ML20093D599

Updated Final Safety Analysis Report, Chapter 4 Tables

ML20093D599
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/11/2019
From:	Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation
Croon G, NRR/DORL/LPL2-1, 415-1023
Shared Package
ML20093B880	List: ML20093C805 ML20093C918 ML20093C987 ML20093D240 ML20093D299 ML20093D333 ML20093D398 ML20093D477 ML20093D505 ML20093D538 ML20093D562 ML20093D599 ML20093D926 ML20093D960 ML20093D989 ML20093E060 ML20093E107 ML20093E148 ML20093E176 ML20093E208 ML20093E239 ML20093E536 ML20093E565 ML20093E584 ML20093E620 ML20093E661 ML20093E698 ML20093G903 ML20093G935 ML20093G955 ML20094N474 ML20094N501 ML20094N531 ML20094N552 ML20094P312 ML20094P348 ML20094P474 ML20094P485 ML20094P506 ML20095H450 ML20095H455 ML20095H463 ML20095H470 ML20095H483 ML20095H496 ML20095H514 ML20095H521 ML20095H529 ML20095H537
References
Download: ML20093D599 (23)

Text

McGuire Nuclear Station UFSAR Appendix 4A. Tables Appendix 4A. Tables

McGuire Nuclear Station UFSAR Table 4-1 (Page 1 of 2)

(09 OCT 2015)

Table 4-1. Reactor Design Comparison Table Thermal And Hydraulic Design Parameters Robust Fuel Assembly

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

13. Average Mass Velocity, 106 lbm/hr-ft2 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

McGuire Nuclear Station UFSAR Table 4-1 (Page 2 of 2)

(09 OCT 2015)

Thermal And Hydraulic Design Parameters Robust Fuel Assembly

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.

McGuire Nuclear Station UFSAR Table 4-2 (Page 1 of 2)

(30 NOV 2012)

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 Modeling Blowdown code, FORCE Finite element structural Fuel Rod Design Fuel Performance Characteristics (temperature, internal pressure, clad strain, etc.)

Semi-empirical thermal model of fuel rod with consideration of fuel density changes, heat transfer, fission gas release, etc.

PAD Nuclear Design

1. Cross Sections and Group Constants Microscopic and Modified ENDF/B library Macroscopic constants for homogenized core regions CASMO-3 or CASMO-4 Group constants for control rods with self-shielding CASMO-3 or CASMO-4

2. X-Y Power Distributions, Fuel Depletion, Critical Boron Concentrations, X-Y Xenon Distributions, Reactivity Coefficients Collapsed 3-D, 2-Group NEM Based Nodal Code SIMULATE-3P or SIMULATE-3 MOX

3. Axial Power Distributions, Control Rod Worths, and Axial Xenon Distribution 2-D and 3-D 2-Group Model Analysis Code SIMULATE-3P or SIMULATE-3 MOX

4. Fuel Rod Power Reconstructed Integral Rod Power SIMULATE-3P or SIMULATE-3 MOX

5. Criticality of Reactor and Fuel Assemblies 1-D, Multi-Group Transport Theory AMPX System of Codes 3-D Monte Carlo KENO-IV Thermal-Hydraulic Design

1. Steady-State Subchannel analysis of local fluid conditions in rod bundles, including inertial and crossflow resistance terms, solution progresses from core-wide to hot assembly to hot channel VIPRE-01

McGuire Nuclear Station UFSAR Table 4-2 (Page 2 of 2)

(30 NOV 2012)

Analysis Technique Computer Code

2. Transient DNB Analysis Subchannel analysis of 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 VIPRE-01

McGuire Nuclear Station UFSAR Table 4-3 (Page 1 of 1)

(14 OCT 2000)

Table 4-3. Deleted Per 1992 Update

McGuire Nuclear Station UFSAR Table 4-4 (Page 1 of 4)

(05 APR 2011)

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 psia)

~2.50 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.

(Typical) 8.426 x 8.426(1)

Fuel Weight (as UO2), lbs. (Typical)(2) 219,819(1)

Zirconium Weight, lbs. (Cladding Surrounding Active Fuel) 41,966(1)

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

McGuire Nuclear Station UFSAR Table 4-4 (Page 2 of 4)

(05 APR 2011)

Robust Fuel Assembly Inner Diameter of Guide Thimbles (upper part), in.

0.442 Outer Diameter of Guide Thimbles (upper part), in.

0.482 Inner Diameter of Guide Thimbles (lower part), in.

0.397 Outer Diameter of Guide Thimbles (lower part), in.

0.439 Inner Diameter of Instrument Guide Thimbles, in.

0.442 Outer Diameter of Instrument Guide Thimbles, in.

0.482 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)

McGuire Nuclear Station UFSAR Table 4-4 (Page 3 of 4)

(05 APR 2011)

Rod Cluster Control Assemblies(Unit 1)

Diameter, in.

Upper 0.341 Lower 0.336 Density, lbs/in. 3 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 Ionitrided Rod Cluster Control Assemblies (Unit 2)

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 Upper Tip 0.294 0.300 Length, in.

Lower Tip Upper Tip 12 28

McGuire Nuclear Station UFSAR Table 4-4 (Page 4 of 4)

(05 APR 2011)

Density, lbs/in3 0.367 Cladding Material Type 136, Cold Worked Stainless Steel, Ionitrided Cladding Thickness

.0385 Number of Full Length Clusters 53 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, Unborated Water)

Variable3 Maximum Core K (Cold, Zero Power, Beginning of Cycle) 1.304 WABAs Material Inside Diameter, in.

