LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Table 4.1-1 Through 4.1-3

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Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Table 4.1-1 Through 4.1-3
ML17046A311
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Issue date: 01/30/2017
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LR-N17-0034
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  • TABLE 4.1-1 THERMAL AND HYDRAULIC DESIGN Reactor Core Heat Output, MWt Reactor Core Heat Output, 106 Btu/hr Heat Generated in the Fuel, % Nominal System Pressure, psia Assumed Initial System Pressure for DNB Transients, psia Minimum DNBR for Design Transients ONE Correlation Coolant Flow 3459 11, 806 97.4 2250 2218 2250 V-5H(3) RFA(7) RFA Total Thermal Design Flow Rate, 106 lb/hr 125.3(S} Effective Flow Rate for Heat Transfer, 106 lb/hr 116.3(S} Effective Flow Area for Heat Transfer, ft2 Average Velocity Along Fuel Rods, ft/sec Average Mass Velocity, 106 lb/hr-tt2 Page 1 of 5 SGS-UFSAR V-5H(3) RFA V-5H(3) RFA V-5H(3) RFA (STDP(l)) (RTDP(2)) 1. 24 (RTDP) 1. 24 (RTDP, Typ) 1. 22 (RTDP I Thm) WRB-1 WRB-2 51.3 51.1 14.1 14.2 2.27(S) 2.28{S) Revision 19 November 19, 2001 I I I
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  • TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Coolant Temperature Nominal Inlet, °F Average Rise in Vessel, °F Average Rise in Core, °F Average in Core, °F Average in Vessel, °F Heat Transfer Active Heat Transfer Surface Area, ft2 Heat Flux Hot Channel Factor, FQ 2 Average Heat Flux, Btu/hr-ft Maximum Heat Flux for Normal Operation, 2 Btu/hr-ft Average Thermal Output, kW/ft Maximum Thermal Output for Normal Operation, kW/ft Peak Linear Power for Determination of Protection Setpoints, kW/ft 542.7(S) 70.4 (5) 582.4(S) 577.9(S) 59,700 2.40 192,470 461,930 5.52 13.3 <22.4 Peak Fuel Center Temperature at Maximum Thermal <4700 Output for Maximum Overpower Trip Point, °F Page 2 of 5 SGS-UFSAR Revision 19 November 19, 2001
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  • TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Fuel Assemblies Design Number of Fuel Assemblies uo2 Rods per Assembly Rod Pitch, in Overall Dimension, in Weight of Fuel (as uo2) in Core, lbs Weight of Zircaloy in Core, lbs Number of Grids per Assembly Loading Technique rue.:. Rods Number in Core Outside Diameter, in Ciametral Gap, in Clad Thickness, in ::lad Material SGS-UFSAR Page 3 of 5 RCC Canless 193 264 0. 4 96 8.426 X 8.426 STD, VSH, V+ 222,739 RFA (9) 217 I 565 All STD All VSH, V+ All RFA 50913 52541 53847 STD 8 Inconel VSH 2 Inconel (Top & Bottom) 6 4 (Mid Grids) V+ 2 Inconel (Top & Bottom) 6 ZirloŽ (Mid Grids) RFA 2 Inconel (Top & Bottom) 1 Inconel (Protective Grid) 6 ZirloŽ (Mid Grids) 3 ZirloŽ (Intermediate Flow Mixing Grids) 3 Region Non-uniform 50,952 0.374 0.0065 0.0225 STD, VSH V+,RFA Zircaloy-4 ZirloŽ Revision 18 April 26, 2000 TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Fuel Pellets Material Density, % of Theoretical Diameter, in RFA Annular Pellet I.D., in Length, in Rod Cluster Control Assemblies Neutron Absorber Cladding Material Clad Thickness, in Number of Clusters Number of Absorbers per Cluster Core Structure Core Barrel, ID I OD, in Thermal Shield, ID I OD, in Nuclear Design Parameters: Structure Characteristics Core Diameter, in (Equivalent) Core Average Active Fuel Height, in Page 4 of 5 SGS-UFSAR uo2 Sintered 95.5 0.3225(lO) 0.155{ll) STD 