ML22340A180
Text
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
16.1 Table:
3.2.1-1 Page:
1 of 3 INITIAL CORE MECHANICAL DESIGN PARAMETERS 1 (ALL DIMENSIONS ARE FOR COLD CONDITIONS.)
Active Portion of the Core Equivalent Diameter, in.
132.7 Active Fuel Height, in.
Region 1 144.0 Region 2 143.4 Region 3 142.8 Length-to-Diameter Ratio 1.09 Total Cross-Section Area, Ft2 96.06 Fuel Assemblies Number 193 Rod Array 15 x 15 Rods per Assembly 204 1 Rod Pitch, in.
0.563 Overall Dimensions 8.426 x 8.426 Fuel Weight, (as UO2), pounds 219,900 Total Weight, pounds 279,000 Number of Grids per Assembly 7
Number of Guide Thimbles 20 Diameter of Guide Thimbles (upper part), in.
0.545 O.D. x 0.515 I.D.
Diameter of Guide Thimbles (lower part), in.
0.484 O.D. x 0.454 I.D.
1 Twenty-one rods are omitted: Twenty provide passage for control rods and one to certain in-core instrumentation.
UFSAR Revision 31.0
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
16.1 Table:
3.2.1-1 Page:
2 of 3 INITIAL CORE MECHANICAL DESIGN PARAMETERS 1 (ALL DIMENSIONS ARE FOR COLD CONDITIONS.)
Fuel Rods Number 39,372 Outside Diameter, in.
0.422 Diametral Gap, in.
Regions 1, 2 & 3 0.0075 Clad Thickness, in.
0.0243 Clad Material Zircaloy-4 Overall Length 149.7 Length of End Cap, overall, in.
0.688 Length of End Cap, inserted in rod, in.
0.250 Fuel Pellets Material UO2 sintered Density (% of Theoretical)
Regions 1, 2 & 3 94/95/95 Feed Enrichments w/o Region 1 2.25 Region 2 2.80 Region 3 3.30 Diameter, in.
Regions 1, 2 & 3 0.3659 Length, in.
0.6000 UFSAR Revision 31.0
IINNDDIIAANNAA AANNDD M MIICCHHIIGGAANN PPOOW WEERR DD.. CC.. CCOOOOKK NNUUCCLLEEAARR PPLLAANNTT UUPPDDAATTEEDD FFIINNAALL SSAAFFEETTYY AANNAALLYYSSIISS RREEPPOORRTT Revision:
16.1 Table:
3.2.1-1 Page:
3 of 3 INITIAL CORE MECHANICAL DESIGN PARAMETERS 1 (ALL DIMENSIONS ARE FOR COLD CONDITIONS.)
Rod Cluster Control Assemblies Neutron Absorber 5% Cd, 15% In. 80% Ag Cladding Material Type 304 SS - Cold Worked Clad Thickness, in.
0.019 Number of Clusters Full Length 53 Part Length 0
Number of Control Rods per Cluster 20 Length of Rod Control, in.
158.45 (overall) 150.57 (insertion length)
Length of Absorber Section, in.
142.00 (full length)
Core Structure Core Barrel, in.
I.D.
148.0 O.D.
152.5 Thermal Shield, in.
I.D.
158.5 O.D.
164.0 Burnable Poison Rods Number 1434 Material Borosilicate Glass Outside Diameter, in.
0.4395 Inner Tube, O.D., in 0.2365 Clad Material S.S.
Inner Tube Material S.S.
Boron Loading, (natural), gm/cm of glass rod.
.0603 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
27.0 Table: 3.3.1-1 Page:
1 of 3 Unit 1 Nuclear Design Data (These data are design values for cycle 1.
Updated design values starting with Cycle 8 operation are listed in Section 3.5.)
Structural Characteristics
- 1.
Fuel Weight (UO2), lbs 216,600
- 2.
Zircaloy Weight, lbs.
44,547
- 3.
Core Diameter, inches 132.7
- 4.
Core Height, inches 144 Reflector Thickness and Composition
- 5.
Top -
Water Plus Steel 10 in.
- 6.
Bottom -
Water Plus Steel 10 in.
- 7.
Side -
Water Plus Steel 15 in.
- 8.
H2O/U, (cold) Core 4.09
- 9.
