NL-16-2280, Vogtle Electric Generating Plant, Units 1 & 2, Updated Final Safety Analysis Report, Table 4.3-1 (Sheet 1 of 3) Through Figure 6-2.1-15 (Sheet 14 of 65)
Text
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-1 (SHEET 1 OF 3)
REACTOR CORE DESCRIPTION
Active core
Equivalent diameter (in.)
132.7 Active fuel height (in.)
143.7 Height-to-diameter ratio 1.08 Total cross section area (ft
- 2) 96.06 H 2 O/U molecular ratio, lattice, cold LOPAR: 2.41 VANTAGE 5: 2.73 Reflector thickness and composition
Top - water plus steel (in.)
10 Bottom - water plus steel (in.)
10 Side - water plus steel (in.)
15 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
Fuel weight, as UO 2 (lb) LOPAR: 222,762 VANTAGE 5: 204,231 (a)
Zircaloy weight (lb) (active core)
LOPAR: 45,296 Zircaloy/ZIRLO weight (lb) (active core) VANTAGE 5: 45,914 Number of grids per assembly LOPAR: 8 R type VANTAGE 5: 2 nonmixing vane type, 6 mixing vane type, 3 IFM 1 protective grid Composition of grids LOPAR: Inconel-718 VANTAGE 5: 2 Inconel-718 end grids 6 Zircaloy-4/ZIRLO spacer grids 3 Zircaloy-4/ZIRLO IFM grids 1 Inconel-718 protective grid Weight of grids in active core (lb) LOPAR: Inconel-718 - 2324 VANTAGE 5: Inconel-718 - 332 Zircaloy-4/ZIRLO 3547 Number of guide thimbles per assembly 24 Composition of guide thimbles Zircaloy-4/ZIRLO VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-1 (SHEET 2 OF 3)
Diameter of guide thimbles, upper part (in.) LOPAR: 0.450 ID x 0.482 OD VANTAGE 5: 0.442 ID x 0.474 OD Diameter of guide thimbles, lower part (in.) LOPAR: 0.397 ID x 0.430 OD VANTAGE 5:0.397 ID x 0.430 OD Diameter of instrument guide thimbles (in.) LOPAR: 0.450 ID x 0.482 OD VANTAGE 5:0.442 ID x 0.474 OD Fuel rods Number 50,952 Outside diameter (in.) LOPAR: 0.374 VANTAGE 5:0.360 Diameter gap (in.) LOPAR: 0.0065 VANTAGE 5:0.0062 Clad thickness (in.)
0.0225 Clad material Zircaloy-4/ZIRLO Fuel pellets (first cycle)
Material UO 2 sintered Density (% of theoretical) 95 First Cycle fuel enrichments (weight percent)
Region 1 2.10 Region 2 2.60 Region 3 3.10 Diameter (in.)
0.3225 Length (in.) Unit 1: 0.530 Unit 2: 0.387 Mass of UO 2 per ft of fuel rod (lb/ft) Unit 1: 0.366 Unit 2: 0.364 Fuel pellets (typical reload)
Material UO 2 sintered Density (% of theoretical) 95 Diameter (in.) LOPAR: 0.3225 VANTAGE 5:0.3088 (non-IFBA)
Length (in.) LOPAR: 0.387 VANTAGE 5:0.370 Axial Blanket Pellet:
0.462/0.500 VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-1 (SHEET 3 OF 3)
Mass of UO 2 per ft of fuel rod (lb/ft) LOPAR: 0.364 VANTAGE 5: 0.334 (a) RCCAs Neutron absorber Hafnium or Ag-In-Cd Diameter (in.)
0.341 Density (lb/in.
0.0185 Number of clusters, full-length 53 Number of absorber rods per cluster 24 BA rods (first cycle)
Number 1518 Material Borosilicate glass OD (in.)
0.381 Inner tube, OD (in.)
0.1805 Clad material SS Inner tube material SS Boron loading (without B 2 O 3 in glass rod) 12.5 Weight of boron-10 per foot of rod (lb/ft) 0.00419 Initial reactivity worth (%) ~7.6 (hot) ~5.5 (cold)
Burnable Absorbers (reload cycles)
Wet Annular Burnable Absorber Rods:
Material A1 2 O 3-B 4 C OD (in.)
0.381 Inner tube, OD (in.)
0.267 Clad material Zircaloy Inner tube material Zircaloy B 10 content (mg/cm) 6.03 Integral Fuel Burnable Absorbers:
Material ZrB 2 Typical B 10 content (mg/in.)
1.50 to 2.25 (1.0X to 1.5X) Excess reactivity (first cycle)
Maximum fuel assembly K (cold, clean, unborated water) 1.39 Maximum core reactivity (cold, zero power, beginning of cycle, zero soluble boron) 1.222
- a. The decrease in fuel weight due to annular axial blanket pellets is not considered.
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-2 (SHEET 1 OF 2)
NUCLEAR DESIGN PARAMETERS (First Cycle)
Core average linear power, including densification effects (kW/ft) 5.45 Total heat flux hot channel factor, F Q 2.32 Nuclear enthalpy rise hot channel factor, F N H 1.55 Reactivity coefficients(a) Design Limits Best Estimate Doppler-only power coefficients, see figure 15.1-5, (pcm/% power)(b)
Upper curve -19.4 to -12.6 -15 to -11 Lower curve -10.2 to -6.7 -13 to -9 Doppler temperature coefficient (pcm/
°F)(b) -2.9 to -1.4 -2.4 to -1.7 Moderator temperature coefficient (pcm/
°F)(b) 0 to -40 -1 to -36 Boron coefficient (pcm/ppm)(b) -12.8 to -7.5 -16 to -7 Rodded moderator density (pcm/g/cm)3(b) 0.43 x 10 5 0.35 x 10 5 Delayed neutron fraction and lifetime
eff BOL, (EOL) 0.0075 (0.0044)(c) *, BOL, (EOL)
µs 19.4 (18.1)
Control rods Rod requirements See table 4.3-3. Maximum bank worth (pcm)
<2000 Maximum ejected rod worth See chapter 15. Bank worth HZP no overlap (pcm)(b) BOL, Xe free EOL Eq. Xe
Bank D 650 750 Bank C 1250 1450 Bank B 1200 1400 Bank A 500 450
Radial factor (BOL to EOL)
Unrodded 1.37 to 1.28 D bank 1.50 to 1.45 D + C banks 1.60 to 1.45 D + C + B banks 1.80 to 1.55
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-2 (SHEET 2 OF 2)
Boron concentrations (ppm)
Zero power, keff = 0.99, cold(d) RCCAs out 1435 Zero power, keff = 0.99, hot(e) RCCAs out 1408 Design basis refueling boron concentration 2000 Zero power, keff 0.95, cold(d) RCCAs in 1327 Zero power, keff = 1.00, hot(e) RCCAs out 1307 Full power, no xenon, keff = 1.0, hot RCCAs out 1178 Full power, equilibrium xenon, keff = 1.0, hot RCCAs out 882 Reduction with fuel burnup First cycle (ppm/GWd/tonne uranium)(f) See figure 4.3-3. Reload cycle (ppm/GWd/tonne uranium)
~100
- a. Uncertainties are given in paragraph 4.3.3.3.
- b. 1 pcm = 10
-5 where is calculated from two statepoint values of keff by 1n (k 1/k 2). c. Bounding lower value used for safety analysis.
- d. Cold means 68
°F, 1 atm.
- e. Hot means 557
°F, 2250 psia.
- f. 1 GWd = 1000 MWd. During the first cycle, fixed BP rods are present which significantly reduce the boron depletion rate compared to reload cycles.
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-3 REACTIVITY REQUIREMENTS FOR ROD CLUSTER CONTROL ASSEMBLIES
Reactivity Effects
(Percent)
BOL (First Cycle)
EOL (First Cycle)
EOL
Representative Equilibrium Cycle)
- 1. Control requirements Fuel temperature, Doppler (%) 1.37 1.21 1.10 Moderator temperature (%)(a) 0.15 1.15 1.15 Redistribution (%) 0.50 0.85 0.98 Rod insertion allowance (%) 0.50 0.50 0.50
- 2. Total control (%) 2.52 3.71 3.73
- 3. Estimated RCCA worth (53 rods)
- a. All full-length assemblies inserted (%) 7.54 7.42 6.76 b. All assemblies but one (highest worth) inserted (%) 6.46 6.39 5.78
- 4. Estimated RCCA credit with 10
-percent adjustment to accommodate uncertainties, item 3b minus 10 percent (%) 5.82 5.75 5.20
- 5. Shutdown margin available, item 4 minus item 2 (%) 3.30 2.04 1.47(b)
- a. Includes void effects.
- b. The design basis minimum shutdown is 1.3 percent.
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-4 DELETED
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-5 AXIAL STABILITY INDEX PRESSURIZED WATER REACTOR CORE WITH A 12-FT HEIGHT
Burnup (MWd/tonne C B Stability Index (h
-1) uranium) F Z (ppm) Exp Calc 1550 1.34 1065 -0.041 -0.032
7700 1.27 700 -0.014 -0.006
5090 (a) -0.0325 -0.0255
Radial Stability Index 2250 (b) -0.068 -0.07
- a. Four-loop plant, 12-ft core in cycle 1, axial stability test.
- b. Four-loop plant, 12-ft core in cycle 1, radial (X-Y) stability test.
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-6 TYPICAL NEUTRON FLUX LEVELS (n/cm 2/s) AT FULL POWER E>1.0 MeV 0.111 MeV < E
<1.0 MeV 0.3 eV E <0.111 MeV <E 0.3 eV
Core center 9.98 x 10 13 1.11 x 10 14 2.17 x 10 14 5.36 x 10 13 Core outer radius
at midheight 4.24 x 10 13 4.85 x 10 13 9.52 x 10 13 2.21 x 10 13 Core top, on axis 2.62 x 10 13 2.13 x 10 13 1.31 x 10 14 4.35 x 10 13 Core bottom, on
axis 2.70 x 10 13 2.25 x 10 13 1.33 x 10 14 4.74 x 10 13 Pressure vessel
ID azimuthal
peak, 2.08 x 10 10 2.83 x 10 10 6.18 x 10 10 1.20 x 10 11 VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-7 COMPARISON OF MEASURED AND CALCULATED DOPPLER DEFECTS
Plant Fuel Type Core Burnup (MWd/tonne
uranium) Measured (pcm)(a) Calculated
(pcm) 1 Air filled 1800 1700 1710 2 Air filled 7700 1300 1440 3 Air and helium filled 8460 1200 1210
- a. pcm = 10 5 x ln (k /k )
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-8 SAXTON CORE II ISOTOPICS ROD MY, AXIAL ZONE 6
Atom Ratio Measured (a) 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 0.73791 +/-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 (b) 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
- a. Reported in reference 34.
- b. Weight ratio.
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-9 CRITICAL BORON CONCENTRATIONS (ppm) HZP, BOL
Plant Type Measured Calculated 2-loop, 121 assemblies, 10-ft
core 1583 1589 2-loop, 121 assemblies, 12-ft
core 1625 1624 2-loop, 121 assemblies, 12-ft
core 1517 1517 3-loop, 157 assemblies, 12-ft
core 1169 1161 3-loop, 157 assemblies, 12-ft
core 1344 1319 4-loop, 193 assemblies, 12-ft
core 1370 1355 4-loop, 193 assemblies, 12-ft
core 1321 1306 VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-10 COMPARISON OF MEASURED AND CALCULATED AG-IN-CD ROD WORTH
2-Loop Plant, 121 Assemblies, 10-ft Core Measured (pcm) Calculated (pcm)
Group B 1885 1893 Group A 1530 1649 Shutdown group 3050 2917
ESADA critical, 0.69-in. pitch (a) 2 w/o PuO 2 , 8% Pu-240, 9 control rods
6.21-in. rod separation 2250 2250 2.07-in. rod separation 4220 4160 1.38-in. rod separation 4100 4019 BENCHMARK CRITICAL EXPERIMENT HAFNIUM CONTROL ROD WORTH
Control Rod Configuration No. of Fuel Rods Measured (b) Worth (ppm B-10)
Calculated (b) Worth (ppm B-10) 9 hafnium rods 1192 138.3 141.0
- a. Reported in reference 35.
- b. Calculated and measured worths are given in terms of an equivalent charge in B-10 concentration.
VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-11 COMPARISON OF MEASURED AND CALCULATED MODERATOR COEFFICIENTS AT HZP, BOL
Plant Type/
Control Bank Configuration Measured iso (a) (pcm/°F) Calculated iso (pcm/°F) 3-loop, 157-assembly, 12-ft 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-assembly, 12-ft 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-assembly, 12-ft core
All Rods Out -0.52 -1.2 D in -4.35 -5.7 D and C in -8.59 -10.0 D, C, and B in
-10.14 -10.55 D, C, B, and A in
-14.63 -14.45
- a. Isothermal coefficients, which include the Doppler effect in the fuel.
iso = FT k kln10 1 2 5° VEGP-FSAR-4 REV 13 4/06 TABLE 4.3-12 BENCHMARK CRITICAL EXPERIMENTS
Description of
Experiments (a) Number of
Experiments LEOPARD k eff Using Experimental Bucklings
UO 2 Al clad 14 1.0012 SS clad 19 0.9963 Borated H 2 O 7 0.9989 Subtotal 40 0.9985 U-Metal
Al clad 41 0.9995 Unclad 20 0.9990 Subtotal 61 0.9993
Total 101 0.9990
- a. Reported in reference 33.
REV 13 4/06 FUEL LOADING ARRANGEMENT (INITIAL CYCLE)
FIGURE 4.3-1 (SHEET 1 OF 2)
REV 13 4/06 FUEL LOADING ARRANGEMENT (TYPICAL RELOAD CYCLE)
FIGURE 4.3-1 (SHEET 2 OF 2)
REV 13 4/06 PRODUCTION AND CONSUMPTION OF HIGHER ISOTOPES FIGURE 4.3-2
REV 13 4/06 BORON CONCENTRATION VERSUS FIRST CYCLE BURNUP WITH AND WITHOUT BA RODS FIGURE 4.3-3 (SHEET 1 OF 2)
REV 13 4/06 BORON CONCENTRATION VERSUS RELOAD CYCLE BURNUP WITH INTEGRAL FUEL BURNABLE ABSORBERS FIGURE 4.3-3 (SHEET 2 OF 2)
REV 13 4/06 ROD ARRANGEMENT WITHIN AN ASSEMBLY (WABA)
FIGURE 4.3-4 (SHEET 1 OF 2)
REV 13 4/06 TYPICAL ROD ARRANGEMENT WITHIN AN ASSEMBLY (IFBA)
FIGURE 4.3-4 (SHEET 2 OF 2)
REV 13 4/06 LOADING PATTERN (TYPICAL DISCRETE BA)
FIGURE 4.3-5 (SHEET 1 OF 2)
REV 13 4/06 LOADING PATTERN (TYPICAL IFBA)
FIGURE 4.3-5 (SHEET 2 OF 2)
REV 13 4/06 NORMALIZED POWER DENSITY DISTRIBUTION NEAR BEGINNING OF LIFE, UNRODDED, HOT FULL POWER, NO XENON FIGURE 4.3-6
REV 13 4/06 NORMALIZED POWER DENSITY DISTRIBUTION NEAR BEGINNING OF LIFE, UNRODDED, HOT FULL POWER, EQUILIBRIUM XENON FIGURE 4.3-7
REV 13 4/06 NORMALIZED POWER DENSITY DISTRIBUTION NEAR BEGINNING OF LIFE, GROUP D AT ROD INSERTION LIMITS, HOT FULL POWER, EQUILIBRIUM XENON FIGURE 4.3-8
REV 13 4/06 NORMALIZED POWER DENSITY DISTRIBUTION NEAR MIDDLE OF LIFE, UNRODDED, HOT FULL POWER, EQUILIBRIUM XENON FIGURE 4.3-9
REV 13 4/06 NORMALIZED POWER DENSITY DISTRIBUTION NEAR END OF LIFE, UNRODDED, HOT FULL POWER, EQUILIBRIUM XENON FIGURE 4.3-10
REV 13 4/06 NORMALIZED POWER DENSITY DISTRIBUTION NEAR END OF LIFE, GROUP D AT ROD INSERTION LIMITS, HOT FULL POWER, EQUILIBRIUM XENON FIGURE 4.3-11
REV 13 4/06 RODWISE POWER DISTRIBUTION IN A TYPICAL ASSEMBLY (G-10) NEAR BEGINNING OF LIFE, HOT FULL POWER, EQUILIBRIUM XENON, UNRODDED CORE FIGURE 4.3-12
REV 13 4/06 RODWISE POWER DISTRIBUTION IN A TYPICAL ASSEMBLY (G-10) NEAR THE END OF LIFE, HOT FULL POWER, EQUILIBRIUM XENON, UNRODDED CORE FIGURE 4.3-13
REV 13 4/06 TYPICAL AXIAL POWER SHAPES OCCURRING AT BEGINNING OF LIFE FIGURE 4.3-14
REV 13 4/06 TYPICAL AXIAL POWER SHAPES OCCURRING AT MIDDLE OF LIFE FIGURE 4.3-15
REV 13 4/06 TYPICAL AXIAL POWER SHAPES OCCURRING AT END OF LIFE FIGURE 4.3-16
REV 13 4/06 COMPARISON OF A TYPICAL ASSEMBLY AXIAL POWER DISTRIBUTION WITH CORE AVERAGE AXIAL DISTRIBUTION, BANK D SLIGHTLY INSERTED FIGURE 4.3-17
REV 13 4/06 FLOW CHART FOR DETERMINING SPIKE MODEL FIGURE 4.3-18
REV 13 4/06 PREDICTED POWER SPIKE DUE TO SINGLE NONFLATTENED GAP IN THE ADJACENT FUEL FIGURE 4.3-19
REV 13 4/06 POWER SPIKE FACTOR AS A FUNCTION OF AXIAL POSITION FIGURE 4.3-20
REV 13 4/06 MAXIMUM FQ = POWER VERSUS AXIAL HEIGHT DURING NORMAL OPERATION FIGURE 4.3-21
REV 13 4/06 PEAK LINEAR POWER DURING CONTROL ROD MALFUNCTION OVERPOWER TRANSIENTS FIGURE 4.3-22
REV 13 4/06 PEAK LINEAR POWER DURING BORATION DILUTION TRANSIENTS FIGURE 4.3-23
REV 13 4/06 COMPARISON BETWEEN CALCULATED AND MEASURED RELATIVE FUEL ASSEMBLY POWER DISTRIBUTION FIGURE 4.3-24
REV 13 4/06 COMPARISON OF TYPICAL CALCULATED AND MEASURED AXIAL SHAPES FIGURE 4.3-25
REV 13 4/06 MEASURED VALUES OF FQ FOR FULL-POWER ROD CONFIGURATIONS FIGURE 4.3-26
REV 13 4/06 TYPICAL DOPPLER TEMPERATURE COEFFICIENT AT BEGINNING OF LIFE AND END OF LIFE FIGURE 4.3-27
REV 13 4/06 TYPICAL DOPPLER-ONLY POWER COEFFICIENTAT BEGINNING OF LIFE AND END OF LIFE FIGURE 4.3-28
REV 13 4/06 TYPICAL DOPPLER-ONLY POWER DEFECT AT BEGINNING OF LIFE AND END OF LIFE FIGURE 4.3-29
REV 13 4/06 TYPICAL MODERATOR TEMPERATURE COEFFICIENT AT BEGINNING OF LIFE, UNRODDED FIGURE 4.3-30
REV 13 4/06 TYPICAL MODERATOR TEMPERATURE COEFFICIENT AT END OF LIFE FIGURE 4.3-31
REV 13 4/06 TYPICAL MODERATOR TEMPERATURE COEFFICIENT AS A FUNCTION OF BORON CONCENTRATION AT BEGINNING OF LIFE, UNRODDED FIGURE 4.3-32
REV 13 4/06 TYPICAL HOT FULL-POWER TEMPERATURE COEFFICIENT VERSUS CYCLE BURNUP FIGURE 4.3-33
REV 13 4/06 TYPICAL TOTAL POWER COEFFICIENT AT BEGINNING OF LIFE AND END OF LIFE FIGURE 4.3-34
REV 13 4/06 TYPICAL TOTAL POWER DEFECT AT BEGINNING OF LIFE AND END OF LIFE FIGURE 4.3-35
REV 13 4/06 ROD CLUSTER CONTROL ASSEMBLY PATTERN FIGURE 4.3-36
REV 13 4/06 TYPICAL ACCIDENTAL SIMULTANEOUS WITHDRAWAL OF TWO CONTROL BANKS AT END OF LIFE, HOT ZERO POWER, BANK D AND B MOVING IN THE SAME PLANE FIGURE 4.3-37
REV 13 4/06 DESIGN TRIP CURVE FIGURE 4.3-38
REV 13 4/06 TYPICAL NORMALIZED ROD WORTH VERSUS PERCENT INSERTION, ALL RODS OUT BUT ONE FIGURE 4.3-39
REV 13 4/06 AXIAL OFFSET VERSUS TIME, PWR CORE WITH A 12-ft HEIGHT AND 121 ASSEMBLIES FIGURE 4.3-40
REV 13 4/06 X-Y XENON TEST THERMOCOUPLE RESPONSE QUADRANT TILT DIFFERENCE VERSUS TIME FIGURE 4.3-41
REV 13 4/06 CALCULATED AND MEASURED DOPPLER DEFECT AND COEFFICIENTS AT BEGINNING OF LIFE, 2-LOOP PLANT, 121 ASSSEMBLIES, 12-ft CORE FIGURE 4.3-42
REV 13 4/06 COMPARISION OF CALCULATED AND MEASURED BORON CONCENTRATION, 2-LOOP PLANT, 121 ASSEMBLIES, 12-ft CORE FIGURE 4.3-43
REV 13 4/06 COMPARISON OF CALCULATED AND MEASURED C B, 2-LOOP PLANT, 121 ASSEMBLIES, 12-ft CORE FIGURE 4.3-44
REV 13 4/06 COMPARISON OF CALCULATED AND MEASURED C B , 3-LOOP PLANT, 157 ASSEMBLIES, 12-ft CORE FIGURE 4.3-45
REV 13 4/06 VOGTLE UNIT 1 BURNUP CREDIT REQUIREMENTS FOR ALL CELL STORAGE FIGURE 4.3-46 01,0002,0003,0004,0005,0006,0007,0008,000 9,0003.03.54.04.55.0ACCEPTABLEUNACCEPTABLEInitial 235U Enrichment (nominal w/o) Fuel Assembly Burnup (MWD\MTU)
REV 13 4/06 VOGTLE UNIT 2 BURNUP CREDIT REQUIREMENTS FOR ALL CELL STORAGE FIGURE 4.3-47
REV 13 4/06 VOGTLE UNITS 1 AND 2 EMPTY CELL CHECKERBOARD STORAGE CONFIGURATIONS FIGURE 4.3-48
REV 13 4/06 VOGTLE UNIT 2 3X3 CHECKERBOARD STORAGE CONFIGURATION FIGURE 4.3-49
REV 13 4/06 VOGTLE UNITS 1 AND 2 INTERFACE REQUIREMENTS (ALL CELL TO CHECKERBOARD STORAGE)
FIGURE 4.3-50
REV 13 4/06 VOGTLE UNIT 2 INTERFACE REQUIREMENTS (CHECKERBOARD STORAGE INTERFACE)
FIGURE 4.3-51
REV 13 4/06 VOGTLE UNIT 2 INTERFACE REQUIREMENTS (3X3 CHECKERBOARD TO ALL CELL STORAGE) FIGURE 4.3-52
REV 13 4/06 VOGTLE UNIT 2 INTERFACE REQUIREMENTS (3X3 TO EMPTY CELL CHECKERBOARD STORAGE)
FIGURE 4.3-53
REV 13 4/06 VOGTLE UNIT 1 IFBA CREDIT REQUIREMENTS FOR ALL CELL STORAGE FIGURE 4.3-54
REV 13 4/06 VOGTLE UNIT 2 BURNUP CREDIT REQUIREMENTS FOR 3-OUT-OF-4 STORAGE FIGURE 4.3-55
REV 13 4/06 VOGTLE UNIT 2 IFBA CREDIT REQUIREMENTS FOR CENTER ASSEMBLY FOR 3X3 STORAGE FIGURE 4.3-56
REV 13 4/06 VOGTLE UNIT 2 BURNUP CREDIT REQUIREMENTS FOR PERIPHERAL ASSEMBLIES FOR 3X3 STORAGE FIGURE 4.3-57
VEGP-FSAR-4 REV 15 4/09 TABLE 4.4-2 VOID FRACTIONS AT NOMINAL REACTOR CONDITIONS
Average Maximum Core (VANTAGE + /
- VANTAGE 5) < 0.01% -
Hot subchannel (VANTAGE + /
VANTAGE 5) 1.8% 7.0%
REV 15 4/09 MEASURED VERSUS PREDICTED CRITICAL HEAT FLUX (WRB-1 CORRELATION)
FIGURE 4.4-1 (SHEET 1 OF 2)
REV 15 4/09 MEASURED VERSUS PREDICTED CRITICAL HEAT FLUX (WRB-2 CORRELATION)
FIGURE 4.4-1 (SHEET 2 OF 2)
REV 14 10/07 TDC VERSUS REYNOLDS NUMBER FOR 26-in. GRID SPACING FIGURE 4.4-2
REV 14 10/07 NORMALIZED RADIAL FLOW AND ENTHALPY DISTRIBUTION AT 4-ft ELEVATION FIGURE 4.4-3
REV 14 10/07 NORMALIZED RADIAL FLOW AND ENTHALPY DISTRIBUTION AT 8-ft ELEVATION FIGURE 4.4-4
