LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Tables 4.3-1 to 4.3-11

Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Tables 4.3-1 to 4.3-11

ML17046A338
Person / Time
Site:	Salem
Issue date:	01/30/2017
From:	Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17046A230	List: LR-N17-0034, Redacted - Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 3.8-1 Through 3.8-56 LR-N17-0034, Redacted - Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 6.3-1A Through 6.3-4 LR-N17-0034, Redacted - Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 9.1-1 Through 9.1-9 LR-N17-0034, Redacted Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 5.5-1 Through 5.5-7 LR-N17-0034, Redacted-Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 2.2, Nearby Industrial, Transportation, and Military Facilities LR-N17-0034, Redacted-Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Section 3.6, Figure 3.6-1 Through Figure 3.6-30 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 15.3A, Generic Small Break Analysis LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 3.6A, Description of Backdraft Damper LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 3.9A, Bolted Connections for Linear Component Support LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 3A, Pse&G Positions on Us NRC Regulatory Guides LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 7A, Program Element Descriptions Resulting from the Implementation of the May 6, 1983 Order LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix a, TMI Lesson Learned LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 15, Figures 15.4-2 to 15.4-111 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 15, Section 15.5, Anticipated Transients Without Scram LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 17, Figures 17.2-1 to 17.2-4 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.2-1 to 4.2-25 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.3-1 to 4.3-35 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.4-1 to 4.4-17 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.4.1 to 4.4-66d LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Figures 4.5-1 to 4.5-4 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Section 4.5, Reload Analyses LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Sections 4.3.1 to 4.3-61 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Sections 5.4.1 to 5.4-14 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Tables 4.2-1 to 4.2-2 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Tables 4.3-1 to 4.3-11 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Tables 4.4-1 to 4.4-4 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 4, Tables 4.5-1 to 4.5-2 (Deleted) LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Figures 5.1-1 to 5.1-13 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Figures 5.2-1 to 5.2-24 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Sections 5.1 to 5.1-8 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Sections 5.2.1 to 5.2-75 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Sections 5.3.1 to 5.3-2 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Table 5.1-1, System Design and Operating Parameters LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Chapter 5, Tables 5.2-1 to 5.2-42 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 14.6-1 Through 14.6-2 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 15.1-1 Through 15.1-9 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 15.2-1 Through 15.2-46 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 15.3-1 Through 15.3-68 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 3.7.1 Through 3.7-13 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 3.9-1 Through 3.9-2 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 3A-1 Through 3A-4 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 7.6-1 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figure 8.2-1 (Intentionally Deleted) and Figure 8.2-2 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 10.2-1 to 10.2-4 Intentionally Deleted LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 10.3-1A Sheets 1 Through 6 of 6 Intentionally Deleted to 10.3-3, Sheets 1 & 2 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 10.4-1 Sheets 1 & 2 of 2, Intentionally Deleted to 10.4-19 LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 11.2-1A Sheets 1 Through 5 of 5 to 11.2-1B Intentionally Deleted LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 11.3-1A & B, Sheets 1, 2 & 3 of 3, Intentionally Deleted. Refer to Plant Drawings 205240 & 205340 in Dcrms LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Figures 11.4-1 to 11.4-9 ... further results
References
LR-N17-0034
Download: ML17046A338 (14)

