909 Steam Jet Air Ejector Accident Range Noble Gas - Detect primary and secondary leakage None SRA-2805 Gland Seal Condenser Exhaust - Normal Range Detection None UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 30.0 Table:
11.3-3 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 primary and secondary leakage None SRA-2909 Steam Jet Air Ejector Accident Range Noble Gas - Detect primary and secondary leakage None Unit Vent Continuous air flow sampler Tritium sampling None 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.
UFSAR Revision 31.0
IINNDDIIAANNAA M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision: 16.1 Table:
11.5-1 Page:
1 of 1 DESIGN AND MEASURED EQUILIBRIUM REACTOR COOLANT FISSION PRODUCT ACTIVITIES FOR OPERATING PWR'S AND CALCULATED VALUES FOR THE D. C. COOK STATIONS GINNA STATION 1 BEZNAU STATION (1)
COOK STATION DESIGN VALUE 2
µc/cc MEASURED VALUE
µc/cc RATIO MEASURED (Design)
DESIGN VALUE (2)
µc/cc MEASURED VALUE
µc/cc RATIO MEASURED (DESIGN)
DESIGN VALUE (2)
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).
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Table:
11.5-2 Page:
1 of 2 Blowdown Treatment System Components Pump Number 1 per unit Fluid Steam generator blowdown Pressure, Suction Atmospheric Temperature 200°F Head 125 ft.
Flow 60 gpm Type 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.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Table:
11.5-2 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 Type 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 UFSAR Revision 31.0