ML21125A590
Text
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 1 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate
- 1. Reactor Core Heat Output, MWt 3,391 3,411 3468
- 2. Reactor Core Heat Output, 106 Btu/hr 11,573.5 11,639 11,833
- 3. Heat Generated in Fuel, % 97.4 97.4 97.4
- 4. System Pressure, Nominal, psia 2,280 2,280 2,280
- 5. System Pressure, Minimum Steady-State, psia 2,250 2,250 2,250
- 6. Minimum Departure from Nucleate Boiling Ratio for Design Transients Typical Flow Channel 1.80 (2) 1.69 (3) 1.69 (3)
Thimble Flow Channel 1.77 (2) 1.61 (3) 1.61 (3) 1 The fresh fuel assemblies for Cycle 21 and beyond will have Optimized ZIRLO clad fuel rods and ZIRLO guide thimbles, instrumentation tubes, mid-grids and IFM grids with balanced vanes. The option to remove thimble plugs will exist for Cycle 13 and beyond. This will increase the bypass flow and cause small changes in the core flow rates and temperatures.
2 These numbers are based on Improved Thermal design Procedure in Reference 2.
3 These numbers are based on Revised Thermal Design Procedure in Reference 3.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 2 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate COOLANT FLOW
- 7. Total Thermal Design Flow Rate, 106 lbm/hr 142.7 134.4 134.7
- 8. Effective Flow Rate for Heat Transfer, 106 lb/hr 136.3 127.5 125.15
- 9. Effective Flow Area for Heat Transfer, ft2 51.1 54.1 54.1
- 10. Average Velocity Along Fuel Rods, ft/sec 16.7 14.6 13.5
- 11. Average Mass Velocity, 106 lbm/hr-ft2 2.72 2.36 2.31 COOLANT TEMPERATURE, °F
- 12. Nominal Inlet 541.3 543.4 (4) 540.8 (4)
- 13. Average Rise in Vessel 61.8 65.3 (4) 66.4 (4)
- 14. Average Rise in Core 63.4 68.4 (4) 71.0 (4)
- 15. Average in Core 574.3 579.3 (4) 578.05 (4) 4 Based on thermal design flow Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 3 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate
- 16. Average in Vessel 572.2 576.0 (4) 574.0 (4)
HEAT TRANSFER
- 17. Active Heat Transfer, Surface Area, ft2 59,700 57,505 57,505
- 18. Average Heat Flux, Btu/hr-ft2 188,700 197,180 200,477
- 19. Maximum Heat Flux for Normal Operation, Btu/hr-ft2 437,800(5) 460,420 468,114
- 20. Average Thermal Output, kW/ft 5.41 5.45 5.54
- 21. Maximum Thermal Output for Normal Operation, kW/ft 12.6 (6) 12.7 12.9 Maximum Thermal Output at Maximum Overpower Trip
- 22. 18.0 (7) 22.5 22.5 Point (118% power), kW/ft
- 23. Heat Flux Hot Channel Factor, FQ 2.32 (8) 2.335 2.335 5
The value of 437,800 Btu/hr-ft2 is associated with a Cycle 1 value of FQ of 2.32.
