ML20309A776
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McGuire Nuclear Station UFSAR Appendix 4A. Tables Appendix 4A. Tables
McGuire Nuclear Station UFSAR Table 4-1 (Page 1 of 2)
Table 4-1. Reactor Design Comparison Table Robust Fuel Assembly Thermal And Hydraulic Design Parameters
- 1. Reactor Core Heat Output, (100%), MWt 3469
- 2. Reactor Core Heat Output, 106 Btu/hr 11836.7
- 3. Heat Generated in Fuel, % 97.4
- 4. System Pressure, Nominal, psia(1) 2280
- 5. System Pressure, Minimum Steady State, psia(1) 2250
- 6. Minimum DNBR at Nominal Conditions Limiting Channel 2.85 (WRB-2M)
- 7. Minimum DNBR at Design Transients Limiting Channel 1.45 (WRB-2M)
- 8. DNB Correlation WRB-2M COOLANT FLOW(3)
- 9. Total Thermal Flow Rate, 106 lbm/hr 145.2
- 10. Effective Flow Rate for Heat Transfer, 106 lbm/hr 136.5
- 11. Effective Flow Area for Heat Transfer, ft2 51.1
- 12. Average Velocity Along Fuel Rods, ft/sec 15.9 6 2
- 13. Average Mass Velocity, 10 lbm/hr-ft 2.67 COOLANT TEMPERATURE, °F(2)
- 14. Nominal Inlet 553.1
- 15. Average Rise in Vessel 61.2
- 16. Average Rise in Core 65.0
- 17. Average in Core 587.3
- 18. Average in Vessel 585.1 HEAT TRANSFER
- 19. Active Heat Transfer, Surface Area, ft2 59,866
- 20. Average Heat Flux, Btu/hr-ft2 192,579
- 21. Maximum Heat Flux for Normal Operation, Btu/hr-ft2 481,447 (09 OCT 2015)
McGuire Nuclear Station UFSAR Table 4-1 (Page 2 of 2)
Robust Fuel Assembly Thermal And Hydraulic Design Parameters
- 22. Average Linear Power, kW/ft 5.53
- 23. Peak Linear Power for Normal Operation, kW/ft(a) 13.8
- 24. Peak Linear Power Resulting from Overpower Transients/Operator Errors (assuming a maximum overpower of 118%), kW/ft(b) 18.0
- 25. Peak Linear Power for Prevention of Centerline Melt, kW/ft >18.0
- 26. Power Density, kW prr Liter of Core 104.5
- 27. Specific Power, kW per kg Uranium(4) 38.8 FUEL CENTRAL TEMPERATURE
- 28. Peak at Peak Linear Power for Prevention of Centerline Melt, °F Burnup Dependent
- 29. Pressure Drop (++)
Across Core, psi 28.8 +/- 2.6 Across Vessel, Including Nozzle psi 51.2 +/- 4.6 Items 30-64 Deleted duplicate and historical information that is in Table 4-4. Moved entries that are not duplicative to Table 4-4. (i.e., Items 30, 33, 54, & 55)
Notes:
- 1. Values used for thermal hydraulic core analysis.
- a. This limit is associated with the value of Fq = 2.50 and includes 2.6% gamma heating.
- b. See Section 4.3.2.2.6
(+) Based on cold dimensions.
(++) Based on best estimate reactor flow as discussed in Section 5.1. RFA pressure drops are based on Reference 98 of Section 4.4.7.
- 2. These values are typical values based on RCS flow of 400,000 gpm and a bypass flow of 6.0%.
- 3. These values are typical values based on RCS flow of 388,000 gpm and nominal inlet temperature of 553.1ºF .
- 4. Typical values. May vary based on reload specific data.
(09 OCT 2015)
McGuire Nuclear Station UFSAR Table 4-2 (Page 1 of 2)
Table 4-2. Analytic Techniques in Core Design Analysis Technique Computer Code Mechanical Design of Core Internals Loads, Deflections, and Stress Analysis Static and Dynamic Blowdown code, Modeling FORCE Finite element structural Fuel Rod Design Fuel Performance Characteristics Semi-empirical thermal PAD (temperature, internal pressure, clad strain, model of fuel rod with etc.) consideration of fuel density changes, heat transfer, fission gas release, etc.
