ML20106E910

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1 to Updated Final Safety Analysis Report, Chapter 4, Appendix 4A, Tables
ML20106E910
Person / Time
Site: Catawba  Duke Energy icon.png
Issue date: 04/02/2020
From:
Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
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ML20106E897 List:
References
RA-19-0423
Download: ML20106E910 (35)
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Catawba Nuclear Station UFSAR Appendix 4A. Tables Appendix 4A. Tables

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

Table 4-1. Reactor Design Comparison Table Thermal and Hydraulic Design Parameters Unit 1 Unit 2

1. Reactor Core Heat Output, (100%), MWt 3469 3411
2. Reactor Core Heat Output, 106 Btu/hr 11836.7 11648.8
3. Heat Generated in Fuel, % 97.4 97.4
4. System Pressure, Nominal, psia(1) 2280 2280 (1)
5. System Pressure, Minimum Steady State, psia 2250 2250
6. Minimum DNBR at Nominal Conditions Limiting Channel 2.9 2.9
7. Minimum DNBR for Design Transients Limiting Channel (1) 1.55 1.45 (2) 1.50
8. DNB Correlation WRB-2M WRB-2M Core Flow(8} Unit 1 Unit 2
9. Total Thermal Flow Rate, 106 lbm/hr 145.5 144.8
10. Effective Flow Rate for Heat Transfer, 106 lbm/hr 134.6 134.0
11. Effective Flow Area for Heat Transfer, ft2 51.1 51.1
12. Average Velocity Along Fuel Rods, ft/sec 15.9 15.9
13. Average Mass Velocity, 106 lbm/hr-ft2 2.63 2.62 Coolant Temperature, °F (7)

Unit 1 Unit 2

14. Nominal Inlet 552.0 554.7
15. Average Rise in Vessel 63.2 62.5 (09 OCT 2016)

Catawba Nuclear Station UFSAR Table 4-1 (Page 2 of 3)

16. Average Rise in Core 67.8 66.4
17. Average in Core 585.9 587.9
18. Average in Vessel 585.1 587.5 Heat Transfer Unit 1 Unit 2 2
19. Active Heat Transfer, Surface Area, ft 59,866 59,866
20. Average Heat Flux, Btu/hr-ft2 192,579 189,360
21. Maximum Heat Flux for Normal Operation, Btu/hr-ft2 481,447 473,399
22. Average Linear Power, kW/ft 5.53 5.44
23. Peak Linear Power for Normal Operation, kW/ft2 13.8 13.58
24. Peak Linear Power Resulting from Overpower Transients/Operator Errors 18.0 18 (assuming a maximum overpower of 118%), kW/ft3
25. Peak Linear Power for Prevention of Centerline Melt, >18.0 >18 kW/ft
26. Power Density, kW per Liter of Core(4) 106.3 104.5
27. Specific Power, kW per kg Uranium 39.4 38.8 Fuel Central Temperature Robust Fuel Assembly
28. Peak at Peak Linear Power for Prevention of Centerline Melt, °F Burnup Dependent
29. Pressure Drop(5, 6)

Across Core, psi 28.8 +/- 2.6 Across Vessel, Including Nozzle psi 51.2 +/- 4.6 Items 30-64 Deleted duplicate information that is in Table 4-4. Moved entries that are not duplicative to Table 4-4. (i.e., Items 30, 33, 54, & 55)

(09 OCT 2016)

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

Notes:

1. Values used for thermal hydraulic core analysis.
2. This limit is associated with the value of FQ = 2.50 and includes 2.6% gamma heating.
3. See Section 4.3.2.2.6
4. Based on cold dimensions
5. Based on best estimate reactor flow rate as discussed in Section 5.1.
6. RFA pressure drops are based on Reference 98 of Section 4.4.7.
7. These values are typical values. Values are based on RCS flow of 388,000 gpm and a bypass flow of 7.5%.
8. Based on a design flow of 388,000 gpm and nominal inlet temperatures.