6 This value of 12.6 kW/ft is associated with a Cycle 1 value of FQ of 2.32.
7 See Section 3.3.2.2.6.
8 The value of FQ = 2.32 was the value of FQ for normal operation reported in the original FSAR.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 4 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate
- 24. Peak Fuel Central Temperature at 100% Power, °F < 4700 < 4700 < 4700 Peak Fuel Central Temperature at Maximum Thermal
- 25. < 4700 < 4700 < 4700 Output for Maximum Overpower Trip Point, °F FUEL ASSEMBLIES
- 27. Number of Fuel Assemblies 193 193 193
- 28. UO2 Rods per Assembly 264 264 264
- 29. Rod Pitch, in 0.496 0.496 0.496
- 30. Overall Dimensions, in 8.426 x 8.426 8.426 x 8.426 8.426 x 8.426
- 31. Fuel Weight (as UO2), lb 222,739 204,200 204,200
- 32. Zircaloy Weight, lb 50,913 45, 914 45, 914 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 5 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate 6 - Flow mixer grids 6 - Flow mixer grids 2 - Non-flow mixer 2 - Non-flow mixer
- 33. Number of Grids per Assembly 8 - Type R grids grids 3 - IFM grids 3 - IFM grids 1 - Protective Grid 1 - Protective Grid Out - In 3 - Region Low 3 - Region Low
- 34. Loading Techniques Checkerboard Leakage Leakage FUEL RODS
- 35. Number 50,952 50,952 50,952
- 36. Outside Diameter, in 0.374 0.360 0.360
- 37. Diametral Gap, in 0.0065 0.0062 0.0062
- 38. Clad Thickness, in 0.0225 0.0225 0.0225 Optimized
- 39. Clad Material Zircaloy-4 Zircaloy-4 ZIRLO starting with Cycle 21 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 6 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate FUEL PELLETS UO2 Sintered UO2 Sintered
- 40. Material UO2 Sintered 0.370 Enriched 0.370 Enriched
- 41. Density (% of Theoretical) 95 95.5 95.5
- 42. Diameter, in 0.3225 0.3088 0.3088
- 43. Length, in 0.530 (4) 0.462 Axial Blankets 0.462 Axial Blankets ROD CLUSTER CONTROL ASSEMBLIES
- 44. Neutron Absorber, Full/Part Length (9) Ag-In-Cd Ag-In-Cd Ag-In-Cd Type 304 Type 304 Type 304
- 45. Cladding Material SS-Cold SS-Cold Worked SS-Cold Worked Worked
- 46. Clad Thickness, in 0.0185 0.0185 0.0185
- 47. Number of Clusters, Full and Part Length (9) 53/0 53/0 53/0 9
Part Length CRDMs were eliminated.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 7 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate
- 48. Number of Absorber Rods per Cluster 24 24 24 CORE STRUCTURE
- 49. Core Barrel, I.D./O.D., in 148.0/152.5 148.0/152.5 148.0/152.5
- 50. Thermal Shield, I.D./O.D., in 158.5/164.0 158.5/164.0 158.5/164.0 STRUCTURE CHARACTERISTICS
- 51. Core Diameter, in (Equivalent) 132.7 132.7 132.7
- 52. Core Height, in (Active Fuel) 144.0 144.0 144.0 REFLECTOR THICKNESS AND COMPOSITION
- 53. Top - Water plus Steel, in 10 10 10
- 54. Bottom - Water plus Steel, in 10 10 10
- 55. Side - Water plus Steel, in 15 15 15
- 56. H2O/U Molecular Ratio Core, Lattice (Cold) 2.41 2.73 2.73 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.1-1 Page: 8 of 8 UPDATED FINAL SAFETY ANALYSIS REPORT REACTOR DESIGN COMPARISON TABLE 1 Typical Cycle Before Typical Cycle After Thermal and Hydraulic Design Parameters Initial Cycle MUR Power Uprate MUR Power Uprate FEED ENRICHMENT, W/O
- 57. Region 1 2.10 4.0/2.6 (10) 4.0/2.6 (10)
- 58. Region 2 2.60 4.0/2.6 (10) 4.0/2.6 (10)
- 59. Region 3 3.10 4.0/2.6 (10) 4.0/2.6 (10) 10 Reload enrichments are cycle-specific, 2.6 w/o value corresponds to the axial blanket.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.4 D. C. COOK NUCLEAR PLANT Table: 3.1-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 ANALYTIC TECHNIQUES IN CORE DESIGN Section Analysis Technique Computer Code Referenced Mechanical Design of Core Internals Loads, Deflections, and Stress Analysis Static and Dynamic Modeling Blowdown code, FORCE, Finite 14.3.3 element structural analysis code, and others Fuel Rod Design Fuel Performance Characteristics (temperature, internal Semi-empirical thermal model of fuel Westinghouse fuel rod design model 3.2.1.3.1 pressure, clad stress, etc.) rod with consideration of fuel density 3.3.3.1 changes, heat transfer, fission gas 3.4.2.2 release, etc. 3.4.3.4.2 Nuclear Design