Number of Fuel Assemblies 193
- 10. UO2 Rods per Assembly 204 Performance Characteristics
- 11. Heat Output, MWt (initial rating) 3,250
- 12. Heat Output, MWt (maximum calculated heat removal rating) 3,391
- 13. Fuel Burnup, MWD/MTU First Cycle First Cycle Enrichments, weight %
16,666
- 14. Region 1 2.25
- 15. Region 2 2.80
- 16. Region 3 3.30
- 17. Equilibrium Enrichment 2.90 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
27.0 Table: 3.3.1-1 Page:
2 of 3 Unit 1 Nuclear Design Data (These data are design values for cycle 1.
Updated design values starting with Cycle 8 operation are listed in Section 3.5.)
- 18. Nuclear Heat Flux Hot Channel Factor, N
Q F
2.711
- 19. Nuclear Enthalpy Rise Hot Channel Factor, H
FN
1.581 Control Characteristics Effective Multiplication (Beginning of Life)
With Burnable Poison Rods in; No Boron
- 20. Cold, No Power, Clean 1.183
- 21. Hot, No Power, Clean 1.154
- 22. Hot, Full Power, Clean 1.132
- 23. Hot, Full Power, Xe and Sm Equilibrium 1.092
- 24. Absorber Material 5% Cd; 15% In; 80% Ag
- 25. Full Length 53
- 26. Part Length 0
- 27. Number of Absorber Rods per RCC Assembly 20
- 28. Total Rod Worth, BOL, %: Boron Concentration for First Core Cycle Loading With Burnable Poison Rods (See Table 3.3.1-3)
- 29. Fuel Loading Shutdown; Rods in (k =.87) 2000 ppm Rods in (k =.90) 1714 ppm
- 30. Shutdown (k =.99) with Rods Inserted Clean, Cold 945 ppm
- 31. Shutdown (k =.99) with Rods Inserted, Clean, Hot 602 ppm
- 32. Shutdown (k =.99) with No Rods Inserted, Clean, Cold 1414 ppm 1 These data are design values for Cycle 1. The current Technical Specification limits are included in the Core Operating Limits Report for the current operating cycle.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
27.0 Table: 3.3.1-1 Page:
3 of 3 Unit 1 Nuclear Design Data (These data are design values for cycle 1.
Updated design values starting with Cycle 8 operation are listed in Section 3.5.)
- 33. Shutdown (k =.99) with No Rods Inserted, Clean, Hot To Maintain k = 1 at Hot Full Power, No Rods Inserted 1385 ppm
- 34. Clean 1152
- 35. Xenon 868 ppm
- 37. Shutdown, All But One Rod Inserted, Clean Cold (k =.99) 1031 ppm
- 38. Shutdown, All But One Rod Inserted, Clean Hot (k =.99) 734 ppm Burnable Poison Rods
- 39. Number and Material 1436 Borosilicate Glass
- 40. Worth Hot Full Power U 9.0%
- 41. Worth Cold U 7.0%
Kinetic Characteristics
- 42. Moderator Temperature Coefficient at Full Power (U/qF)
-0.3 x 10-4 to 3.2 x 10-4
- 43. Moderator Pressure Coefficient (U/psi)
+ 0.3 x 10-6 to 4.0 x 10-6 44 Moderator Density Coefficient, U/gm/cm3
-0.1 x 10-5 to 0.8 x 10-5
- 45. Doppler Coefficient (U/qF)
-1.0 x 10 1.7 x 10-5
- 46. Delayed Neutron Fraction, %
0.51 to 0.70
- 47. Prompt Neutron Lifetime, sec 1.4 x 10-5 to 2.0 x 10-5
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 16.4 Table:
3.3.1-2 Page:
1 of 1 REACTIVITY REQUIREMENTS FOR CONTROL RODS Requirements Per Cent Beginning of Life End of Life Control Power Defect 1.70 3.05 Rod Insertion Limit 0.70 0.50 Total Control 2.40 3.55 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 16.4 Table:
3.3.1-3 Page:
1 of 1 Unit 1 CALCULATED ROD WORTHS U Core Condition Rod Configuration Worth Less 10%
Design Reactivity Requirement Shutdown Margin BOL, HFP 53 rods in 8.15%
BOL, HZP 52 rods in; Highest Worth Rod Stuck Out 6.65%
5.98%
2.40%
3.58%
EOL, HFP 53 rods in 7.96%
(3rd Cycle)
EOL, HZP 52 rods in; Highest Worth Rod Stuck Out 6.16%
5.54%
3.55%
1.99%
BOL = Beginning of Life EOL = End of Life HFP = Hot Full Power HZP = Hot Zero Power
Calculated rod worth is reduced by 10% to allow for uncertainties.