REV 14 10/07 NORMALIZED RADIAL FLOW AND ENTHALPY DISTRIBUTION AT 12-ft ELEVATION, CORE EXIT FIGURE 4.4-5
REV 14 10/07 THERMAL CONDUCTIVITY OF UO 2 (DATA CORRECTED TO 95% THEORETICAL DENSITY)
FIGURE 4.4-6
DELETED
REV 15 4/09 DELETED FIGURE 4.4-7
REV 14 10/07 RCS TEMPERATURE-PERCENT POWER MAP FIGURE 4.4-8
REV 14 10/07 DISTRIBUTION OF INCORE INSTRUMENTATION FIGURE 4.4-10
REV 14 10/07 LOOSE PARTS MONITORING SYSTEM BLOCK DIAGRAM FIGURE 4.4-11
VEGP-FSAR-4
4.5-1 REV 16 10/10 4.5 REACTOR MATERIALS 4.5.1 CONTROL ROD DRIVE SYSTEM STRUCTURAL MATERIALS 4.5.1.1 Materials Specifications All parts exposed to reactor coolant are made of metals which resist the corrosive action of the water. Three types of metals are used exclusively: stainless steels, nickel-chromium-iron, and
cobalt-based alloys. In the case of stainless steels, only austenitic and martensitic stainless
steels are used. For pressure boundary parts, martensitic stainless steels are not used in the
heat-treated conditions which cause susceptibility to stress-corrosion cracking or accelerated
corrosion in Westinghouse pressurized water reactor chemistry. Pressure boundary
parts/components are made of type 304 or equivalent material.
Internal latch assembly parts are fabricated of heat-treated martensitic stainless steel. Heat treatment is such that susceptibility to stress-corrosion cracking is not initiated. A. Pressure Vessel All pressure-containing materials comply with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code and are fabricated from austenitic (type 304) stainless steel. B. Coil Stack Assembly The coil housings require a magnetic material. Both low carbon cast steel and ductile iron have been successfully tested for this application. The choice, made on the basis of cost, indicates that ductile iron will be specified on the control rod
drive mechanism (CRDM). The finished housings are zinc plated or flame
sprayed to provide corrosion resistance. Coils are wound on bobbins of molded Dow Corning type 302 material, with double glass insulated copper wire. Coils are then vacuum impregnated with
silicon varnish. A wrapping of mica sheet is secured to the coil outside diameter.
The result is a well-insulated coil capable of sustained operation at 200
°C. C. Latch Assembly Magnetic pole pieces are fabricated from type 410 stainless steel. All nonmagnetic parts, except pins and springs, are fabricated from type 304 stainless steel. Haynes-25 is used to fabricate link pins. Springs are made from nickel-chromium-iron alloy (Inconel-X). Latch arm tips are clad with Stellite-6 to
provide improved wearability. Hard chrome plate and Stellite-6 are used
selectively for bearing and wear surfaces. D. Drive Rod Assembly The drive rod assembly utilizes a type 410 stainless steel drive rod. The coupling is machined from type 403 stainless steel. Other parts are type 304 stainless steel with the exception of the springs, which are nickel-chromium-iron alloy; the
locking button, which is Haynes-25; and the belleville washers, which are
Inconel-718. Several small parts (screws and pins) are Inconel-600.
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4.5-2 REV 16 10/10 4.5.1.2 Fabrication and Processing of Austenitic Stainless Steel Components The discussions provided in subsection 5.2.3, concerning the processes, inspections, and tests
on austenitic stainless steel components to ensure freedom from increased susceptibility to
intergranular corrosion caused by sensitization, and the discussions provided in subsection
5.2.3, concerning the control of welding of austenitic stainless steels especially control of delta
ferrite, are applicable to the austenitic stainless steel pressure-housing components of the
CRDM. 4.5.1.3 Contamination Protection and Cleaning of Austenitic Stainless Steel The CRDMs are cleaned prior to delivery in accordance with the guidance of American National Standards Institute (ANSI) 45.2.1. Process specifications in packaging and shipment are
discussed in subsection 5.2.3. Westinghouse personnel conduct surveillance of these
operations to ensure that manufacturers and installers adhere to appropriate requirements as
discussed in subsection 5.2.3. 4.5.1.4 Other Materials Haynes-25 is used in small quantities to fabricate link pins. The material is ordered in the solution-treated, cold-worked condition. Stress-corrosion cracking has not been observed in
this application over the last 15 years.
The CRDM springs are made from nickel-chromium-iron alloy (Inconel-750) ordered to MIL-S-23192 or MIL-N-24114 Class A No. 1 temper-drawn wire. Operating experience has
shown that springs made of this material are not subject to stress-corrosion cracking. 4.5.2 REACTOR INTERNALS MATERIALS 4.5.2.1 Materials Specifications All the major material for the reactor internals is type 304 stainless steel. Parts not fabricated from type 304 stainless steel include bolts and dowel pins, which were fabricated from type 316
stainless steel, and radial support key bolts, which were fabricated of Inconel-750. Radial
support clevis inserts are Inconel-600, and the holddown spring is type 403 stainless steel.
These materials are listed in table 5.2.3-2. There are no other materials used in the reactor
internals or core support structures which are not otherwise included in the ASME Code,Section III, Appendix I. 4.5.2.2 Controls on Welding The discussions provided in subsection 5.2.3 are applicable to the welding of reactor internals and core support components.
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4.5-3 REV 16 10/10 4.5.2.3 Nondestructive Examination of Tubular Products and Fittings The nondestructive examination of wrought seamless tubular products and fittings is in
accordance with Section III of the ASME Code. 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel Components The discussions provided in subsection 5.2.3 and section 1.9 verify conformance of reactor internals and core support structures with Regulatory Guide 1.44.
The discussions provided in subsection 5.2.3 and section 1.9 verify conformance of reactor internals and core support structures with Regulatory Guide 1.31.
The discussion provided in section 1.9 verifies conformance of reactor internals with Regulatory Guide 1.34.
The discussion provided in section 1.9 verifies conformance of reactor internals and core support structures with Regulatory Guide 1.71. 4.5.2.5 Contamination Protection and Cleaning of Austenitic Stainless Steel The discussions provided in subsection 5.2.3 and section 1.9 are applicable to the reactor internals and core support structures and verify conformance with ANSI 45 specifications and
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4.6-1 REV 14 10/07 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4.6.1 INFORMATION FOR CONTROL ROD DRIVE SYSTEM The control rod drive system (CRDS) is described in paragraph 3.9.4.1. Figures 3.9.4-1 and 3.9.4-2 provide the details of the control rod drive mechanisms, and figure 4.2-8 provides the layout of the CRDS. No hydraulic system is associated with its functioning. The
instrumentation and controls for the reactor trip system are described in section 7.2, and the
reactor control system is described in section 7.7. 4.6.2 EVALUATIONS OF THE CRDS The CRDS has been analyzed in detail in the failure mode and effects analysis.
(1) This study and the analyses presented in chapter 15 demonstrate that the CRDS performs its intended safety function, a reactor trip, by putting the reactor in a subcritical condition when a safety
system setting is reached, with any assumed cr edible failure of a single active component. The essential elements of the CRDS (those required to ensure reactor trip) are isolated from
nonessential portions of the CRDS (the rod control system) as described in section 7.2. The
essential portion of the CRDS is protected from the effects of postulated moderate- and
high-energy line breaks.
Despite the extremely low probability of a common mode failure impairing the ability of the reactor trip system to perform its safety function, analyses have been performed in accordance with the requirements of WASH-1270. These analyses, documented in references 2 and 3, have demonstrated that acceptable safety crit eria would not be exceeded even if the CRDS were rendered incapable of functioning during a reactor transient for which its function would
normally be expected.
The design of the control rod drive mechanism (CRDM) is such that failure of the CRDM cooling system will, in the worst case, result in an individual control rod trip or a full reactor trip (section
7.2). 4.6.3 TESTING AND VERIFICATION OF THE CRDS The CRDS is extensively tested prior to its operation. These tests may be subdivided into five categories:
- Prototype tests of components.
- Prototype CRDS tests.
- Production tests of components following manufacture and prior to installation.
- Onsite preoperational and initial startup tests.
- Periodic inservice tests.
These tests, which are described in paragraph 3.9.4.4, sections 4.2 and 14.2, and chapter 16, are conducted to verify the operability of the CRDS when called upon to function.
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4.6-2 REV 14 10/07 4.6.4 INFORMATION FOR COMBINED PERFORMANCE OF REACTIVITY SYSTEMS As is indicated in chapter 15, the only postulated events which assume credit for reactivity control systems, other than a reactor trip to render the plant subcritical, are the steam line
break, feedwater line break, and loss-of-coolant a ccident (LOCA). The reactivity control systems for which credit is taken in these accidents are the reactor trip system and the safety injection
system (SIS). Additional information on the CRDS is presented in subsection 3.9.4 and on the
SIS in section 6.3. Note that no credit is taken for the boration capabilities of the chemical and
volume control system (CVCS) as a system in the analysis of transients presented in chapter
- 15. Information on the capabilities of the CVCS is provided in subsection 9.3.4. The adverse
boron dilution possibilities due to the operation of the CVCS are investigated in subsection
15.4.6. Prior proper operation of the CVCS has been presumed as an initial condition to
evaluate transients, and appropriate technical specifications and requirements in the Technical
Requirements Manual have been prepared to ensure the correct operation or remedial action. 4.6.5 EVALUATION OF COMBINED PERFORMANCE The evaluation of the steam line break, the feedwater line break, and the LOCA, which presumes the combined actuation of the reactor trip system to the CRDS and the SIS, is
presented in subsections 15.1.5, 15.2.8, and 15.6.5. Reactor trip signals and safety injection
signals for these events are generated from functionally diverse sensors and actuate diverse
means of reactivity control, i.e., control rod insertion and injection of soluble poison.
Nondiverse but redundant types of equipment are utilized only in the processing of the incoming sensor signals into appropriate logic which initiates the protective action. This equipment is
described in detail in sections 7.2 and 7.3. In particular, note that protection from equipment
failures is provided by redundant equipment and periodic testing. Effects of failures of this
equipment have been extensively investigated.
(4) The failure mode and effects analysis described in reference 4 verifies that any single failure will not have a deleterious effect upon
the engineered safety features actuation system. Adequacy of the emergency core cooling
system and SIS performance under faulted conditions is verified in section 6.3. 4.
6.6 REFERENCES
- 1. Shopsky, W. E., "Failure Mode and Effects Analysis of the Solid State Full-Length Rod Control System," WCAP-8976 , September 1977. 2. "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330 , August 1974. 3. Gangloff, W. C., and Loftus, W. D., "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients," WCAP-7706-L (Proprietary) and WCAP-7706 (Nonproprietary), February 1971. 4. Eggleston, F. T., Rawlins, D. H., and Petrow, J. R., "Failure Mode and Effects Analysis of the Engineering Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Nonproprietary), April 1976.
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4A-1 REV 14 10/07 APPENDIX 4A RESPONSE TO NUREG-0737, II.F.2, INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING 4A.1 THE INADEQUATE CORE COOLING (ICC) MONITORING SYSTEM INSTALLED AT VEGP WILL INCLUDE THE FOLLOWING:
- Core exit thermocouple (T/C) monitoring.
- Core subcooling margin monitoring.
- Reactor vessel level monitoring.
A detailed electrical and layout description of each of the above ICC monitoring subsystems is
given below. A. Core Exit Thermocouple System The core exit thermocouple monitoring system consists of two redundant independent
trains that monitor all operable core exit thermocouples (each train monitoring up to 25 channels). A layout sketch of the system is shown in figure 4A.1. The core exit
thermocouples are mounted at the top of the core support plate. They are then routed to four upper head core exit thermocouple nozzle assemblies (CETNA) penetrations. After exiting the CETNA penetrations, the thermocouple wires proceed through a swagelok and then to qualified connectors to facilitate disconnection during removal of the upper head. Upon exiting the reactor vessel cavity, the cables are routed in a manner consistent with the requirements of Regulatory Guide 1.75 to the in-containment
qualified reference junction boxes. Each reference junction box includes three
redundant platinum resistance temperature detectors (RTDs) imbedded in a block of
copper to reflect the temperature at the junction of the chromel alumel to copper wire.
The uncompensated core exit thermocouple signals (up to 25 per train) and the
reference junction box temperatures (3) are routed to remote processing units (RPUs)
A3 and B3. The signals from both RPUs are routed to both display processing units (DPUs) for calculation of the compensated core exit thermocouple value. The value
chosen for the reference junction box temperature is a function of the data quality of
each of the RTD signals. Following the calculation of all operable compensated
thermocouple values, the information from both DPUs are transmitted to both seismically
qualified flat panel plant safety monitoring sy stem (PSMS) displays. The displays are located on section D of the Plant Vogtle control board. (See figure 18.1-1.) DPU-A and
display A are powered by train A and DPU-B, and display B are powered by train B. The
cabling between the RPUs, DPUs, and displays meets the requirements of Regulatory
Guide 1.75. B. Core Subcooling Margin Monitor The inputs to the core subcooling margin monitor include the following:
- Wide range reactor coolant system (RCS) pressure (4 channels)
- Core exit compensated thermocouple values (50 channels, excluding invalid channels)
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4A-2 REV 14 10/07
- Reference junction box RTD values (6 channels)
The electrical layout of the subcooling margin monitor is shown in figure 4A-2. One channel of wide range RCS pressure is input into each RPU channel (A2, A3, B2, and
B3). Also the uncompensated thermocouple channels and the corresponding three
reference junction box RTD signals are input into RPUs A3 and B3. The outputs of each
of the RPUs are routed to each DPU. The RCS subcooling margin is then calculated
based upon the wide range RCS pressure and compensated core exit thermocouple
readings. The value of RCS pressure utilized in the calculation is a function of the
quality of the pressure readings. The value of core exit thermocouple temperature is
based upon the auctimeered high quadrant thermocouple average reading. The
subcooling margin calculated values are routed to both displays (A and B). The cable
routing from sensor input to display meet the requirements of Regulatory Guide 1.75.
The PSMS displays are the same display panels utilized in displaying the core exit
thermocouple information. C. Reactor Vessel Level Instrumentation System The reactor vessel level instrumentation system (RVLIS) consists of two redundant
independent trains that monitor the water level in the reactor vessel.
The wide range RVLIS reading provides an indication of reactor vessel water level from the bottom of the vessel to the top of the vessel during natural circulation conditions.
The narrow range RVLIS reading provides an indication of reactor vessel water level
from the middle of the hot leg pipe to the top of the reactor vessel head during natural
circulation conditions. The dynamic head RVLIS reading provides an indication of
reactor core, internals, and outlet nozzle pressure drop for any combination of operating
reactor coolant pumps. Comparison of the measured pressure drop with the normal, single phase pressure drop provides an approximate indication of the relative void
content of the circulating fluid. The inputs to the RVLIS system include the following: 1. RCS hot leg wide range RTD's (2 channels)
- 2. Wide range RCS pressure (4 channels)
- 3. Differential pressure (6 channels)
- 4. Reference leg temperature values (14 channels)
- 5. Reactor coolant pump status (4 channels)
A fluid diagram of one train of the VEGP RVLIS system is shown in figure 4A-3 for the inputs associated solely with the RVLIS system. The electrical block diagram associated
with the RVLIS system is shown in figure 4A-4.