- 1. Cross Sections and Group Constants Microscopic data Modified ENDF/B-V or ENDF/B-VI 3.3.3.2 Macroscopic constants for homogenized library 3.3.3.2 core regions PHOENIX-P Group constants for control rods with PHOENIX-P 3.3.3.2 self-shielding Nuclear Design (Continued)
- 2. X-Y Power Distributions, Fuel Depletion, Critical 3D, 2-Group Nodal Expansion Method ANC 3.3.3.3 Boron Concentrations, X-Y Xenon Distributions, Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.4 D. C. COOK NUCLEAR PLANT Table: 3.1-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 ANALYTIC TECHNIQUES IN CORE DESIGN Section Analysis Technique Computer Code Referenced Reactivity Coefficients
- 3. Axial Power Distributions, Control Rod Worths, 1-D, 2-Group Diffusion Theory APOLLO 3.3.3.3 and Axial Xenon Distribution
- 4. Fuel Rod Power Integral Transport Theory LASER 3.3.3.1 Effective Resonance Temperature Monte Carlo Weighting Function REPAD Thermal-Hydraulic Design
- 1. Steady-State Subchannel analysis of local fluid THINC-IV 3.4.3.4.1 conditions in rod bundles, including inertial and crossflow resistance terms, solution - progresses from core-wide to hot assembly to hot channel
- 2. Transient Departure from Nucleate Boiling Subchannel analysis of local fluid THINC-I (THINC-III) 3.4.3.4.1 Analysis conditions in rod bundles during transients by including accumulation terms in conservation equations solution progresses from core-wide to hot assembly to hot channel Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.4 D. C. COOK NUCLEAR PLANT Table: 3.1-3 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 DESIGN LOADING CONDITIONS FOR REACTOR CORE COMPONENTS
- 1. Fuel Assembly Weight
- 2. Fuel Assembly Spring Forces
- 3. Internals Weight
- 4. Control Rod Trip (equivalent static load)
- 5. Differential Pressure
- 6. Spring Preloads
- 7. Coolant Flow Forces (static)
- 8. Temperature Gradients Differences In Thermal Expansion
- 9. a. Due to temperature differences
- b. Due to expansion of different materials
- 10. Interference Between Components
- 11. Vibration (mechanically or hydraulically induced)
- 12. One Or More Loops Out Of Service
- 13. Operational Transients
- 14. Pump Overspeed
- 15. Seismic Loads (operating basis earthquake and design basis earthquake)
- 16. Blowdown Forces (due to cold and hot leg break)
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.2-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 Max imum Deflections Allowed For Reactor Internal Support Structure No Loss- of- Function Component Allowab le Deflections ( in)
Deflections ( in)
Upper Barrel Radial inward 4.1 8.2 Radial outward 1.0 1.0 Upper Package 0.10 0.15 Rod Cluster Guide Tubes 1.00 1.75 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 4 Reactor Core Description Activ e Core Equivalent Diameter, in 132.7 Active Fuel Height, First Core, in 144.0 Height-to-Diameter Ratio 1.09 Total Cross Section Area, ft2 96.06 H2O/U Molecular Ratio, lattice (Cold) 2.73 Reflector Thick ness And Composition Top - Water plus Steel, in 10 Bottom - Water plus Steel, in 10 Side - Water plus Steel, in 15 Fuel Assemb lies Number 193 Rod Array 17 x 17 Rods per Assembly 264 Rod Pitch, in 0.496 Overall Transverse Dimensions, in 8.426 x 8.426 Fuel Weight (as UO2), lb - per assembly 1058 Zircaloy Weight, lb - per assembly 238 Number of Grids per Assembly 2-R 6-Z 3-IFM 1-P Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 4 Reactor Core Description Composition of Grids R-Inconel 718 Zircaloy 4/ZIRLO IFM - Zircaloy 4/ZIRLO P-Debris Resistant- Inconel 718 Weight of Grids (Effective in Core), lb - per assembly 20.10 Number of Guide Thimbles per Assembly 24 Composition of Guide Thimbles Zircaloy 4/ZIRLO Diameter of Guide Thimbles (upper part), in 0.442 I.D. x 0.474 O.D.
Diameter of Guide Thimbles (lower part), in 0.397 I.D. x 0.429 O.D.
Diameter of Instrument Guide Thimbles, in 0.440 I.D. x 0.474 O.D.