The design basis minimum shutdown is 1.3%.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT LANT LANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 24.0 Table:
3.4.1-1 Page:
1 of 2 UNIT 1 THERMAL AND HYDRAULIC DESIGN PARAMETERS Cycle 1 Design Parameters1 Total Heat Output, MWt 3250 Total Heat Output, Btu/hr 11,090 x 106 Heat Generated in Fuel, %
97.4 Maximum Thermal Overpower, %
112 Nominal System Pressure, psia 2250 Hot Channel Factors Heat Flux Nuclear, Fq 2.60 Engineering, E
qF 1.03 Total 2.80 Enthalpy Rise Nuclear N
F 1.55 Coolant Flow Total Flow Rate, lbs/hr 135.6 x 106 Average Velocity along Fuel Rods, ft/sec 15.5 Average Mass Velocity, lb/hr-ft2 2.53 x 106 Coolant Temperature, °F Design Nominal Inlet 536.32 Average Rise in Vessel 63.0 Average Rise in Core 65.7 Average in Core 570.3 Average in Vessel 567.8 Nominal Outlet of Hot Channel 667.5 Heat Transfer Active Heat Transfer Surface Area, ft2 52,200 Average Heat Flux, Btu/hr-ft2 207,000 Maximum Heat Flux, Btu/hr-ft2 579,600 Maximum Thermal Output, kw/ft 18.8 Maximum Clad Surface Temperature BOL at Nominal Pressure, °F 657 Maximum Average Clad Temperature BOL at Rated Power, °F 720 1 See Table 3.5.3-1 for current cycle design parameters.
2 Best Estimate Nominal Inlet Temperature is 533.0 °F UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT LANT LANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 24.0 Table:
3.4.1-1 Page:
2 of 2 UNIT 1 THERMAL AND HYDRAULIC DESIGN PARAMETERS Fuel Central Temperatures (Region 3-BOL) or nominal fuel rod dimensions, °F Maximum at 100% Power 4250 Maximum at 112% Power 4500 Design Parameters DNB Ratio Minimum DNB Ratio at nominal operating conditions 1.97 Pressure Drop, psi Across Core 32 Across Vessel, including nozzles 51 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 16.4 Table:
3.4.1-2 Page:
1 of 1 ENGINEERING HOT CHANNEL FACTOR E
q F
Pellet Diameter, Density Enrichment, and Eccentricity Rod Diameter, (Pitch and Bowing) 1.03 Pellet Diameter, Density, Enrichment 1.08 E
H F
Rod Diameter, Pitch and Bowing Inlet Flow Maldistribution 1.01 Flow Redistribution 1.03 Flow Mixing 0.90
Resulting E
H F
1.01
To point of minimum DNB ratio.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 16.1 Table:
3.4.1-3 Page:
1 of 1 SENSITIVITY ANALYSIS EFFECT OF VARYING THE POWER DISTRIBUTION Power Distribution Power
% of Rated Tin Pressure Statistical Number Of Fuel Rods Which May Experience DNB Design 112 536.3 2250 0.54 Best Estimate 112 536.3 2250 0.09 EFFECT OF VARYING POWER LEVELS Power Distribution Power
% of Rated Tin Pressure Statistical Number of Fuel Rods Which May Experience DNB Best Estimate 100 536.3 2250 0.01 Best Estimate 112 536.3 2250 0.09 EFFECT OF VARYING FLOW RATE AT 12%
POWER Power Distribution Flow
% of Rated Tin Pressure Statistical Number of Fuel Rods Which May Experience DNB Best Estimate 100 536.3 2250 0.09 Best Estimate 95 536.3 2250 0.18 Best Estimate 90 536.3 2250 0.42
The statistical number of rods which could experience DNB takes into account the distribution of the experimental data from which the W-3 DNB correlation was developed and the distribution of the power in the core.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
25.0 Table:
3.5.1-1 Page:
1 of 2 Unit 1 Westinghouse 15x15 Fuel Design Parameters Parameter 15x15 W Upgrade Fuel Assembly Design Fuel Assembly Length, in.
159.975 Fuel Rod Length, in.
152.88 Assembly Envelope, in.
8.426 Compatible with Core Internals Yes Fuel Rod Pitch, in.
0.563 Number of Fuel Rods/Assembly 204 Number of Guide Thimbles/Assembly 20 Number of Instrumentation Tube Assembly 1
Compatible with Moveable In-Core Yes Detector System Fuel Tube Material ZIRLO¥ Fuel Rod Clad OD, in.
