ML21125A596

From kanterella
Jump to navigation Jump to search
0 to Updated Final Safety Analysis Report, Chapter 1, Tables
ML21125A596
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 04/19/2021
From:
Indiana Michigan Power Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML21126A238 List: ... further results
References
AEP-NRC-2021-19
Download: ML21125A596 (46)
v • d • e


Text

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.0-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 3 UFSAR Revision History Revision Number Submittal Date Submittal Description 0 July 1982 Original UFSAR 1 July 1983 1983 Update 2 July 1984 1984 Update 3 July 1985 1985 Update 4 July 1986 1986 Update 5 July 1987 1987 Update 6 July 1988 1988 Update 7 July 1989 1989 Update 8 July 1990 1990 Update 9 July 1991 1991 Update 10 July 1992 1992 Update 11 July 1993 1993 Update 12 July 1994 1994 Update 13 July 1995 1995 Update 14 July 1996 1996 Update 15 July 1997 1997 Update 16.0 July 1999 1999 Update 16.1 10/20/1999 Minor Version Update 16.2 11/24/1999 Minor Version Update 16.3 06/23/2000 Minor Version Update 16.4 08/21/2000 Minor Version Update 16.5 10/20/2000 Minor Version Update

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.0-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 3 UFSAR Revision History Revision Number Submittal Date Submittal Description 16.6 02/09/2001 Minor Version Update 17.0 06/14/2001 Major Revision 17.1 10/31/2001 Minor Version Update 17.2 04/16/2002 Minor Version Update 17.3 09/09/2002 Minor Version Update 18.0 12/07/2002 Major Revision 18.1 07/31/2003 Minor Version Update 18.2 03/02/2004 Minor Version Update 19.0 06/18/2004 Major Revision 19.1 09/29/2004 Minor Version Update 19.2 12/15/2004 Minor Version Update 19.3 03/24/2005 Minor Version Update 20.0 08/19/2005 Major Revision 20.1 11/30/2005 Minor Version Update 20.2 06/01/2006 Minor Version Update 21.0 04/05/2007 Major Revision 21.1 08/28/2007 Minor Version Update 21.2 04/02/2008 Minor Version Update 22.0 09/12/2008 Major Revision 22.1 08/14/2009 Minor Version Update 23.0 09/10/2010 Major Revision 24.0 03/17/2012 Major Revision

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.0-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 3 UFSAR Revision History Revision Number Submittal Date Submittal Description 25.0 09/092013 Major Revision 26.0 03/03/2015 Major Revision 27.0 07/20/2016 Major Revision 28.0 05/25/2018 Major Revision 29.0 10/24/2019 Major Revision

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 1 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 1 Total Primary Heat Output, MWt 3250 3250 2758 1518.5 2200 2 Total Core Heat Output, Btu/hr 11,090 x 106 11,090 x 106 9413 x 106 5181 x 106 7479 x 106 3 Heat Generated in Fuel, % 97.4 97.4 97.4 97.4 97.4 4 Maximum thermal Overpower 12% 12% 12% 12% 12%

5 System Pressure, Nominal, psia 2250 2250 2250 2250 2250 System Pressure, Minimum Steady 6 2220 2220 2220 2220 2220 State, psia Hot Channel Factors 7 Heat Flux, Fq 2.79 2.79 3.23 2.80 3.23 8 Enthalphy Rise, FH 1.60 1.60 1.77 1.60 1.77

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 2 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report DNB Ratio at Nominal Operating 9 1.97 2.02 2.00 2.11 1.81 Conditions 10 Minimum DNBR for Design Transients 1.30 1.30 1.30 1.30 1.30 Coolant Flow 11 Total Flow Rate, lb/hr 135.6 x 106 133.0 x 106 136.3 x 106 66.7 x 106 101.5 x 106 Effective Flow Rate for Heat Transfer, 12 129.5 x 106 128.9 x 106 130 x 106 63.6 x 106 97.0 x 106 lb/hr Effective Flow Area for Heat Transfer, 13 51.4 x 103 51.4 x 103 51.4 x 103 51.4 x 103 51.4 x 103 ft2 Average Velocity Along Fuel Rods, 14 15.5 15.3 15.4 15.0 14.3 ft/sec 15 Average Mass Velocity, lb/hr-ft2 2.53 x 106 2.52 x 106 2.53 x 106 2.37 x 106 2.32 x 106 Coolant Temperature, °F Design 16 536.3 530.2 543 552.5 546.2 Nominal Inlet

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 3 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Maximum Inlet Due to Instrumentation 17 540.3 534.2 547 556.5 550.2 Error and Deadband, °F 18 Average Rise in Vessel, °F 63.0 64.1 53.0 57.6 55.9 19 Average Rise in Core 65.7 66.8 55.5 60.0 58.3 20 Average in Core 570.3 564.8 571.0 582.5 575.4 21 Average in Vessel 567.8 563.2 569.5 581.3 574.2 22 Nominal Outlet of Hot Channel 667.5 631.7 633.5 642.9 642 23 Average Film Coefficient, Btu/hr-ft2-F 5850 5800 5790 5600 5400 Average Film Temperature Difference, 24 35.4 35.6 30.3 31.0 31.8

°F

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 4 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Heat Transfer at 100% Power 25 Active Heat Transfer Surface Area, ft2 52,200 52,200 52,200 28,715 42,460 26 Average Heat Flux, Btu/hr-ft2 207,900 207,900 175,600 175,800 171,600 27 Maximum Heat Flux, Btu/hr-ft2 579,600 579,600 567,300 491,000 554,200 28 Average Thermal Output, kw/ft 6.7 6.7 5.7 5.7 5.5 29 Maximum Thermal Output, kw/ft 18.8 18.8 18.4 16.0 17.0 Maximum Clad Surface Temp at 30 657 657 657 657 657 Nominal Pressure, °F

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 5 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Fuel Central Temperature, °F 31 Maximum at 100% Power 4250 4250 4090 3750 4030 32 Maximum at Overpower 4500 4500 4380 4000 4300 Thermal Output, kw/ft at Maximum 33 21.1 21.1 20.6 17.9 20.0 Overpower Core Mechanical Design Parameters Fuel Assemblies RCC Canless RCC Canless RCC Canless RCC Canless RCC Canless 34 Design 15x15 15x15 15x15 14x14 15x15 35 Rod Pitch, in. 0.563 0.563 0.563 0.556 0.563 36 Overall Dimensions, In. 8.426 x 8.426 8.426 x 8.426 8.426 x 8.426 7.763 x 7.763 8.426 x 8.426 37 Fuel Weight (as UO2), pounds 216, 600 216, 600 216, 000 120, 130 176, 200

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 6 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 38 Total Weight, pounds 276, 000 276, 000 276, 000 154, 519 226, 200 39 Number of Grids per Assembly 7 7 9 7 7 Fuel Rods 40 Number 39,372 39,372 39,372 21,659 32,028 41 Outside Diameter, in. 0.422 0.422 0.422 0.422 0.422 42 Diametral Gap, in. (Region 1, 2) 0.0075 0.0075 0.0065 0.0065 0.0065 (Region 3) 0.0085 0.0085 43 Clad Thickness, in 0.0243 0.0243 0.0243 0.0243 0.0243 44 Clad Material Zircaloy Zircaloy Zircaloy Zircaloy Zircaloy