Nuclear Design
- 1. Cross Sections and Group Constants Microscopic and Modified ENDF/B library Macroscopic constants CASMO-3 for homogenized core or CASMO-4 regions Group constants for CASMO-3 control rods with self- or CASMO-4 shielding
- 2. X-Y Power Distributions, Fuel Depletion, Collapsed 3-D, 2- SIMULATE-3P Critical Boron Concentrations, X-Y Xenon Group NEM Based or SIMULATE-3 MOX Distributions, Reactivity Coefficients Nodal Code
- 3. Axial Power Distributions, Control Rod 2-D and 3-D 2-Group SIMULATE-3P Worths, and Axial Xenon Distribution Model Analysis Code or SIMULATE-3 MOX
- 4. Fuel Rod Power Reconstructed Integral SIMULATE-3P Rod Power or SIMULATE-3 MOX
- 5. Criticality of Reactor and Fuel Assemblies 1-D, Multi-Group AMPX System of Transport Theory Codes 3-D Monte Carlo KENO-IV Thermal-Hydraulic Design
- 1. Steady-State Subchannel analysis of VIPRE-01 local fluid conditions in rod bundles, including inertial and crossflow resistance terms, solution progresses from core-wide to hot assembly to hot channel (30 NOV 2012)
McGuire Nuclear Station UFSAR Table 4-2 (Page 2 of 2)
Analysis Technique Computer Code
- 2. Transient DNB Analysis Subchannel analysis of VIPRE-01 local fluid conditions in rod bundles during transients by including accumulation terms in conservation equations; solution progresses from core-wide to hot assembly to hot channel (30 NOV 2012)
McGuire Nuclear Station UFSAR Table 4-3 (Page 1 of 1)
Table 4-3. Deleted Per 1992 Update (14 OCT 2000)
McGuire Nuclear Station UFSAR Table 4-4 (Page 1 of 5)
Table 4-4. Reactor Core Description (Units 1 and 2)
Robust Fuel Assembly Active Core Design RCC Canless Equivalent Diameter, in. 132.7 Core Average Active Fuel Height, in. 144.0 Height-to-Diameter Ratio 1.09 Total Cross-Section Area, ft2 96.06 H2O/U Molecular Ratio, Lattice (68°F, 2250 ~2.50 psia)
Reflector Thickness and Composition Top - Water plus Steel, in. ~10 Bottom - Water plus Steel, in. ~10 Side - Water plus Steel, in. ~15 Core Structure Core Barrel, ID/OD, in. 148.0/152.0 Thermal Shield Neutron Pad Design Fuel Assemblies 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(1)
(Typical)
Fuel Weight (as UO2), lbs. (Typical)(2) 219,819(1)
Zirconium Weight, lbs. (Cladding 41,966(1)
Surrounding Active Fuel)
Number of Grids per Assembly 12 Composition of grids INC718 Protective Grid, 2 INC718 End Grids, 6 ZIRLO Spacer Grids, 3 ZIRLO IFM Grids Weight of Grids (Effective in Core) lbs INC-1066, ZIRLO -2280 Number of Guide Thimbles per Assembly 24 Composition of Guide Thimbles ZIRLO (13 APR 2020)
McGuire Nuclear Station UFSAR Table 4-4 (Page 2 of 5)
Robust Fuel Assembly Inner Diameter of Guide Thimbles (upper 0.442 part), in.
Outer Diameter of Guide Thimbles (upper 0.482 part), in.
Inner Diameter of Guide Thimbles (lower 0.397 part), in.
Outer Diameter of Guide Thimbles (lower 0.439 part), in.
Inner Diameter of Instrument Guide 0.442 Thimbles, in.
Outer Diameter of Instrument Guide 0.482 Thimbles, in.