(09 OCT 2016)

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

Table 4-2. Analytical Techniques In Core Design Analysis Technique Computer Code Section Referenced Fuel Rod Design Fuel Performance Semiempirical thermal model PAD 4.2.3.1, Characteristics of fuel rod with consideration 4.2.3.2 (temperature, internal of fuel density changes, heat pressure,clad strain, etc.) transfer, fission gas release, 4.2.4.1 etc. 4.2.4.3 Nuclear Design

1. Cross Sections and Group Microscopic data Modified ENDF/B library 4.3.3 Constants Macroscopic constants for CASMO-3 or CASMO-4 4.3.3 homogenized core regions Group constants for control CASMO-3 or CASMO-4 4.3.3 rods with self-shielding
2. X-Y Power Diffusion Theory 4.3.3 Distributions,Fuel 3D, 2-Group SIMULATE -3 or SIMULATE-Depletion, Critical Boron Nodal Code 3 MOX Concentrations, X-Y Xenon Distributions, Reactivity Coefficients
3. Axial Power Distributions 4.3.3 Control Rod Worths, and 3D 2-Group Nodal Analysis SIMULATE-3 or SIMULATE-3 Axial Xenon Distribution Code MOX
4. Fuel Rod Power Reconstructed Integral Rod SIMULATE-3 or SIMULATE-3 4.3.4 Power MOX
5. Criticality of Reactor and 2-D, Multi-group Transport CASMO-3 4.3.2.6 Fuel Assemblies Theory 3-D Monte Carlo KENO-IV (09 OCT 2016)

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

Analysis Technique Computer Code Section Referenced Thermal-Hydraulic Design

1. Steady-state Subchannel analysis of local VIPRE-01 4.4.4.5 fluid conditions in rod bundles, including the inertial and crossflow resistance terms
2. Transient Departure from Subchannel analysis of local VIPRE-01 4.4.4.5.4 Nucleate Boiling Analysis fluid conditions in rod bundles during transients by including accumulation terms in conservation equations (09 OCT 2016)

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

Table 4-3. 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. Temperatures Gradients
9. Differences in Thermal Expansion
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 (operation basis earthquake and safe shutdown earthquake)
16. Blowdown Force (due to cold and hot leg break)

(22 OCT 2001)

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

Table 4-4. Reactor Core Description Active Core Robust Fuel Assembly Design RCC Canless Equivalent Diameter, in. 132.7 Core Average Active Fuel 144.0 Height, in.

Height-to-Diameter Ratio 1.09 Total Cross-Section Area, ft2 96.06 H2 O/U Molecular Ratio, ~2.5 Lattice (68°F, 2250 psi)

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 Barrell, 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 8.426 x 8.426(1)

Dimensions, in. (Typical)

Fuel Weight (as UO2), lbs. 220,012(1)

Zirconium Weight, lbs. 41,966(1)

(Cladding Surrounding Active Fuel)(3) 12 Composition of grids INC718 Protective Grid, 2 INC718 End Grids, 6 ZIRLO Spacer Grids, 3 ZIRLO IFM Grids Weight of Grids (Effective in INC-1066, ZIRLO-2820 Core) lbs.

(09 OCT 2019)

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

Number of Guide Thimbles per 24 Assembly Composition of Guide ZIRLO Thimbles Diameter of Guide Thimbles 0.442 I.D. x 0.482 O.D.

(upper part), in.

Diameter of Guide Thimbles 0.397 I.D. x 0.439 O.D.

(lower part), in.

Diameter of Instrument Guide 0.442 I.D. x 0.482 O.D.