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 7 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Fuel Pellets 45 Material UO2Sintered UO2Sintered UO2Sintered UO2Sintered UO2Sintered 46 Density (% of Theoretical) 94-93-92 94-93-92 94-92-91 94-92-91 94-92-91 47 Diameter Gap, in. (Region 1, 2) 0.3659 0.3659 0.3669 0.3669 0.3669 (Region 3) 0.3649 0.3649 48 Length, in. 0.6000 0.6000 0.6000 0.6000 0.6000 Rod Cluster Control Assemblies 5% Cd-15% In- 5% Cd-15% 5% Cd-15% 5% Cd-15% 5% Cd-15%

49 Neutron Absorber 80%Ag In-80%Ag In-80%Ag In-80%Ag In-80%Ag Type 304 SS- Type 304 SS- Type 304 SS- Type 304 SS- Type 304 SS-50 Cladding Material Cold Worked Cold Worked Cold Worked Cold Worked Cold Worked

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 8 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 51 Clad Thickness, In. 0.019 0.019 0.019 0.019 0.019 52 Number of Cluster 53 53 53 37 53 53 Number of Control Rods per Cluster 20 20 20 16 20 Core Structure 133.875/

54 Core Barrel I.D./O.D., in. 148.0/15215 148.0/152.5 148.0/152.5 109.0/112.5 137.875 55 Thermal Shield I.D./O.D., in. 158.5/164.0 158.5/164.0 158.5/164 115.3/122.5 Final Nuclear Design Data Structural Characteristics 56 Fuel Weight (As UO2), lbs 216,600 216,600 216,000 120,130 176,200 57 Clad Weight, lbs 44,547 44,547 44,600 24,260 36,300

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 9 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 58 Core Diameter, in (Equivalent) 132.7 132.7 132.5 96.5 119.5 59 Core Height, in. (Active Fuel) 144 143.4 144 144 144 Reflector Thickness and Composition 60 Top - Water plus Steel, in. 10 10 10 10 10 61 Bottom - Water plus Steel, in. 10 10 10 10 10 62 Side - Water plus Steel, in. 15 15 15 15 15 63 H2 O/U, (Cold volume Ratio) 4.09 4.09 4.18 4.20 4.18 64 Number of Fuel Assemblies 193 193 193 121 157 65 UO2 Rods per Assembly 204 204 204 179 204

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 10 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Performance Characteristics 3 region, 3 region, 3 region, 3 region, 3 region, 66 Loading Technique non-uniform non-uniform non-uniform non-uniform non-uniform Fuel Discharge Burnup, MWD/MTU 67 Average First Cycle 14,000 14,000 14,200 15,100 14,500 68 Equilibrium Core Average 21,800 21,800 24,700 33,000 33,000 Feed Enrichments, weight %

69 Region 1 2.25 2.25 2.2 2.27 1.85 70 Region 2 2.80 2.80 2.7 3.03 2.55 71 Region 3 3.30 3.30 3.2 3.40 3.10 Equilibrium 3.2 3.2 - 3.40 3.10

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 11 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Control Characteristics Effective Multiplication (Beginning of Life) 72 Cold, No Power, Clean 1.183 1.183 1.257 1.211 1.180 73 Hot, No Power, Clean 1.154 1.154 1.999 1.167 1.38 Hot, Full Power, Xe and Sm 74 1.092 1.092 1.152 1.113 1.077 Equilibrium Rod Cluster Control Assemblies 5% Cd-15% In- 5% Cd-15% 5% Cd-15% 5% Cd-15% 5% Cd-15%

75 Material 80% Ag In-80% Ag In-80% Ag In-80% Ag In-80% Ag 76 Number of RCC Assemblies 53 53 53 53 53 Number of Absorber Rods per RCC 77 20 20 20 20 20 Assembly

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 12 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report See Table See Table See Table See Table See Table 78 Total Rod Worth 3.2.1-3 3.2.1-3 3.2.1-3 3.2.1-3 3.2.1-3 Boron Concentration To shut reactor down with no rods 79 inserted, 1408/1265 1408/1265 1480/1370 1598/1676 1250/1210 Clean (keff = .99) Cold/Hot, ppm/ppm To control at power with no rods 80 1168/850 1168/850 1200/780 1465/1007 1000/920 inserted, clean/equilibrium xenon and samarium, ppm/ppm 1% k/k/ 1% k/k/ 1% k/k/ 1% k/k/

81 Boron worth, Hot 7.3 k/k 85 ppm 85 ppm 89 ppm 130 ppm 1% k/k/ 1% k/k/ 1% k/k/ 1% k/k/

82 Boron worth, Cold 5.6 k/k 70 ppm 70 ppm 72 ppm 98 ppm

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 13 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Kinetic Characteristics Moderator Temperature, Coefficient, -0.3 x 10-4 to -0.3 x 10-4to -0.3 x 10-4 to +0.3 x 10-4 to +0.3 x 10-4 to 83 k/k/°F -3.2 x 10-4 -3.2 x 10-4 -3.0 x 10-4 -3.5 x 10-4 -3.5 x 10-4 Moderator Pressure Coefficient, +0.3 x 10-6 to +0.3 x 10-6 to -0.3 x 10-6 to -0.3 x 10-6 to -0.3 x 10-6 to 84 k/k/psi +4.0 x 10-6 +4.0 x 10-6 +3.0 x 10-6 3.5 x 10-6 3.5 x 10-6 Moderator Density Coefficient -0.1 x 10-5 to -0.1 x 10-5 to +0.5 x 10-3 to 85 +0.03 to -0.30 -0.10 to -0.30 k/k/g/cm3 -0.8 x 10-5 -0.8 x 10-5 -2.5 x 10-3

-1.0 x 10-5 to -1.0 x 10-5 to -1 1 x 10-5 to -1 x 10-5 to -1 x 10-5 to 86 Doppler Coefficient, k/k/°F

-1.7 x 10-5 -1.7 x 10-5 -1 8 x 10-5 -1.6 x 10-5 -1.6 x 10-5 Reactor Coolant System Code Requirements Component ASME III ASME III ASME III ASME III ASME III 87 Reactor Vessel Class A Class A Class A Class A Class A Steam Generator

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 14 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Tube Side ASME III ASME III ASME III ASME III ASME III 88 Class A Class A Class A Class A Class A ASME III ASME III ASME III ASME III ASME III 89 Shell Side Class C Class C* Class C* Class C* Class C*

ASME III ASME III ASME III ASME III ASME III 90 Pressurizer Class A Class A Class A Class A Class A ASME III ASME III ASME III ASME ASME 91 Pressurizer Relief Tank Class C Class C Class C Class C Class C 92 Pressurizer Safety Valves ASME III ASME III ASME III ASME III 93 Reactor Coolant Piping USAS B31.1 USAS B31.1 USAS B31.1 USAS B31.1 USAS B31.1

The shell side of the steam generator conforms to the requirements for Class A vessels and is so stamped as permitted under the rules of Section III.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 15 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Principal Design Parameters Of The Reactor Coolant System 94 Reactor Primary Heat Output, MWt 3250 3250 2758 1518.5 2200 95 Reactor Primary Heat Output, Btu/hr 11,090 x 106 11,090 x 106 9413 x 106 5181 x 106 7508 x 106 96 Operating Pressure, psig 2235 2235 2235 2235 2235 97 Reactor Inlet Temperature 536.3 530.2 543 552.5 546.2 98 Reactor Outlet Temperature 599.3 594.3 596.0 610.0 602.1 99 Number of Loops 4 4 4 2 3 100 Design Pressure, psig 2485 2485 2485 2485 101 Design Temperature, °F 650 650 650 650 650 102 Hydrostatic Test Pressure (Cold), psig 3107 3107 3110 3110 3110

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 16 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Coolant Volume, including pressurizer, 103 12,612 12,710 12,600 6450 9088 cu. Ft.