Text

TABLE 4.3-1 REACTOR CORE DESCRIPTION Active Core Equivalent Diameter, in. Core Average Active Fuel Height, First Core (Hot), in. Height-to-Diameter Ratio Total Cross Section Area, ft2 H20/U Molecular Ratio, Lattice (Cold) Reflector Thickness and Composition Top -Water plus Steel, in. Bottom -Water plus Steel, in. Side -Water plus Steel, in. Fuel Assemblies Number Rod Array Rods per Assembly Rod Pitch, in. Overall Transverse Dimensions, in. Fuel Weight (as uo2), lb Zircaloy Weight, lb Number of Grids per Assembly Weight of Grids (Effective in Core), lb Number of Guide Thimbles per Assembly Composition of Guide Thimbles 1 of 3 SGS-UFSAR 132.7 143.7 1.09 96.06 2. 41 -vlQ 15 193 17 X 17 264 0.496 8.426 X 8.426 222,739{V5H,V+) 217,565 (RFA) 52,541 (V5H, V+) 53,847 (RFA) V5H 2 Inconel (Top & Bottom) 6 Zircaloy-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) 2324 (VSH, V+) 3248 (RFA) 24 Zircaloy-4 (V5H) Z. 1 TM o (V+, RFA) Revision 20 May 6, 2003 TABLE 4.3-1 (Cont.) REACTOR CORE DESCRIPTION Dia. of Guide Thimbles (upper ) , in. Dia. of Guide Thimbles (lower part), in. Dia. of Instrument Guide Thimbles, in. Fuel Rods Number Outside Diameter, in. Diameter Gap, in. Clad Thickness, in. Clad Material Fuel Pellets Material * (1) Fuel Enrichments w/o(l) Region 1 Region 2 Region 3A Region 3B Diameter, in. RFA Annular Pellet I.D., in. (2) Length, in. Mass of U02 Per Foot of Fuel Rod, lb/ft Rod Cluster Control Assemblies Neutron Absorber Composition, percent Diameter, in. Density, lb/in. Clad Material Clad Thickness, in. Number of Clusters, full length Number of Absorber Rods per Cluster Full Length Assembly Weight (dry), lb Burnable Absorber Rods ( 1} Material (PYREX) Outside Diameter, in. Inner Tube, OD, in. SGS-UFSAR 2 of 3 0.442 ID X 0.474 OD (V5H, V+) 0.442 ID X 0.482 OD (RFA) 0.397 ID X 0.429 OD V5H, V+) 0.397 ID X 0.439 OD (RFA) 0.442 ID X 0.474 OD V5H, V+) 0.442 ID X 0.482 OD (RFA) 50,952 0. 374 0.0065 0.0225 Zircaloy-4 (V5H) ZirloŽ (V+, RFA) uo2 Sintered 95,5 4.6 4.2 4.6 0.3225 0.155 0.530 (STD) 0.387 (VSH, V+) 0.387 (RFA solid) '21 0. 462 or 0.500 (RFA annular) (2) o. 364 (VSH, V+) 0.355 (RFA) Ag-In-Cd 80, 15, 5 0.381 0.367 Type 316L, Ionnitride Surface 0.0185 53 24 149 Borosilicate Glass 0.381 0.1815 Revision 23 October 17, 2007 Clad Material Inner Tube Material TABLE 4.3-1 (Cont.) REACTOR CORE DESCRIPTION Boron Loading (w/o B2o3 in glass rod) Weight of Boron -10 per foot of rod1 lb/ft Material (WABA) B4C Density (Fraction of Theoretical) Absorber I.D.1 in. Absorber O.D.1 in. BA Clad Material Inner Clad Thickness/ in. Inner Clad O.D., in. Outer Clad Thickness/ in. Outer Clad 0. D. I in. Gap Material Material Content Excess Reactivity Maximum Core Reactivity (Cold, Zero .. . f c 1 ) (3) Power, o yc e Stainless Steel Stainless Steel 12.5 0.00419 Al2o3 -B4C Compound 0.7 0.278 0.318 Zircaloy 0.021 0. 267 0. 026 0.381 Helium ZrB2 1.570 to 2.355 mg B10Jin. (1) 1. 