0.422 Fuel Rod Clad Thickness, in.
0.0243 Fuel/Clad Gap, mil 7.5 Fuel Pellet Diameter, in 0.3659 Guide Thimble Material ZIRLO¥ Guide Thimble ID, in.1 0.499 Structural Material - Five Inner Grids ZIRLO¥ Structural Material - Two End Grids Inconel 1 Above dashpot UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
25.0 Table:
3.5.1-1 Page:
2 of 2 Unit 1 Parameter 15x15 W Upgrade Fuel Assembly Design Grid height, in.
1.90 (Inner Grids)
Valley-to-Valley, in.
1.522 (End Grids) 0.875 (IFMs) 0.972 (Robust Protective Grid)
Bottom Nozzle Reconstitutable Top Nozzle Holddown Springs 3-leaf UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT LANT LANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 24.0 Table:
3.5.1-2 Page:
1 of 1 COMPARISON OF BURNABLE ABSORBER RODS DESIGN PARAMETERS Parameter W
Wet Annular Burnable Absorber (BA)
W Borosilicate Glass BA 15x15 Fuel Assembly (FA)
Overall Length, in 150.00
152 Absorber Length, in 134.00
142.7 Absorber Material A12O3-B4C B2O3 Absorber Form Annular Pellet Glass Tube Outer Clad O.D., in
.381
.439 Absorber Clad Material Zircaloy Stainless Absorber Thickness, in
.020
.077 Guide Thimble I.D., in
.499
.499
Typical length which can be changed to accommodate specific plant fuel cycle application.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT LANT LANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 29.0 Table:
3.5.1-3 Page:
1 of 1 UNIT 1
SUMMARY
OF FUEL ASSEMBLY NORMALIZED GRID IMPACT FORCES SSE/DBE1 and LOCA OBE GRID TYPE force/limit = ratio force/limit = ratio 15Upgrade Mid 32.5%
44%
IFM 45.2%
49%
1 SSE (Safe Shutdown Earthquake) is DBE (Design Basis Earthquake) for D.C. Cook.
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT LANT LANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 29.0 Table: 3.5.1-3A Page:
1 of 1 UNIT 1 Maximum Thimble Tube and Fuel Rod Stresses Maximum Thimble Tube Stress, - Ksi Maximum Fuel Rod Stress, Ksi Grid Type SSE +
LOCA Limit Ratio 15 Upgrade Pm 13.35 24.35 54.8%
21.82 45.96 47.5%
Pm + Pb 22.87 36.53 62.6%
24.52 68.95 35.6%
Maximum Thimble Tube Stress, - Ksi Maximum Fuel Rod Stress, Ksi Grid Type OBE Limit Ratio OBE Limit Ratio 15 Upgrade Pm 6.101 11.6 53%
20.463 21.89 93%
Pm + Pb 17.309 17.4 99%
23.059 32.84 70%
UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 24.0 Table:
3.5.2-1 Page:
1 of 1 FUEL ASSEMBLY DESIGN PARAMETERS COOK NUCLEAR PLANT UNIT 1 - CYCLE 22 Region 23A 23B 24A 24B 24C 24D Enrichment (w/o U235) 3.715
4.200*
3.800 4.200 1.500 1.600 Density (percent theoretical) 95.548*
95.411*
95.50 95.50 95.50 95.50 Number of Assemblies 51 32 48 36 25 1
Approximate Burn up at Beginning of Cycle 22 (MWD/MTU)
22,105 21,364 0
0 0
0 Approximate Burn up at End of Cycle 22 (MWD/MTU)
40,301 43,331 23,653 23,964 8,619 7,891
All values are as-built.