104 Total Reactor Flow, gpm 350,000 350,000 178,000 268,500 104A Total Reactor Flow lb/sec 37,765 31,765 Principal Design Parameters Of The Reactor Vessel SA-302 SA-302 SA-302 Grade B, low Grade B, low Grade B, low Same as alloy steel, alloy steel, alloy steel, Same as others others 105 Material internally clad internally clad internally clad See Table 4.2-1 See Table with with with 4.2-1 austenitic austenitic austenitic stainless steel stainless steel stainless steel 106 Design Pressure, psig 2485 2485 2485 2485 2485 107 Design Temperature, °F 650 650 650 650 650

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 17 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 108 Operating Pressure, psig 2235 2235 2235 2235 2235 109 Inside Diameter of Shell, in. 173 173 173 132.0 155.5 110 Outside Diameter Across Nozzles, in. 262-7/16 262-7/16 262-7/16 244-1/16 236 43-9-23/32 (Unit 1) 111 Minimum Clad Thickness, in. 43-9-11/16 43-9-11/16 39-0 41-6 43-9 15/16 (Unit 2)

Overall Height of Vessel & Enclosure 112 5/32 5/32 5/32 5/32 5/32 Head, ft-in.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 18 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Principal Design Parameters Of The Steam Generators 113 Number of Units 4 4 4 2 3 Vertical U- Vertical U- Vertical U- Vertical U-Vertical U-Tube Tube with Tube with Tube with Tube with with integral-114 Type integral- integral- integral- integral-moisture moisture moisture moisture moisture separator separator separator separator separator 115 Tube Material Inconel Inconel Inconel Inconel Inconel 116 Shell Material Carbon Steel Carbon Steel Carbon Steel Carbon Steel Carbon Steel 117 Tube Side Design Pressure, psig 2485 2485 2485 2485 2485 118 Tube Side Design Temperature, °F 650 650 650 650 650 119 Tube Side Design Flow, lb/hr 33.9 x 106 33.8 x 106 34.1 x 106 33.4 x 106 33.9 x 106

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 19 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 120 Shell Side Design Pressure, psig 1085 1085 (design) 1085 1085 1085 121 Shell Side Design Temperature, °F 600 1/4 556 556 556 Operating Pressure, Tube Side, 122 2235 3107 2235 2235 2235 Nominal psig Operating Pressure, Shell Side, 123 1085 (design) 1085 (design) 1105.3 1020 1020 Max, psig Maximum Moisture at Outlet at Full 124 1/4 1/4 1/4 1/4 1/4 Load, %

Hydrostatic Test Pressure, Tube Side 125 3107 3107 3110 3110 3110 (cold), psig

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 20 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Principal Design Parameters Of The Reactor Coolant Pumps 126 Number of Units 4 4 4 4 4 Vertical, Vertical, Vertical, Vertical, Vertical, single single stage single stage single stage single stage stage radial flow radial flow radial flow radial flow radial flow with bottom 127 Type with bottom with bottom with bottom with bottom suction and suction and suction and suction and suction and horizontal horizontal horizontal horizontal horizontal discharge discharge discharge discharge discharge 128 Design Pressure, psig 2485 2485 2485 2485 2485 129 Design Temperature, °F 650 650 650 650 650 Operating Pressure, 130 2235 2235 2235 2235 2235 Nominal, psig 131 Suction Temperature, °F 539 539 556 551.5 546.5

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 21 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report 132 Design Capacity, gpm 88,500 87,500 80,000 80,000 88,500 133 Design Head, ft. 277 277 252 259 261 134 Hydrostatic Test Pressure (cold), psig 3107 3107 3110 3110 3110 AC Induction AC Induction AC Induction AC Induction AC Induction 135 Motor Type single speed single speed single speed single speed single speed air cooled air cooled air cooled 136 Motor Rating (nameplate) 6000 HP 6000 HP 6000 HP 6000 HP 6000 HP

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 1.2-1 Page: 22 of 22 UPDATED FINAL SAFETY ANALYSIS REPORT Comparison Of Design Parameters**

    • This table is retained for historical purpose only. It compares original Cook Plant parameters to other similar nuclear plants Donald C. Cook Zion Station Indian Point Point Beach H. B.

Reference Thermal And Hydraulic Design Nuclear Plant Units 1 & 2 #2 Units 1 & 2 Robinson #2 Line No. Parameters Units 1 & 2 Final Report Final Report Final Report Final Report Final Report Principal Design Parameters Of The Reactor Coolant Piping See Table See Table 137 Material Austenitic SS Austenitic SS Austenitic SS 4.2-1 4.2-1 138 Hot Leg - I.D., in. 29 29 29 29 29 139 Cold Leg - I.D., in. 27-1/2 27-1/2 27-1/2 27-1/2 27-1/2 Between Pump and Steam generator -

140 31 31 31 31 31 I.D., in.

137 Design Pressure, psig 2485 2485 2485 2485 2485

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 21 REFERENCES I. Emergency Core Cooling System (ECCS)

1. WCAP-7498-L, R. M. Hunt (editor), "Safety Related Research and Development for Westinghouse Pressurized Water Reactors - Program Summaries, Spring, 1970", May, 1970.
2. WCAP-7396-L, R. M. Hunt (editor), "Safety Related Research and Development for Westinghouse Pressurized Water Reactors - A Program Outline, Fall, 1969", November, 1969.
3. WCAP-7304-L, R. M. Hunt (editor), "Safety Related Research and Development for Westinghouse Pressurized Water Reactors - A Program Outline, Spring, 1969", April, 1969.
4. WCAP-7379-L, Vol. I and Vol. II Topical Report, Performance on Zircaloy Clad Fuel Rods During a LOCA, J. B. Roll.
5. WCAP-7422, "Westinghouse PWR Core Behavior Following LOCA", January 1970.
6. WCAP-7435, "PWR FLECHT Group 1 Test Report", January 1970; J. O. Cermak, et al.
7. WCAP-7495-L, Vol. I and Vol. II, Topical Report, Performance of Zircaloy Clad Fuel Rods During a Simulated Loss of Coolant Accident, Multi-Rod Tests, R. Schrieber, et al. (WNES Proprietary)
8. WCAP-7437, "LOCTA-R2 program - Loss of Coolant Transient Analysis", January 1970, W. A. Bazella, et al.
9. WCAP-7503, "Design Pipe Break for Westinghouse PWR Coolant System",. October 1970, R. Salvatori, et al.
10. WCAP-7401, "Comparison Between BLODWN-2 Results and Test Data"., November 1969, S. Fabic
11. WCAP-7544, "PWR FLECHT Group II Test Report", September 1970, F. F. Cadek, et al.
12. WCAP-7665, "PWR FLECHT Final Report". April 1971, F.F. Cadek, et al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 21 I. Emergency Core Cooling System (ECCS)