Fuel Rods Number 50,952 Outside Diameter, in 0.360 Diameter Gap, in 0.0062 Clad Thickness, in 0.0225 Zircaloy-4 / ZIRLO /
Clad Material Optimized ZIRLO Fuel Pellets Material UO2 Sintered Density (percent of Theoretical) Approx. 95.5 Maximum Fuel Enrichments w/o 4.95 Diameter, in 0.3088 Length, in 0.370 Enriched Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 4 Reactor Core Description 0.462 Axial Blanket Mass of UO2 per Foot of Fuel Rod, lb/ft 0.336 1 Rod Cluster Control Assemb lies Neutron Absorber Ag-In-Cd Composition 80%, 15%, 5%
Diameter, in 0.341 Density, lb/in3 0.367 Type 304, Cold Worked Stainless Cladding Material Steel Clad Thickness, in 0.0185 Number of Clusters Full Length 53 Number of Absorber Rods per cluster 24 Full Length Assembly Weight (dry), lb 149 Ex cess Reactiv ity Maximum Fuel Assembly k 1.476 2 (Cold, Clean, Unborated Water)
Maximum Core Reactivity 1.224 2 (Cold, Zero Power Beginning of Cycle) 1 Based on fuel at 95.5% theoretical density 2
Typical values Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.3-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 4 Reactor Core Description Integral Fuel Burnab le Ab sorb er Number ~ 8640 2 Material ZrB2 Coating Thickness, mil ~ 0.2 Boron 10 Loading, mg/in 2.25 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.3-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 2 Nuclear Design Parameters
( Best- Estimate V alues Are Representativ e Of A Typical Cycle)
Core Average Linear Power, kW/ft, including densification effects1 5.45 / 5.54 Total Heat Flux Hot Channel Factor, FQ 2.335 N
Nuclear Enthalpy Rise Hot Channel Factor, F 1.61 [ 1+ 0.3(1-P)]
Reactiv ity Coefficients Design Limits Best Estimate Doppler-only Power, Coefficients, pcm/°F Upper Curve -19.4 to -12.224 -12.4 to -7.9 Lower Curve -9.55 to -5.818 -10.9 to -7.5 Doppler Temperature Coefficient, pcm/°F -3.20 to -0.91 -1.9 to -1.3 3
Moderator Temperature Coefficient, pcm/°F + 5 to -38 + 5.04to -29.543 Boron Coefficient, pcm/ppm -10.9 to -7.6 Rodded Moderator Density, pcm/gm/cc 0.54 x 10 E 05 0.40 x 10 E 05 Radial Assemb ly Peak ing Factor Design Limits Best Estimate Radial Assembly Peaking Factor5 Unrodded 1.36 to 1.49 D bank 1.51 to 1.58 D+ C 1.61 to 1.70 Boron Concentrations ( ppm) Design Limits Best Estimate Zero Power, K eff = 0.99, Cold, Rod Cluster Control 1804 Assemblies Out Zero Power, K eff = 0.99, Hot, Rod Cluster Control 1930 Assemblies Out Design Basis Refueling Boron Concentration 2400 1855 1
Before and After MUR power uprate values listed.
2 Uncertainties are referenced in Section 3.3.3.3.
3 Design limit dependent on vessel average moderator temperature. Value reported is for Cycle 14 temperature of 574.0 °F.
4 Administrative rod withdrawal limits are required if an MTC violation is observed during startup physics testing, as specified by an action statement in Technical Specification 3.1.3.A.1.
5 Typical values.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.3-2 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 2 Nuclear Design Parameters
( Best- Estimate V alues Are Representativ e Of A Typical Cycle)
Zero Power, K eff = 0.95, Cold, Rod Cluster Control 1754 Assemblies In Zero Power, K eff = 1.00, Hot, Rod Cluster Control 1795 Assemblies Out Full Power, No X enon, K eff = 1.0, Hot, Rod Cluster 1648 Control Assemblies Out Full Power, Equilibrium X enon, K eff Hot, Rod 1316 Cluster Control Assemblies Out Reduction with Fuel Burnup Reload Cycle,
~ 84 ppm/GWD/MTU6 Delayed Neutron Fraction and Lifetime Design Limits Best Estimate eff BOL, (EOL 0.0075, (0.0040) 0.0062, (0.0050)
BOL, (EOL) sec 5 20.1, (22.3)
Control Rods Best Estimate Best Estimate Rod Requirements See Table 3.3-3 Maximum Bank Worth, pcm 1380 Maximum Ej ected Rod Worth See Chapter 14 Bank Worth, pcm7 BOL, X e free HZP EOL, X e free HZP Bank D 1135 1380 Bank C 966 1222 Bank B 851 1259 Bank A 572 617 6
Gigawatt Day (GWD) = 1000 Megawatt Day (1000 MWD).