As discussed, the RCS hot leg wide range RTD signals are input to RPUs A2 and B1.
Also, one wide range RCS pressure channel is input into each RPU (A2, A3, B2, and
B3). In addition, one of two sets of three differential pressure signals (wide range, narrow range, and dynamic head) are input into RPU A3 and B3, respectively. Also seven
reference leg compensating temperature inputs from each train of RVLIS are input into
RPUs A3 and B3. Finally, to determine the appropriate RVLIS indication, the running
status of each reactor coolant pump is input into the non-1E RPU N1.
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4A-3 REV 14 10/07 4A.2 Several analyses have been performed to verify the design of the RVLIS system described in item 4a.1c. The results of these are discussed in the following documents: A. Summary Report, Westinghouse Reactor Ve ssel Level Instrumentation System for Monitoring Inadequate Core Cooling, December 1980, submitted to the NRC via T. M.
Anderson to Darrell G. Eisenhut, NS-TMA-2358 dated December 23, 1980. B. Responses to NRC Request for Additional Information on the Westinghouse RVLIS, Summary Report. C. Supplemental Information on the Westinghouse RVLIS, submitted to the NRC via E. P. Rahe to L. E. Phillips, NS-EPR-2579 dated March 19, 1982.
In addition to the analyses conducted in the three references above, the hydraulic components
of the RVLIS system were installed at the Semiscale Test Facility in Idaho so that transient
response characteristics could be obtained during small-break loss-of-coolant accident (LOCA)
and other accident conditions. A description of the tests conducted and a discussion of the test
results are presented in the following documents: D. Westinghouse Evaluation of RVLIS Performance at the Semiscale Test Facility, December 1981, submitted to the NRC via E. P. Rahe to L. E. Phillips, NS-EPR-2526 dated December
8, 1981. E. Westinghouse Evaluation of RVLIS Performance at the Semiscale Test Facility for Test S-UT-8, January 1982, submitted to the NRC via E. P. Rahe to L. E. Phillips, NS-EPR-2542
dated January 13, 1982. F. Westinghouse Evaluation of RVLIS performance at the Semiscale Test Facility for Test S-IB-7 submitted to the NRC via E. P. Rahe to L. E. Phillips, SED-SA-00081 dated June 28, 1982. 4A.3 A description of the tests conducted on the Westinghouse RVLIS system and the results of the tests are presented in references D, E, and F listed above.
Hardware (from sensor to computer inputs) similar to that installed on VEGP is currently
functioning at several operating plants for monitoring inadequate core cooling. The algorithms
for computing the core exit temperature, core subcooling margin, and reactor vessel level
utilized in hardware at the operating plants is similar to that implemented at VEGP. 4A.4 Response to II.F.2, Attachment I, (a) Design and Qualification Criteria for Pressurized Water Reactor Incore Thermocouples A. Attachment I provides design of the display package on the PSMS. The display package hierarchy, as summarized from Attachment I, includes the following: 1. Top level plant status summary a Westinghouse copyrighted 1985, not included as part of the FSAR.
VEGP-FSAR-4
4A-4 REV 14 10/07 2. Four lower level graphic displays a. Core temperature map b. Pressure-temperature operating limits
- d. Nuclear power 3. Four pages of menu display a. Primary Data Trend Menu
- b. Secondary Data Trend Menu
- c. Containment Data Trend Menu
- d. Detailed Data Menu 4. Four multi-page sets of data a. Six-page set of primary data trends
- b. Five-page set of secondary data trends
- c. Two-page set of containment data trends
- d. Eight-page set of detailed data B. The following provides a top down display of the core exit thermocouple information: 1. a. Maximum core exit thermocouple temperature. b. Quadrant core exit thermocouple maximum, average and maximum temperature. Also provides a comparison between the RCS hot leg RTDs and the quadrant T/C data. c. Spatially oriented core exit thermocouple map showing each thermocouple temperature. d. Alphanumeric listing of core exit thermocouple location, tag designation, and temperature reading per quadrant. e. A 2-h trend history of the four core exit thermocouple quadrant maximum temperatures. 2. The core exit thermocouple display pages are designed such that any numeric thermocouple readout greater than 1200~F will be flashed at a
frequency of 1 hertz. C. The following provides a summary of the top down display of the core subcooling margin (based upon core exit thermocouples):
VEGP-FSAR-4
4A-5 REV 14 10/07 1. a. Core subcooling margin based upon core exit thermocouples. b. RCS pressure-temperature plot exhibiting plant approach to saturation.
- c. Alphanumeric listing of both trains of core subcooling margin.
- d. A 2-h trend history of the core subcooling margin.
- 2. The core subcooling margin will indicate "SUBCOOL" when the maximum core exit thermocouple temperature is at or below the RCS coolant saturation point. "SUPERHEAT" and the appropriate numeric value in degrees F will be
displayed in reverse video when the maximum core exit thermocouple
temperature exceeds the coolant saturation temperature. D. The following provides a summary of the top down display of the RVLIS system. 1. Displays appropriate RVLIS narrow and wide range and dynamic head readings depending upon RCP status. 2. Mimic of analog meters indicating RVLIS narrow, wide, and dynamic readings with respect to reactor vessel. Only displays appropriate ranges based upon RCP status. 3. Alphanumeric listing of appropriate ranges for both trains of RVLIS system. 4. A 2-h trend history of all three RVLIS ranges. Also presents a trend of RCP status. E. Since the VEGP PSMS display system features two redundant independent displays, one display console is considered the primary display and the other display console is considered the backup display. As such, the backup display
console for ICC monitoring is also a qualified display. F. The ranges of the ICC instrumentation are given in table 7.5.2-1. 4A.5 Response to II.F.2, Appendix B, Design and Qualification Criteria for Accident Monitoring Instrumentation A. Equipment Qualification 1. Core Exit Thermocouple Monitoring Listed below are the appropriate documents indicating the qualification tests conducted on the PSMS subsystems.
Subsystem Document Al-Ch Connectors ESE-43B,C
Reference junction box ESE-44A Microprocessors ESE-53 VEGP-FSAR-4
4A-6 REV 14 10/07
Plasma display ESE-63B 2. Core Subcooling Margin Monitoring Subsystem Document Wide range RCS pressure ESE-2 Core exit thermocouples See item above
Microprocessors ESE-53 Plasma display ESE-63B 3. RVLIS Monitoring System Subsystem Document Wide range RCS pressure ESE-1A Differential pressure ESE-4 Wide range RTDs ESE-6 High volume pressure sensor ESE-48A Hydraulic isolator ESE-49A Reference leg RTDs ESE-42A Microprocessors ESE-53 Plasma display ESE-63B B. Single Failure Criteria A detailed discussion of the Regulatory Guide 1.97, Post Accident Monitoring
Design Basis, is presented in section 7.5 of the VEGP FSAR. Included in the
discussion is a justification for the number of channels selected and the diverse
variable identified where necessary. Discussed in FSAR section 7.5 is a detailed
description of the characteristics associated with each ICC monitoring system. C. RPUs A1 and A2, DPU-A, and Display A are provided by inverter power bus I.
RPUs B1 and B2, DPU-B, and display B are powered by inverter power bus II.
RPU A3 is powered by inverter power bus III, and RPU B3 is powered by inverter
power bus IV.
A sketch of signal flows between the protection channels, RPUs, DPUs, and displays is shown in figure 4A-5.
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4A-7 REV 14 10/07 4A.6 VEGP is adopting the format and content of the Westinghouse Owners Group (WOG)
Emergency Response Guidelines for writing t he plant specific procedures. Attachment II illustrates the generic WOG. These procedures provide for a critical safety function
status tree for monitoring the status of plant core cooling. All variables necessary to
implement the core cooling status tree ar e provided by the VEGP ICC instrumentation system. The functional restoration guideline, to which the operator is directed based
upon the logic dictated by the tree, also utilizes the information provided by the ICC
instrumentation. 4A.7 Evaluation of the acceptability of the location of the plant safety monitoring system (PSMS) displays will be included as part of the detailed control room design review.
REV 14 10/07 CORE EXIT THERMOCOUPLE MONITORING SYSTEM (LAYOUT SKETCH)
FIGURE 4A-1
REV 13 4/06 CORE SUBCOOLING MARGIN SYSTEM FIGURE 4A-2
REV 13 4/06 REACTOR VESSEL LEVEL INSTRUMENTATION FLUID LAYOUT DRAWING FIGURE 4A-3
REV 13 4/06 REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM FIGURE 4A-4
REV 13 4/06 VOGTLE PSMS ELECTRICAL INTERCONNECTION FIGURE 4A-5
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5.1-1 REV 19 4/15 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1
SUMMARY
DESCRIPTION This section describes the reactor coolant syst em (RCS) and includes a schematic flow diagram (figure 5.1.2-1), a piping and instrumentation diagram (drawings 1X4DB111, 2X4DB111, 1X4DB112, 2X4DB112, 1X4DB113, and 2X4DB113), and an elevation drawing (drawings 1X4DL4A17 and 2X4DL4A17). 5.1.1 DESIGN BASES The performance and safety design bases of the RCS and its major components are interrelated. These design bases are listed below: A. The RCS has the capability to transfer to the steam and power conversion system the heat produced during power operation and when the reactor is subcritical, including the initial phase of plant cooldown. B. The RCS has the capability to transfer to the residual heat removal system the heat produced during the subsequent phase of plant cooldown and cold
shutdown. C. The RCS heat removal capability under power operation and normal operational transients, including the transition from forced to natural circulation, ensures no
fuel damage within the operating bounds permitted by the reactor control and
protection systems. D. The RCS provides the water used as the core neutron moderator and reflector and as a solvent for the neutron absorber used as chemical shim control. E. The RCS maintains the homogeneity of the soluble neutron poison concentration and the rate of change of the coolant temperature so that uncontrolled reactivity
changes do not occur. F. The RCS pressure boundary is capable of accommodating the temperatures and pressures associated with operational transients. G. The reactor vessel supports the reactor core and control rod drive mechanisms. H. The pressurizer maintains the system pressure during operation and limits pressure transients. During the reduction or increase of plant load, the pressurizer accommodates volume changes in the reactor coolant. I. The reactor coolant pumps supply the coolant flow necessary to remove heat from the reactor core and transfer it to the steam generators. J. The steam generators provide high-quality steam to the turbine. The tube and tube sheet boundary are designed to prevent the transfer of radioactivity
generated within the core to the secondary system.
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5.1-2 REV 19 4/15 K. The RCS piping contains the coolant under operating temperature and pressure conditions and limits leakage (and activity release) to the containment
atmosphere. The RCS piping contains demineralized borated water that is
circulated at the flowrate and temperature consistent with achieving the reactor
core thermal and hydraulic performance. L. The RCS is monitored for loose parts, as described in subsection 4.4.6. M. Applicable industry standards and equipment classifications of RCS components are identified in table 3.2.2-1. N. The reactor vessel is provided with a head vent that meets the requirements of TMI Action Item II.B.1. (See subsections 5.4.7 and 5.4.15.) O. Unisolable sections of safety injection, normal and alternate charging, and auxiliary spray lines interconnected with the reactor coolant system, two 12-in.
residual heat removal suction lines attached to the reactor coolant loop, and the
pressurizer surge line are instrumented with resistance temperature detectors (RTDs) strapped on the pipe to detect thermal stratification. (See paragraph
5.4.3.3.4.) 5.1.2 DESIGN DESCRIPTION The reactor coolant system (RCS), shown in drawings 1X4DB111, 2X4DB111, 1X4DB112, 2X4DB112, and 1X4DB113, consists of four similar heat transfer loops connected in parallel to the reactor pressure vessel. Each loop contains a reactor coolant pump, steam generator, and
associated piping and valves. In addition, the system includes a pressurizer, pressurizer relief
and safety valves, interconnecting piping, and instrumentation necessary for operational control.
All the above components are located in the containment building.
During operation, the RCS transfers the heat generated in the core to the steam generators, where steam is produced to drive the turbine-generator. Borated demineralized water is
circulated in the RCS at a flowrate and temperature consistent with achieving the reactor core
thermal-hydraulic performance. The water also acts as a neutron moderator and reflector and
as a solvent for the neutron absorber used in chemical shim control.
The RCS pressure boundary provides a barrier against the release of radioactivity generated within the reactor and is designed to ensure a high degree of integrity throughout the life of the
plant. The RCS pressure is controlled by the use of the pressurizer where water and steam are maintained at saturation conditions by electrical heaters and water sprays. Steam can be
formed (by the heaters) or condensed (by the pressurizer spray) to minimize pressure variations
due to contraction and expansion of the reactor coolant. Spring-loaded safety valves and power-operated relief valves connected to the pressurizer provide for steam discharge from the
RCS. Discharged steam is piped to the pressurizer relief tank (pressurizer relief discharge
system), where the steam is condensed and cooled by mixing with water.
The extent of the RCS is defined as:
- The reactor vessel, including control rod drive mechanism housings.
- The portion of the steam generators containing reactor coolant.
- The reactor coolant pumps.
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5.1-3 REV 19 4/15
- The pressurizer.
- The safety and relief valves.
- The head vent.
- The interconnecting piping, valves, and fittings between the principal components listed above.
- The piping, fittings, and valves leading to connecting auxiliary or support systems.
The RCS schematic flow diagram is shown in figure 5.1.2-1. Included with this figure is a
tabulation of principal pressures, temperatures, and flowrates of the system under normal
steady-state, full-power operating conditions. These parameters are based on the best-
estimate flow at the pump discharge. The RCS volume under these conditions is presented in
table 5.1.2-1. A piping and instrumentation diagram of the RCS is shown in drawings 1X4DB111, 2X4DB111, 1X4DB112, 2X4DB112, and 1X4DB113. This diagram shows the extent of the systems located within the containment and the points of separation between the RCS and the secondary (heat utilization) system. Drawings 1X4DE312, 1X4DE313, 1X4DE314, 1X4DE317, 1X4DE320, 1X4DE322, 1X2D48E007 and 1X2D48E008 provide plan and elevation views of the containment, and drawings 1X4DL4A17 and 2X4DL4A17 show plan and section of the reactor
coolant loops. These figures show principal dimensions of RCS components in relationship to
supporting and surrounding steel and concrete structures and demonstrate the protection
provided to the RCS by its physical layout. 5.1.3 SYSTEM COMPONENTS The major components of the reactor coolant system are as follows: A. Reactor Vessel The reactor vessel is cylindrical and has a welded, hemispherical bottom head and a removable, flanged, hemispherical upper head. The vessel contains the core, core support structures, control rods, and other parts directly associated with the
core. The vessel has inlet and outlet nozzles located in a horizontal plane just below the reactor vessel flange but above the top of the core. Coolant enters the vessel
through the inlet nozzles and flows down the core barrel-vessel wall annulus, turns
at the bottom, and flows up through the core to the outlet nozzles. B. Steam Generators The steam generators are vertical shell and U-tube evaporators with integral
moisture-separating equipment. The reactor coolant flows through the inverted
U-tubes, entering and leaving through the nozzles located in the hemispherical
bottom head of the steam generator. Steam is generated on the shell side and
flows upward through the moisture separators to the outlet nozzle at the top of the vessel.
VEGP-FSAR-5
5.1-4 REV 19 4/15 C. Reactor Coolant Pumps The reactor coolant pumps (RCPs) are single-speed centrifugal units driven by
water/air-cooled, three-phase induction motors. The shaft is vertical with the motor
mounted above the pump. A flywheel on the shaft above the motor provides
additional inertia to extend pump coastdown. The flow inlet is at the bottom of the
pump, and the discharge is on the side. D. Piping The reactor coolant piping is seamless, stainless steel piping. The hot leg is
defined as the piping between the reactor vessel outlet nozzle and the steam
generator. The mechanical stress improvement process (MSIP) has been applied to the Unit 1 and Unit 2 hot legs near the reactor vessel outlet nozzle. The cold leg is defined as the piping between the RCP outlet and the reactor vessel. The
crossover leg is defined as the piping between the steam generator outlet and the
RCP inlet. E. Pressurizer The pressurizer is a vertical, cylindrical vessel with hemispherical top and bottom
heads. The pressurizer is connected to the hot leg of one of the coolant loops by
a surge line. Electrical heaters are installed through the bottom head of the
vessel. The spray nozzle and relief and safety valve connections are located in
the top head of the vessel. F. Safety and Relief Valves The pressurizer safety valves are of the totally enclosed pop type. The valves are
spring loaded and self-activated with backpressure compensation. The
power-operated relief valves are solenoid-operated valves. They are operated
automatically or by remote manual cont rol. Remotely operated gate valves are provided to isolate the inlet to the power-operated relief valves if excessive
leakage occurs. Position-indicating lights are provided in the control room for
these valves. 5.1.4 SYSTEM PERFORMANCE CHARACTERISTICS Design and performance characteristics of the reactor coolant system are provided in table 5.1.2-1. A. Reactor Coolant Flow The reactor coolant flow, a major parameter in the design of the system and its
components, is established by a detailed design procedure supported by operating
plant performance data and component hy draulics experimental data. The procedure establishes a best-estimate flow as well as conservatively high and low
flows for the applicable mechanical and thermal design considerations. In
establishing the range of design flows, the procedure accounts for the
uncertainties in the component flow resistances and the pump head-flow
capability, established by analysis of the available experimental data. The
procedure also accounts for the uncertainties in the technique used to measure
flow in the operating plant.
Definitions of the three reactor coolant flows applied in various plant design considerations are presented in the following paragraphs.
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5.1-5 REV 19 4/15 B. Best-Estimate Flow The best-estimate flow is considered to be the most likely value for the plant
operating condition. This flow is based on the best estimate of the reactor vessel, steam generator, and piping flow resistances and on the best estimate of the
reactor coolant pump (RCP) head-flow capability, with no known uncertainties
assigned to either the system flow resistance or the pump head. The
best-estimate flow provides the basis for the establishment of the other design
flows required for the system and component design. System pressure losses based on best-estimate flow are presented in table 5.1.2-1.
The best-estimate flow analysis has been based on extensive experimental data, including accurate flow and pressure drop data from one operating plant, flow
resistance measurements from several fuel assembly hydraulics tests, and hydraulic performance measurements from several pump impeller model tests.
Since operating plant flow measurements have been shown to be in close
agreement with the calculated best-estimate flows, the flows established with this
design procedure can be applied to the plant design with a high level of
confidence.
Although the best-estimate flow is the most likely value to be expected in operation, more conservative flowrates are applied in the thermal and mechanical
designs. C. Thermal Design Flow Thermal design flow is the flowrate used as a basis for the reactor core thermal
performance, the steam generator thermal performance, and the nominal plant
parameters used throughout the design. The thermal design flow accounts for the
uncertainties in flow resistances (reactor vessel, steam generator, and piping),
RCP head, and the methods used to measure flowrate. The thermal design flow is
approximately 5.8% less than the best-estimate flow with 10% equivalent steam
generator plugging. The thermal design flow is confirmed when the plant is placed
in operation. Tabulations of important design parameters based on the thermal
design flow are provided in table 5.1.2-1. D. Mechanical Design Flow Mechanical design flow is a conservatively high flow used in the mechanical
design of the reactor vessel internals and fuel assemblies. The mechanical design
flow is based on a reduced system resistance and on increased pump head
capability. The mechanical design flow is approximately 1.7% greater than the
best-estimate flow with 0% equivalent steam generator tube plugging.
Pump overspeed due to a turbine generator overspeed of 20% results in a peak reactor coolant flow of 120% of the mechanical design flow. The overspeed
condition is applicable only to operating conditions when the reactor and turbine
generator are at power.
REV 14 10/07 RCS PROCESS FLOW DIAGRAM FIGURE 5.1.2-1 (SHEET 1 OF 3)
REV 14 10/07 RCS PROCESS FLOW DIAGRAM FIGURE 5.1.2-1 (SHEET 2 OF 3)
REV 14 10/07 RCS PROCESS FLOW DIAGRAM FIGURE 5.1.2-1 (SHEET 3 OF 3)
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5.2-1 REV 19 4/15 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY This section discusses the measures employed to provide and maintain the integrity of the
reactor coolant pressure boundary (RCPB) for the plant design lifetime
- a. Section 50.2 of 10 CFR 50 defines the RCPB as extending to the outermost containment isolation valve in
system piping which penetrates the containment and is connected to the reactor coolant system (RCS). This section is limited to a description of the components of the RCS, as defined in
section 5.1, unless otherwise noted. Components which are part of the RCPB (as defined in 10
CFR 50) but are not described in this section are described in the following sections: A. Section 6.3 - RCPB components which are part of the emergency core cooling system. B. Subsection 9.3.4 - RCPB components which are part of the chemical and volume control system. C. Subsection 3.9.N.1 - Design loading, stress limits, and analyses applied to the RCS and American Society of Mechanical Engineers (ASME) Code Class 1
components. D. Subsection 3.9.N.3 - Design loadings, stress limits, and analyses applied to ASME Code Class 2 and 3 components.