Based upon the Nominal EOC 21 burn up of 18,351 MWD/MTU
Assumes EOC burn up of 19,960 MWD/MTU UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 24.0 Table:
3.5.2-2 Page:
1 of 1 KINETICS CHARACTERISTICS COOK NUCLEAR PLANT UNIT 1 CYCLE 22 WITH W 15X15 UPGRADE FUEL Most Positive Moderator Temperature Coefficient (pcm/qF)1
+5.0 < 70% RTP2 linear ramp to 0.0 from 70 to 100% RTP Doppler Temperature Coefficient (pcm/qF)
-0.9 to -3.2 Least negative Doppler - Only Power Coefficient, Zero to Full Power (pcm/% power)
-9.55 to -6.11 Most Negative Doppler - Only Power Coefficient, Zero to Full Power (pcm/% power)
-19.4 to -12.79 Delayed Neutron Fraction, Eeff (%)
0.40 to 0.70 Eeff (%) minimum (BOL rod ejection only)
> 0.5 Maximum Differential Rod Worth of Two Banks Moving Together at HZP with 100% overlap (pcm/sec)
< 75 ARO Shutdown Boron (ppm) NOXE, PKSM, K=0.99 For Most Reactive Time in Life
<2,200 Worth of Most Reactive Rod (ppm) 155.1 1 1 pcm = 1.0 x 10-5 U 2 RTP = Rated Thermal Power UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT NUCLEAR PLANT NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT REPORT REPORT Revision: 24.0 Table:
3.5.2-3 Page:
1 of 1 SHUTDOWN REQUIREMENTS AND MARGINS COOK NUCLEAR PLANT UNIT 1 - CYCLE 22 Control Rod Worth (%)
BOC EOC Available Rod Worth Less Worst Stuck Rod 6.158 6.222 (A) Less 10%
5.542 5.600 Control Rod Requirements (%)
Reactivity Defects (Doppler, Tavg, RIA, Redistribution) 1.400 2.113 Void Allowance 0.050 0.050 (B) Total requirements 1.450 2.163 (C) Shutdown Margin [(A)-(B)] (%)
4.092 3.437 (D) Required Shutdown Margin (%)
1.300 1.300 Excess Shutdown Margin [(C) - (D)] (%)
2.792 2.137 UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
26.0 Table:
3.5.3-1 Page:
1 of 3 Unit 1 Cook Nuclear Plant Unit 1 Thermal-Hydraulic Design Parameters For Upgrade Fuel THERMAL AND HYDRAULIC PARAMETERS DESIGN PARAMETERS 1 Reactor Core Heat Output, MWt 3,304 Reactor Core Heat Output, 106 Btu/hr 11,273 Heat Generated in Fuel, %
97.4%
Core System Pressure, Nominal, psia 2,115 Pressurizer Pressure, Nominal Steady-State, psia 2,100 Minimum DNBR at Nominal Conditions Typical Flow Channel 2.18 1 Thimble (Cold Wall) Flow Channel 2.09 1 Safety Analysis DNBR for Design Transients Typical Flow Channel 1.55 1 Thimble Flow Channel 1.55 1 DNB Correlation WRB-1 Coolant Conditions Minimum Measured Flow, 103 gpm 362.9 Effective Flow Area for Heat Transfer, ft2 51.5 Average Inlet Velocity along Fuel Rods, ft/sec 15.35 2 Average Mass Velocity, 106 lbm/hr-ft2 2.606 2 Nominal Vessel/Core Inlet Temperature, qF 543.6 2 1 Based upon Revised Thermal Design Procedure (RTDP) 2 Based on Thermal Design Flow = 354,000 gpm UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
26.0 Table:
3.5.3-1 Page:
2 of 3 Unit 1 Cook Nuclear Plant Unit 1 Thermal-Hydraulic Design Parameters For Upgrade Fuel THERMAL AND HYDRAULIC PARAMETERS DESIGN PARAMETERS 1 Vessel Average Temperature, qF 575.4 3 Core Average Temperature, qF 579.2 2 Vessel Outlet Temperature, qF 607.2 2 Average Temperature Rise in Vessel, qF 63.6 2 Average Temperature Rise in Core, qF 67.9 2 Average Enthalpy Rise in Core, Btu/lbm 91.52 2 Heat Transfer Active Heat Transfer, Surface Area, ft2 52,100 Average Heat Flux, Btu/hr-ft2 210,900 Maximum Heat Flux for Normal Operation, Btu/hr-ft2 4 489,300 Average Linear Power, kW/ft 6.83 Peak Linear Power for Normal Operation, kW/ft 4 15.85 Maximum Clad Surface Temperature, qF 653 Fuel Centerline Temperature Temperature at Peak Linear Power for Prevention of Centerline Melt, qF 4700 Calculational Factors Engineering Heat Flux Factor 1.000 Fuel Densification Factor (axial) 1.002 3 Evaluations have been performed utilizing available DNB margin to support a maximum vessel average temperature of 576.3 ºF.
4 Based upon 2.32 FQ Peaking Factor UFSAR Revision 31.0
INDIANA MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revised:
26.0 Table:
3.5.3-1 Page:
3 of 3 Unit 1 Cook Nuclear Plant Unit 1 Thermal-Hydraulic Design Parameters For Upgrade Fuel THERMAL AND HYDRAULIC PARAMETERS DESIGN PARAMETERS 1 Radial Peaking Factor Design Nuclear Enthalpy Rise Hot Channel Factor 1.545 Pressure Drop Across Core, psi (Best Estimate Flow) 25.4 5 5 Includes the effect of IFMs and thimble plug removal UFSAR Revision 31.0