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, Optimized ZIRLO Fuel Pellets Material UO2 Sintered Density (percent of 95.5 Theoretical)

Fuel Enrichments w/o 0.711-5.0 Diameter, in. 0.3225 Length, in. 0.387 (chamfered) (enriched);

0.400-0.600 (chamfered) (axial blanket)

Mass of UO2 per Foot of Fuel 0.360 (1)

Rod, lb/ft Hybrid Enhanced Performance Rod Cluster Control Assemblies (2)

Neutron Absorber B4C Diameter, in. 0.294 Density, lbs/in3 0.064 Tip Material Ag-In-Cd Composition 80 percent, 15 percent, 5 percent (Ag-In-Cd)

Diameter, in. 0.301 Length, in. 40 Density, lbs/in3 0.367 (Ag-In-Cd)

(09 OCT 2019)

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

Type 304 & 316, Cold Worked Stainless Cladding Material Steel Clad Thickness, in. 0.0385 Number of Clusters Full Length 53 Number of Absorber Rods per Cluster 24 Full Length Assembly Weight (dry), lb. 94 Chrome Coated Next Generation Rod Cluster Control Assemblies 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 Number of Absorber Rods per Cluster 24 Full Length Assembly Weight (dry), lb. 94 Hybrid Ionitrided Rod Cluster Assemblies Neutron Absorber B4C Diameter, in. 0.294 Length, in. 102 Density, lbs/in3 0.064 Tip Material Ag-In-Cd Composition 80 percent, 15 percent, 5 percent (Ag-In-Cd)

Diameter, in.

Lower Tip 0.294 Upper Tip 0.300 (09 OCT 2019)

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

Length, in.

Lower Tip 12 Upper Tip 28 Density, lbs/in3 0.367 (Ag-In-Cd)

Cladding Material Type 316 Cold Worked Stainless Steel Number of Full Length Clusters Unit 1 53 Unit 2 53 Number of Absorber Rods per Cluster 24 Full Length Assembly Weight (dry), lb. 94 Burnable Poison Rods Material Al2O3-B4C Outside Diameter, in. 0.381 Clad Material Zircaloy-4 Boron Loading Proprietary WABAs Material Al2O3-B4C Inside Diameter, in. 0.225 Outside Diameter, in. 0.381 Clad Material Zircaloy-4 Boron Loading Proprietary Notes:

1. Not exact for every core. Total weight will vary as region UO2 varies. See region specific data for the most current values.
2. Information regarding the Westinghouse Hybrid EP-RCCAs has been retained for historical purposes. These RCCAs will be retained as potential spare RCCAs.
3. The values indicated are typical Mark-BW and RFA fuel assemblies.

(09 OCT 2019)

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

Table 4-5. Nuclear Design Parameters Design Limits Core Average Linear Power, kW/ft Specified in Table 4-1, Item 22 (based on 2.6% direct moderator heating)

Total Heat Flux Hot Channel Factor, FQ Specified in the COLR Reactivity Coefficients Doppler-only Power Coefficients, pcm/%

power (See Figure 15-3) Upper Curve -19.4 to -12.6 Lower Curve -9.5 to -6.0 Fuel Temperature Coefficient, (BOL) -0.9 pcm/°F (EOL) -1.2 Moderator Temperature Coefficient, pcm/°F Most pos BOL (0-70% FP) 7.0 Most pos BOL (70-100% FP) -0.233 Most pos EOL HFP -24 Most pos EOL HZP -10 Most neg EOL HFP > -51 Boron Coefficient, pcm/ppm -5 Delayed Neutron Fraction and Lifetime l BOL - (min) µsec > 16 l BOL - (max) µsec < 22 l EOL - (min) µsec 18 l BOL - (max) µsec < 32 eff BOL - (min) > 0.0055 eff BOL - (max) < 0.0070 eff EOL - (min) > 0.0040 eff EOL - (max) < 0.0060 Control Rods Rod Worths See Table 4-7 Maximum Bank Worth, pcm See Chapter 15 Maximum Ejected Rod Worth See Chapter 15 (09 OCT 2016)

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

Note:

1. 1 pcm = (percent mille rho) = 10-5 where is calculated from two statepoint values of Keff by ln (K2/K1)

(09 OCT 2016)

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

Table 4-6. Nuclear Design Parameters. HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED Boron Concentrations (ppm) (First Cycle)