The abbreviation RCS, as used in this section, is as defined in section 5.1. When the term
RCPB is used in this section, its definition is that of Section 50.2 of 10 CFR 50. 5.2.1 COMPLIANCE WITH CODES AND CODE CASES 5.2.1.1 Compliance with 10 CFR 50.55a RCS components are designed and fabricated in accordance with 10 CFR 50, Section 50.55a, Codes and Standards. The addenda of the ASME code applied in the design of each
component are listed in table 5.2.1-1. 5.2.1.2 Applicable Code Cases Regulatory Guides 1.84 and 1.85 are discussed in section 1.9. The following discussion addresses only unapproved or conditionally approved code cases (per Regulatory Guides 1.84
and 1.85) used on Class 1 components.
Code Case 1528 (SA-508, Class 2a) material has been used in the manufacture of the VEGP steam generators and pressurizers. Regulatory Guide 1.85 presently reflects a conditional
Nuclear Regulatory Commission (NRC) approval of Code Case 1528. Westinghouse has
conducted a test program which demonstrates the adequacy of Code Case 1528 material. The a The operating licenses for both VEGP units hav e been renewed and the original licensed operating terms have been extended by 20 years. In acco rdance with 10 CFR Part 54, appropriate aging management programs and activities have been initiated to manage the detrimental effects of aging to maintain functionality during the peri od of extended operation (see chapter 19).
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5.2-2 REV 19 4/15 results of the test program are documented in reference 1. Reference 1 and a request for
approval (reference 2) of the use of Code Case 1528 have been submitted to the NRC.
5.2.1.3 References 1. Letter NS-CE-1228, dated October 4, 1976, C. Eicheldinger (Westinghouse) to J. F.
Stolz (NCR). 2. Letter NS-CE-173, dated March 17, 1978, C. Eicheldinger (Westinghouse) to J. F. Stolz (NRC). 5.2.2 OVERPRESSURE PROTECTION Reactor coolant system (RCS) overpressure prot ection is provided by the pressurizer safety valves, steam generator safety valves, and t he reactor protection system and associated equipment. Combinations of these systems ensure compliance with the overpressure
requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure
Vessel Code,Section III, Paragraphs NB-7300 and NC-7300, for pressurized water reactor systems. The only portions of an auxiliary system connected to the RCS that are utilized for overpressure protection of the RCS are the liquid relief valves of the residual heat removal system (RHRS).
These valves protect the RCS at low temperatures when the RHRS is in operation. 5.2.2.1 Design Bases Overpressure protection is provided for the RCS by the pressurizer safety valves. This protection is afforded for the following events which envelope those credible events that could lead to overpressure of the RCS if adequate overpressure protection is not provided:
- Loss of electrical load and/or turbine trip.
- Uncontrolled rod withdrawal at power.
- Loss of reactor coolant flow.
- Loss of normal feedwater.
- Loss of offsite power to the station auxiliaries.
The sizing of the pressurizer safety valves is based on analysis of a complete loss of steamflow
to the turbine with the reactor operating at 102 percent of engineered safeguards design power.
In this analysis, feedwater flow is also assumed to be lost, and no credit is taken for operation of
the pressurizer power-operated relief valves (PORV), pressurizer level control system, pressurizer spray system, rod control system , steam dump system, or steam line PORV. The reactor is maintained at full power (no credit for direct reactor trip on turbine trip), and steam
relief through the steam generator safety valves is considered. The total pressurizer safety
valve capacity is required to be at least as large as the maximum surge rate into the pressurizer
during this transient.
This sizing procedure results in a safety valve capacity well in excess of the capacity required to prevent exceeding 110 percent of system desi gn pressure for the events listed above.
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5.2-3 REV 19 4/15 Overpressure protection for the steam system is provided by steam generator safety valves.
The steam system safety valve capacity is based on providing enough relief to remove the engineered safeguards design steamflow. This must be done while limiting the maximum steam
system pressure to less than 110 percent of the steam generator shell side design pressure.
Blowdown and heat dissipation systems of the nuc lear steam supply system (NSSS) connected to the discharge of these pressure relieving devices are discussed in subsection 5.4.11.
Steam generator blowdown systems are discussed in subsection 10.4.8. 5.2.2.2 Design Evaluation A description of the pressurizer safety valves performance characteristics along with the design description of the incidents, assumptions made, method of analysis, and conclusions are
discussed in chapter 15.
The relief capacities of the pressurizer and steam generator safety valves are determined from the postulated overpressure transient conditions in conjunction with the action of the reactor protection system. An evaluation of the functional design of the overpressure protection system
and an analysis of the capability of the system to perform its function for a typical plant are
presented in reference 1. The report describes in detail the types and number of pressure relief
devices employed, relief device description, locati ons in the systems, reliability history, and the details of the methods used for relief device sizing based on typical worst-case transient
conditions and analysis data for each transient condition. An overpressure protection report
specifically for the VEGP is prepared in accordance with Article NB-7300 of Section III of the
ASME Code. The description of the analytical model used in the analysis of the overpressure
protection system and the basis for its validity are discussed in reference 2.
The capacities of the pressurizer safety and relief valves are discussed in subsection 5.4.13.
The setpoints and reactor trip signals which occur during overpressure transients are discussed
in subsection 5.4.10. 5.2.2.3 Piping and Instrumentation Diagrams Overpressure protection for the RCS is provi ded by the pressurizer safety and relief valves shown in drawing 1X4DB112. These valves discharge to the pressurizer relief tank through a common manifold.
The steam system safety valves are discussed in section 10.3 and are shown in drawings 1X4DB159-1, 1X4DB159-2, and 1X4DB159-3. 5.2.2.4 Equipment and Component Description The operation, significant design parameters, number and types of operating cycles, and environmental conditions of the pressurizer safety valves are discussed in section 3.11.N and
subsections 3.9.N.1 and 5.4.13.
Section 10.3 contains a discussion of the equipment and components of the steam system overpressure protection features.
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5.2-4 REV 19 4/15 5.2.2.5 Mounting of Pressure Relief Devices The design and installation of the pressure relief devices for the RCS are described in
subsection 5.4.11. The design basis for the assumed loads for the primary and secondary side
pressure relief devices are described in section 3.9.N. Subsection 10.3.2 provides a discussion
of the main steam safety valves and the pow er-operated atmospheric steam relief valves. 5.2.2.6 Applicable Codes and Classification The requirements of the ASME Boiler and Pressure Vessel Code,Section III, Paragraphs
NB-7300 (Overpressure Protection Report) and NC-7300 (Overpressure Protection Analysis),
are met. Piping, valves, and associated equipment used for overpressure protection are classified in accordance with American National Standards Institute (ANSI) N18.2a-1975, Nuclear Safety
Criteria for the Design of Stationary Pressurized Water Reactor Plants. These safety class designations are delineated in table 3.2.2-1 and shown in drawings 1X4DB111, 2X4DB111, 1X4DB112, 2X4DB112, and 1X4DB113. 5.2.2.7 Material Specifications Refer to subsection 5.2.3 for a description of material specifications.
5.2.2.8 Process Instrumentation Each pressurizer safety valve discharge line incorporates a control board temperature indicator and alarm to notify the operator of steam discharge due to either leakage or actual valve
operation. For a further discussion on process instrumentation associated with the system, refer to chapter 7. 5.2.2.9 System Reliability The reliability of the pressure relieving devices is discussed in section 4 of reference 1.
5.2.2.10 RCS Pressure Control During Low-Temperature Operation a An important aspect of RCS overpressure protection at low temperatures is the use of administrative controls which are discussed in some detail in paragraph 5.2.2.10.2. Although specific alarms do not exist to invoke specific administrative procedures, annunciation is
provided to alert the operator to arm the cold overpressure mitigation system. Operating procedures maximize the use of a pressurizer steam bubble, since a steam bubble reduces the
maximum pressure reached for some transients, and slows the rate of pressure increase for
others, and aids the operator in controlling RCS pressure during low temperature operation.
a The cold overpressure protection system setpoi nt calculation was evaluated as a time-limited aging analysis (TLAA) for license renewal in accordance with 10 CFR Part 54. The results of this evaluation are provided in paragraph 19.4.6.4.
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5.2-5 REV 19 4/15 When the RCS is at temperatures below approximately 350
°F, it is opened to the RHRS for the purposes of removing residual heat from the core, providing a path for letdown to the purification
subsystem, and controlling the RCS pressure when the plant is operating in a water solid mode.
The RHRS is provided with self-actuated water relief valves to prevent overpressure in this relatively low design pressure system caused either within the system itself or from transients
transmitted from the RCS. The RHRS relief valves mitigate pressure transients originated in the
RCS to maximum pressure values determined by the relief valve set pressure.
The low design pressure RHRS is normally isolated from the high design pressure RCS during reactor power operation at temperatures above approximately 350
°F by two isolation valves in series. Therefore, the RHRS can be inadvertently isolated from the RCS by these same isolation valves. The PORVs and associated logic provide overpressure mitigation for those
transients which might occur if the RHRS isolation valves were inadvertently closed. The PORV
logic is manually armed at the system setpoint.
Two pressurizer PORVs are each supplied with actuation logic. The logic for each PORV continuously monitors RCS temperature and pressure, converts an auctioneered RCS
temperature to the Appendix G allowable pressure, and then compares the allowable pressure
to the actual RCS pressure. As the actual RCS pressure approaches the allowable pressure, a
main control board alarm is annunciated. If the RCS pressure continues to increase, an
actuation signal is transmitted to a PORV and the valve opens to mitigate the transient.
As described in subsection 5.4.13, the VEGP PORVs are safety related. They were designed in accordance with the ASME Code and are qualified via the Westinghouse pump and valve
operability program which is described in paragraph 3.10.N.2.2.
The hardware and logic associated with this function will operate following an operating basis earthquake. Offsite power is not required for the system to function. The actuation logic in the
system is testable. However, the PORVs and RHR relief valves are not exercised with the
reactor at power. They are capable of being tested as required by the ASME Code and the
VEGP Technical Specifications. 5.2.2.10.1 Transient Evaluation Potential overpressurization transients to the RCS, while at relatively low temperatures, can be caused by either of two types of events to the RCS; i.e., mass input or heat input. Both types result in more rapid pressure changes when the RCS is water solid.
Anticipated mass and heat input transients are evaluated to demonstrate conformance with Appendix G. The most limiting heat input transient is an inadvertant reactor coolant pump startup in a loop where the steam generator secondary temperature is 50
°F higher than the primary temperature in any loop. The most limiting mass input transient is a charging-letdown mismatch where two emergency core cooling system (ECCS) centrifugal charging pumps and one normal centrifugal charging pump are charging water into the reactor coolant system with
the letdown path isolated.
It should be noted that the following transient is also addressed. With the plant in a cooled down and depressurized condition in which the cold overpressure protection system is required
to be operable and with charging and letdown established and RHRS open to RCS, a dc vital
bus fails. This failure causes normal letdown to isolate and also results in the loss of one of the
two PORVs. However, RHRS relief valves mitigate the transient.
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5.2-6 REV 19 4/15 5.2.2.10.2 Administrative Controls During plant operation the following precautions are observed: A. At least one RHR inlet line from the reactor coolant loop is not isolated unless there is a steam bubble in the pressurizer. This precaution ensures that there is a relief path from the reactor coolant loop to a RHR suction line relief valve when
the RCS is at low temperature and is water solid. B. Whenever the plant is water solid and the reactor coolant pressure is being maintained by the low pressure letdown control valve, letdown should include
flow from the operating RHR loop through the RHRS cleanup to the letdown heat
exchanger valve. C. One RCP should normally be running anytime RCS temperature is changed by more than 10
°F in 1 h. Additionally, RCPs should not be started if steam generator secondary water temperature is greater than 10
°F above the RCS temperature. D. During a typical plant cooldown, operable steam generators should be connected to the steam header to ensure a uniform cooldown of the reactor coolant loops. E. To preclude inadvertent emergency core cooling system (ECCS) actuation during heatup and cooldown, blocking of the high pressurizer pressure, and low steam
line pressure safety injection, signal actuation logic at 1970 psig is required.
These manual blocking features are further discussed in paragraph 7.3.1.2.2.6. During further cooldown, closure and power lockout of the accumulator isolation valves is performed at 1000 psig and power lockout to the safety injection pumps is performed at approximately 220
°F in the RCS. F. Periodic ECCS pump performance testing requires the testing of the pumps during normal power operation or at hot shutdown conditions. This precludes any potential for developing a cold overpressurization transient.
Should cold shutdown testing of the pumps be required, the test is done when
the vessel is open to the atmosphere, again precluding overpressurization
potential.
If cold shutdown testing with the vessel closed is necessary, the procedures
require only one pump to be tested with ECCS discharge valve closure and
RHRS alignment to both isolate potential ECCS pump input and to provide
backup benefit of the RHRS relief valves.
The SI signal circuitry testing, if performed during cold shutdown, also requires
RHRS alignment and safety injection pumps power lockout to preclude
developing cold overpressurization transients. G. A steam bubble will be maintained in the pressurizer when the RCS temperature is greater than 220
°F. 5.2.2.11 Consequences of a Postulated Loss of a dc Bus Coupled with a Single Failure Disabling a PORV Allowing a Cold Pressurization Event This discussion addresses the following postulated event:
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5.2-7 REV 19 4/15 With the plant in a cooled down and depressurized condition in which the cold overpressure protection system is required to be operable and with charging and
letdown established, a dc vital bus fails. This failure causes normal letdown to
isolate and also results in the loss of one of the two PORVs. In addition to the dc bus failure, an additional random failure of the second PORV is postulated to occur. This sequence of events places the plant in a condition in
which letdown is isolated, the automatic cold overpressure protection system is inoperable, and charging flow is filling the pressurizer, increasing system
pressure towards the Appendix G limits.
To begin this discussion the limitations placed on plant operation by the VEGP Technical
Specifications are addressed. A. With RCS temperature below 200
°F (i.e., cold shutdown) one RHR pump is required to be in operation. This requirement ensures that at least one RHR
suction relief valve is available for overpressure protection of the RCS. This
valve will relieve the flow from two ECCS centrifugal charging pumps and one
normal centrifugal charging pump at the valve lift setting pressure. B. Whenever the cold overpressure protection system is required to be operable, only the two ECCS centrifugal charging pumps and the normal centrifugal
charging pumps are allowed to be operable. The safety injection pumps will be
inoperable. This ensures that the maximum charging letdown mismatch will be
that stated in 5.2.2.10.1. C. Considering these requirements, anytime RHR is in operation and the RCS is in a condition requiring the cold overpressure protection to be operable, there is no
overpressure event as a result of the prescribed event. Assuming the event as
described (a) did occur, the RHR relief valve would prevent RCS pressure from reaching the Appendix G limit by relieving all charging flow.
Typically at least one RHR train is in operation, or at a minimum one RHR loop suction valve is
open, providing an open path from the RCS to a RHR suction relief valve, whenever RCS temperature is below 350
°F. For this reason an overpressure event resulting from the prescribed event is very unlikely; however, the discussion is extended to the infrequent case where the RHRS is isolated from the RCS, and t he cold overpressure protection system is required to be operable.
To gain a better understanding of the results of the event, it is necessary to address the functions of some of the chemical and volume control system control valves. As stated earlier, the letdown valves fail closed on loss of dc power isolating letdown. Between the charging
pumps and the normal charging isolation valves ar e two normally throttled valves which receive their power from the process and control racks powered by the essential ac instrument buses.
These valves are unaffected by a dc bus failure and continue to work normally during the event.
One of these valves is the charging flow c ontrol valve (1-FCV-121), which automatically regulates flow to maintain a prescribed pressurizer level. Assuming this valve continues to function normally, as pressurizer level rises, charging flow is reduced until the charging flow is
limited to that required for seal injection (32 gal/min) plus a minimal amount (15 gal/min)
required for regenerative heat exchanger cooling. At this flowrate, ample time is provided (as discussed below) to allow appropriate operator action. If valve control is in manual, the valve
position remains unchanged. The other valve is the charging flow backpressure regulator (a) As additional information, during RHR operation letdow n is typically taken from the discharge of the RHR pumps and is not isolated by the dc bus failure.
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5.2-8 REV 19 4/15 (1-HCV-182), which is manually positioned to regulate flow to the seal. This valve remains in its
initial position. The effect of these two valves is to limit charging flow to its value at the
beginning of the event. Assuming maximum letdown at the initiation of the event, total flow (charging plus seal injection) to the RCS is limited to approximately 130 gal/min.
An additional consideration is that, with the plant in the hot shutdown condition and RHR isolated from the RCS, normal operation is to have a steam bubble in the pressurizer of
approximately 1350 ft
- 3. At a maximum charging rate of 130 gal/min, it would take in excess of 30 min to reach the Appendix G limit at 200
°F, the temperature corresponding to the coldest RCS temperature at which RHR is permitted to be isolated. As an extreme case, with a bubble
of only half the normal size, the corresponding time available for appropriate action would be in
excess of 15 min. To summarize: A. The postulated event is unlikely to occur since the dc buses have a battery as an emergency power supply, and should the dc bus fail, it must be coupled with the additional failure of the second PORV for overpressurization. B. In the unlikely event that the prescribed event did occur, RHR would normally be online and capable of mitigating any potential overpressure resulting from two
ECCS centrifugal charging pumps and one normal centrifugal charging pump. C. In the highly unlikely event that the prescribed event should occur when RHR is isolated from the RCS, the operator would have sufficient time to mitigate the
event. D. The Appendix G curves are excessively conservative for their intended purpose of ensuring vessel integrity during cold shutdown.
No further action is necessary to address this postulated event, and the existing plant design
and operational techniques result in successful event mitigation. 5.2.2.12 Testing and Inspection Testing and inspection of the overpressure protection components are discussed in paragraph 5.4.13.4 and chapter 14. Testing capabilities of the overpressurization protection
system are consistent with testing principles for systems' electronics in paragraph 7.2.2.2.3.
Operational surveillance procedures will demons trate the operability of the overpressure protection system when system operation is requi red. Inservice inspection is performed in accordance with Section XI of the ASME Code. Relief requests, augmented examinations, and
alternate techniques will be utilized as appropriate. 5.2.2.13 References 1. Cooper, L., Miselis, V., and Starek, R. M., "Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June 1972 (also letter NS-CE-622 dated April 16, 1975, C. Eicheldinger (Westinghouse) to D. B. Vassallo (NRC), Additional Information on WCAP-7769, Revision 1). 2. Burnett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907, October 1972.
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5.2-9 REV 19 4/15 5.2.3 REACTOR COOLANT PRESSURE BOUNDARY (RCPB) MATERIALS 5.2.3.1 Material Specifications Typical material specifications used for the principal pressure-retaining applications in Class 1 primary components and for Class 1 and 2 auxiliary components in systems required for reactor shutdown and for emergency core cooling are listed in table 5.2.3-1. Typical material
specifications used for the reactor internals required for emergency core cooling, for any mode
of normal operation or under postulated accident conditions, and for core structural load bearing
members are listed in table 5.2.3-2.
Tables 5.2.3-1 and 5.2.3-2 may not be totally inclusive of the material specifications used in the listed applications; however, the listed specifications are representative. Identification of actual
materials is available in VEGP quality assurance records.
The materials utilized conform to the applicable American Society of Mechanical Engineers (ASME) code rules.
The welding materials used for joining the ferritic base materials of the RCPB conform to or are equivalent to ASME Material Specifications SFA 5.1, 5.2, 5.5, 5.17, 5.18, and 5.20. They are
qualified to the requirements of the ASME Code,Section III.
The welding materials used for joining the austenitic stainless steel base materials of the RCPB conform to ASME Material Specifications SFA 5.4 and 5.9. They are qualified to the
requirements of the ASME Code,Section III.