0.530 V-5H(3) 0.387 RFA Solid 0.387 RFA Annular 0.462 or 0.500(12} Ag-In-Cd Type 316L Ionnitride Surface 0.0185 53 24 148.0 I 152.5 158.5 I 164.0 132.7 143.7 Revision 20 May 6, 2003 I TABLE 4.1-1 (Continued) THERMAL AND HYDRAULIC DESIGN Reflector Thickness and Composition Top -Water Plus Steel, in Bottom -Water Plus Steel, in -10 Side -Water Plus Steel, in -15 H20/U, Molecular Ratio, Lattice (cold) 2.41 (1) Standard Thermal Design Procedure. (2) Revised Thermal Design Procedure. (3) Also valid for V+ assemblies without Intermediate Flow Mixing Grids. (4) Deleted (5) For analyses where high average core temperature is bounding. (6) Deleted (7) With Intermediate Flow Mixing Grids. (8) Deleted (9) With annular axial blankets. (10) Applicable to solid or annular pellets. (11) Top and bottom 6n of RFA fuel stack height. (12) Starting with Unit 1 Region 17 and Unit 2 Region 15. Page 5 of 5 SGS-UFSAR Revision 20 May 6, 2003 Analysis Mechanical Design of core Internals Loads, Deflections, and Stress Analysis Fuel Rod Design Fuel Performance Characteristics (temperature, internal pressure, clad stress, etc.) Nuclear Design 1) Cross Sections and Group Constants 2) X-Y and X-Y-Z Power Distributions, Fuel Depletion, Critical Boron Concentrations, x-y and X-Y-Z Xenon Distributions, Reactivity Coefficients 3) Axial Power Distributions Control Rod Worths, and Axial Xenon Distribution SGS-UFSAR TABLE 4.1-2 ANALYTIC TECHNIQUES IN CORE DESIGN Technique Static and Dynamic Modeling Semi-empirical thermal model of fuel rod with consideration of fuel density changes, heat transfer, fission gas release, etc. Microscopic data Macroscopic constants for homogenized core regions Group constants for control rods with self-shielding 2-Group Diffusion Theory 1-D, 2-Group Diffusion Theory 1 of 2 Computer code Blowdown code, FORCE, Finite element structural analysis code, and others Westinghouse fuel rod design model Modified ENDF/B library LEOPARD/CINDER type or PHOENIX-P HAMMER-AIM or PHOENIX-P TURTLE (2-D) or ANC(2-D or 3-D) PANDA or APOLLO Section Referenced 4.2.1.3.1 4.3.3.1 4.4.2.2 4.4.3.4.2 4.3.3.2 4.3.3.2 4.3.3.2 4.3.3.3 4.3.3.3 Revision 17 October 16, 1998
  • Analysis 4) Fuel Rod Power Effective Resonance Temperature Thermal-Hydraulic Design 1) Steady-state 2) Transient DNB Analysis SGS-UFSAR
  • TABLE 4.1-2 (Cont) Technique Integral Transport Theory Monte Carlo Weighting Function Subchannel analysis of local fluid conditions in rod bundles, including inertial and crossflow resistance terms, solu-tion progresses from core-wide to hot assembly to hot channel 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 2 of 2 Computer Code LASER REP AD THINC-IV THINC-I (THINC-III)
  • Section Referenced 4.3.3.1 4.4.3.4.1 4.4.3.4.1 Revision 6 February 15, 1987 TABLE 4.1-3 DESIGN LOADING CONDITIONS FOR REACTOR CORE COMPONENTS
  • 1. Fuel Assembly Weight 2. Fuel Assembly Spring Forces 3. Internals Weight 4. Control Rod Scram (equivalent static load) 5. Differential Pressure 6. Spring Preloads 7. Coolant Flow Forces (static) 8. Temperature 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. 13. 14. 15. One or more loops out of service All operational transients listed in Table 5.1-10. Pump overspeed Seismic loads (operation basis earthquake and design basis earthquake) 16. Blowdown forces (due to cold and hot leg break) SGS-UFSAR 1 of 1 Revision 6 February 15, 1987
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