13. WCAP-7805, "Performance of Zircaloy Rods During Simulated LOCA -Single Rod Test". December 1971
14. WCAP-7808, "Performance of Zircaloy Rods During Simulated LOCA -Multirod Tests." December 1971, R. E. Schreiber, et al.
15. WCAP-7950, "Fuel Assembly Safety Analysis for Combined Seismic and LOCA Loads" July 1972, L. T. Gesinski
16. WCAP-8170, Calculational Model for Core Reflooding After a LOCA - WREFLOOD Code June 1974, G. Collier, et al.
17. WCAP-8200, WFLASH-A- A Fortran IV Program for Simulation of Transients in a Multi-loop PWR,. June 1974, V. J. Esposito, et al.
18. WCAP-8305, LOCTA IV program for Loss of Coolant Transient Analysis., June 1974, F. M. Bordelon, et al.
19. WCAP-8306, SATAN VI Program June 1974, F. M. Bordelon, et al.
20. WCAP-8339, Westinghouse ECCS Evaluation Model, June 1974, F. M. Bordelon, et al.
21. WCAP-8341, Westinghouse ECCS Evaluation Model Sensitivity Studies, July 1974, Safeguards Engineering Department.
22. WCAP-8342, Westinghouse ECCS Evaluation Model Sensitivity Studies, July, 1974
23. WCAP-8356, Westinghouse ECCS-Plant Sensitivity Studies, 1974, R. Salvatori
24. WCAP-8410, FLECHT - Phase B System Design Description W. F. Cleary et al.
25. WCAP-8431, PWR FLECHT - Phase B1 Data Report; December, 1974; J. P. Waring, et al.
26. WCAP-8471, Westinghouse ECCS Evaluation Model: Supplementary Information; January 1975, F. M. Bordelon, et al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 21 I. Emergency Core Cooling System (ECCS)

27. WCAP-8472, Westinghouse ECCS Evaluation Model: Supplementary Information, January 1975, F. M. Bordelon, et al.
28. WCAP-8566-A, Westinghouse ECCS Four Loop Plant (17x17) Sensitivity Studies, 1975, W.J. Johnson et al.
29. WCAP-8622, Westinghouse ECCS Evaluation Model, November 1975, Nuclear Safety Department.
30. WCAP-8651, FLECHT Low Flooding Rate Cosine Test Series Data Report, December 1975, E. R. Rosal, et al.
31. WCAP-8838, FLECHT Low Flooding Rate Test Series Evaluation Report, March 1977, G. P. Lilly, et al.
32. WCAP-8971-A, Westinghouse Core Cooling System Small Break, October 1975 Model, 1977, R. J. Skwarek. Et al.
33. WCAP-9005, "Post DNB Heat Transfer During Blowdown", September 1975, R. F.

Farman, et al.

34. WCAP-9183, "PWR FLECHT Skewed Profile Low Flooding Rate Test Series Evaluation Report", November 1977, G. P. Lilly, et al.
35. WCAP-9220, "Westinghouse ECCS Evaluation Model", February 1978, Nuclear Safety Department.
36. WCAP-9221-P-A, "Westinghouse ECCS Evaluation Model", 1981 Version, E. P. Rake
37. WCAP-9279, "Westinghouse ECCS Evaluation Model", March 1978, W. T. Bogard, et al.
38. WCAP-9584, "Analysis of Delayed RCP Trip During Small LOCA for Westinghouse NSSS.", August 1979, Nuclear Safety Department.
39. WCAP-9587, "Study of Two Phase Natural Circulation Following Small LOCA Using the NOTRUMP Code", August 1979, K. Kesavan, et al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 21 I. Emergency Core Cooling System (ECCS)

40. WCAP-9600, "Small Break Accidents for Westinghouse NSSS", June 1979, Nuclear Safety Department.
41. WCAP-9628, "Asymmetric LOCA Loads Evaluation Phase B, Class 2", November 1979.
42. WCAP-9658, "PWR FLECHT SEASET 21-Rod Bundle Flow Blockage Task".
43. WCAP-9662, "Asymmetric LOCA Load Evaluation, Phase B, Class 3" January 1980.
44. WCAP-9744, "Loss of Feedwater Induced Loss of Coolant Accident Report"
45. WCAP-9748, "Asymmetric LOCA Loads Evaluation, Phase C, Class 2"; June 1980.
46. WCAP-9753, "Inadequate Core Cooling Studies of Scenarios With Feedwater Available Using the NOTRUMP Code".
47. WCAP-9765, "Documentation of the Westinghouse Core Uncovery Tests is the Small Break Evaluation Model"., July 1980, Nuclear Safety Department.
48. WCAP-9891, "PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Data Evaluation and Analysis", November 1981.
49. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," N. Lee, et. al., August 1985.
50. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," P. E. Meyer, August 1985.
51. WCAP-11145, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," S.D. Rupprecht, et. al.
52. XN-73-1 Rev 2, "GASPRX Calculation Procedure for Internal Gas Pressure Due to Fission Gas Release", March 1974, K. R. Merckx
53. XN-73-25 "GAPEXX a Computer Program for Predicting Pelltt-To-Cladding Heat Transfer Coefficients". August 1973, K. G. Galbraith

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 5 of 21 I. Emergency Core Cooling System (ECCS)

54. XN-74-27 Rev 2, "Bulger a Computer Code to Determine the Deformation and the Onset of Bulging of Zircaloy Fuel Rod Cladding", December 1974, K. R. Merckx
55. XN-75-1, "Computational Simulations of Pressurized Water Reactor Full Length Emergency Cooling Heat Transfer Experiments" March 1975, F. Lang
56. XN-75-6, "Flow Blockage Model for LOCA Analyses", R. E. Collingham, et al.
57. XN-75-19, "Carryout Rate Fraction Correlation for Pressurized Water Reactors", March 1975, F. Lang
58. XN-75-19 Supp 1, "Statistical Evaluation of the Carryout Rate Fraction", June 1975, F.

Lang et al.

59. XN-75-41 Vol. I, "Exxon Nuclear Company WREM-BASED Generic PWR ECCS Evaluation Model", July 1975, L. Steves
60. XN-75-41 Vol. III Rev. 2, "Small Break Model", August 1975, J. Kahn
61. XN-75-43, "Core Physics Methods and Data used as Input to LOCA Analysis", F. B. S.