Outside Diameter, in.

Clad Material Boron Loading A12O3-B4 0.225 0.381 Zircaloy-4 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.

McGuire Nuclear Station UFSAR Table 4-5 (Page 1 of 2)

(14 APR 2000)

Table 4-5. Nuclear Design Parameters [HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED]

Core Average Linear Power, kW/ft, including densification effects and gamma heating effects 5.44 Total Heat Flux Hot Channel Factor, FQ 2.50 Nuclear Enthalpy Rise Hot Channel Factor, H

N F

Variable limit based on the magnitude and location of the axial peak, Fz.

Reactivity Coefficients (Reload Cycles)

Tech Spec/Safety Analysis Design Limits Best estimate Least-negative Doppler-only power coefficient, pcm/%

Power

-9.5 to -6.0

-17.5 to -8.3 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 uncertainty included 1504 2000 Zero Power, Keff = 1.00, Hot, ARO, 1 percent uncertainty included 1406 2100 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

McGuire Nuclear Station UFSAR Table 4-5 (Page 2 of 2)

(14 APR 2000)

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).

McGuire Nuclear Station UFSAR Table 4-6 (Page 1 of 1)

(14 OCT 2000)

Table 4-6. Reactivity Requirements For Rod Cluster Control Assemblies [HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED]

Reactivity Effects, percent Beginning of Life (First Cycle)

End of Life (First Cycle)

End of Life (Typical 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 percent adjustment to accommodate uncertainties (3.b.-10 percent), %

5.83 6.47 5.70 6.32 5.30

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%

4. Includes allowances for transient xenon effects

McGuire Nuclear Station UFSAR Table 4-7 (Page 1 of 5)

(11 NOV 2006)

Table 4-7. UO2 Benchmark Critical Experiments UO2 Benchmark Critical Experiments for CASMO-3, TABLES-3 and SIMULATE-3 Methodology No.

Ref.

General Description Enrichment (w/o U235)

Reflector Separating Material Characterizing 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 19 37 UO2 Rod Lattice 2.46 634 ppm Water 0.100% B/A1(1) 1.64 1.0002+/-0.0010 UO2 Benchmark Critical Experiments for SCALE 4.4 Methodology No.

Ref.

General Description Enrichment (w/o U235)

Poison Material Poison Thickness (cm)

Critical Separation (cm)

Critical No. of 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

McGuire Nuclear Station UFSAR Table 4-7 (Page 2 of 5)

(11 NOV 2006)

UO2 Benchmark Critical Experiments for SCALE 4.4 Methodology No.

Ref.

General Description Enrichment (w/o U235)

Poison Material Poison Thickness (cm)

Critical Separation (cm)

Critical No. of 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

McGuire Nuclear Station UFSAR Table 4-7 (Page 3 of 5)

(11 NOV 2006)

UO2 Benchmark Critical Experiments for SCALE 4.4 Methodology No.

Ref.

General Description Enrichment (w/o U235)

Poison Material Poison Thickness (cm)

Critical Separation (cm)

Critical No. of 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 No.

Ref.

General Description Enrichment (w/o U235)

Non-fuel Pins Pin Lattice Spacing (cm)

Lattice Width (rods)

Critical No. of 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

McGuire Nuclear Station UFSAR Table 4-7 (Page 4 of 5)

(11 NOV 2006)

No.

Ref.

General Description Enrichment (w/o U235)

Poison Material Distance from SS plate to Fuel Cluster (cm)

Length by widths of Array Critical Spacing Between 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 No.

Ref.

General Description Enrichment (w/o U235)

Poison Material Poison Thickness Distance from SS plate to Fuel Cluster (cm)

Length by Width of Array Critcal Spacing Between 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 No.

Ref.

General Description Enrichment (w/o U235)

Boral Poison Loading (g B/cm2)

Flux Trap Width (cm)

Flux Trap to Fuel Separation (cms)

Critical No. of Rods X

Y 214 62 Neutron Flux Traps 4.31 0.36 3.73 0.295 0.295 952

McGuire Nuclear Station UFSAR Table 4-7 (Page 5 of 5)

(11 NOV 2006)

No.

Ref.

General Description Enrichment (w/o U235)

Boral Poison Loading (g B/cm2)

Flux Trap Width (cm)

Flux Trap to Fuel Separation (cms)

Critical No. of Rods X

Y 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:

1. Percentages refer to weight percent boron content

McGuire Nuclear Station UFSAR Tables 4 4-11 (Page 1 of 1)

(14 OCT 2000)

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

McGuire Nuclear Station UFSAR Table 4-12 (Page 1 of 1)

(14 OCT 2000)

Table 4-12. Axial Stability Index Pressurized Water Reactor Core With a 12 Foot Height Burnup (MWD/MTU)

FZ CB (ppm)

Stability Index (hr-1)

Exp Calc 1550 1.34 1065

-0.041

-0.032 7700 1.27 700

-0.014

-0.006 Difference:

+0.027

+0.026

McGuire Nuclear Station UFSAR Table 4 4-20 (Page 1 of 1)

(14 OCT 2000)

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

McGuire Nuclear Station UFSAR Table 4-21 (Page 1 of 1)

(14 OCT 2000)

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

McGuire Nuclear Station UFSAR Table 4-22 (Page 1 of 1)

(14 APR 2005)

Table 4-22. Statistically Combined Uncertainty Factors for Fq, FDeltaH, and Fz Uncertainty Factor MODEL Uncertainty 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.