Zero Power, Keff = 1.00, Cold, Rod Cluster Control Assemblies Out, 1 percent uncertainty included 1650 Zero Power, Keff = 1.00, Hot, Rod Cluster Control Assemblies Out, 1 percent uncertainty included 1500 Design Basis Refueling Boron Concentration 2000 Zero Power, Keff = 1.00, Cold, Rod Cluster Control Assemblies In, 1 percent uncertainty included 1000 Zero Power, Keff = 1.00, Hot, Rod Cluster Control Assemblies Out 1400 Full Power, No Xenon, Keff = 1.0, Hot, Rod Cluster Control Assemblies Out 1350 Full Power, Equilibrium Xenon, Keff = 1.0, Hot Rod Cluster Control Assemblies Out 1050 Reduction with Fuel Burnup First Cycle, ppm/GWD/MTU1 See Figure 4-20.

Reload Cycle, ppm/GWD/MTU ~100 Note:

1. Gigawatt Day (GWD) = 1000 Megawatt Day (1000 MWD). During the first cycle, fixed burnable poison rods are present which significantly reduce the boron depletion rate compared to reload cycles.

(22 OCT 2001)

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

Table 4-7. Reactivity Requirements For Rod Cluster Control Assemblies. HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED End of Life Reactivity Effects, Beginning of Life End of Life (Equilibrium Cycle) percent (First Cycle) (First Cycle) (Preliminary)

1. Control requirements Fuel temperature (Doppler), percent 1.28 1.10 1.10 Moderator temperature, percent .10 0.80 1.10 Void, percent .01 .05 .05 Redistribution, percent .50 .85 .95 Rod Insertion Allowance, percent .50 .50 .50
2. Total Control, percent 2.39 3.30 3.70
3. Estimated Hybrid Rod Cluster Control Assembly Worth (53 Rods)
a. All full length assemblies inserted, percent 8.53 8.03 7.65
b. All but one (highest worth) assemblies inserted, percent 7.23 6.90 6.49
4. Estimated Rod Cluster Control Assembly credit with 10 percent 6.51 6.21 5.84 adjustment to accommodate uncertainties (3b - 10 percent), percent
5. Shutdown margin available (4-2), percent 4.12 2.91 2.14(1)

Note:

1. The design basis minimum shutdown is 1.3%

(22 OCT 2001)

Catawba Nuclear Station UFSAR Table 4-8 (Page 1 of 4)

Table 4-8. UO2 Benchmark Critical Experiments Deleted Per 2007 Update.

UO2 Critical Experiments for SCALE 4.4 Methodology Critical Separation Enrichment (CM)

Poison Thickness Critical No. of No. Ref. General Description W% U235 Poison Material (cm) X Y Rods 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 Multiple Fuel Clusters 4.31 None - 2.83 19.81 432 57 60 Multiple Fuel Clusters 4.31 None - 2.83 13.64 360 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 Critical Separation Enrichment (CM)

Poison Thickness Critical No. of No. Ref. General Description W% U235 Poison Material (cm) X Y Rods 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 (15 NOV 2007)

Catawba Nuclear Station UFSAR Table 4-8 (Page 2 of 4) 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 Mutiple Fuel Clusters 4.31 SS-304 .302 2.83 8.10 396 71 60 Multiple 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 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 Critical Separation Enrichment Poison Thickness (CM)

Critical No. of No. Ref. General Description W% U235 Poison Material (cm) X Y Rods 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 (15 NOV 2007)

Catawba Nuclear Station UFSAR Table 4-8 (Page 3 of 4)

Enrichment Pin Lattice Spacing Lattice Width Critical No. of No. Ref. General Description W% U235 Non-Fuel Pins (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 Distance from SS Critical Spacing Enrichment plate to Fuel Length by Width of Between Clusters No. Ref. General Description W% U235 Poison Material Cluster(cm) Array (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 plate to Length by Critical Spacing Enrichment Fuel Cluster General Poison Poison Width of Between No. Ref. Description W% U235 Material Thickness (cm) 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 (15 NOV 2007)