Kogen

62. XN-76-8, "RODEX: Fuel Rod Thermal-Mechanical Response Evaluation Code",

February 1977, K. Merckx

63. XN-76-27, "Exxon Nuclear Company WREM-BASED Generic PWR ECCS Evaluation Model Update ENC-WREM II", July 1976, L. Worley et al.
64. XN-76-36, "Exxon Nuclear Co. WREM-BASED Generic PWR ECCS Evaluation Model (ENC-WREM - II) 4 Loop PWR With Ice Condenser Large Break Example Problem", August 1976, L. H. Steves
65. XN-76-47 (P), "Combined Seismic - LOCA Mechanical Evaluation for Exxon Nuclear 15 x 15 Reload Fuel for Westinghouse PWR'S", April 1977. C. A. Brown
66. XN-76-51, "D. C. Cook Unit 1 LOCA Analyses Using the ENC WREM-BASED PWR ECCS Evaluation Model (ENC-WREM-II)". October 1976. L. H. Steves

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 6 of 21 I. Emergency Core Cooling System (ECCS)

67. XN-78-30, "Exxon Nuclear Company WREM-BASED Generic PWR ECCS Evaluation Model Update ENC WREM-IIA", August 1978, S. E. Jensen et. al.

68 XN-NF-82-20(P), Revision 1, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates", August, 1982, W. V. Kyser.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 7 of 21 II. Ice Condenser

69. WCAP-2951, "Ice Condenser Reactor Containment", June 1966, S. J. Weems, et al.
70. WCAP-7040, "Ice Condenser Reactor Containment", April 1967, S. J. Weems, et al.
71. WCAP-7079, "Preliminary Design and Evaluation of the Ice Condenser Reactor Containment and Associated Engineered Safeguards", July 1967. S. J. Weems (MPR),

W. L. Bottinger, S. N. Ehrenpreis, F. P. Green, N. P. Grimm, W. G. Lyman, and J.

Stevenson.

72. WCAP-7079 Supplement 1, "Supplementary Information to WCAP-7079, Preliminary Design and Evaluation of the Ice Condenser Reactor Containment and Associated Engineered Safeguards", September 1967. S. J. Weems (MPR), S. N. Ehrenpreis, N. P.

Grimm, and W. G. Lyman

73. WCAP-7079 Supplement 2, "Supplementary Information to WCAP-7079, Preliminary Design and Evaluation of the Ice Condenser Reactor Containment and Associated Engineered Safeguards", December 1967. W. J. Weems (MPR), N. P. Grimm, and W.

G. Lyman

74. WCAP-7183 "Design and Performance Evaluation of the Ice Condenser Reactor Containment System for the Donald C. Cook Nuclear Plant", March 1968. W. J.

McCurdy (MPR), S. J. Weems (MPR, W. L. Boettinger, F. M. Bordelon, J. W. Dorrycott, N. P. Grimm, A. J. F. Iredale, W. G. Lyman, R. R. Oft, and J. R. van Seuren

75. WCAP-7183 Supplement 1 "Supplementary Information to WCAP-7183, Design and Performance Evaluation of the Ice Condenser Reactor Containment System for the Donald C. Cook Nuclear Plant", July 1968. W. J. MCCurdy (MPR), S. J. Weems (MPR),

W. L. Boettinger, J. W. Dorrycott, N. P. Grimm, A. J. F. Iredale, and W. G. Lyman

76. WCAP-7183-L Supplement 2 "Topical Report - Supplementary Information to WCAP-7183, Design and Performance Evaluation of the Ice Condenser Reactor Containment System", August 1969, Proprietary. H. W. Mc Curdy (MPR), S. J. Weems, (MPR), F.

M. Bordelon, N. P. Grimm, E. J. Kilpela, W. G. Lyman

77. Atomic Safety and Licensing Board Hearing Record, Docket Number 50-327 and 50-328, Sequoyah Nuclear Plant Units 1 and 2, Chattanooga, Tennessee, April 23, 1970.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 8 of 21 II. Ice Condenser

78. WCAP-7426, "Iodine Removal in the Ice Condenser System", April 1970. D. D.

Malinowski

79. WCAP-7611, "Design and Performance Evaluation of the Ice Condenser Inlet Door".
80. WCAP-8077, "Ice Condenser Containment Pressure Transient Analysis Method",

March 1973.

81. WCAP-8110, "Test Plans and Results for Ice Condenser System".
82. WCAP-8355, "Long Term Ice Condenser Containment LOTIC Code Supplement 1",

Hoiech, T. and Raymond, M., July 1974.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 9 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters

83. WCAP-6069, "Burnup Physics of Heterogeneous Reactor Lattices", June 1965, C. A.

Poncelet

84. WCAP-6076, "Effects of Fuel Burnup on Reactivity and Reactivity Coefficients in Yankee Core I", October 1965, C. G. Poncelet
85. WCAP-6065, "Melting Point of Irradiated Uranium Dioxide", February 1965, J. A.

Christensen, et al.

86. WCAP-7208, "Power Distribution Control of Westinghouse PWR's", September 1968, R. F. Barry, et al.
87. WCAP-7308, Evaluation of Nuclear Hot Channel Factor Uncertainties, April 1969, F.

L. Langford, et al.

88. WCAP-7407, Power Maldistribution Investigations January 1970, R. F. Barry et al.
89. WCAP-7411, Rod Bundle Axial Nonuniform Heat Flux DNB Tests, May 1970, E. R.

Rosal

90. WCAP-7703, A Review of Fuel Rod Integrity at the Ginna Reactor.
91. WCAP-7911, Core Physics Characteristics of the D. C. Cook Nuclear Power Plant Unit 1, Cycle 1., September 1973, T. R. Freeman
92. WCAP-7912, Topical Report on Power Peaking Factors, March 1972, A. F.

McFarlane, et al.

93. WCAP-8296, Effect of 17x17 Fuel Assembly Geometry on DNB, March 1974, K. W.

Hill, et al.

94. WCAP-8385, Topical Report on Power Distribution Control and Load Following Procedure, September 1974, T. Morita, et al.
95. WCAP-8688, Summary Report of the Startup Nuclear Test Results for D. C. Cook Unit 1, Cycle 1, December 1975, J. F. Nelson et al.
96. WCAP-8567, Improved Thermal Design Procedure, July 1975, H. Chelemer, et al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 10 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters

97. WCAP-8785, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations, October 1976, J. V. Miller (editor)
98. WCAP-9000, Nuclear Design of Westinghouse PWRs with B.P. Rods, June 1969, R.

F. Barry, et al.

99. WCAP-9002, Use of Internally Pressurized Fuel Rods in Westinghouse PWRs, February 1969, H. M. Ferrari, et al.

100. WCAP-9118, The Core Physics Characteristics of the D. C. Cook Unit 2 Nuclear Power Plant, Cycle 1, June 1977, A. L. Casadei, et al.

101. WCAP-9436, Summary Report of the Startup Nuclear Test Results for D. C. Cook Unit 2, Cycle 1,.

102. WCAP-9556, Nuclear Design and Core Management of the D. C. Cook Nuclear Plant Unit 2, Cycle 2, August 1979, B. F. Cooney, et al.

103. WCAP-9828, Nuclear Design and Core Management of the D. C. Cook Nuclear Plant Unit 2, Cycle 3, December 1980, J. R. Secker, et al.