The welding materials used for joining nickel-chromium-iron alloy in similar base material combination and in dissimilar ferritic or austenitic base material combination conform to ASME
Material Specifications SFA 5.11 and 5.14. They are qualified to the requirements of the ASME
Code,Section III. 5.2.3.2 Compatibility with Reactor Coolant 5.2.3.2.1 Chemistry of Reactor Coolant a The reactor coolant system (RCS) chemistry specifications are given in table 5.2.3-3.
The RCS water chemistry is selected to minimi ze corrosion. Routinely scheduled analyses of the coolant chemical composition are performed to verify that the reactor coolant chemistry meets the specifications.
The chemical and volume control system (CVCS) provides a means for adding chemicals to the RCS which perform the following functions:
- Control the pH of the coolant during prestartup testing and subsequent operation.
- Scavenge oxygen from the coolant during heatup.
- Control radiolysis reactions involving hydrogen, oxygen, and nitrogen during all power operations subsequent to startup.
a The Water Chemistry Control Program is credi ted as a license renewal aging management program (see subsection 19.2.28).
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5.2-10 REV 19 4/15 The normal limits for chemical additives and reactor coolant impurities for power operation are
shown in table 5.2.3-3.
The pH control chemical utilized is lithium hy droxide monohydrate, enriched in the lithium-7 isotope to 99.9 percent. This chemical is chosen for its compatibility with the materials and water chemistry of borated water/stainless steel/zirconium/ Inconel systems. In addition, lithium-7 is produced in solution from the neutron irradiation of the dissolved boron in the
coolant. The lithium-7 hydroxide is introduced into the RCS via the charging flow. The solution
is prepared in the laboratory and transferred to the chemical additive tank. Reactor makeup
water is then used to flush the solution to the suction header of the charging pumps. The
concentration of lithium-7 hydroxide in the RCS is maintained in the range specified for pH
control. If the concentration exceeds this range, the cation bed demineralizer is employed in the
letdown line in series operation with the mixed bed demineralizer.
During reactor startup from the cold condition, hydrazine is employed as an oxygen scavenging agent. The hydrazine solution is introduced into the RCS in the same manner as described above for the pH control agent.
The reactor coolant is treated with dissolved hydrogen to control the net decomposition of water by radiolysis in the core region. The hydrogen also reacts with oxygen and nitrogen introduced
into the RCS as impurities under the impetus of core radiation. Sufficient partial pressure of
hydrogen is maintained in the volume control tank so that the specified equilibrium
concentration of hydrogen is maintained in the reactor coolant. A self-contained pressure
control valve maintains a minimum pressure in the vapor space of the volume control tank.
This can be adjusted to provide the correct equilibrium hydrogen concentration.
Boron, in the chemical form of boric acid, is added to the RCS for long-term reactivity control of the core.
A soluble zinc compound may be added to the reactor coolant as a means to reduce radiation fields within the primary system. The zinc used may be either natural zinc or zinc depleted of 64 Zn. When used, the target system zinc concentration is normally maintained to a concentration no greater than 40 ppb.
Suspended solid (corrosion product particulates) and other impurity concentrations are maintained below specified limits by controlling the chemical quality of makeup water and
chemical additives and by purification of the reactor coolant through the CVCS. 5.2.3.2.2 Compatibility of Construction Materials with Reactor Coolant All of the ferritic low-alloy and carbon steels which are used in principal pressure-retaining applications have corrosion-resistant cladding on all surfaces that are exposed to the reactor
coolant. The corrosion resistance of the cladding material is at least equivalent to the corrosion
resistance of types 304 and 316 austenitic stainless steel alloys or nickel-chromium-iron alloy, martensitic stainless steel, and precipitation-hardened stainless steel. These corrosion-
resistant cladding materials may be subjected to the ASME code required postweld heat
treatment for ferritic base materials.
Ferritic low-alloy and carbon steel nozzles have safe ends of either stainless steel wrought materials, stainless steel weld metal analysis A-7 (designated A-8 in the 1974 edition of the
ASME code), or nickel-chromium-iron alloy weld metal F-Number 43. The latter buttering
material requires further safe ending with austenitic stainless steel base material after
completion of the postweld heat treatment when the nozzle is larger than a 4-in. nominal inside
diameter and/or the wall thickness is greater than 0.531 in.
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5.2-11 REV 19 4/15 All of the austenitic stainless steel and nickel-chromium-iron alloy base materials with primary
pressure-retaining applications are used in the solution annealed condition. These heat
treatments are as required by the material specifications.
During subsequent fabrications, these materials are not heated above 800
°F other than locally by welding operations. The solution-annealed surge line material is subsequently formed by hot bending followed by a resolution annealing heat treatment.
Components employing stainless steel sensit ized in the manner expected during component fabrication and installation operate satisfactorily under normal plant chemistry conditions in
pressurized water reactor (PWR) systems because chlorides, fluorides, and oxygen are
controlled to very low levels. 5.2.3.2.3 Compatibility with External Insulation and Environmental Atmosphere In general, all of the materials listed in table 5.2.3-1 which are used in principal pressure-retaining applications and are subject to elevat ed temperature during system operation are in contact with thermal insulation that covers their outer surfaces.
The thermal insulation used on the RCPB is either reflective stainless steel type or made of compounded materials which yield low leachable chloride and/or fluoride concentrations. The
compounded materials in the form of blocks, boards, cloths, tapes, adhesives, cements, etc.,
are silicated to provide protection of austenitic stainless steels against stress corrosion which
may result from accidental wetting of the insulation by spillage, minor leakage, or other
contamination from the environmental atmosphere. Section 1.9 indicates the degree of
conformance with Regulatory Guide 1.36, Nonmetallic Thermal Insulation for Austenitic
Stainless Steel.
In the event of coolant leakage, the ferritic materials will show increased general corrosion rates. Where minor leakage is anticipated from service experience, such as valve packing, pump seals, etc., only materials that are compatible with the coolant are used. These are as
shown in table 5.2.3-1. Ferritic materials exposed to coolant leakage can be readily observed
as part of the inservice visual and/or nondestructive inspection program to ensure the integrity
of the component for subsequent service. 5.2.3.3 Fabrication and Processing of Ferritic Materials 5.2.3.3.1 Fracture Toughness The fracture toughness properties of the RCPB components meet the requirements of the ASME Code,Section III, Paragraphs NB, NC, and ND-2300, as appropriate.
The fracture toughness properties of the reactor vessel materials are discussed in section 5.3.
Limiting steam generator and pressurizer reference temperature for a nil ductility transition (RT NDT) temperatures are guaranteed at 60
°F for the base materials and the weldments. These materials meet the 50-ft-lb absorbed energy and 35-mils lateral expansion requirements
of the ASME Code,Section III, at 120
°F. The actual results of these tests are provided in the ASME material data reports which are supplied for each component and submitted to the owner at the time of shipment of the component.
VEGP-FSAR-5
5.2-12 REV 19 4/15 Calibration of temperature instruments and Charpy impact test machines are performed to meet the requirements of the ASME Code,Section III, Paragraph NB-2360.
Westinghouse has conducted a test program to determine the fracture toughness of low-alloy ferritic materials with specified minimum yield strengths greater than 50,000 psi to demonstrate
compliance with Appendix G of the ASME Code,Section III. In this program, fracture toughness
properties were determined and shown to be adequate for base metal plates and forgings, weld
metal, and heat-affected zone metal for higher strength ferritic materials used for components of
the RCPB. The results of the program are documented in reference 1, which has been
submitted to the Nuclear Regulatory Commission (NRC) for review. 5.2.3.3.2 Control of Welding All welding is conducted utilizing procedures qualified according to the rules of Sections III and IX of the ASME Code. Control of welding variables, as well as examination and testing, during
procedure qualification and production welding is performed in accordance with ASME Code
requirements.
Westinghouse practices for storage and handling of welding electrodes and fluxes comply with ASME Code,Section III, Paragraph NB-2400.
Section 1.9 indicates the degree of conformance of the ferritic materials components of the RCPB with Regulatory Guides 1.34, Control of Electroslag Weld Properties, 1.43, Control of
Stainless Steel Weld Cladding of Low-Alloy Steel Components, 1.50, Control of Preheat
Temperature for Welding of Low-Alloy Steel, and 1.71, Welder Qualification for Areas of Limited
Accessibility. 5.2.3.4 Fabrication and Processing of Austenitic Stainless Steel Paragraphs 5.2.3.4.1 through 5.2.3.4.5 address Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel, and present the methods and controls utilized by Westinghouse to
avoid sensitization and prevent intergranular attack (IGA) of austenitic stainless steel
components. Also, section 1.9 indicates the degree of conformance with Regulatory
Guide 1.44. 5.2.3.4.1 Cleaning and Contamination Protection Procedures It is required that all austenitic stainless steel materials used in the fabrication, installation, and testing of nuclear steam supply components and systems be handled, protected, stored, and
cleaned according to recognized and accepted methods which are designed to minimize
contamination which could lead to stress corrosion cracking. The rules covering these controls
are stipulated in Westinghouse process specifications. As applicable, these process
specifications supplement the equipment specifications and purchase order requirements of
every individual austenitic stainless steel co mponent or system which Westinghouse procures for the VEGP nuclear steam supply systems (NSSSs), regardless of the ASME Code
classification. Westinghouse process specifications are also given to Bechtel, the architect-
engineer, and to Georgia Power Company for recommended use within their scope of supply
and activity.
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5.2-13 REV 19 4/15 The process specifications that define these requirements and that follow the guidance of the
American National Standards Institute (ANSI) N-45 committee specifications include the
following:
Number Process Specification 82560HM Requirements for Pressure Sensitive Tapes for Use on Austenitic Stainless Steels. 83336KA Requirements for Thermal Insulation Used on Austenitic Stainless Steel Piping and Equipment. 83860LA Requirements for Marking of Reactor Plant Components and Piping. 8435OHA Site Receiving Inspection and Storage Requirements for Systems, Material, and Equipment. 8435lNL Determination of Surface Chloride and Fluoride on Austenitic Stainless Steel Materials. 853l0QA Packaging and Preparing Nuclear Components for Shipment and Storage. 292722 Cleaning and Packaging Requirements of Equipment for Use in the NSSS. 597756 Pressurized Water Reactor Auxiliary Tanks Cleaning Procedures. 597760 Cleanliness Requirements During Storage Construction, Erection, and Startup Activities of Nuclear Power System.
Section 1.9 indicates the degree of conformance of the austenitic stainless steel components of
the RCPB with Regulatory Guide 1.37, Quality Assurance Requirements for Cleaning of Fluid
Systems and Associated Components of Water-Cooled Nuclear Power Plants. 5.2.3.4.2 Solution Heat Treatment Requirements The austenitic stainless steels listed in tables 5.2.3-1 and 5.2.3-2 are utilized in the final heat-treated condition required by the respective ASME Code,Section II materials specification for
the particular type or grade of alloy. 5.2.3.4.3 Material Testing Program Westinghouse practice is that austenitic stainless steel materials of product forms with simple shapes need not be corrosion tested provided that the solution heat treatment is followed by
water quenching. Simple shapes are defined as all plates, sheets, bars, pipe, and tubes, as
well as forgings, fittings, and other shaped products that do not have inaccessible cavities or
chambers that would preclude rapid cooling when water quenched. When testing is required, the tests are performed in accordance with ASTM A 262, Practice A or E, as amended by
Westinghouse Process Specification 84201MW.
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5.2-14 REV 19 4/15 5.2.3.4.4 Prevention of Intergranular Attack of Unstabilized Austenitic Stainless Steels Unstabilized austenitic stainless steels are subject to IGA provided that three conditions are
present simultaneously. These are:
- A sensitized steel.
- A high temperature.
If any one of the three conditions described above is not present, IGA will not occur. Since high
temperatures cannot be avoided in all components in the NSSS, reliance is placed on the elimination of the other two conditions to prevent IGA on wrought stainless steel components.
This is accomplished by:
- Control of primary water chemistry to ensure a benign environment.
- Utilization of materials in the final heat-treated condition and the prohibition of subsequent heat treatments in the 800
°F and 1500
°F temperature range.
- Control of welding processes and procedures to avoid heat-affected zone sensitization.
- Confirmation that the welding procedures used for the manufacture of components in the primary pressure boundary and the reactor internals do not result in the sensitization of heat-affected zones.
Further information on each of these steps is provided in the following paragraphs.
The water chemistry in the RCS is controlled to prevent the intrusion of aggressive species. In particular, the maximum permissible oxygen and chloride concentrations are 0.1 ppm and 0.15
ppm, respectively. Table 5.2.3-3 lists the recommended reactor coolant water chemistry
specifications. The precautions taken to prevent the intrusion of chlorides into the system
during fabrication, shipping, and storage are stipulated in the appropriate process specifications.
The use of hydrogen overpressure precludes the presence of oxygen during operation. The
effectiveness of these controls has been demonstrated by both laboratory tests and operating
experience. The long-term exposure of severely sensitized stainless steels to reactor coolant
environments in early Westinghouse PWRs has not resulted in any sign of IGA. Reference 2
describes the laboratory experimental findings and reactor operating experience. The additional
years of operations since the issuing of reference 2 have provided further confirmation of the
earlier conclusions that severely sensitized stainless steels do not undergo any IGA in
Westinghouse PWR coolant environments.
Although there is no evidence that PWR coolant water attacks sensitized stainless steels, Westinghouse considers it good metallurgical practice to avoid the use of sensitized stainless
steels in the NSSS components. Accordingly, measures are taken to prohibit the purchase of
sensitized stainless steels and to prevent sensitization during component fabrication. Wrought
austenitic stainless steel stock is used for components that are part of:
- The RCPB.
- Systems required for reactor shutdown.
- Systems required for emergency core cooling.
VEGP-FSAR-5
5.2-15 REV 19 4/15
- Reactor vessel internals relied upon to permit adequate core cooling for normal operation or under postulated accident conditions.
The wrought austenitic stainless steel stock is utilized in one of the following conditions:
- Solution annealed and water quenched.
- Solution annealed and cooled through the sensitization temperature range within less than approximately 5 min.
It is generally accepted that these practices will prevent sensitization. Westinghouse has
verified this by performing corrosion tests on wrought material as it was received.
The heat-affected zones of welded components must, of necessity, be heated into the sensitization temperature range, 800
°F to 1500°F. However, severe sensitization (i.e., continuous grain boundary precipitates of chromium carbide, with adjacent chromium depletion) can be avoided by controlling welding parameters and welding processes. The heat input and
associated cooling rate through the carbide precipitation range are of primary importance.
Westinghouse has demonstrated this by corrosion testing a number of weldments.
Heat input is calculated according to the formula: H = ()()()SIE 60 where: H = joules/in.
E = volts.
I = amperes.
S = travel speed (in./min).
Of 25 production and qualification weldments tested, representing all major welding processes and a variety of components and incorporating base metal thicknesses from 0.10 to 4.0 in., only
portions of two were severely sensitized. Of these, one involved a heat input of 120,000 J, and
the other involved a heavy socket weld in relatively thin-walled material. In both cases, sensitization was caused primarily by high-heat inputs relative to the section thickness. In only
the socket weld did the sensitized condition exist at the surface, where the material is exposed
to the environment. The component has been redesigned, and a material change has been
made to eliminate this condition.
The heat input in all austenitic pressure boundary weldments has been controlled by:
- Prohibiting the use of block welding.
- Limiting the maximum interpass temperature to 350
°F.
- Exercising approval rights on all welding procedures. 5.2.3.4.5 Retesting Unstabilized Austenitic Stainless Steels Exposed to Sensitization Temperatures As described in the previous section, it is not normal Westinghouse practice to expose unstabilized austenitic stainless steels to the sensitization range of 800
°F to 1500°F during fabrication into components. If during the course of fabrication, the steel is inadvertently
exposed to the sensitization temperature range, 800
°F to 1500°F, the material may be tested in VEGP-FSAR-5
5.2-16 REV 19 4/15 accordance with ASTM A 262, as amended by Westinghouse Process Specification 84201MW, to verify that it is not susceptible to IGA, except that testing is not required for: A. Cast metal or weld metal with a ferrite content of 5 percent or more. B. Material with a carbon content of 0.03 percent or less that is subjected to temperatures in the range of 800
°F to 1500°F for less than 1 h. C. Material exposed to special processing, provided the processing is properly controlled to develop a uniform product and provided that adequate documentation exists of service experi ence and/or test data to demonstrate that
the processing will not result in increased susceptibility to intergranular stress
corrosion.
If it is not verified that such material is not susceptible to IGA, the material is resolution
annealed and water quenched or rejected. 5.2.3.4.6 Control of Welding The following paragraphs address Regulatory Guide 1.31, Control of Ferrite Content in
Stainless Steel Weld Metal, and present the methods used, and the verification of these
methods, for austenitic stainless steel welding.
The welding of austenitic stainless steel is controlled to mitigate the occurrence of microfissuring or hot cracking in the weld. Although published data and experience have not
confirmed that fissuring is detrimental to the quality of the weld, it is recognized that such
fissuring is undesirable in a general sense. Also, it has been well documented in the technical
literature that the presence of delta ferrite is one of the mechanisms for reducing the
susceptibility of stainless steel welds to hot cracking. However, there is insufficient data to
specify a minimum delta ferrite level below which the material will be prone to hot cracking. It is
assumed that such a minimum lies somewhere between 0- and 3-percent delta ferrite.
The scope of these controls discussed herein encompasses welding processes used to join stainless steel parts in components designed, fabricated, or stamped in accordance with the
ASME Code,Section III, Class 1 and 2, and core support components. Delta ferrite control is
appropriate for the above welding requirements, except where no filler metal is used or where
for other reasons such control is not applicable. These exceptions include electron beam
welding, autogenous gas shielded tungsten arc welding, explosive welding, and welding using
fully austenitic welding materials.
The fabrication and installation specifications require welding procedure and welder qualification in accordance with Section III and include the delta ferrite determinations for the austenitic
stainless steel welding materials that are used for welding qualification testing and for
production processing. Specifically, the undiluted weld deposits of the "starting" welding
materials are required to contain a minimum of 5-percent delta ferrite (the equivalent ferrite
number may be substituted for percent delta fe rrite) as determined by chemical analysis and calculation using the appropriate weld metal constitution diagrams in Section III. When new
welding procedure qualification tests are evaluated for these applications, including repair
welding of raw materials, they are performed in accordance with the requirements of Sections III and IX.
The results of all the destructive and nondestructive tests are reported in the procedure qualification record in addition to the information required by Section III.
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5.2-17 REV 19 4/15 The starting welding materials used for fabrication and installation welds of austenitic stainless
steel materials and components meet the requirements of Section III. The austenitic stainless
steel welding material conforms to ASME weld metal analysis A-7 (designated A-8 in the 1974
edition of the ASME Code), type 308 or 308L for all applications. Bare weld filler metal, including consumable inserts, used in inert gas welding processes conform to ASME SFA 5.9, and are procured to contain not less than 5-percent delta ferrite according to Section III. Weld
filler metal materials used in flux shielded welding processes conform to ASME SFA 5.4 or 5.9
and are procured in a wire-flux combination to be capable of providing not less than 5-percent
delta ferrite in the deposit according to Section III. Welding materials are tested using the
welding energy inputs to be employed in production welding.
Combinations of approved heat and lots of starting welding materials are used for all welding processes. The welding quality assurance program includes identification and control of
welding material by lots and heats as appropriate. All of the weld processing is monitored
according to approved inspection programs which include review of starting materials, qualification records, and welding parameters. Welding systems are also subject to:
- Quality assurance audit including calibration of gauges and instruments.
- Identification of starting and completed materials.
- Welder and procedure qualifications.
- Availability and use of approved welding and heat-treating procedures.
- Documentary evidence of compliance with materials, welding parameters, and inspection requirements.
Fabrication and installation welds are inspected using nondestructive examination methods
according to Section III rules.
To ensure the reliability of these controls, Westinghouse has completed a delta ferrite verification program, described in reference 3. This program has been approved as a valid
approach to verify the Westinghouse hypothesis and is considered an acceptable alternative for
conformance with the NRC Interim Position on Regulatory Guide 1.31. The Regulatory Staff's
acceptance letter and topical report evaluation were received on December 30, 1974. The
program results, which do support the hypothesis presented in reference 3, are summarized in
reference 4.