200 ( 1) Typical reload values. Current values are given in the appropriate unit (2} (3) and Core Loading Plan. Robust Fuel Assembly (RFA) uses annular pellets at the top & bottom 6" of the fuel stack height. Middle 132" of fuel stack height is solid pellets. The RFA annular pellet length starting with Salem Unit 1 Region 17 and Salem Unit 2 Region 15 is 0.500 inches. Typical reload value. This parameter is function of energy requirements and number of burnable absorbers used. 3 of 3 SGS-UFSAR Revision 24 May 11, 2009 I TABLE 4.3-2 NUCLEAR DESIGN PARAMETERS Core Average Linear Power1 kW/ft1 including densification effects Total Heat Flux Hot Channel Factor, FQ Nuclear Enthalpy Rise Hot Channel Factor1 (1) Reactivity Coefficients 5.52 2. 4 0 RFA 1. 65, V5H 1. 57 Doppler Coefficient See Figures 4.3-17 and 4.3-18 ( 1) Moderator Temperature Coefficient at Operating Conditions, pcm/°F(2) Boron Coefficient in Primary Coolant1 pcm/ppm Rodded Moderator Density Coefficient at Operating Conditions, pcm/gm/cc 0 to 16 to -6 5 ::; +0.52 X 10 Cycle-specific values based on accident analysis. The Core Operating Limits Report (COLR) contains the current cycle limits. (2) -5 Note: 1 pcm (percent mille) 10 Ap where Ap is calculated from two statepoint values of keff by ln (k2/k1l. 1 of 2 SGS-UFSAR Revision 24 May 11, 2009 TABLE 4.3-2 (Cont'd) NUCLEAR DESIGN PARAMETERS eff BOL, (EOL) 0.0075 (0.0044) Control Rod Worths Rod Requirements Maximum Bank Worth, pcm Maximum Ejected Rod Worth . (3) Boron Concentratlons Refueling CB, ARI (K<0.95) To control at HZP, ARO, (K=l.O) To control at HFP, ARO, (K=l.O): 0 MWD/MTU, No Xenon 150 MWD/MTU, Eq Xenon 1000 MWD/MTU, Eq Xenon See Table 4.3-3 < 2000 See Chapter 15 2 2050 1700 1950 1400 -1700 1000 -1250 1000 -1250 (3) Typical reload values. Current cycle values are given in the appropriate NOR or COLR. 2 of 2 SGS-UFSAR Revision 24 May 11, 2009 TABLE 4.3-3 REACTIVITY REQUIREMENTS FOR ROD CLUSTER CONTROL ASSEMBLIES Reactivity Effects, Beginning-of-Life (IJ 2. Control requirements Fuel temperature (Doppler), %Ap Moderator temperature, %Ap Void, %Ap Redistribution, %Ap Rod Insertion Allowance, %Ap Rod Misalignment Relaxation Penalty, %Ap Total Control, %Ap 3. Estimated Rod Cluster Control 4. Assembly Worth (53 Rods) a. All full length assemblies inserted, %Ap b. All but one (highest worth) assemblies inserted, %Ap Estimated Rod Cluster Control Assembly credit with 10 percent adjustment to accommodate uncertainties (3b -10 percent) , %Ap 5. Shutdown margin available (4-2), %Ap 1. 32 0.11 0.01 0.50 0.50 0.12 2.56 8.595 6.85 6.17 3.61 End-of Life (lJ 1. 30 1. 25 0.05 0.85 0.50 0.12 4.07 8.00 6.30 5.67 1.60(2) Typical reload values. See Plant Operations Package Table 10 for limiting End-of-Life cycle specific rack-up. (2) The design basis minimum shutdown is 1.3 percent. 1 of 1 I SGS-UFSAR Revision 24 May 11, 2009 Burnup Fz (MWD/T) 1550 1. 34 7700 1.27 SGS-UFSAR TABLE 4.3-4 AXIAL STABILITY INDEX PWR CORE WITH A 12-FOOT HEIGHT CB (ppm) 1065 700 Difference: 1 of 1 Stability Index (hr-1) Exp -0.041 -0.014 +0.027 Calc -0.032 -0.006 +0.026 Revision 6 February 15, 1987