104. WCAP-3680-20, Xenon-Induced Spatial Instabilities in Large Pressurized Water Reactors, March 1968, C. G. Poncelet, et al.

105. WCAP-3269-40, An Experimental Evaluation of the Power Coefficient in Slightly Enriched PWR Cores, April 1965, W.T. Sha.

106. WCAP-8185, Reference Core Report 17 x 17 June 1, April 1974 107. WCAP-8288, Safety Analysis of the 17x17 Fuel Assembly for a Combined Seismic and Loss-of-Coolant Accident, December 1973 108. WCAP-8279, Hydraulic Flow Tests of the 17x17 Fuel Assembly, February 1974 109. WCAP-8299, The Effect of 17x17 Geometry on Interchannel Thermal Mixing, March 1974 110. WCAP-8449, 17x17 Drive Line Components Test - Phase I, II, III, D-Loop-Drop and Deflection, December 1974

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 11 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 111. WCAP-8692, Revision 1, Fuel Rod Bowing, July 1979.

112. WCAP-9500, Reference Core Report - 17x17 Optimized Fuel Assembly.

113. WCAP-9273, Westinghouse Reload Safety Evaluation Methodology, March 1978, Berdelon, F. M. et al.

114. WCAP-7956, THINC IV - An Improved Program for Thermal-Hydraulic Analysis of Rod Bundle Cores, June 1973, Chelemer, H., et al.

115. WCAP-8054, Application of THINC IV Program to PWR Design, September 1973, Hodreiter, L. E., et al.

116. WCAP-8762, New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids, July 1976, Mothey, F. E., et al.

117. WCAP-8971-A, Westinghouse Core Cooling System Small Break, October 1975 Model, 1977, Skwarek, R. J. et al.

118. WCAP-8746, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, March 1977, Ellenberger, S. L., et al.

119. WCAP-7908, FACTRAN - AFORTRAN IV Code for Thermal Transients in a UO2 Fuel Rod, June 1972, Hargrove, H. G.

120. WCAP-7907, LOFTRAN Code Description, October 1972, Burnett, T. W. T., et al.

121. WCAP-3269-26, LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094, September 1963, Barry, R. F.

122. WCAP-7758-A, The TURTLE 24.0 Diffusion Depletion Code, February 1975, Barry, R. F., and Altmore, S.

123. WCAP-8028-A, TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code, January 1975, Risher, D. H. Jr., and Barry, R. F.

124. WCAP-9227, Reactor Core Response to Excessive Secondary Steam Releases, January 1978, Hollingsworth, S. D. and Wood, D. C.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 12 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 125. WCAP-7588, An Evaluation of the Rod Ejection Accident of Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods, Revision 1A, Risher, D.

H., Jr.

126. WCAP-7413, Use of Burnable Poison Rods in Westinghouse Reactors, October 1967, Wood P.M., et al.

127. WCAP-7811, Power Distribution Control in Westinghouse Pressurized Water Reactors, December 1971, Moore, J. S.

128. WCAP-2759, The Revised LEOPARD Code - A Spectrum Depending Non-Spatial Depletion Program, March 1965, Barry, R. F.

129. WCAP-7267-L, Core Power Capabilities in Westinghouse PWRs, October 1969, McFarlane, A. F.

130. WCAP-10376, Core Physics Characteristics of the Donald C. Cook Station Nuclear Plant (Unit 1 Cycle 8), July 1983, Hubbard, B. Y., et al.

131. WCAP-10021-P-A, Westinghouse Wet Annular Burnable Absorber Evaluation Report, Revision 1, October 1983, Skaritka, J., et al.

132. WCAP-9719, Properties of Fuel and Core Component Materials, Revision 1, July 1978, including Appendix B, Al2O3 B4C Pellets, October 1980, and Revisions, September 1982, Beaumont, M. D., et al.

133. WCAP-9402-A, Verification Testing and Analysis of the 17x17 Optimized Fuel Assembly, August 1981, Davidson, S. L. and Iorii, J. A.

134. WCAP-8963, Safety Analysis for the Revised Fuel Rod Internal Pressure Design Bases, November 1976, Risher, D. H.

135. WCAP-8381, Revised Clad Flattening Model, July 1974, George, R.A., et al.

136. WCAP-8720, Addendum 1, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations Application for Transient Analyses, September 1979, Leech, W. J.

137. WCAP-7048, The PANDA Code, April 1967, Barry, R. F., et al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 13 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 138. WCAP-6086, Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analysis of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium through Curium, August 1969, Nodvick, R. J., et al.

139. WCAP-7015, Subchannel Thermal Analysis of Rod Bundle Cores, Revision 1, January 1969, Chelemer, H., et al.

140. WCAP-7959-A, Effect of Axial Spacing on Interchannel Thermal Mixing with R Mixing Vane Grid, January 1975, Cadek, F. F., et al.

141. WCAP-8720-Addenda 2, Revised PAD Code Thermal Safety Model, October 1982, Leech, W. J., et al.

142. XN-74-5 Rev 1, Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR), May 1975, J. D. Kahn 143. XN-NF-82-21(P), Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, March 1982, T. R. Lindquist.

144. XN-NF-621(P), Revision 1, Exxon Nuclear DNB Correlation for PWR Fuel Designs, April 1982, R. B. Macduff.

145. XN-74-15, H. B. Robinson Bunkle Bowing Study, March 1974, K. R. Merck 146. XN-74-21 Rev 2 XTHETA: Multi-Rode Heatup Code for Single Channel Transient Analysis April 1975, F. Lang et al.

147. XN-74-27 SUPP 2 Rev 2, BULGEX an XTHETA Subroutine to Calculate Mechanical Cladding Response During a PWR Loss-Of-Coolant Accident, January 1975, T. A.

Bjornard et al.

148. XN-74-27 (A) Rev 2 BULGEX: a Computer Code to Determine the Deformation And the Onset of Bulging of Zircaloy Fuel Rod Cladding (Applicable for Loss of Coolant Accident Conditions), December 1974, K. R. Merckx 149. XN-74-44, Single Phase Hydraulic Performance of Westinghouse and Exxon Nuclear H. B. Robinson Fuel Assemblies, October 1974, J. Yates 150. XN-74-56, Analysis of the Ginna RCC Drop Test, December 1974, L. C. Worley

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 14 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 151. XN-75-10, PWR Upper Tie Plate Locking Mechanism Tensile Test for Flared Guide Tubes, March 1975, K. L. Ford 152. XN-75-17, Fretting Corrosion Analysis of Exxon Nuclear H. B. Robinson # 2 Fuel Design, March 1975, W. C. Gallaugher 153. XN-75-21, XCOBRA-IIIC A Computer Code to determine the Distribution of Coolant During Steady-State And Transient Core Operation, K. P. Galbraith 154. XN-75-27, Exxon Nuclear Neutronic Design Methods for Pressurized Water Reactors, June 1975, F. B. Skogen 155. XN-75-27 SUPP 2, Exxon Nuclear Neutronic Design Methods for Pressurized Water Reactors, September 1976, F. B. Skogen 156. XN-75-27 (P) SUPP 3, Exxon Nuclear Neutronics Design Methods for PWR, November 1980. M. R. Killgore 157. XN-75-32, "Computation Procedures for Evaluating Fuel Rod Bowing (AX1BOW)",

April 1975, K. R. Merckx 158. XN-75-39, "Generic Fuel Design for 15x15 Reload Assemblies for Westinghouse Plants", September 1975. W. C. Gallaugher 159. XN-75-42, "PWR Thermal-Hydraulic Hot Channel Calculations", July 1975, T. Pattern et. al.