Fuel Rods Number 50,592 Outside Diameter, in. 0.374 Diameter Gap, in. 0.0065 Clad Thickness, in. 0.0225 Clad Material ZIRLO Fuel Pellets Material UO2 Sintered Density (percent of Theoretical) 95.5 Fuel Enrichments w/o(5)
Reload Regions 0.711-5.00 Diameter, in. 0.3225 Length, in. 0.387 (chamfered) (enriched);
0.400 - 0.600 (chamfered) (axial blanket)
Mass of UO2 per Foot of Fuel Rod, lb/ft 0.360(1)
Rod Cluster Control Assemblies(Unit 1)
Westinghouse Enhanced Performance (EP) RCCAs Neutron Absorber 80%, 15%, 5%
Composition (Ag,In,Cd)
(13 APR 2020)
McGuire Nuclear Station UFSAR Table 4-4 (Page 3 of 5)
Rod Cluster Control Assemblies(Unit 1)
Diameter, in.
Upper 0.341 Lower 0.336 3
Density, lbs/in. 0.367 Cladding Material Type 304 Cold Worked Stainless Steel, Chrome Plated Number of Full Length Clusters 53 Number of Absorber Rods per Cluster 24 Full Length Assembly Weight, (dry), lb. 149 AREVA AIC HARMONI RCCAs Neutron Absorber 80%, 15%, 5%
Composition (Ag,In,Cd)
Diameter, in.
Upper 0.341 Lower 0.336 Density, lbs/in. 3 0.367 Cladding Material Type 304 Cold Worked Stainless Steel, Ion-nitrated Number of Full Length Clusters 53 Number of Absorber Rods per Cluster 24 Full Length Assembly Weight, (dry), lb. 149 Hybrid Chrome Coated Next Generation Rod Cluster Control Assemblies (Unit 2)
Neutron Absorber B4C Diameter, in. 0.294 Length, in. 102 Tip Material (Ag-In-Cd)
Diameter, in.
Lower Tip 0.296 Upper Tip 0.301 Length, in.
Lower Tip 18 Upper Tip 22 Cladding Material Type 304L Stainless Steel Number of Full Length Clusters 53 (13 APR 2020)
McGuire Nuclear Station UFSAR Table 4-4 (Page 4 of 5)
Number of Absorber Rods per Cluster 24 Full Assembly Weight (dry), lb. 94 Hybrid Enhanced Performance Rod Cluster Control Assemblies (Unit 2) (6)
Neutron Absorber B4C Diameter, in. 0.294 Length, in. 102 Tip Material (Ag-In-Cd)
Diameter, in. 0.301 Length, in. 40 Cladding Material Type 304L & 316, Cold Worked Stainless Steel Clad Thickness, in. 0.0385 Number of Full Length Clusters 53 Number of Absorber Rods per Cluster 24 Full Assembly Weight (dry), lb. 94 Hybrid Ionitrided Rod Cluster Control Assemblies (Unit 2)(6)
Neutron Absorber B4C Diameter, in. 0.294 Length, in. 102 Density, lbs/in3 0.064 Tip Material (Ag-In-Cd)
Composition 80%, 15%, 5%
(Ag,In,Cd),
Diameter, in.
Lower Tip 0.294 Upper Tip 0.300 Length, in.
Lower Tip 12 Upper Tip 28 Density, lbs/in3 0.367 Cladding Material Type 136, Cold Worked Stainless Steel, Ionitrided Cladding Thickness .0385 Number of Full Length Clusters 53 (13 APR 2020)
McGuire Nuclear Station UFSAR Table 4-4 (Page 5 of 5)
Full Assembly Weight (dry), lb. 94 Burnable Poison Rod Loading & Initial Reactivity Worth Weight of Boron - 10 per foot of rod, lb/ft Variable Initial Reactivity Worth, % (hot) 0.0-~3.0 (typical)
Initial Reactivity Worth, %p (cold) 0.0-~2.2 (typical)
Excess Reactivity Maximum Fuel Assembly K (Cold, Clean, Variable3 Unborated Water)
Maximum Core K (Cold, Zero Power, 1.304 Beginning of Cycle)
WABAs Material A12O3-B4 Inside Diameter, in. 0.225 Outside Diameter, in. 0.381 Clad Material Zircaloy-4 Boron Loading Proprietary Note:
- 1. The values indicated are typical, for 17 x 17 Robust Fuel Assemblies, or Mk-BW fuel assemblies.