Catawba Nuclear Station UFSAR Table 4-8 (Page 4 of 4) 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 Enrichment Separation (CM)

Boral Poison Flux Trap Width No. Ref. General Description W% U235 Loading (g B/cm2) (cm) X Y Critical No. of Rods 214 62 Neutron Flux Traps 4.31 0.36 3.73 0.295 0.295 952 223 62 Neutron Flux Traps 4.31 0.36 3.73 4.077 4.077 858 224 62 Neutron 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 (15 NOV 2007)

Catawba Nuclear Station UFSAR Table 4-9 (Page 1 of 1)

Table 4-9. Axial Stability Index Pressurized Water Reactor Core With A 12 Foot Height Stability Index (hr-1)

Burnup CB (MWD/MTU) FZ (ppm) Exp Calc 1550 1.34 1065 -0.041 -0.032 7700 1.27 700 -0.014 -0.006 Difference: +0.027 +0.026 (22 OCT 2001)

Catawba Nuclear Station UFSAR Table 4-10 (Page 1 of 1)

Table 4-10. Typical Neutron Flux Levels (n/cm2-sec) At Full Power 5.53 Kev <E 1.0 6.25 ev E <5.53 E > 1.0 Mev Mev Kev E < .625 ev (nv)0 CORE CENTER 6.51 x 1013 1.12 x 1014 8.50 x 1013 3.00 x 1013 CORE OUTER RADIUS AT MIDHEIGHT 3.23 x 1013 5.74 x 1013 4.63 x 1013 8.60 x 1012 CORE TOP, ON AXIS 1.53 x 1013 2.42 x 1013 2.10 x 1013 1.63 x 1013 CORE BOTTOM, ON AXIS 2.36 x 1013 3.94 x 1013 3.50 x 1013 1.46 x 1013 PRESSURE VESSEL INNER WALL, 2.77 x 1010 5.75 x 1010 6.03 x 1010 8.38 x 1010 AZIMUTHAL PEAK, CORE MIDHEIGHT (27 MAR 2003)

Catawba Nuclear Station UFSAR Table 4-11 (Page 1 of 1)

Table 4-11. Deleted Per 1998 Update (22 OCT 2001)

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

Table 4-12. Deleted Per 2001 Update (22 OCT 2001)

Catawba Nuclear Station UFSAR Table 4-13 (Page 1 of 1)

Table 4-13. Saxton Core II Isotopics Rod My, Axial Zone 6 Atom Ratio Measured(1) 2 Precision (%) Leopard Calculation U-234/U 4.65 x 10-5 +/-29 4.60 x 10-5 U-235/U 5.74 x 10-3 +/-0.9 5.73 x 10-3 U-236/U 3.55 x 10-4 +/-5.6 3.74 x 10-4 U-238/U 0.99386 +/-0.01 0.99385 Pu-238/Pu 1.32 x 10-3 +/-2.3 1.222 x 10-3 Pu-239/Pu 0.73971 +/-0.03 0.74497 Pu-240/Pu 0.19302 +/-0.2 0.19102 Pu-241/Pu 6.014 x 10-2 +/-0.3 5.74 x 10-2 Pu-242/Pu 5.81 x 10-3 +/-0.9 5.38 x 10-3 Pu/U(2) 5.938 x 10-2 +/-0.7 5.970 x 10-2 Np-237/U-238 1.14 x 10-4 +/-15 0.86 x 10-4 Am-241/Pu-239 1.23 x 10-2 +/-15 1.08 x 10-2 Cm-242/Pu-239 1.05 x 10-4 +/-10 1.11 x 10-4 Cm-244/PU-239 1.09 x 10-4 +/-20 0.98 x 10-4 Notes:

1. Reported in Reference 29
2. Weight ratio (22 OCT 2001)

Catawba Nuclear Station UFSAR Table 4-14 (Page 1 of 1)