7 Note: For two statepoint values of keff, k1 and k2, the reactivity change in pcm (percent milli) is given by In (k2/k1) x105.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 25.0 D. C. COOK NUCLEAR PLANT Table: 3.3-3 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 SHUTDOWN REQ UIREMENTS AND MARGINS Typical V alues BOC EOC Control Rod Worth ( pcm)
Available Rod Worth Less Worst Stuck Rod 4856 5879 (A) less 10% 4371 5291 Control Rod Req uirements ( pcm)
Reactivity Defects (Doppler, Tavg, RIA, Redistribution) 1431 2764 Void Allowance 50 50 (B) Total Requirements 1481 2814 (C) Available Shutdown Margin [ (A) - (B)] (pcm) 2890 2477 (D) Required Shutdown Margin (pcm) 1300 1300 Excess Shutdown Margin [ (C) - (D)] (pcm) 1590 1177 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.4 D. C. COOK NUCLEAR PLANT Table: 3.3-4 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 AXIAL STABILITY INDEX PRESSURIZED WATER REACTOR CORE WITH A 12 FOOT HEIGHT Stability Index (hr-1)
Burnup Difference FZ CB (ppm) Exp Calc (MWD/MTU) (Exp-Calc) 1550 1.34 1065 -0.041 -0.032 -0.009 7700 1.27 700 -0.014 -0.006 -0.008 Difference: +0.027 +0.026 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 19.1 D. C. COOK NUCLEAR PLANT Table: 3.3-6 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 1 COMPARISON OF MEASURED AND CALCULATED DOPPLER DEFECTS Core Burnup Measured Calculated Plant Fuel Type (MWD/MTU) (pcm)1 (pcm) 1 Air-filled 1800 1700 1710 2 Air-filled 7700 1300 1440 3 Air and helium-filled 8460 1200 1210 1
1 pcm = 10-5 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.4-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 1 of 4 Cook Nuclear Plant Unit 2 R eactor Design Comparison Tab le 1 Typical Typical Thermal and Hydraulic Cycle Cycle After Initial Cycle Before Design Parameters MUR Power MUR Power Uprate Uprate Reactor Core Heat Output, MWt 3391 3411 3468 Reactor Core Heat Out, 106 BTU/hr 11,573.5 11,639 11,833 Heat Generated in Fuel, % 97.4 97.4 97.4 System Pressure, Nominal, psia 2 2280 2280 2280 System Pressure, Minimum Steady-State, psia 2250 2250 2250 Minimum DNBR at Nominal Conditions Typical Flow Channel 3.03 3 2.42 2.66 Thimble (Cold Wall) Flow Channel 2.70 3 2.28 2.49 Design DNBR for Design Transients Typical Flow Channel 1.80 4 1.69 5 1.69 5 Thimble Flow Channel 1.77 4 1.615 1.61 5 1 The fresh fuel assemblies for Cycle 21 and beyond will have Optimized ZIRLOclad fuel rods and ZIRLO guide thimbles, instrumentation tubes, mid-grids and IFM grids with balanced vanes. The option to remove thimble plugs will exist for Cycle 13 and beyond. This will increase the bypass flow and cause small changes in the core flow rates, temperatures and pressure drops.