- 2. Not exact for every core. Total weight will vary as region UO2 varies. See region specific data for the most current values.
- 3. Maximum Fuel Assembly k-infinities for cold clean unborated water are dependent upon the fuel assembly enrichment.
- 4. Variable, depending on cycle length and BA loading.
- 5. The fuel enrichments for the first core are 2.10w/o (Region 1), 2.60w/o (Region 2), 3.10w/o (Region 3) per Ref. 19 in Section 4.2.4.
- 6. Information regarding the Hybrid EP-RCCAs and Hybrid Ionitrided RCCAs has been retained for historical purposes. These RCCAs will be retained for potential spare RCCAs.
(13 APR 2020)
McGuire Nuclear Station UFSAR Table 4-5 (Page 1 of 2)
Table 4-5. Nuclear Design Parameters [HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED]
Core Average Linear Power, kW/ft, including 5.44 densification effects and gamma heating effects Total Heat Flux Hot Channel Factor, FQ 2.50 Nuclear Enthalpy Rise Hot Channel Factor, FHN Variable limit based on the magnitude and location of the axial peak, Fz.
Tech Spec/Safety Analysis Best Reactivity Coefficients (Reload Cycles) Design Limits estimate Least-negative Doppler-only power coefficient, pcm/% -9.5 to -6.0 -17.5 to -8.3 Power Distributed Doppler Temperature Coefficient, pcm/°F -3.50 to -0.9 -2.0 to -1.2 Moderator Temperature Coefficient, pcm/°F <+7 at 0 P 7 1<0 at P = 1.0 +5 to -38 Rodded Moderator Density, pcm/gm/cc <0.43 x 105 0.38 x 105 Delayed Neutron Fraction and Lifetime First Cycle Reload Cycle eff BOL 0.0075 0.0062 eff EOL 0.0044 0.0052 l BOL, µ sec 19.4 17 l EOL, µ sec 18.1 21 Control Rods Rod Requirements See Table 4-6 See Table 4-6 Maximum Bank Worth, pcm2 <2000 ~1250 Maximum Ejected Rod Worth See 15.0 See 15.0 Boron Concentrations First Cycle Reload Cycle Zero Power, Keff = 1.00, Cold, ARO, 1 percent 1504 2000 uncertainty included Zero Power, Keff = 1.00, Hot, ARO, 1 percent 1406 2100 uncertainty included Design Basis Refueling Boron Concentration 2000 2875 Zero Power, Keff = 1.00, Hot, ARO 1292 2000 Full Power, No Xenon, Keff = 1.0, Hot, ARO 1177 1800 Full Power, Equilibrium Xenon, Keff = 1.0 Hot, ARO 879 1330 Reduction with Fuel Burnup, ppm/GWD/MTU 3 See Figure 4-33 Notes:
- 1. See Figure 4-72 (14 APR 2000)
McGuire Nuclear Station UFSAR Table 4-5 (Page 2 of 2)
- 2. Note: 1 pcm = (percent mille rho) = 10-5 where is calculated from two statepoint values of Keff by 1n (k2/K1).
- 3. Gigawatt Day (GWD) = 1000 Megawatt Day (1000 MWD).