• TABLE 4.3-5 TYPICAL NEUTRON FLUX LEVELS (n/cm2-sec) AT FULL POWER E > 1.0 Mev 5.53 Kev < E 0.625 ev < E < 1.0 Mev < 5.53 Kev -Core Center 6.51 X 1013 1.12 X 1014 8.50 X 1013 Core Outer Radius 3.23 X 1013 5.74 X 1013 4.63 X 1013 at Midheight Core Top, on Axis 1.53 X 1013 2.42 X 1013 2.10 X 1013 Core Bottom, on Axis 2.36 X 1013 3.94 X 1013 3.50 X 1013 Pressure Vessel 2. 77 X 1010 5. 75 X 1010 6.03 X 1010 Inner Wall, Azimuthal Peak, Core Midheight 1 of 1 SGS-UFSAR

• E < .625 ev (nv)0 3.00 X 1013 8.60 X 1012 1.63 X 1013 1.46 X 1013 8.38 X 1010 Revision 6 February 15, 1987

• *

• TABLE 4.3-6 COMPARISON OF MEASURED AND CALCULATED DOPPLER DEFECTS Core Burnup Plant Fuel TyPe (MWD/MTU) 1 Air-filled 1800 2 Air-filled 7700 3 Air and 8460 helium filled (1) pcm = 105 x ln k /k 1 2 Measure 1700 1300 1200 1 of 1 SGS-UFSAR (pcm) (1) Calculated (pcm) 1710 1440 1210 Revision 6 February 15, 1987 Al Clad SS Clad Borated Total C-Metal Al Clad Unclad Total Grand Total (1) SGS-UFSAR TABLE 4.3-7 BENCHMARK CRITICAL EXPERIMENTS of (1) No. of 14 19 H2o 7 40 41 20 61 101 in Reference 25. 1 of 1 LEOPARD Keff Experimental Sucklings 1. 0012 0.9963 0.9989 0.9985 0.9995 0.9990 0.9993 0.9990 Revision 25 October 26, 2010

• *

• TABLE 4.3-8 SAXTON CORE II ISOTOPICS ROD MY, AXIAL ZONE 6 Atom Ratio Measured(!) 2a Precision U-234/U 4.65 X 10-5 +/- 29 U-235/U 5. 74 X 10-3 +/- 0.9 U-236/U 3.55 X 10 -4 +/- 5.6 U-236/U 0.99386 +/- 0.01 Pu-238/Pu 1.32 X 10 -3 +/- 2.3 Pu-239/Pu 0.73971 +/- 0.03 Pu-240/Pu 0.19302 +/- 0.2 Pu-241/Pu 6.014 X 10 -2 +/- 0.3 Pu-242/Pu 5.81 X 10 -3 +/- 0.9 Pu/U(2) 5.938 X 10 -2 +/- 0.7 Np-237/U-238 1.14 X 10-4 +/- 15 Am-241/Pu-239 1.23 X 10-2 +/- 15 Cm-242/Pu-239 1.05 X 10-4 +/- 10 Cm-244/Pu-239 1.09 X 10-4 +/- 20 (1) Reported in Reference 26 (2) Weight ratio . 1 of 1 SGS-UFSAR (%) LEOPARD Calculation 4.6o x 10-s 5.73 X 10-3 3.74 X 104 0.99385 1.222 X 10-3 0.74497 0.19102 5.74 X 10 -2 5.38 X 10 -3 5.970 X 10 -2 0.86 X 10 -4 1.08 X 10-2 1.11 X 10-4 0.98 X 10-4 Revision 6 February 15, 1987 TABLE 4.3-9

• CRITICAL BORON CONCENTRATIONS, HZP, Plant Type Measured 2-Loop, 121 Assemblies 10 foot core 1583 2-Loop, 121 Assemblies 12 foot core 1625 2-Loop, 121 Assemblies 12 foot core 1517 3-Loop, 157 Assemblies 12 foot core 1169 *

• 1 of 1 SGS-UFSAR BOL Calculated 1589 1624 1517 1161 Revision 6 February 15, 1987

• *

• TABLE 4.3-10 COMPARISON OF MEASURED AND CALCULATED ROD WORTH 2-Loop Plant, 121 Assemblies, 10-foot Core Group B Group A Shutdown Measured (pcm) Calculated (pcm) 1885 1893 1530 1649 3050 2917 ESADA Critical (1), 0.69 inch Pitch, 240 2 w/o Puo2, 8 percent Pu , 9 Control Rods 6.21 in. rod separation 2.07 in. rod separation 1.38 in. rod separation (1) Reported in Reference 27 . SGS-UFSAR 2250 4220 4100 1 of 1 2250 4160 4010 Revision 6 February 15, 1987

• *

• TABLE 4. 3-11 COMPARISON OF MEASURED AND CALCULATED MODERATOR COEFFICIENTS AT HZP, BOL Plant Type/ Control Bank Configuration Measured a. (1) (pcm/oF) lSO 3-Loop, 157 Assemblies 12 foot core D at 160 steps D in, C at 190 steps D in, C at 28 steps B, C, and D in 2-Loop, 121 Assemblies 12 foot core D at 180 steps D in, C at 180 steps C and D in, B at 165 B, C, and D in A at 174 steps -0.50 -3.01 -7.67 -5.16 +0.85 -2.40 steps -4.40 -8.70 Calculated a. (pcmfOF) lSO -0.50 -2.75 -7.02 -4.45 +1.02 -1.90 -5.58 -8.12 (1) Isothermal coefficients, which include the Doppler effect in the fuel. aiso = 105 In k2/kl I aT°F 1 of 1 SGS-UFSAR Revision 6 February 15, 1987