Section 1.9 indicates the degree of conformance of the austenitic stainless steel components of the RCPB with Regulatory Guides 1.34, Control of Electroslag Properties, and 1.71, Welder
Qualification for Areas of Limited Accessibility. 5.2.3.5 References 1. Logsdon, W. A., Begley, J. A., and Gottshall, C. L., "Dynamic Fracture Toughness of ASME SA-508 Class 2a and ASME SA-533 Grade A Class 2 Base and Heat-Affected
Zone Material and Applicable Weld Metals," WCAP-9292, March 1978. 2. Golik, M. A., "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," WCAP-7477-L (Proprietary), March 1970, and WCAP-7735 (Nonproprietary), August 1971. 3. Enrietto, J. F., "Control of Delta Ferrite in Austenitic Stainless Steel Weldments," WCAP-8324-A, June 1975.
VEGP-FSAR-5
5.2-18 REV 19 4/15 4. Enrietto, J. F., "Delta Ferrite in Production Austenitic Stainless Steel Weldments," WCAP-8693, January 1976. 5.2.4 INSERVICE INSPECTION AND TESTING OF REACTOR COOLANT PRESSURE BOUNDARY a Inservice inspection and testing of Class 1 pressure-retaining components such as vessels, piping, pumps, valves, bolting, and supports within the reactor coolant pressure boundary shall be performed in accordance with Section XI of the American Society of Mechanical Engineers (ASME) Code including any applicable addenda in accordance with 10 CFR 50.55a(g)(4)(ii)
(specific edition and any applicable addenda of the code will be delineated in each program),
with certain exceptions whenever specific written relief is granted by the Nuclear Regulatory
Commission (NRC) in accordance with 10 CFR 50.55a(a)3 and 10 CFR 50.55a(g)(6)(i). The
inservice testing of pumps and valves in accordance with the requirements of Articles IWP and
IWV of the code is discussed in subsection 3.9.6. Class 2 and 3 components examinations are
addressed in section 6.6.
The preservice inspection program requirements for each unit were completed prior to the commercial operation date for each of the respective units. The preservice inspection program for Unit 1 complied with the ASME Code,Section XI, 1980 Edition including addenda through
Winter 1980, except that reactor pressure vessel examinations were performed using the 1980
Edition including addenda through Winter 1981. The preservice inspection program for Unit 2 complied with the ASME Code,Section XI, 1983 Edition including addenda through Summer
1983, except that reactor pressure vessel examinations were performed using the 1980 Edition
including addenda through Winter 1981. Certain preservice inspection requirements of the ASME Code,Section XI were determined to be impractical and relief requests were granted by
the NRC, pursuant to 10 CFR 50.55a(g)(i). The relief requests were supported by information
pursuant to 10 CFR 50.55a (a) (3). The inservice inspection program and inservice test
program were submitted to the NRC prior to commercial operation. These programs comply
with applicable inservice inspection provisions of 10 CFR 50.55a(g) and the NRC guidelines
attached as an appendix to section 121.0 of review questions entitled "Guidance for Preparing
Preservice and Inservice Inspection Programs and Relief Requests Pursuant to 10 CFR
50.55a(g)." Where compliance with code requirements is not practical, relief requests have
been submitted to the NRC for review and approval. The inservice programs detail the areas
subject to examination and method, extent, and frequency of examinations. Additionally, component supports and snubber testing requirements are included in the inspection programs. 5.2.4.1 System Boundary Subject to Inspection In addition to the reactor pressure vessel, all Class 1 components such as vessels, piping, pumps, valves, bolting, and supports shall be inspected to the extent practical, in accordance with Article IWB of ASME Code,Section XI. Class 1 pressure-retaining components and their
specific boundaries are identified in the inspection plan documents.
a The Inservice Inspection Program is credited as a license renewal aging management program (see subsection 19.2.13).
VEGP-FSAR-5
5.2-19 REV 19 4/15 5.2.4.2 Arrangement and Accessibility The physical arrangement of components was designed to allow personnel and equipment
access to the extent practical to perform t he required inservice examinations. Removable insulation and shielding was provided on those piping systems requiring volumetric and surface
examination. Temporary or permanent working platforms, scaffolding, and ladders are provided to facilitate access to piping welds.
An inservice inspection design review was undertaken to identify exceptions to the access requirements of the code with subsequent design modifications and/or inspection technique
development to ensure code compliance to the ext ent practical. Additional exceptions may be identified and reported to the NRC after plant operation, as specified in 10 CFR 50.55a(g)(5)(iv).
Space has been provided to handle and store insulation, structural members, shielding, and other materials related to the inspection. Suitable hoists and other handling equipment, lighting, and sources of power for inspection equipment were installed at appropriate locations. The
reactor pressure vessel (RPV) inspections are performed primarily from the vessel internal surfaces. Other areas of the RPV such as the closure head are accessible from the outer
surfaces of the vessel for inspection. Closure studs, nuts, and washers are removed to a dry
location for direct inspection. 5.2.4.3 Examination Techniques and Procedures The visual, surface, and volumetric examination techniques, procedures, and special
techniques are in accordance with the requirements of subarticle IWA-2200 and table IWB-2500-1 of the ASME Code,Section XI except where compliance with code requirements is not
practical and relief has been requested from the NRC. Liquid penetrant methods and/or
magnetic particle methods are used for surfac e examinations. Radiography and/or ultrasonic techniques, whether manual or remote, are used for volumetric examinations. A special vessel
inspection tool is used to inspect the RPV welds from the vessel internal surfaces. Welds
located in the reactor vessel beltline region are examined to meet the requirements of
Regulatory Guide 1.150 to the extent practicable. Other RPV welds are examined to meet the
requirements of Regulatory Guide 1.150, with the exception of the near surface examination, to
the extent practicable. Other examination techniques may be used provided that the results are
demonstrated to be equivalent or superior to the above techniques. 5.2.4.4 Inspection Intervals Inspection intervals are as defined in subarticles IWA-2400 and IWB-2400 of ASME Code,Section XI. The interval may be extended by as much as 1 year to permit inspections to be
concurrent with plant outages. It is intended that inservice examinations be performed during
normal plant outages such as refueling shutdowns or maintenance shutdowns occurring during
the inspection interval. 5.2.4.5 Examination Categories and Requirements The examination categories and requirements are in accordance with subarticle IWB-2500 and table IWB-2500-1 of ASME Code,Section XI. The preservice examinations complied with IWB-
2200.
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5.2-20 REV 19 4/15 5.2.4.6 Evaluation of Examination Results Examination results are evaluated in accordance with IWB-3000, with flaw indications in
accordance with IWB-3400 and table IWB-3410. Repair procedures are in accordance with IWB-4000 of ASME Code,Section XI. 5.2.4.7 System Leakage and Hydrostatic Pressure Tests System pressure tests comply with IWA-5000 and IWB-5000 of ASME Code,Section XI. 5.2.5 DETECTION OF LEAKAGE THROUGH REACTOR COOLANT PRESSURE BOUNDARY The reactor coolant pressure boundary (RCPB) leakage detection systems monitor leaks from the reactor coolant and associated systems. T hese systems provide information which permit the plant operators to take corrective action if a leak is evaluated as detrimental to the safety of
the facility. 5.2.5.1 Design Bases The leak detection systems are designed in accordance with the requirements of 10 CFR 50
and the general design criterion 30 to provide a means of detecting and, to the extent practical, identifying the source of the reactor coolant leakage. The systems conform with Regulatory
Guide 1.45. Main systems that monitor the environmental condition of the containment include the sump level monitoring system, the airborne par ticulate radioactivity monitoring systems, and the containment fan cooler condensate measuri ng system. In addition to the above systems, the humidity, temperature, pressure, and radiogas monitors provide indirect indication of
leakage to the containment.
Associated systems and components connected to the reactor coolant system have intersystem leakage monitoring devices.
These leakage detection systems are qualified for all seismic events not requiring a shutdown.
The airborne radioactivity monitoring system is qualified for a safe shutdown earthquake (SSE). 5.2.5.1.1 Leakage Classification RCPB leakage is classified as either identified or unidentified leakage. Identified leakage
includes: leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered, and conducted to a sump or collecting tank; or leakage into the
containment atmosphere from sources that are both specifically located and known either not to
interfere with the operation of unidentified leakage monitoring systems or not to be from a flaw
in the RCPB; or leakage into auxiliary systems and secondary systems. Unidentified leakage is all other leakage. 5.2.5.1.2 Limits for Reactor Coolant Leakage Limits for reactor coolant leakage are identified in the Technical Specifications.
VEGP-FSAR-5
5.2-21 REV 19 4/15 5.2.5.2 Identified Intersystem Leakage Detection Unidentified leakage into closed primary systems is directed to the reactor coolant drain tank or
pressurizer relief tank. Identified leakage, such as pump seal or valve packing leakage, is
directed to the reactor coolant drain tank where it is monitored by tank pressure, temperature, level, and flow instrumentation on the reactor coolant drain tank discharge lines.
Identified leakage, such as leakage past the pressurizer safety valves or power-operated relief valves (PORVs), is directed to the pressurizer relief tank. This leakage is monitored by
temperature instrumentation in the piping syst em and tank pressure, temperature, and level instrumentation. Leakage collected in the pressurizer relief tank is directed to the reactor
coolant drain tank for subsequent treatment and discharge.
An important identified leakage path for reactor coolant into other systems is through the steam generator tubes into the secondary side of the steam generator. Identified leakage to the steam
generators is detected by means of the steam generator sample liquid or condenser air ejector
radiation monitors. Two additional primary-to-secondary leak detection systems are also provided: a noble gas detector and a system utiliz ing N16 as the detection medium. The N16 detector is installed in the turbine building main steam pipe chase, between the two main steam
pipes. The N16 leak monitor is independent of the primary loop fission and corrosion product
radioactivity, so when the tube leak rates increase without considerable changes in the
secondary side radioactivity levels, the system can still detect small leaks. The noble gas
detector is located in the condenser steam jet air ejector discharge header immediately prior to
the filtration unit. For details of these radiation monitors, see subsection 11.5.2.
Auxiliary systems connected to the RCPB incor porate design and administrative provisions that serve to limit leakage. These provisions incl ude isolation valves designed for low seat leakage, periodic testing of RCPB check valves (paragraph 6.3.4.2), and inservice inspection (subsection
5.2.4 and section 6.6). Leakage is detected by in creasing auxiliary system level, temperature, and pressure indications or lifting of relief valves accompanied by increasing values of
monitored parameters in the relief valve dischar ge path. These systems are isolated from the RCS by normally closed valves and/or check valves. 5.2.5.2.1 Description and Operation of Identified Leak Detection System A. Residual Heat Removal System (RHRS) (Suction Side)
The RHRS is isolated from the RCS on the suction side by motor-operated valves HV-8701A/B and HV-8702A/B. Leakage past these valves is detected by
lifting of relief valves PSV-8708A or PSV-8708B, accompanied by increasing
pressurizer relief tank level, pressure, and temperature indications and alarms on
the main control board. B. Safety Injection System (SIS)/Accumulators The accumulators are isolated from the RCS by check valves 1204-U6-083
through -086 and 1204-U6-079 through -082. Leakage past these valves and
into the accumulator subsystem is detected by redundant control room
accumulator pressure and level indications and alarms. C. SIS/RHR Discharge Subsystem The RHR pump portion of the SIS is isolated from the RCS by check valves
1204-U6-083 through -086, 1204-U6-147 through -150, 1204-U6-125 and -126, 1204-U6-128 and -129, and normally closed motor-operated valve HV-8840.
VEGP-FSAR-5
5.2-22 REV 19 4/15 Leakage past these valves will eventually pressurize the RHR discharge header
and the pump suction header through the normally open pump miniflow isolation
valves FV-610 or FV-611. A continued increase in RHR pump discharge
pressure will be indicated in the control room and ultimately result in lifting relief
valves PSV-8708A and PSV-8708B in the suction header. D. SIS/Safety Injection Pump Subsystem The safety injection pump portion of the SIS is isolated from the RCS by check valves 1204-U4-083 through -086, 1204-U4-143 through -146, 1204-U6-124
through -127, 1204-U4-120 through -123, and normally closed motor- operated
valves HV-8802A/B. Leakage past these valves will pressurize the safety
injection pump discharge header, resulting in control room indication of
increasing pressure and, eventually, lifting of relief valve PSV-8851 or PSV-
8853A/B. E. SIS/Centrifugal Charging Pump Subsystem The charging pump subsystem of the SIS is isolated from the RCS by check valves 1204-U4-026 through -029, 1204-U6-013, and motor-operated valves HV-
8801A/B. Leakage past valves HV-8801A/B is not possible, since the valve
inlets are pressurized by the operating charging pump. F. Head Gasket Monitoring Connections The reactor vessel flange and head are sealed by two metallic O-rings. These gaskets are of the hollow self-energizing type in which pressure of the fluid being
sealed enters the interior of the gasket. The O-rings are fastened to the closure
head by a mechanical connection to facilitate removal.
Seal leakage is detected by means of two leak-off connections: one between the inner and outer ring, and one outside the outer O-ring. A manual isolation valve
is installed just outside the missile barrier of each leak-off line. Downstream of
these valves the lines are headered before being routed to the reactor coolant
drain tank in the waste processing system. An air-operated isolation valve, actuated from the control board, is installed in the common line. During normal
plant operation, the leak-off piping is aligned such that leakage across the inner
O-ring passes through valves 1201-U4-088 and HV-8032 into the drain tank. A
surface mounted, resistance temperature detector, installed on the bottom of the
common pipe, signals leakage at an alarm setpoint. A blind flanged branch line
containing isolation valve 1201-U4-089 is provided to confirm and establish the
magnitude of the leakage.
Once inner O-ring leakage is discovered, valve 1201-U4-087 should be opened and valve 1201-U4-088 closed so that possible leakage across the second
O-ring would be monitored.
In addition, during plant refueling operations both the inner and outer reactor vessel flange leak-off valves are closed. This prevents possible gas leakage
from the reactor coolant drain tank to the containment atmosphere. Refer to drawings 1X4DB111 and 2X4DB111 for the flow diagram representation.
The reactor vessel is the only flanged vessel within the RCPB that is provided with leak-off collection provisions.
Leakage past the reactor vessel head gaskets results in temperature indication and alarm in the control room.
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5.2-23 REV 19 4/15 G. Component and Auxiliary Com ponent Cooling Water Systems Leakage from the RCS to the component cooling water (CCW) and auxiliary
component cooling water (ACCW) systems, which service all RCPB associated
components that require cooling, is detected by the CCW and ACCW
radioactivity monitoring system (subsection 11.5.2) and/or increasing surge tank
level. Components serviced by these aux iliary cooling systems include: reactor coolant pump thermal barriers, RHR heat exchangers, letdown line heat
exchangers, reactor coolant pump seal water heat exchangers. 5.2.5.3 Unidentified Leakage Detection Normally, unidentified leakage from the RCS is very low. The RCS is an all-welded system, with the exception of the connections on the pressurizer safety valves, reactor vessel head, pressurizer and steam generator manways, and reactor vessel head vent, which are flanged.
In general, valves in the RCS that are 2 in. and under are of the packless type. All valves larger than 2 in. have dual packing with a leak-off connection to the reactor coolant drain tank between
the two packings or a reduced packing configuration with the valve stem leakoff line capped.
Primary indications of unidentified coolant leakage to the containment are provided by air particulate radioactivity monitors, gaseous radi oactivity monitors, fan cooler condensate flow monitors, and containment sump level monitors.
In normal operation, these primary monitors show a background level that is indicative of the normal level of unidentified leakage inside the containment. Variations in airborne radioactivity
or specific humidity above the normal level signify an increase in unidentified leakage rates and
signal to the plant operators that corrective action may be required. Similarly, increases in
containment sump level signify an increase in unidentified leakage.
RCS unidentified leakage may also be indicated by increasing charging pump flowrate compared with normal RCS inventory changes and by unscheduled increases in reactor
makeup water usage.
Reactor coolant inventory monitoring provides an indication of system leakage. Net level changes in the pressurizer and volume control tank are indicative of system leakage, since the
chemical and volume control system is a closed loop connected to the RCS. Monitoring net
makeup to the chemical and volume control system, as well as net collected leakage, provides
an important method of obtaining information for use in establishing a water inventory balance.
An abnormal increase in makeup water requirements or a significant change in the water
inventory balance can be indicative of increased system leakage.
The sensitivity and response time of the detection equipment for unidentified leakage is such that a leakage rate, or its equivalent, of 1 gal/min can be detected in approximately 1 h.
The above methods are supplemented by visual and ultrasonic inspections of the RCPB during plant shutdown periods, in accordance with the inservice inspection program (subsection 5.2.4). 5.2.5.3.1 Description and Operation of Main Unidentified Leak Detection Systems Systems employed for detecting leakage to the c ontainment from unidentified sources are:
- Containment airborne particulate radioactivity monitor.
- Containment gaseous radioactivity monitor.
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5.2-24 REV 19 4/15
- Containment air cooler condensate flow monitor.
- Containment sump level monitor. Additionally, humidity, temperature, and pressure monitoring of the containment atmosphere are
used for alarms and indirect indication of leakage to the containment. A. Containment Airborne Particulate Radioactivity Monitoring System An air sample is drawn outside the containment into a closed system by a
sample pump and is then consecutively passed through a particulate filter with
detector and a gaseous monitor chamber with detector. The filter collects 99
percent of the particulate matter greater than 1 mm in size. The sample
transport system includes:
- A pump to obtain the air sample.
- A flow control valve to provide flow adjustment.
- A flow meter to indicate the flowrate.
- A flow alarm assembly to provide high- and low-flow alarm signals. The particulate filter is continuously monitored by a scintillation crystal with a photomultiplier tube that provides an output signal proportional to the activity
collected on the filter. The particulate monitor has a minimum detectable concentration of 10
-11 µCi/cm 3 and a range of 10
-11 to 10-6 µCi/cm 3. More details concerning the particulate monitors can be found in subsection 11.5.2. Particulate activity can be correlated with the coolant fission and corrosion product activities. Any increase of more than two standard deviations above the count rate for background would indicate a possible leak. The total particulate
activity concentration above background, due to an abnormal leak and natural
decay, increases almost linearly with time for the first several hours after the
beginning of a leak. As shown in figure 5.2.5-1, with 0.01-percent failed fuel, containment background airborne particulate radioactivity equivalent to 10
-3 percent/day, and a partition factor equal to 0.001, a 1-gal/min leak would be
detected in approximately 1 h. Larger leaks would be detected in proportionately
shorter times (exclusive of sample transport time, which remains constant). The
detection capabilities and response times are shown in figure 5.2.5-1. The activity is indicated on displays and electronically recorded. High-activity alarm indications are displayed on the radiation monitoring cabinets. Local
alarms provide operational status of supporting equipment such as pumps, motors, and flow and pressure controllers. The leakage flowrate can be determined by performing a water inventory balance of the reactor coolant system when the count rate indicates a possible leak as
explained above. B. Containment Gaseous Radioac tivity Monitoring System The containment gaseous radioactivity monitor determines gaseous radioactivity in the containment by monitoring continuous air samples from the containment
atmosphere. After passing through the gas monitor, the sample is returned via
the closed system to the containment atmosphere.
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5.2-25 REV 19 4/15 Each sample is continuously mixed in a fixed, shielded volume where its activity is monitored. The monitor has a range of 10
-7 to 10-2 µCi/cm 3 and a minimum detectable concentration of 5 x 10
-7 µCi/cm 3. The containment gaseous radioactivity monitors are described in subsection 11.5.2. Gaseous radioactivity can be correlated with the gaseous activity of the reactor coolant. Any increase more than two standard deviations above the count rate
for background would indicate a possible leak. The total gaseous activity level
above background increases almost linearly for the first several hours after the
beginning of the leak. As specified in figure 5.2.5-1, with 0.01-percent failed fuel, containment background airborne gaseous radioactivity equivalent to 1
percent/day, and a partition factor equal to 1, a 1-gal/min leak would be detected
in approximately 1 h. Larger leaks would be detected in proportionately shorter
times (exclusive of the sample transport time, which remains constant). The
detection capabilities and response times are shown in figure 5.2.5-1. The detector outputs are transmitted to the radiation monitoring system cabinets in the control room, where the activity is indicated by displays and electronically recorded. High-activity alarm indications are displayed on the control board
annunciator in addition to the radiation monitoring system cabinets. Local alarms
annunciate the operational status of the supporting equipment. The leakage flowrate can be determined by performing a water inventory balance of the reactor coolant system when the count rate indicates a possible leak as
explained above. The containment purge system radioactivi ty monitors (subsection 11.5.2) serve as backup to the containment air particulate and gaseous airborne radioactivity
monitoring system while the purge is in operation. The containment purge monitors function in the same manner as the containment air particulate and gaseous radioactivity monitors, except that the
purge monitors sample from the containment purge exhaust line. C. Containment Air Cooler Condensate Monitoring System The condensate monitoring system permi ts measurements of the liquid runoff from the containment cooler units. It consists of a containment cooler drain
collection header, a vertical standpipe, valving, and standpipe level
instrumentation. The condensation from the containment coolers flows via the
collection header to the vertical standpipe. A differential pressure transmitter
provides standpipe level signals. The sy stem provides measurements of low leakages by monitoring standpipe level increase versus time. The condensate flowrate is a function of containment humidity, nuclear service cooling water (NSCW) temperature, and containment purge rate. The water
vapor dispersed by a 1-gal/min leak is usually greater than the water vapor brought in with the outside air. Air brought in from the outside is heated to 60
°F before it enters the containment. After the air enters the containment, it mixes with the containment atmosphere and is heated to between 100
°F and 120°F while the relative humidity drops.