160. XN-75-48, "Definition and Justification of Exxon Nuclear Company DNB Correlation for PWR's", October 1975, K. Garbraith et al.

161. XN-75-52, "Lateral Core Seismic Analysis for Exxon Nuclear 15x15 Reload Fuel for Westinghouse PWR'S", October 1975, C. A. Brown 162. XN-76-7 Rev 1, "Evaluation of Zircaloy Fuel Rod Autoclaving", June 1976, L. Van Swam 163. XN-76-25, "Donald C. Cook Unit 1 Cycle 2 Reload Fuel Licensing Data Submittal",

July 1976. F. Skogen

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 15 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 164. XN-76-35 SUPP 1, "Assumptions Used in the Plant Transient Analysis for the D. C.

Cook Unit 1 Nuclear Power Plant". November 1976, J. Kahn 165. XN-76-40, "Exxon Nuclear Power Distribution Control for PWR'S September 1976, J.

Hom et al.

166. XN-76-58, "D. C. Cook Unit 1 Reference Cycle 2 Design, November 1976, R. J.

Burnside 167. XN-NF-77-3, "D. C. Cook Unit Cycle 2 Start-up Predictions and Nuclear Data for Operations", February 1977, R. J. Burnside 168. XN-NF-77-10, "Neutronics Reanalysis of D. C. Cook Unit 1, Cycle 2" May 1977, F. B.

Skogen 169. XN-NF-77-36, " D. C. Cook Cycle 3 Fuel Management Analysis". August 1977. R. J.

Burnside 170. XN-NF-78-4, "Procedure for Monitoring Exposure Dependent FQ Limit in Westinghouse PWR'S", January 1978, R. B. Stoat 171. XN-NF-78-9 "D. C. Cook Cycle 3 Startup Predictions and Nuclear Data for Operations",

March 1978, R. J. Burnside 172. XN-NF-78-44, " A Generic Analysis of the Control Rod Ejection", January 1979, R. J.

Burnside et al.

173. XN-NF-79-6 (P) Rev 1, "Exxon Nuclear Analysis of Power Distribution Measurement Uncertainty for Westinghouse PWR'S. July 1979, J. S. Holm 174. XN-NF-79-10 "D. C. Cook Unit 1 Cycle 4 Safety Analysis Report", February 1979, M.

R. Killgore 175. XN-NF-79-17 "Plant Transient Reanalysis for the D. C. Cook Unit 1 Nuclear Power Plant", June 1979, R. H. Kelley 176. XN-NF-79-46, "D. C. Cook Unit 1 Cycle 4 Startup and Operations Report", June 1979, R. L. Feunbacher

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 16 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 177. XN-NF-79-76, "Examination of Exxon Nuclear Company Fuel Irradiated at D. C. Cook Unit No. 1, April 78, 79", December 1979, J. R. Tandy 178. XN-NF-80-10, "D. C. Cook Unit 1 Cycle 5 Safety Analysis Report", March 1980. M.

R. Killgore 179. XN-NF-81-26, "Single Phase Hydraulic Performance of Westinghouse 17x17 Fuel Assembly", April 1980. J. Yates 180. XN-NF-81-60, " D. C. Cook Unit 2 Primary Design Parameters for ECCS&RTS Analysis" August 1981, S. E. Jensen 181. XN-NF-81-64, "D. C. Cook Cycle 6 Fuel Management Analysis", August 1981, M. R.

Killgore 182. XN-NF-81-73 "Turbulent Mixing in Rod Bandles" October 1981, R. B. Macduff 183. XN-NF-81-90, "D. C. Cook Unit 1, Cycle 7 Fuel Management Analysis", November 1981, M. E. Finch et. al.

184. XN-NF-82-37, Revision 1, "D. C. Cook Unit 2, Cycle 4 Safety Analysis Report",

December 1982, P. D. Wimpy, et. al.

185. XN-NF-82-36(P), "D. C. Cook Unit 2, Cycle 4 Fuel Management Report", April 1982, P. D. Wimpy, et. al.

186. XN-NF-82-74(P), Revision 1, "D. C. Cook Unit 2, Cycle 4 Startup and Operations Report", February 1983, P. D. Wimpy.

187. XN-CC-21(A) Rev 2, "XPOSE the Exxon Nuclear Revised Leopard" April 1975, F. A.

Skogen 188. XN-CC-26 Rev 1 "XPIN the Exxon Nuclear Revised Hambur User Manual", December 1975, W. W. Parath et al.

189. XN-CC-28 Rev 4 XTG: A Two-Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing And Users Manual (PWR Version), July 1976, R. B. Stout

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 17 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 190. XN-CC-32 XTRAN-PWR: a Computer Code for the Calculation of Rapid Transients in Pressurized Water Reactors with Moderator and Fuel Temperature Feedback, September 1975, J. R. Morgan 191. XN-CC-33 (A) Rev 1, Rev of XN-73-34 (Rev 2), HUXY: a Generalized Multiord Heatup Code with 10CFR50 Appendix K Heatup Options Users Manual. November 1975, L. S. Steves et. al.

192. XN-CC-39, ICECON: a Computer Program used to Calculate Containment Back Pressure for LOCA Analysis (Including ICE Condenser Plants), July 1976 Energy Inc.

193. XN-CC-41, RODEX: Code Manual for Fuel Rod Thermomechanical Evaluation, August 1977, K. R. Merckx 194 BART-A1: A computer code for the Best Estimate Analysis of Reflood Transients, WCAP-9561, January 1980, G. Collier, et. al.

195. XN-NF-83-61, D. C. Cook Unit 1 LOCA-ECCS Analysis for Extended Exposure, August 1983, T. Tahvile, et. al.

196. XN-NF-84-21(P), Revision 1: Donald C. Cook Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break LOCA/ECCS Analysis, Revision 1, May 1984, M. J.

Ades, et. al.

197. XN-NF-84-21, Revision 2, Supplement 1: Donald C. Cook Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break LOCA/ECCS Analysis: K(Z) curve, April 1985, T. Tahvile, et. al.

198. XN-NF-84-21, Revision 2, Supplement 2: Donald C. Cook Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break LOCA/ECCS Analysis: K(Z) curve, April 1985, T. Tahvile, et. al.

199. XN-NF-85-20(P): Modification of the EXEM/PWR FLECHT Based Reflood Quench and Heat Transfer Correlations for D. C. Cook Unit 2, April 1985, B. Vaishnavi, et. al.

200. XN-NF-84-25(P), Mechanical Design Report Supplement for D. C. Cook Unit 1 Extended Burnup Fuel Assemblies, April 1984, N. L. Garner, et. al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 18 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 201. NUREG/CR 3988, MARCH-2, Meltdown Accident Response Characteristic Code Description and Users Manual, BMI-2115, Battelle Columbus Laboratories, September 1984, R. O.Wooten, et. al.