Table 4-14. Critical Boron Concentrations, HZP, BOL Plant Type Measured Calculated 2-Loop, 121 Assemblies 10 foot core, ppm 1583 1589 2-Loop, 121 Assemblies 12 foot core, ppm 1625 1624 2-Loop, 121 Assemblies 12 foot core, ppm 1517 1517 3-Loop, 157 Assemblies 12 foot core, ppm 1169 1161 3-Loop, 157 Assemblies 12 foot core, ppm 1344 1319 4-Loop, 193 Assemblies 12 foot core, ppm 1370 1355 4-Loop, 193 Assemblies 12 foot core, ppm 1321 1306 (27 MAR 2003)

Catawba Nuclear Station UFSAR Table 4-15 (Page 1 of 1)

Table 4-15. Benchmark Critical Experiments B4C Control Rod Worth WREC Critical No. Of Fuel Rods No. Of Control Measured1 Calculated Worth Experiment Rods Worth, %  %

2A 888 12 .395 O.D. B4C 8.20 8.37 3B 888 12 . 232 O.D. B4C 4.81 4.82 4B 884 16 . 232 O.D. B4C 6.57 6.35 5B 945 16 . 232 O.D. B4C 5.98 5.83 AG-IN-CD Comparison of Measured and Calculated Rod Worth 4-Loop Plant, 193 Assemblies, 12-foot core Measured (pcm) Calculated (pcm)

Bank D 1403 1366 Bank C 1196 1154 All Rods In Less One 6437 6460 ESADA Critical2, 0.69 Inch Pitch, 2 w% PuO2, 8% Pu240 9 Control Rods 6.21 inch rod separation 2250 2250 2.07 inch rod separation 4220 4160 1.38 inch rod separation 4100 4019 Line Item Deleted Per 2001 Update Note:

1. The measured worth was derived from the calculated value of ln k1/k2, where k1 and k2 were calculated with the measured buckling before and after insertion of the control rods, which replace fuel rods in arrays at the center of the experiment. The standard deviation in the measured worth is about 0.3% based on the uncertainties in the measured axial buckling.
2. Reported in Reference 30.

(27 MAR 2003)

Catawba Nuclear Station UFSAR Table 4-16 (Page 1 of 1)

Table 4-16. Comparison Of Measured And Calculated Moderator Coefficients At HZP, BOL Plant Type/ Control Bank Measured iso(1) Calculated iso (2)

Configuration (pcm/°F) (pcm/°F) 3-Loop, 157 Assemblies, 12 foot core D at 160 steps -0.50 -0.50 D in, C at 190 steps -3.01 -2.75 D in, C at 28 steps -7.67 -7.02 B, C, and D in -5.16 -4.45 2-Loop, 121 Assemblies, 12 foot core D at 180 steps +0.85 +1.02 D in, C at 180 steps -2.40 -1.90 C and D in, B at 165 steps -4.40 -5.58 B, C, and D in A at 174 steps -8.70 -8.12 4-loop, 193 assemblies, 12 foot core ARO -0.52 -1.2 D in -4.35 -5.7 D + C in -8.59 -10.0 D + C + B in -10.14 -10.55 D + C + B + A in -14.63 -14.45 Notes:

1. Isothermal coefficients, which include the Doppler effect in the fuel.

k2

2. iso = 10 5 ln / T° F k1 (27 MAR 2003)

Catawba Nuclear Station UFSAR Table 4-17 (Page 1 of 1)

Table 4-17. Deleted Per 2000 Update (22 OCT 2001)

Catawba Nuclear Station UFSAR Table 4-18 (Page 1 of 1)

Table 4-18. Deleted Per 1993 Update (22 OCT 2001)

Catawba Nuclear Station UFSAR Table 4-19 (Page 1 of 1)

Table 4-19. Void Fractions At Nominal Reactor Conditions With Design Hot Channel Factors HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED Average Maximum Core 0.0 Hot Subchannel 1.5 3.5 (22 OCT 2001)