2 Pressure in the core. See Reference (1).
3 Based on Improved Thermal Design Procedure, Reference (84).
4 Including 31.1 percent rod bow penalty.
5 Value used in DNB analyses (RTDP Transients).
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.4-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 2 of 4 Cook Nuclear Plant Unit 2 R eactor Design Comparison Tab le 1 Typical Typical Thermal and Hydraulic Cycle Cycle After Initial Cycle Before Design Parameters MUR Power MUR Power Uprate Uprate DNB Correlation WRB-1 WRB-2 WRB-2 Coolant Flow 6 Total Thermal Design Flow Rate, 106 142.7 134.3 134.7 lbm/hr Best Estimate Flow, 106 lbm/hr 148.4 145.2 145.2 Mechanical Design Flow, 106 lbm/hr 154.3 154.5 154.5 Minimum Effective Flow Rate for Heat 136.3 127.4 125.15 Transfer, 106 lbm/hr Effective Flow Area for Heat Transfer, 51.1 54.1 54.1 ft2 Average Velocity Along Fuel Rods, 16.7 14.6 13.5 ft/sec Average Mass Velocity, 106 lbm/hr 2.72 2.35 2.31 Coolant Temperature 6 Nominal Inlet, °F 541.3 543.4 540.8 Average Rise in Vessel, °F 61.8 65.3 66.4 Average Rise in Core, °F 63.4 68.4 71.0 Average in Core, °F 574.3 579.3 578.05 6 Based on Thermal Design Flow.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.4-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 3 of 4 Cook Nuclear Plant Unit 2 R eactor Design Comparison Tab le 1 Typical Typical Thermal and Hydraulic Cycle Cycle After Initial Cycle Before Design Parameters MUR Power MUR Power Uprate Uprate Average in Vessel, °F 572.2 7 576.0 574.0 Heat Transfer Active Heat Transfer, Surface Area, ft2 59,700 57,505 57,505 Average Heat Flux, BTU/hr-ft2 188,700 197,180 200,477 Maximum Heat Flux for Normal 437,800 8 460,420 9 468,114 9 Operation, BTU/hr-ft2 Average Linear Power, kW/ft 5.41 5.45 5.54 Peak Linear Power for Normal 12.6 8 12.7 9 12.9 9 Operation, kW/ft Peak Linear Power Resulting from Overpower Transients/Operator Errors, 18.0 10 22.5 22.0 (assuming a maximum overpower of 118%), kW/ft Peak Linear Power for Prevention of 18.0 > 22.5 > 22.5 Centerline Melt, kW/ft 11 Fuel Central Temperature Peak at Peak Linear Power for 4700 4700 4700 Prevention of Centerline Melt, °F 7 The vessel average temperature was increased to 573.8°F as per amendment 19 of May 13, 1980.
8 This limit is associated with the value of FQ = 2.32.
9 This limit is associated with the value of FQ = 2.335.
10 See Section 3.3.2.2.6.
11 See Section 3.4.2.2.6.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revised: 26.0 D. C. COOK NUCLEAR PLANT Table: 3.4-1 UPDATED FINAL SAFETY ANALYSIS REPORT Page: 4 of 4 Cook Nuclear Plant Unit 2 R eactor Design Comparison Tab le 1 Typical Typical Thermal and Hydraulic Cycle Cycle After Initial Cycle Before Design Parameters MUR Power MUR Power Uprate Uprate Pressure Drop 12 Across Core, psi 23.3 + 2.3 27.0 + 2.7 27.0 2.7 Across Vessel, including noz z les, psi 43.2 4.3 50.1 5.0 50.1 5.0 12 Based on Best Estimate Flow as discussed in 3.4.2.6.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 18.2 D. C. COOK NUCLEAR PLANT Table: 3.4-2 UPDATED FINAL SAFETY ANALY SIS Page: 1 of 1 REPORT V OID FRACTIONS AT NOMINAL REACTOR CONDITIONS 1 WITH DESIGN HOT CHANNEL FACTORS Ave rage Maxi mum Core 0.2% -------
Hot Subchannel 0.9% 2.1%
1 Based upon Minimum Measured Flow.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 18.2 D. C. COOK NUCLEAR PLANT LANT Table: 3.4-3 UPDATED FINAL SAFETY ANALY SIS Page: 1 of 2 REPORT SYSTEM DESIGN AND OPERATING PARAMETERS
( TYPICAL CYCLES BEFORE AND AFTER MUR POWER UPRATE) 1 At 7 0 ° At Hot2 Approximate total RCS volume (including pressuriz er and surge line), with 12,470 12,845 0% steam generator tube plugging. (ft.3)
Approximate system liquid volume, (including pressuriz er water) at 12,019 3 maximum guaranteed power with 0% steam generator tube plugging. (ft.3)
SYSTEM THERMAL AND HYDRAULIC DATA
( BASED ON THERMAL DESIGN FLOW)
Typical Cycle Before Typical Cycle After MUR Power Uprate MUR Power Uprate4 NSSS Power, MWt 3423 3480 Reactor Power, MWt 3411 3468 Thermal Design Flows, gpm Active Loop 88,500 88,500 Reactor 354,000 354,000 Total Reactor Flow, 10 lb/hr 6
134.4 134.4 Temperatures, °F Reactor Vessel Outlet 606.4 611.1 Reactor Vessel Inlet 541.2 545.1 Steam Generator Outlet 541.0 544.8 Steam Generator Steam 521.1 524.0 Feedwater 431.0 444.1 Steam Pressure, psia 820.0 840.9 Total Steam Flow, 106 lb/hr 14.78 15.37 1
The option to remove thimble plugs will exist for Cycle 13 and beyond. This will increase bypass flow and cause small changes in the core flow rates and temperatures.