(14 APR 2000)
McGuire Nuclear Station UFSAR Table 4-6 (Page 1 of 1)
Table 4-6. Reactivity Requirements For Rod Cluster Control Assemblies [HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED]
Beginning of Life End of Life (First End of Life (Typical Reactivity Effects, percent (First Cycle) Cycle) Reload Cycle)
- 1. Control requirements
- a. Power Defect, % 2.072 3.082 2.854
- b. Rod Insertion Allowance, % 0.50 0.50 0.354
- 2. Total Control, % 2.57 3.58 3.20
- 3. Estimated Rod Cluster Control Assembly Worth (53 Rods) Unit 1 Unit 2 Unit 1 Unit 2 Typical
- a. All full length assemblies inserted, % 7.67 8.51 7.49 8.31 6.77
- b. All but one (highest worth) assemblies inserted, 6.48 7.19 6.33 7.02 5.89
- 4. Estimated Rod Cluster Control Assembly credit with 10 5.83 6.47 5.70 6.32 5.30 percent adjustment to accommodate uncertainties (3.b.-10 percent), %
- 5. Shutdown margin available (4-2), % 3.263 3.903 2.123 2.743 2.101 Note:
- 1. The design basis minimum shutdown is 1.3%.
- 2. Includes Void Effects
- 3. The design basis minimum shutdown for Cycle 1 was 1.6%
McGuire Nuclear Station UFSAR Table 4-7 (Page 1 of 5)
Table 4-7. UO2 Benchmark Critical Experiments UO2 Benchmark Critical Experiments for CASMO-3, TABLES-3 and SIMULATE-3 Methodology Enrichment Separating Characterizing No. Ref. General Description (w/o U235) Reflector Material Separation (cm) keff 2 37 UO2 Rod Lattice 2.46 1037 ppm Water - - 1.0001+/-0.0005 3 37 UO2 Rod Lattice 2.46 764 ppm Water - 1.64 1.0000+/-0.0006 9 37 UO2 Rod Lattice 2.46 Water - 6.54 1.0030+/-0.0009 10 37 UO2 Rod Lattice 2.46 143 ppm Water - 4.91 1.0001+/-0.0009 11 37 UO2 Rod Lattice 2.46 514 ppm Water SS 1.64 1.0000+/-0.0006 13 37 UO2 Rod Lattice 2.46 15 ppm Water 1.614% B/A1(1) 1.64 1.0000+/-0.0010 14 37 UO2 Rod Lattice 2.46 92 ppm Water 1.257% B/A1(1) 1.64 1.0001+/-0.0010 15 37 UO2 Rod Lattice 2.46 395 ppm Water 0.401% B/A1(1) 1.64 0.9998+/-0.0016 17 37 UO2 Rod Lattice 2.46 487 ppm Water 0.242% B/A1(1) 1.64 1.0000+/-0.0010 (1) 19 37 UO2 Rod Lattice 2.46 634 ppm Water 0.100% B/A1 1.64 1.0002+/-0.0010 UO2 Benchmark Critical Experiments for SCALE 4.4 Methodology Critical Enrichment Poison Poison Thickness Separation (cm) Critical No. of No. Ref. General Description (w/o U235) Material (cm) Rods X Y 51 60 Multiple Fuel Clusters 4.31 None - 4.72 4.72 253.8 53 60 Multiple Fuel Clusters 4.31 None - 6.61 6.61 432.7 55 60 Multiple Fuel Clusters 4.31 None - 2.83 14.98 396 56 60 Mutliple Fuel Clusters 4.31 None - 2.83 19.81 432 57 60 Multiple Fuel Clusters 4.31 None - 2.83 13.64 360 (11 NOV 2006)
McGuire Nuclear Station UFSAR Table 4-7 (Page 2 of 5)