202. NUREG-75/057, TOODEE2: A Two-Dimensional Time dependent Fuel Element Thermal Analysis Program, May 1975, G. N. Lauben.

203. XN-76-51, Supplement 1, Flow Blockage and Exposure Sensitivity Study for D. C.

Cook Unit 1 Reload Fuel Using ENC WREM-II Model, January 1977, K. P. Galbraith et. al.

204. XN-76-51, Supplement 2, Flow Blockage and Exposure Sensitivity Study for ENC D.

C. Cook Unit 1 Reload Fuel Using ENC WREM-2 Model, January 1978, G. C. Cooke.

205. XN-76-51, Supplement 3, Flow Blockage and Exposure Sensitivity Study for ENC D.

C. Cook Unit 1 Reload Fuel Using ENC WREM-2 Model, March 1978, R. E.

Collingham et. al.

206. XN-NF-78-30, Amendment 1, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA: Response to NRC Request for Additional Information, February 1979, S. E. Jensen et. al.

207. XN-NF-81-07, LOCA ECCS Reanalysis for D. C. Cook Unit 1 Using the ENC WREM-IIA PWR ECCS Evaluation Model, February 1981, S. E. Jensen et. al.

208. XN-NF-81-58(P), Revision 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, January, 1983, K. R. Merckx, Ed.

209. XN-NF-82-07(P), Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, March 1982, W. V. Kayser.

210. XN-NF-82-20(P), Supplement 2, Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates: Large Break Example Problem for 4-Loop PWR with Ice Condenser, March 1982, T. Tahvili.

211. XN-NF-86-16(P), Revision 1, and all supplements, PWR 17 x 17 Fuel Cooling Test Program, Reflood Quench, Carryover, and Heat Transfer Correlations, Exxon Nuclear Company, Inc., Richland, WA 99352, January 1986.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 19 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 212. XN-NF-85-16(P), Volume 1, and all Supplements, PWR 17 x 17 Fuel Cooling Test Program, Sensitivity Studies, Exxon Nuclear Company, Inc., Richland, WA, 99352, January 1986.

213. XN-NF-85-28(P), Rev. 1, Supp. 1, D. C. Cook Unit 2, Cycle 6 Safety Analysis Report:

Disposition of Standard Review Plan Chapter 15 Events, Exxon Nuclear Company, Richland, WA 99352, October 1986.

214. XN-NF-85-64(P), Plant Transient Analysis for D. C. Cook Unit 2 with 10% Steam Generator Tube Plugging, Exxon Nuclear Company, Inc., Richland, WA 99352, November 1985.

215. XN-NF-85-68(P), Rev. 1, Donald C. Cook Unit 2 Limiting Break LOCA/ECCS Analysis, 10% Steam Generator Tube Plugging, and K(Z) Curve, Exxon Nuclear Company, Inc., Richland, WA 99352, August 1986.

216. XN-NF-87-31(P), Steam Line Break Analysis for D. C. Cook Unit 2, Exxon Nuclear Plant, Inc., Richland, WA 99352, May 1987.

217. ANF-87-91(P), D. C. Cook 2 Debris-Resistant Lower Tie Plate Pressure Drop Test Report, June 1987, J. Yates.

218. WCAP-11902, Reduced Temperature and Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 Licensing Report, October 1988, D. L. Cecchett and D. B.

Augustine.

219. WCAP-10444-P-A, Reference Core Report VANTAGE 5 Fuel Assembly, September 1985, S. L. Davidson and W. R. Kramer, ed.

220. WCAP-11596-P-A, Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, June 1988, T. Q. Nguyen, et. al.

221. WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989, A. J. Friedland and S. Ray.

222. WCAP-7956-P-A, THINC-IV - An Improved Program for Thermal Hydraulic Analysis of Rod Bundle Cores, February 1989, H. Chelemer et. al.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 20 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 223. WCAP-8054-P-A, Application of the THINC-IV Program to PWR Design, February 1989, L. E. Hochreiter and H. Chelemer.

224. WCAP-10965-P-A, ANC: Westinghouse Advanced Nodal Computer Code, September 1986: S. L. Davidson (ed) et. al.

225. WCAP-10125-P-A, Extended Burnup Evaluation of Westinghouse Fuel, December 1985, S. L. Davidson (ed) et. al.

226. WCAP-10851-P-A, "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," August 1988, R. A. Weiner et. al.

227. WCAP-12568 (Proprietary), WCAP-12569 (Non Proprietary), "Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology for American Electric Power D. C. Cook Unit 1 Nuclear Power Station," Revision 1, August 1993, C.F. Ciocca.

228. WCAP-12078 (Proprietary), "Input and Output Parameters for the Accident Analyses Performed for Reduced Temperature and Pressure Operation for Donald C. Cook Nuclear Plant Unit 1," December 1988.

229. WCAP-12901 (Proprietary), "Input and Output Parameters for the Accident Analyses Performed for Vantage 5 Fuel Transition for Donald C. Cook Nuclear Plant Unit 2,"

May 1991.

230. 77-5002104-01 (MD-1-SGRP-005-N) Replacement Steam Generator Report for AEP Donald C. Cook Plant Unit One, prepared by Framatome Technologies, Inc.

231. 222-7803-PR-01 (MD-1-SGRP-040-N) Thermal-Hydraulics Performance Report for Unit 1 replacement steam generator, prepared by Framatome Technologies Inc.

232. 222-7803-PR-02 (MD-1-SGRP-041-N) Three Dimensional Thermal Hydraulics Analysis Report for Unit 1 replacement steam generator, prepared by Framatome Technologies Inc.

233. 222-7803-PR-03 (MD-1-SGRP-061-N) RELAP5 Thermal-Hydraulics Simulations for Unit 1 replacement steam generator, prepared by Framatome Technologies Inc.

UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 30.0 D. C. COOK NUCLEAR PLANT Table: 1.6-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 21 of 21 III. Nuclear, Thermal-Hydraulic And Mechanical Design Parameters 234. 222-7803-FIV-01 (MD-1-SGRP-038-N) Flow-Induced Vibration Analysis Report for Unit 1 replacement steam generator, prepared by Framatome Technologies Inc.

235. 222-7803-FIV-02 (MD-1-SGRP-039-N) Tube Wear Analysis Report for Unit 1 replacement steam generator, prepared by Framatome Technologies Inc.

236. WCAP 15302, "Donald C. Cook Nuclear Plant Units 1 and 2 - Modifications to the Containment Systems Westinghouse Safety Evaluation (SECL 99-076, Revision 3),"

September 1999.

237. WCAP 14285, "Donald C. Cook Nuclear Plant Unit 1 - Steam Generator Tube Plugging Program Licensing Report," Revision 1, May 1995.

238. WCAP 14286, "Donald C. Cook Nuclear Plant Unit 1 - Steam Generator Tube Plugging Program Engineering Report," December 1995.

Retrieved from "https://kanterella.com/w/index.php?title=ML21125A596&oldid=3415444"