2 This includes a 3% volume increase (1.3% for thermal expansion and 1.7% for pipe connections to the reactor coolant loops, volume in the rod drive mechanisms and calculation inaccuracies). Refer to Westinghouse letters AEP-97-151, AEP-98-078, AEP-98-082, AEP-98-161, and the Westinghouse IMP database SEC-SAI-4824-CO.
3 Total RCS Volume (12,845 ft.3 ) - Pressuriz er steam volume at full power (826 ft. 3).
4 Based upon reactor loop average temperature of 578.1°F.
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 18.2 D. C. COOK NUCLEAR PLANT LANT Table: 3.4-3 UPDATED FINAL SAFETY ANALY SIS Page: 2 of 2 REPORT SYSTEM FLOW
SUMMARY
Thermal Minimum Best Mechanical Flows, gpm Design 5 Measured6 Estimate Design 4 Pumps Running, each loop 88,500 91,600 95,500 101,600 5
Fixed value analyses (non-RTDP transients).
6 DNB analyses values (RTDP transients).
Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.4 D. C. COOK NUCLEAR PLANT Table: 3.4-4 UPDATED FINAL SAFETY ANALY SIS REPORT Page: 1 of 2 COMPARISON OF THINC- IV AND THINC- I PREDICTIONS WITH DATA FROM REPRESENTATIV E WESTINGHOUSE TWO AND THREE LOOP REACTORS Improv ement ( oF) for Power % Full rms( oF ) ( oF )
Reactor Measured Inlet THINC- IV ov er THINC-( MWt) Power THINC- I THINC- IV Temp ( oF) I Ginna 847 65.1 543.7 1.97 1.83 0.14 854 65.7 544.9 1.56 1.46 0.10 857 65.9 543.9 1.97 1.82 0.15 947 72.9 543.8 1.92 1.74 0.18 961 74.0 543.7 1.97 1.79 0.18 1091 83.9 542.5 1.73 1.54 0.19 1268 97.5 542.0 2.35 2.11 0.24 1284 98.8 540.2 2.69 2.47 0.22 1284 98.9 541.0 2.42 2.17 0.25 1287 99.0 544.4 2.26 1.97 0.29 1294 99.5 540.8 2.20 1.91 0.29 1295 99.6 542.0 2.10 1.83 0.27 Robinson 1427.0 65.1 548.0 1.85 1.88 0.03 1422.6 64.9 549.4 1.39 1.39 0.00 Unit 2
UFSAR Revision 30.0 INDIANA MICHIGAN POWER Revision: 16.4 D. C. COOK NUCLEAR PLANT Table: 3.4-4 UPDATED FINAL SAFETY ANALY SIS REPORT Page: 2 of 2 COMPARISON OF THINC- IV AND THINC- I PREDICTIONS WITH DATA FROM REPRESENTATIV E WESTINGHOUSE TWO AND THREE LOOP REACTORS Improv ement ( oF) for Power % Full rms( oF ) ( oF )
Reactor Measured Inlet THINC- IV ov er THINC-( MWt) Power THINC- I THINC- IV Temp ( oF) I 1529.0 88.0 550.0 2.35 2.34 0.01 2207.3 100.7 534.0 2.41 2.41 0.00 2213.9 101.0 533.8 2.52 2.44 0.08 Unit 2