The most important factor in condensing the water vapor is the temperature of
the NSCW which is supplied to the containment coolers. This water is assumed to vary in temperature between 35
°F and 95°F.
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5.2-26 REV 19 4/15 Drainage flowrate from the units due to normal condensation is calculated for the ambient (background) atmospheric conditions present within the containment.
With the initiation of an additional or abnormal leak, the containment atmosphere
humidity and condensation runoff rate both begin to increase, the water level
rises in the vertical pipe, and the high condensate flow alarm is actuated. Level
changes of as little as 0.25 in. in the cooler condensate standpipes can be
detected. Figure 5.2.5-1 shows the detection capabilities of the system for various conditions. Normal background leakage increases containment humidity to the
point where the condensation rate will increase, which improves the detection
capabilities of this system. As shown in figure 5.2.5-1, a sensitivity of 1 gal/min in
approximately 1 h can be achieved when supplying cold NSCW to the
containment coolers or with the initial background leakage. The rate of leakage can be determined when the precise NSCW, outside air, and containment air temperatures, and the outside relative humidity are known. D. Containment Sump Level Monitoring System Since a leak in the primary system would result in reactor coolant flowing into the containment normal or reactor cavity sumps, leakage would be indicated by a level increase in the sump. Indication of increasing sump level is transmitted from
the sump to the control room level indicator by means of a sump level
transmitter. The system provides meas urements of low leakages by monitoring level increase versus time. The detection capabilities of the containment normal sump and reactor cavity sump are shown in figure 5.2.5-1, assuming that the water from the leak is
collected in the sump. The actual reactor coolant leakage rate can be established from the increase above the normal rate of change of sump level. A check of other instrumentation
would be required to eliminate possible leakage from nonradioactive systems as a cause of an increase in sump level. The leakage rate can also be determined
from the frequency of sump pump operation. Under normal conditions, the containment normal sump pumps operate infrequently and reactor cavity sump pump operates very infrequently. Gross
leakage can be surmised from unusual frequency of pump operation. Sump level
and pump running indication are provided in the control room to alert the
operators. 5.2.5.3.2 Additional Unidentified Leakage Detection Methods Other methods available for detecting leakage are: A. Charging Pump Operation During normal operation, one of the charging pumps is in operation. If a gross increase in reactor coolant leakage occurs, the flowrate of the charging pump would increase, indicating leakage from the RCS. This leakage must be
sufficient to cause a decrease in pressurizer or volume control tank level that is
within the sensitivity range of the level indicators. The flowrate of the charging VEGP-FSAR-5
5.2-27 REV 19 4/15 pump would automatically increase to try to maintain pressurizer level. Charging
pump discharge flow indication is provided in the control room. The leakage rate can be determined by the amount that the charging pump flowrate increases above the letdown flowrate to maintain constant pressurizer
level. Any significant increase in the charging flowrate is a possible indication of
a leak. B. Containment Humidity Monitoring System The containment humidity system, ut ilizing temperature-compensated humidity detectors, is provided to determine the water-vapor content of the containment atmosphere. An increase in the humidity of the containment atmosphere
indicates release of water within the containment. The range of the containment humidity measuring system is 5- to 99-percent relative humidity at 80
°F with a temperature range of 40 to 120
°F. The accuracy of the humidity detectors is
+/-3 percent. The response of the containment humidity under various outside air conditions and no leakage falls within the extremes shown in figure 5.2.5-1. The humidity monitor supplements the condensate monitor. It is most sensitive under
conditions when there is no condensation. A rapid increase of humidity over the background level by more than 10 percent can be taken as a probable indication of a leak. The leakrate can be determined when the outside air temperature and humidity and the containment atmosphere temperature are known. C. Liquid Inventory The operators can surmise gross leakage from changes in the reactor coolant inventory. Noticeable decreases in the pressurizer level not associated with known changes in operation are investigated. Likewise, makeup water usage
information which is available from the plant computer is checked frequently for
unusual makeup rates not due to plant operations. 5.2.5.4 Safety Evaluation The leak detection system has no safety desi gn basis; however, the containment atmosphere radioparticulate and radiogas monitors are qualified for an SSE per the recommendation of
Regulatory Guide 1.45. 5.2.5.5 Tests and Inspections Periodic testing of leakage detection systems is conducted to verify the operability and
sensitivity of detector equipment. These tests include installation calibrations and alignments, periodic channel calibrations, frictional tests, and channel checks. A description of calibration
and maintenance procedures and frequencies for the containment radioactivity monitoring
system is presented in subsection 11.5.2.
The humidity detector and condensate measuring syst em are also periodically tested to ensure proper operation and verify sensitivity.
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5.2-28 REV 19 4/15 Inservice inspection criteria, the equipment used, procedures involved, the frequency of testing, inspection, surveillance, and examination of the structural and leaktight integrity of RCPB
components are described in detail in subsection 5.2.4. 5.2.5.6 Instrumentation Applications The following indications are provided in the control room to allow operating personnel to
monitor for leakage: A. Containment air particulate monitor - air particulate activity. B. Containment gaseous activity monitor - gaseous activity.
C. Containment cooler condensate monitoring system - standpipe level.
D. Containment humidity measuring system - containment humidity. E. Containment normal sump level and reactor cavity sump level.
F. Gross leakage detection methods - charging pump flowrate, letdown flowrate, pressurizer level, and reactor coolant temperatures are available for the charging pump flow method. Containment sump levels and pump operation are available
for the sump pump operation method. Total makeup waterflow is available from
the plant computer for liquid inventory.
REV 13 4/06 PRIMARY COOLANT LEAK DETECTION RESPONSE TIME FIGURE 5.2.5-1
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5.3-1 REV 19 4/15 5.3 REACTOR VESSEL 5.3.1 REACTOR VESSEL MATERIALS 5.3.1.1 Material Specifications Material specifications are in accordance with the American Society of Mechanical Engineers (ASME) Code requirements and are given in subsection 5.2.3. All ferritic reactor vessel
materials comply with the fracture toughness requirements of Section 50.55a and Appendices G
and H of 10 CFR 50.
The ferritic materials of the reactor vessel beltline are restricted to the following maximum limits of copper and phosphorus to reduce sensitivity to irradiation embrittlement in service:
Element Base Metal (percent)
As Deposited Weld
Metal (percent)
Copper 0.10 (ladle) 0.12 (check) 0.10 Phosphorus 0.012 (ladle) 0.017 (check) 0.020 5.3.1.2 Special Processes Used for Manufacturing and Fabrication A. The vessel is Safety Class 1. Design and fabrication of the reactor vessel is carried out in strict accordance with ASME Code,Section III, Class 1
requirements. The vessel head, flanges, and nozzles are manufactured as
forgings. The cylindrical portion of the vessel is made of several shells, each
consisting of formed plates joined by full penetration longitudinal and girth weld
seams. The hemispherical heads are made from dished plates. The reactor
vessel parts are joined by welding, using the single or multiple wire submerged
arc and the shielded metal arc processes. B. The use of severely sensitized stainless steel as a pressure boundary material has been prohibited and has been eliminated by either a select choice of material
or by programming the method of assembly. C. The control rod drive mechanism (CRDM) head adapter threads and surfaces of the guide studs are chrome plated to prevent possible galling of the mated parts. D. At all locations in the reactor vessel where stainless steel and Inconel are joined, the final joining beads are Inconel weld metal in order to prevent cracking. E. The location of full penetration weld seams in the upper closure head and vessel bottom head are restricted to areas that permit accessibility during inservice
inspection. F. The stainless steel clad surfaces are sampled to ensure that composition requirements are met.
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5.3-2 REV 19 4/15 G. Freedom from underclad cracking is ensured by special evaluation of the procedure qualification for cladding applied on low-alloy steel (SA-508, Class 2).
a H. Minimum preheat requirements have been established for pressure boundary welds using low-alloy material. The preheat is maintained until either an
intermediate postweld heat treatment or a full postweld heat treatment is
completed or until the completion of welding. I. A field weld is made, after the reactor vessel has been set, to install the permanent reactor vessel cavity seal ring. This stainless steel filler weld joins the
seal ring to the reactor vessel seal ledge. A minimum preheat is specified for this
weld in compliance with the ASME Code requirements. 5.3.1.3 Special Methods for Nondestructive Examination The nondestructive examination (NDE) of the reactor vessel and its appurtenances is conducted in accordance with ASME Code,Section III requirements; also, numerous examinations are
performed in addition to ASME Code,Section III requirements. The NDE of the vessel is
discussed in the following paragraphs, and the reactor vessel quality assurance program is
given in table 5.3.1-1. 5.3.1.3.1 Ultrasonic Examination A. In addition to the required ASME Code straight beam ultrasonic examination, angle beam inspection over 100 percent of one major surface of plate material is performed during fabrication to detect discontinuities that may be undetected by
the straight beam examination. B. In addition to the ASME Code,Section III NDE, all full penetration ferritic pressure boundary welds in the reactor vessel are ultrasonically examined during
fabrication. This test is performed upon completion of the welding and
intermediate heat treatment but prior to the final postweld heat treatment. C. After hydrotesting, all full penetration ferritic pressure boundary welds in the reactor vessel, as well as the nozzle to safe end welds, are ultrasonically
examined. These inspections are also performed in addition to the ASME Code,Section III NDE. 5.3.1.3.2 Penetrant Examinations The partial penetration welds for the CRDM head adapters and the bottom instrumentation tubes are inspected by dye penetrant after the root pass, in addition to code requirements.
Core support block attachment welds are inspected by dye penetrant after the first layer of weld
metal and after each 1/2 in. of weld metal. All clad surfaces and other vessel and head internal
surfaces are inspected by dye penetrant after the hydrostatic test.
a Underclad cracking of the reactor pressure ve ssel was evaluated as a time-limited aging analysis (TLAA) for license renewal in accordance with 10 CF R Part 54. The results of this evaluation are provided in paragraph 19.4.6.5.
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5.3-3 REV 19 4/15 5.3.1.3.3 Magnetic Particle Examination The magnetic particle examination requirements bel ow are in addition to the magnetic particle examination requirements of Section III of the ASME Code.
All magnetic particle examinations of materials and welds are performed in accordance with the following:
- Prior to the final postweld heat treatment, only by the prod, coil, or direct contact method.
- After the final postweld heat treatment, only by the yoke method.
The following surfaces and welds are examined by magnetic particle methods. The acceptance
standards are in accordance with Section III of the ASME Code. A. Surface Examinations 1. Magnetic particle examination of all exterior vessel and head surfaces after the hydrostatic test. 2. Magnetic particle examination of all exterior closure stud surfaces and all nut surfaces after final machining or rolling. Continuous circular and
longitudinal magnetization is used. 3. Magnetic particle examination of all inside diameter surfaces of carbon and low alloy steel products that have their properties enhanced by
accelerated cooling. This inspection is performed after forming and
machining and prior to cladding. B. Weld Examination Magnetic particle examination of the welds attaching the closure head lifting lugs and refueling seal ledge to the reactor vessel after the first layer and each 1/2 in.
of weld metal is deposited. All pressure boundary welds are examined after
back-chipping or back-grinding operations. 5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels Welding of ferrite steels and austenitic stainless steels is discussed in subsection 5.2.3.
Subsection 5.2.3 includes discussions which indicate the degree of conformance with
Regulatory Guide 1.44. Section 1.9 discusses the degree of conformance with Regulatory
Guides 1.43, 1.50, 1.71, and 1.99. 5.3.1.5 Fracture Toughness a Assurance of adequate fracture toughness of ferritic materials in the reactor vessel (ASME Code,Section III, Class 1 component) is provided by compliance with the requirements for
fracture toughness testing included in NB-2300 to Section III of the ASME Code and Appendix
G of 10 CFR 50.
The initial Charpy V-notch minimum upper shelf fracture energy levels for the reactor vessel beltline (including welds) are 75 ft-lb, as required by Appendix G of 10 CFR 50. The vessel a Reactor vessel neutron embrittlement was evaluated as a TLAA for license renewal in accordance with 10 CFR 54.21 (see subsection 19.4.1).
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5.3-4 REV 19 4/15 fracture toughness data for Units 1 and 2 are given in tables 5.3.1-2 and 5.3.1-3, respectively.
The end-of-life RT NDT and upper shelf energy projections estimated using Regulatory Guide 1.99 for the end-of-life neutron fluence at the 1/4 T and ID reactor vessel locations for Units 1
and 2 are given in tables 5.3.3-2 and 5.3.3-3. 5.3.1.6 Material Surveillance a The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties as required by 10 CFR Part 50, Appendix H, in accordance with the schedule in FSAR tables 5.3.1-8 and 5.3.1-9. The results of these examinations shall be used to update figures in the Pressure Temperature Limits Report (PTLR).(16) In the surveillance program, the evaluation of radiation damage is based on preirradiation testing of Charpy V-notch and tensile specimens and post irradiation testing of Charpy V-notch, tensile, and 1/2-T compact tension (CT) fracture mechanics test specimens.
The program is directed toward evaluation of the effect of radiation on the fracture toughness of
reactor vessel steels based on the transition temperature approach and the fracture mechanics
approach. The program conforms to American Society of Testing Materials (ASTM) E-185-82, Conducting Surveillance Tests for Light-Water-Cooled Nuclear Reactor Vessels, and 10 CFR
50, Appendix H. [HISTORICAL] The reactor vessel surveillance program uses six specimen capsules. The capsules are located in guide baskets welded to the outside of the neutron shield pads and positioned directly opposite the center portion of the core. The capsules can be removed when the vessel head is removed and can be replaced when the internals are removed. The six capsules contain reactor vessel steel specimens, oriented both parallel and normal (longitudinal and transverse) to the principal rolling direction of the limiting base material located in the core region of the reactor vessel and associated weld metal and weld heat-affected zone metal. The six capsules contain 54 tensile specimens, 360 Charpy V-notch specimens (which include weld metal and weld heat-affected zone material), and 72 CT specimens. Archive material sufficient for two additional capsules is retained at Westinghouse. The surveillance program withdrawal schedule, lead factor, test samples, and materials in the reactor vessel are given in tables 5.3.1-
7 and 5.3.1-8. [HISTORICAL]
Removal of these specimen capsules was completed for Unit 1 at refueling outage 1R14 and for Unit 2 at refueling outage 2R14. Refer to paragraph 5.3.1.6.1.4 for a description of the external
neutron monitoring system also known as exte rnal vessel neutron dosimetry system (EVNDS) used following removal of the last specimen capsules.
Dosimeters, as described below, are placed in filler blocks drilled to contain them. The dosimeters permit evaluation of the flux seen by the specimens and the vessel wall. In addition, thermal monitors made of low melting point alloys are included to monitor the maximum
temperature of the specimens. The specimens are enclosed in a tight-fitting stainless steel
sheath to prevent corrosion and ensure good thermal conductivity. The complete capsule is
helium leak tested. As part of the surveillance program, a report of the residual elements in
weight percent to the nearest 0.01 percent is made for surveillance material and as deposited
weld metal. Each of the six capsules contains the following specimens:
a The Reactor Vessel Surveillance Program is credited as a license renewal aging management program (see subsection 19.2.25).
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5.3-5 REV 19 4/15 Material Number of
Charpys Number of
Tensiles Number of
CTs Limiting base material a 15 3 4 Limiting base material b 15 3 4 Weld metal (c) 15 3 4 Heat-affected zone 15 - - The following dosimeters and thermal monitors are included in each of the six capsules:
A. Dosimeters
- 3. Nickel.
- 4. Cobalt-aluminum (0.15-percent cobalt).
- 5. Cobalt-aluminum (cadmium shielded).
- 6. Uranium-238 (cadmium shielded).
- 7. Neptunium-237 (cadmium shielded). B. Thermal Monitors 1. 97.5-percent lead, 2.5-percent silver, (579
°F melting point). 2. 97.5-percent lead, 1.75-percent silver, 0.75-percent tin (590
°F melting point).
The fast neutron exposure of the specimens occurs at a faster rate than that experienced by the vessel wall, with the specimens being located between the core and the vessel. Since these specimens experience accelerated exposure and are actual samples from the materials used in
the vessel, the transition temperature shift measurements are representative of the vessel at a
later time in life. Data from CT fracture toughness specimens are expected to provide additional
information for use in determining allowable stresses for irradiated material.
Correlations between the calculations and measurements of the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are
described in paragraph 5.3.1.6.1. The anticipated degree to which the specimens perturb the
fast neutron flux and energy distribution is considered in the evaluation of the surveillance
specimen data. Verification and possible readjustment of the calculated wall exposure is made
by the use of data on all capsules withdrawn. The schedule for removal of the capsules for
postirradiation testing conforms with ASTM E-185-82 and Appendix H of 10 CFR 50. 5.3.1.6.1 Measurement of Integrated Fast Neutron (E > 1.0 MeV) Flux at the Irradiation Samples The use of passive neutron sensors such as those included in the internal surveillance capsule dosimetry sets does not yield a direct measur e of the energy dependent neutron flux level at the a Specimens oriented in the major rolling or working direction.
b Specimens oriented normal to the major rolling or working direction.
(c) Weld metal to be selected in accordance with ASTM E-185-82.
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5.3-6 REV 19 4/15 measurement location. Rather, the activation or fission process is a measure of the integrated
effect that the time- and energy-dependent neutron flux has on the target material over the
course of the irradiation period. An accurate assessment of the average flux level and, hence, time integrated exposure (fluence) experienc ed by the sensors may be developed from the measurements only if the sensor characteristics and the parameters of the irradiation are well
known. In particular, the following variables are of interest:
- The measured specific activity of each sensor.
- The physical characteristics of each sensor.
- The operating history of the reactor.
- The energy response of each sensor.
- The neutron energy spectrum at the sensor location.
This section describes the procedures used to det ermine sensor specific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating
history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates. 5.3.1.6.1.1 Determination of Sensor Reaction Rates. The specific activity of each of the radiometric sensors is determined using established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor is determined by means of a high
purity germanium gamma spectrometer. In the case of the surveillance capsule multiple foil
sensor sets, these analyses are performed by direct counting of each of the individual wires or, as in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution
and chemical separation of cesium from the sensor.
The irradiation history of the reactor over its operating lifetime is determined from plant power generation records. In particular, operating data are extracted on a monthly basis from reactor
startup to the end of the capsule irradiation period. For the sensor sets utilized in the
surveillance capsule irradiations, the half-lives of the product isotopes are long enough that a
monthly histogram describing reactor operation has proven to be an adequate representation for
use in radioactive decay corrections for the reactions of interest in the exposure evaluations.
Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation are determined
from the following equation:
d t j teeCYFN A R j j=1 P P ref j 0 where: A = measured specific activity provided in terms of disintegrations per second per gram of target
material (dps/grn).
R = reaction rate averaged over the irradiation
period and referenced to operation at a core
power level of P ref expressed in terms of reactions per second per nucleus of target
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5.3-7 REV 19 4/15 isotope (rps/nucleus).
N 0 = number of target element atoms per gram of sensor. F = weight fraction of the target isotope in the sensor material. Y = number of product atoms produced per reaction.
P j = average core power level during irradiation
period j (MW).
P ref = maximum or reference core power level of the reactor (MW).
C j = calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period. = decay constant of the product isotope (sec
-1). t j = length of irradiation period j (sec).
t d = decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals comprising the total
irradiation period.