UO2 Benchmark Critical Experiments for SCALE 4.4 Methodology Critical Enrichment Poison Poison Thickness Separation (cm) Critical No. of No. Ref. General Description (w/o U235) Material (cm) Rods X Y 58 60 Multiple Fuel Clusters 4.31 None - 2.83 12.02 288 59 60 Multiple Fuel Clusters 4.31 None - 2.83 11.29 252 60 60 Multiple Fuel Clusters 4.31 None - 2.83 10.86 234 61 60 Multiple Fuel Clusters 4.31 None - 2.83 8.38 225 62 60 Multiple Fuel Clusters 4.31 None - 2.83 0 219.2 64 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 2.83 247.1 65 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 4.54 270 66 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 3.38 252 67 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 6.49 342 68 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 9.96 432 69 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 11.55 450 6D 60 Multiple Fuel Clusters 4.31 None - 2.83 2.83 221.3 70 60 Multiple Fuel Clusters 4.31 SS-304 .302 2.83 8.10 396 71 60 Mulriple Fuel Clusters 4.31 SS-304 .485 2.83 2.83 271.8 72 60 Multiple Fuel Clusters 4.31 SS-304 .485 2.83 4.47 306 73 60 Multiple Fuel Clusters 4.31 SS-304 .485 2.83 8.36 432 83 60 Multiple Fuel Clusters 4.31 Boraflex .452 2.83 2.83 642.5 84 60 Multiple Fuel Clusters 4.31 Boraflex .452 2.83 6.61 669.8 85 60 Multiple Fuel Clusters 4.31 Boraflex .452 2.83 8.5 675.9 94 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 8.5 663.3 (11 NOV 2006)
McGuire Nuclear Station UFSAR Table 4-7 (Page 3 of 5)
UO2 Benchmark Critical Experiments for SCALE 4.4 Methodology Critical Enrichment Poison Poison Thickness Separation (cm) Critical No. of No. Ref. General Description (w/o U235) Material (cm) Rods X Y 95 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 4.72 633.5 96 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 3.6 616 97 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 2.83 601 98 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 2.83 597.9 100 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 4.72 631.2 101 60 Multiple Fuel Clusters 4.31 Boraflex .226 2.83 6.61 650.8 105 60 Multiple Fuel Clusters 4.31 Boraflex .452 2.83 2.83 643.1 106 60 Multiple Fuel Clusters 4.31 Boraflex .452 2.83 4.94 660 107 60 Multiple Fuel Clusters 4.31 Boraflex .452 2.83 6.61 672.2 131 60 Multiple Fuel Clusters 4.31 None - 12.27 N/A 3-12x16 Enrichment Pin Lattice Lattice Width Critical No. of No. Ref. General Description (w/o U235) Non-fuel Pins Spacing (cm) (rods) Rods 43 60 Single Lattice 4.31 None 1.892 17 218.6 45 60 Single Lattice 4.31 None 1.892 14 216.2 46 60 Single Lattice 4.31 None 1.892 12 225.8 47 60 Single Lattice 4.31 25 water holes 1.892 14 167.6 48 60 Single Lattice 4.31 25 Al clad voids 1.892 14 203.0 4C 60 Single Lattice 4.31 None 1.892 18 223.0 96 60 Single Lattice 2.35 None 1.684 23 523.9 97 60 Single Lattice 2.35 25 water holes 1.684 23 485.8 (11 NOV 2006)
McGuire Nuclear Station UFSAR Table 4-7 (Page 4 of 5)
Critical Distance from SS Spacing Enrichment plate to Fuel Length by widths Between No. Ref. General Description (w/o U235) Poison Material Cluster (cm) of Array Clusters (cm) 14 61 3 x 1 Arrays 2.35 None - 20 x 16 8.42 15 61 3 x 1 Arrays 2.35 None - 20 x 17 11.92 21 61 3 x 1 Arrays 2.35 None - 20 x 14 4.46 Distance from SS Critcal Spacing Enrichment Poison Poison plate to Fuel Length by Width Between No. Ref. General Description (w/o U235) Material Thickness Cluster (cm) of Array Clusters (cm) 26 61 3 x 1 Arrays 2.35 SS-304 0.302 4.04 20 x 16 7.76 27 61 3 x 1 Arrays 2.35 SS-304 0.302 0.64 20 x 16 7.42 34 61 3 x 1 Arrays 2.35 SS-304 0.302 0.64 20 x 17 10.44 35 61 3 x 1 Arrays 2.35 SS-304 0.302 4.04 20 x 17 11.47 5 61 3 x 1 Arrays 2.35 SS-304 0.485 2.73 20 x 16 7.64 28 61 3 x 1 Arrays 2.35 SS-304 0.485 0.64 20 x 16 6.88 29 61 3 x 1 Arrays 2.35 SS-304 0.485 4.04 20 x 16 7.51 Flux Trap to Fuel Separation Boral (cms)