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

Table 4-20. Measurements Required In The Calculation Of Reactor Flow Using A Calorimetric Technique Parameter Instrument Function

1. Feedwater venturi Rosemount P gauge and feedwater flow1 pressure differential compatible readout
2. Feedwater temperature Continuous lead feedwater enthalpy and thermocouple density1 venturi thermal expansion
3. Steam pressure Transducer and process steam enthalpy computer readout
4. Reactor coolant Thot Narrow range RTD and data RCS hot leg enthalpy acquisition system or DVM readout
5. Reactor coolant Tcold Narrow range RTD and data RCS cold leg enthalpy acquisition system or DVM RCS specific volume readout
6. Reactor coolant pressure Transducer and process RCS enthalpy and specific computer readout volume Other information required for the calculation is as follows:
7. Feedwater venturi coefficient from vendor calibration.
8. Primary system heat losses and pump heat input obtained from calculations.

Notes:

1. In addition to the originally-installed venturi flow nozzle instruments, ultrasonic flow meters were later installed on Unit 1 to provide more precise feedwater measurement. These ultrasonic flowmeters measure both feedwater flow and temperature, and provide input to the core power calorimetric calculation.

(09 OCT 2016)

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

Table 4-21. Statistically Combined Uncertainty Factors for Fq, FH, and Fz Uncertainty Uncertainty Factor MODEL 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 FSCUF 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 FSCUF excludes engineering hot channel factor uncertainty.

(24 APR 2006)

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

Table 4-22. Elbow Tap Coefficients Unit 1 Unit 2 Loop A Tap I 0.29773 0.30680 Loop A Tap II 0.29348 0.29606 Loop A Tap III 0.29515 0.30382 Loop B Tap I 0.30410 0.30313 Loop B Tap II 0.30803 0.28601 Loop B Tap III 0.30444 0.30689 Loop C Tap I 0.28915 0.31712 Loop C Tap II 0.28489 0.29659 Loop C Tap III 0.29097 0.30389 Loop D Tap I 0.30331 0.29936 Loop D Tap II 0.29932 0.29929 Loop D Tap III 0.31051 0.30137 Note:

Do not delete table. Elbow tap coefficients are committed to be included in UFSAR by Duke Letter to the NRC dated February 26, 2003 and NRC Issuance of Amendment 199 dated March 19, 2003.

(27 MAR 2003)

Catawba Nuclear Station UFSAR Table 4-23 (Page 1 of 1)

Table 4-23. Fuel Assembly Design Information for Current Demonstration Programs Parameter NGF (1) MOX (1)

Total Number of Assemblies in Test 8 4 Program Overall Transverse Dimensions, in. 8.434 8.437 (Typical)

Rod Cladding Material Optimized ZIRLO TM M5 TM Rod Length, in. 152.80 152.40 Rod Outside Diameter, in. 0.3740 0.3740 Rod Pitch, in. 0.496 0.496 Fuel Density (percent of Theoretical 95.5 95.0 Fuel Pellet Material UO2 MOX Fuel Weight (as UO2/MOX), lbs. 1139 1157 (2)

Composition of Guide Thimbles Optimized ZIRLO TM M5 TM Notes:

All values are typical or reference values for the design.

Includes plutonium and uranium dioxide.

(24 APR 2006)

Catawba Nuclear Station UFSAR Table 4-24 (Page 1 of 1)

Table 4-24. Mechanical and Thermal Hydraulic Analysis Methods for Current Demonstration Programs NGF Demonstration Program The NGF assemblies are analyzed with the same methods as those contained in UFSAR Section 4.2.3 and 4.4.1.

MOX Demonstration Program BAW-10231P-A, Rev. 1, COPERNIC Fuel Rod Design Computer Code, January 2004.

DPC-NE-2005P-A, Rev. 3, Thermal-Hydraulic Statistical Core Design Methodology, September 2002.

(24 APR 2006)

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