Enrichment Poison Loading Flux Trap Width Critical No. of No. Ref. General Description (w/o U235) (g B/cm2) (cm) X Y Rods 214 62 Neutron Flux Traps 4.31 0.36 3.73 0.295 0.295 952 (11 NOV 2006)
McGuire Nuclear Station UFSAR Table 4-7 (Page 5 of 5)
Flux Trap to Fuel Separation Boral (cms)
Enrichment Poison Loading Flux Trap Width Critical No. of No. Ref. General Description (w/o U235) (g B/cm2) (cm) X Y Rods 223 62 Neutron Flux Traps 4.31 0.36 3.73 4.077 4.077 858 224 62 Nuetron Flux Traps 4.31 0.36 3.73 2,186 2.186 874 229 62 Neutron Flux Traps 4.31 0 3.81 0.295 0.295 308 230 62 Neutron Flux Traps 4.31 0.05 3.75 0.295 0.295 855 Note:
McGuire Nuclear Station UFSAR Tables 4 4-11 (Page 1 of 1)
Table 4-8. Deleted Per 1996 Update Table 4-9. Deleted Per 1996 Update Table 4-10. Deleted Per 1996 Update Table 4-11. Deleted Per 1996 Update (14 OCT 2000)
McGuire Nuclear Station UFSAR Table 4-12 (Page 1 of 1)
Table 4-12. Axial Stability Index Pressurized Water Reactor Core With a 12 Foot Height Stability Index (hr-1)
Burnup (MWD/MTU) FZ CB (ppm) Exp Calc 1550 1.34 1065 -0.041 -0.032 7700 1.27 700 -0.014 -0.006 Difference: +0.027 +0.026 (14 OCT 2000)
McGuire Nuclear Station UFSAR Table 4 4-20 (Page 1 of 1)
Table 4-13. Deleted Per 1998 Update Table 4-14. Deleted Per 1998 Update Table 4-15. Deleted Per 1992 Update Table 4-16. Deleted Per 1992 Update Table 4-17. Deleted Per 1992 Update Table 4-18. Deleted Per 1992 Update Table 4-19. Deleted Per 2000 Update Table 4-20. Deleted Per 1993 Update (14 OCT 2000)
McGuire Nuclear Station UFSAR Table 4-21 (Page 1 of 1)
Table 4-21. Void Fractions at Nominal Reactor Conditions With Design Hot Channel Factors Average Maximum Core 0.0 -
Hot Subchannel 0.3 1.0 (14 OCT 2000)
McGuire Nuclear Station UFSAR Table 4-22 (Page 1 of 1)
Table 4-22. Statistically Combined Uncertainty Factors for Fq, FDeltaH, and Fz Uncertainty MODEL Uncertainty Factor Factor Value Fq-SCUF CASMO-3/SIMULATE-3P 1.071 FH-SCUF CASMO-3/SIMULATE-3P 1.040 Fz-SCUF CASMO-3/SIMULATE-3P 1.053 Low Enriched Uranium (LEU) fuel Fq-SCUF CASMO-4/SIMULATE-3 MOX 1.0735 FH-SCUF CASMO-4/SIMULATE-3 MOX 1.04 (SCD) 1.032 (non-SCD)(2)
Fz-SCUF CASMO-4/SIMULATE-3 MOX 1.049 Mixed Oxide (MOX) Fuel Fq-SCUF CASMO-4/SIMULATE-3 MOX 1.078 FH-SCUF CASMO-4/SIMULATE-3 MOX 1.04 (SCD) 1.035 (non-SCD)(2)
Fz-SCUF CASMO-4/SIMULATE-3 MOX 1.049 Note:
- 1. The CASMO-4/SIMULATE-3 MOX uncertainties are based on values in DPC-NE-1005-P-A, the values shown above have been increased to ensure that they remain bounding.
- 2. Non-SCD FH-SCUF excludes engineering hot channel factor uncertainty.
(14 APR 2005)