In the above equation, the ratio P j/P ref accounts for month-by-month variation of power level within a given fuel cycle. The ratio C j is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cy cle. Since the neutron flux at the measurement locations within the surveillance capsules is dominated by neutrons produced in the peripheral
fuel assemblies, the change in the relative power in these assemblies from fuel cycle to fuel
cycle can have a significant impact on the activation of neutron sensors. For a single-cycle
irradiation, C j = 1.0. However, for multiple-cycle irradiations, particularly those employing low-leakage fuel management, the additional C j correction must be utilized in order to provide accurate determinations of the decay-corrected reaction rates for the dosimeter sets contained
in the surveillance capsules. 5.3.1.6.1.2 Corrections to Reaction Rate Data. Prior to using the measured reaction rates in the least squares adjustment procedure discussed in paragraph 5.3.1.6.1.3, additional corrections are made to the U-238 measurements to account for the presence of U-235
impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the
course of the irradiation.
In addition to the corrections made for the presence of U-235 in the U-238 fission sensors, corrections are also made to both the U-238 and Np-237 sensor reaction rates to account for
gamma ray induced fission reactions occurring over the course of the irradiation.
VEGP-FSAR-5
5.3-8 REV 19 4/15 5.3.1.6.1.3 Least Squares Adjustment Procedure. Least squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation
resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best
Estimates for key exposure parameters such as neutron fluence (E > 1.0 MeV) or iron atom
displacements (dpa) along with their uncertainties are then easily obtained from the adjusted
spectrum. The use of measurements in combination with the analytical results reduces the
uncertainty in the calculated spectrum and acts to remove biases that may be present in the
analytical technique.
In general, the least squares methods, as applied to pressure vessel fluence evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, ()()+/-+/-=+/-g g g igRii ig R relates a set of measured reaction rates, R i , to a single neutron spectrum, g, through the multigroup dosimeter reaction cross-section, ig , each with an uncertainty . The use of least squares adjustment methods in light water reactor (LWR) dosimetry evaluations is not new. ASTM has addressed the use of adjustment codes in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance," and
many industry workshops have been held to discuss the various applications. For example, the ASTM-EURATOM Symposia on Reactor Dosime try holds workshops on neutron spectrum unfolding and adjustment techniques at each of its biannual conferences.
The primary objective of the least squares ev aluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement. The analytical method alone may be deficient because it inherently contains uncertainty due to the input assumptions to the
calculation. Typically these assumptions incl ude parameters such as the temperature of the water in the peripheral fuel assemblies, bypass region and downcomer regions, component
dimensions, and peripheral core source. Industry consensus indicates that the use of the
calculation alone results in overall uncertainties in the neutron exposure parameters in the
range of 15-20% (1). By combining the calculated results with avail able measurements, the uncertainties associated with the key neutron exposure parameters can be reduced. Specifically ASTM Standard E 944 states, "The algorithims of the adjustment codes tend to decrease the variances of the adjusted
data compared to the corresponding input values.
The least squares adjustment codes yield estimates for the output data with minimum variances , that is, the "best estimates." This is the primary reason for using these adjustment procedures." ASTM E 944 provides a
comprehensive listing of available adjustment codes.
The FERRET least squares adjustment code (1) was initially developed at the Hanford Engineering Development Laboratory (HEDL) and has had extensive use in both the Liquid Metal Fast Breeder (LMFBR) program and the NRC Sponsored Light Water Reactor Dosimetry
Improvement Program (LWR-PV-SDIP). As a re sult of participation in several cooperative efforts associated with the LWR-PV-SDIP, the FERRET approach was adopted by
Westinghouse in the mid 1980's as the preferred approach for the evaluation of LWR
surveillance dosimetry. The least squares me thodology was judged superior to the previously employed spectrum averaged cross-section approac h that is totally dependent on the accuracy of the shape of the calculated neutron spectrum at the measurement locations.
The FERRET code is employed to combine the results of plant specific neutron transport calculations and multiple-foil reaction-rate measurements to determine best estimate values of exposure parameters in terms of both neutron fluence greater than 1.0 MeV, (( E > 1.0 MeV),
VEGP-FSAR-5
5.3-9 REV 19 4/15 and iron atom displacements, (dpa), along with associated uncertainties in the measurement locations.
The application of the least squares methodology requires the following input:
- The calculated neutron energy spectrum and associated uncertainties at the measurement location.
- The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.
- The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
For a given application, the calculated neutron spectrum is obtained from the results of plant-specific neutron transport calculations applicable to the irradiation period experienced by the
dosimetry sensor set. This calculation is performed using the benchmarked transport
calculational methodology described in paragraph 5.3.1.6.2. The sensor reaction rates are
derived from the measured specific activities obtained from the counting laboratory using the
specific irradiation history of the sensor set to perform the radioactive decay corrections. The dosimetry reaction cross-sections and uncertain ties are obtained from the SNLRML dosimetry
cross-section library (2). The SNLRML library is an evaluat ed dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E1018, "Application
of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)." There are no additional data
or data libraries built into the FERRET code system. All of the required input is supplied
externally at the time of the analysis.
The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum are input to the least squares procedure in the form of variances
and covariances. The assignment of the input uncertainties also follows the guidance provided
in ASTM Standard E 944. 5.3.1.6.1.4 External Neutron Monitoring Syst em. The external neutron monitoring system also known as the external vessel neutron dosim etry system (EVNDS) provides for continuing neutron fluence measurement after sufficient specimen material exposure has been achieved and even after the last of the six internal surveillance capsules has been removed from the
reactor vessel. It enables verification of fast neutron exposure distributions within the reactor
vessel wall beltline region and establishes a mechanism to enable long term monitoring of this
portion of the reactor vessel as required per 10 CFR 50 Appendix H. These fluence data can
also support potential license renewal activities.
The EVNDS is located external to the reactor vessel, allowing for ease of dosimetry removal and replacement. It is installed in the annular air gap between the reactor vessel insulation and
the primary concrete shield wall. The EVNDS is a passive system consisting of six aluminum dosimeter capsules containing radiometric moni tors and four stainless steel gradient chains, which are bead chains connecting and supporting the dosimeter capsules. The bead chains are
in turn supported by an arrangement of stainless steel hardware-tubular brackets on a support
bar suspended by chains from bracket plate assembly, which is welded to the ventilation port
liner plate under a banana cover. The bead chains are mechanically secured to the concrete floor below the reactor vessel. The system is shown on drawings 1X6AB03-00020, 1X6AB03-00021, and 1X6AB03-00022 for Unit 1 and is shown on drawings 2X6AB03-00022, 2X6AB03-00023, and 2X6AB03-00024 for Unit 2.
VEGP-FSAR-5
5.3-10 REV 19 4/15 The EVNDS measures fluence for approximately 1/8 of the vessel wall circumference, positioned relative to well known reactor features. Neutron transport calculations then
determine the fluence for the entire vessel beltline wall. The system assists in the evaluation of
radiation damage to the reactor vessel beltline region by measuring the fluence to this region, which can be used to predict the shift in the reference nil ductility transition temperature (RT NDT). When used in conjunction with previously removed dosimetry from the internal surveillance
capsules and with the results of neutron transport calculations, the external vessel neutron
measurements allow the projection of embrittlement gradients through the reactor vessel wall
with minimum uncertainty. Minimizing the uncer tainty in the neutron exposure projections will help to assure that the reactor can be operated in the least restrictive mode possible with
respect to:
- 10 CFR 50 Appendix G pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system,
- Emergency Response Guideline (ERG) pre ssure / temperature limit curves, and
Comprehensive sensor sets are employed at discrete locations within the reactor cavity to
characterize the neutron energy spectrum variations axially and azimuthally over the beltline
region of the reactor vessel. In addition, the stainless steel gradient chains are used in
conjunction with the encapsulated dosimeters to complete the mapping of the neutron
environment between the discrete locations chosen for spectrum determinations.
The first replacement of irradiation dosimetry is at refueling outage 1R15 (Sept. 2009). The first replacement of irradiation dosimetry is at refueling outage 2R14 (March 2010). An irradiation
interval of five fuel cycles between replacements is typical. 5.3.1.6.2 Calculation of Integrated Fast (E > 1.0 MeV) Exposure at the Irradiation Samples and Reactor Vessel Wall Discrete ordinates transport calculations are performed on a fuel cycle-specific basis to
determine the neutron and gamma ray environment within the reactor geometry. The specific
methods applied have been benchmarked according to the guidelines of Regulatory
Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron
Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001, and have been approved by the NRC staff for general application to pressurized
water reactor (PWR) analysis. A description of the transport methodology along with the SER
documenting NRC staff approval of the method and computer codes are provided in
Reference 13.
In the application of this methodology to the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, a series of two-dimensional plant-specific transport
calculations are carried out and then synthesized to generate a three-dimensional neutron flux distribution, (r,,z) throughout the geometry of interest using the procedures outlined in Regulatory Guide 1.190. These three-dimensional mappings of the neutron environment are completed for each operating fuel cycle and then integrated to determine the neutron fluence
experienced by the surveillance test specimens and the pressure vessel wall.
In the approved analysis methodology, the transport calculations are completed using the DORT discrete ordinates code Version 3.1 (3) and the BUGLE-96 cross-section library (10). The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set VEGP-FSAR-5
5.3-11 REV 19 4/15 produced specifically for LWR application. In these analyses, anisotropic scattering is treated
with a P 5 legendre expansion and the angular discretization is modeled with an S 16 order of angular quadrature.
Energy- and space-dependent core power distributions, as well as system operating temperatures, are treated on a fuel cycle-specific basis. The spatial variation of the neutron
source is obtained from a burnup-weighted average of the respective power distributions from
individual fuel cycles including pinwise gradients for all fuel assemblies located on the periphery
of the core. The energy distribution of the source is determined on a fuel assembly-specific
basis and includes the effects of fissioning in both uranium and plutonium isotopes.
The results of the transport calculations are validated on a plant-specific basis by comparison with the results of surveillance capsule dosimetry developed using the procedures described in
paragraph 5.3.1.6.1. These comparisons are used to demonstrate that the plant-specific
application is consistent with the uncertainty evaluations provided in Reference 13 and to
establish that the 20% uncertainty criterion listed in Regulatory Guide 1.190 is met. These
comparisons are not used to modify or bias the results of the transport calculations. 5.3.1.6.2.1 Reference Forward Calculation. The forward transport calculation for the reactor is carried out in r, geometry using the DORT two-dimensional discrete ordinates code (3) and the BUGLE-96 cross-section library (10). The BUGLE-96 library is a 47 neutron group, ENDFB-VI based, data set produced specifically for LWR applications. In these analyses, anisotropic scattering is treated with a P 3 expansion of the scattering cross-sections and the angular discretization is modeled with an S 8 order of angular quadrature. The reference forward calculation is normalized to a core midplane power density characteristic of operation at the
stretch rating for the reactor. The spatial core power distribution utilized in the reference forward calculation is derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in the
development of this reference core power distribution is the use of an out-in fuel management
strategy; i.e., fresh fuel on the core periphery. Furthermore, a 2 uncertainty derived from the statistical evaluation of plant-to-plant and cycle-to-cycle variations in peripheral power is used
for the peripheral fuel assemblies.
Due to the use of this bounding spatial power distribution, the results from the reference forward calculation establish conservative exposure projec tions for reactors of this design operating at the stretch rating. Since it is unlikely that actual reactor operation would result in the
implementation of a power distribution at the nominal +2 level for a large number of fuel cycles and, further, because of the widespread implementation of low-leakage fuel management strategies, the fuel cycle-specific calculations for this reactor will result in exposure rates well
below these conservative predictions. 5.3.1.6.2.2 Cycle-Specific Adjoint Calculations. All adjoint analyses are also carried out using an S 8 order of angular quadrature and the P 3 cross-section approximation from the BUGLE-96 library. Adjoint source locations are chosen at several key azimuths on the pressure vessel inner radius. In
addition, adjoint calculations were carried out for sources positioned at the
geometric center of all surveillance capsules. Again, these calculations are
run in r, geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, (E > 1.0 MeV).
VEGP-FSAR-5
5.3-12 REV 19 4/15 The importance functions generated from these indi vidual adjoint analyses provide the basis for all absolute exposure projections and compar ison with measurement. These importance functions, when combined with cycle-specific neutron source distributions, yield absolute predictions of neutron exposure at the locations of interest for each of the operating fuel cycles
and establish the means to perform similar pr edictions and dosimetry evaluations for all subsequent fuel cycles.
Having the importance functions and appropriate core source distributions, the response of interest can be calculated as:
()()()dE ddrrE,,rSE,,r,R r Eoo= where: (R 0 ,0) = Neutron flux (E > 1.0 MeV) at radius R 0 and azimuthal angle 0. I(r,,E) = Adjoint importance function at radius r, azimuthal angle , and neutron source energy E.
S(r,,E) = Neutron source strength at core location r, and energy E.
It is important to note that the cycle-specific neutron source distributions, S(r,,E), utilized with the adjoint importance functions, l(r,,E), permit the use not only of fuel cycle-specific spatial variations of fission rates within the reactor core, but also allow for the inclusion of the effects of the differing neutron yield per fission and the variation in fission spectrum introduced by the
build-in of plutonium isotopes, as the burnup of individual fuel assemblies increases. 5.3.1.7 Reactor Vessel Fasteners The reactor vessel closure studs, nuts, and washers are designed and fabricated in accordance
with the requirements of the ASME Code,Section III. The closure studs are fabricated of SA-
540, Class 3, Grade B24. The closure stud material meets the fracture toughness requirements
of the ASME Code,Section III, and 10 CFR 50, Appendix G. Conformance with Regulatory
Guide 1.65, Materials and Inspections for Reactor Vessel Closure Studs, is discussed in section 1.9. Nondestructive examinations are performed in accordance with the ASME Code, Section
III. Bolting material properties for Units 1 and 2 are given in tables 5.3.1-4 and 5.3.1-5, respectively.
Refueling procedures require that the reactor vessel studs, nuts, and washers are lifted part of the way out of their respective holes and a stud support collar be put in place prior to the lift of
the integrated head assembly during preparation for refueling. In this way the studs are lifted
with and stored on the head. An alternative method of the procedures is that the reactor vessel
studs, nuts, and washers may be removed from the reactor closure and placed in storage racks
during preparation for refueling. In this method, the storage racks are removed from the
refueling cavity and stored at convenient locations on the containment operating deck prior to
removal of the reactor closure head and refueling cavity flooding. In either case, the reactor
closure studs are not exposed to the borated refueling cavity water. Additional protection
against the possibility of incurring corrosion effects is ensured by the use of a manganese base
phosphate surfacing treatment.
The stud holes in the reactor flange are sealed with special plugs before removing the reactor closure, thus preventing leakage of the borated refueling water into the stud holes.
VEGP-FSAR-5
5.3-13 REV 19 4/15 5.3.1.8 References 1. Schmittroth, E. A., "FERRET Data Analysis Code," HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979. 2. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium," Radiation Shielding Information Center, Oak Ridge National
Laboratory, July 1994. 3. RSICC Computer Code Collection CCC-650, "DOORS 3.1 One-, Two-, and Three-Dimensional Discrete Ordinates Neutron/
Photon Transport Code System," Radiation Shielding Information Center, Oak Ridge National Laboratory, August 1996. 4. Singer, L. R., "GPC Alvin W. Vogtle Unit No. 1 Reactor Vessel Radiation Surviellance Program," WCAP-11011, February 1986. 5. Singer, L. R., "GPC Alvin W. Vogtle Unit No. 2 Reactor Vessel Radiation Surviellance Program," WCAP-11381, April 1986. 6. Yanichko, S. E., et al., "Analysis of Capsule U From the GPC Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-12256, May 1989. 7. Terek, E., et al., "Analysis of Capsule U From the GPC VEGP Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-13007, August 1991. 8. Malone, M. J., et al., "Analysis of Capsule Y From the GPC Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-13931, February 1994. 9. Grendys, P. A., et al., "Analysis of Capsule Y From the GPC VEGP Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-14532, February 1996. 10. RSIC Data Library Collection DLC-185, "BUGLE 96 Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived From ENDF/B-VI for LWR Shielding and
Pressure Vessel Dosimetry Applications," Radiation Shielding Information Center, Oak
Ridge National Laboratory, March 1996. 11. Laubham, T. J., et al., "Analysis of Capsule V from the Georgia Power Vogtle Electric Generating Plant Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-15067, September 1998. 12. Laubham, T. J., et al., "Analysis of Capsule X from the Southern Nuclear Vogtle Electric Generating Plant Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-15159, March 1999. 13. J. D. Andrachek, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," WCAP-14040-A, Revision 4, May 2004. 14. RSIC Computer Code Collection PSR-145, "FERRET: Least-Squares Solution to Nuclear Data and Reactor Physics Problems," Radiation Shielding Information Center, Oak Ridge National Laboratory, January 1980. 15. Westinghouse LTR-REA-06-136, "Alvin W. Vogtle Units 1 and 2 Ex-Vessel Neutron Dosimetry System Description and Safety Evaluation Factors," (AX6AB03-00019). 16. NRC Letter to C. K. McCoy, "Vogtle Electric Generating Plant, Units 1 and 2 -
Acceptance for Referencing of Pressure Temperature Limits Report," February 12, 1996.
VEGP-FSAR-5
5.3-14 REV 19 4/15 5.3.2 PRESSURE-TEMPERATURE LIMITS 5.3.2.1 Limit Curves Startup and shutdown operating limitations are based on the properties of the reactor pressure vessel beltline materials. Actual material property test data are used. The methods outlined in Appendix G Section XI of the American Society of Mechanical Engineers (ASME) Code are
employed for the shell regions in the analysis of protection against nonductile failure. The initial
operating curves are calculated, assuming a period of reactor operation such that the beltline
material will be limiting. The heatup and cooldown curves are given in the Pressure and
Temperature Limits Report as required by the Technical Specifications. Beltline material
properties degrade with radiation exposure, and this degradation is measured in terms of the
adjusted reference nil ductility temperature, which includes a reference nil ductility temperature shift (RT NDT). The reference temperature, RT NDT , for materials in the reactor vessel closure flange region and the beltline regions are shown in tables 5.3.1-2 and 5.3.1-3. Tables 5.3.2-2 through 5.3.2-5 give the properties for the vessel beltline materials and data for the C v curve (energy vs. temperature).
Predicted RT NDT values are derived using guidance provided in Regulatory Guide 1.99 Revision 2. For a selected time of operation, this shift is assigned a sufficient magnitude so that no unirradiated ferritic materials in other components of the reactor coolant system (RCS) will
be limiting in the analysis.
The operating curves including pressure-temperature limitations shown in the Pressure and Temperature Limits Report are calculated in accordance with 10 CFR 50, Appendices G and H, and ASME Code,Section XI, Appendix G requirements, Code Case N-640 and WCAP-16142, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units 1 and
2, Revision 1."
The results of the material surveillance program described in paragraph 5.3.1.6 will be used to verify that the RT NDT predicted from the effects of the fluence, copper content curve is appropriate if RT NDT determined from the surveillance program is greater than the predicted RT NDT. Temperature limits for inservice leak and hydrotests will be calculated in accordance with ASME Code,Section XI, Appendix G C onformance with Regulatory Guide 1.99, Revision 2, is discussed in section 1.9. 5.3.2.2 Operating Procedures The transient conditions that are considered in the design of the reactor vessel are presented in paragraph 3.9.1.1. These transients are representative of the operating conditions that should
prudently be considered to occur during plant operation. The transients selected form a
conservative basis for evaluation of the RCS to ensure the integrity of the RCS equipment.
Those transients listed as upset condition transients are given in table 3.9.N.1-1. None of these transients will result in pressure-temperature changes which exceed the heatup and cooldown
limitations, as described in the Pressure and Temperature Limits Report.
VEGP-FSAR-5
5.3-15 REV 19 4/15 5.3.3 REACTOR VESSEL INTEGRITY 5.3.3.1 Design The reactor vessel is cylindrical with a welded hemispherical bottom head and a removable, bolted, flanged, and gasketed hemispherical upper head. The reactor vessel flange and head
are sealed by two hollow metallic O-rings. Seal leakage is detected by means of two leakoff connections, one between the inner and outer ring and one outside the outer O-ring. The
vessel contains the core, core support structures, control rods, and other parts directly
associated with the core. The reactor vessel closure head contains head adapters. These
head adapters are tubular members, attached by partial penetration welds to the underside of
the closure head. The upper end of these adapters contains acme threads for the assembly of
control rod drive mechanisms (CRDMs) or instru mentation adapters. The seal arrangement at the upper end of these adapters consists of a welded flexible canopy seal. Mechanical
assemblies may be used to fix or prevent leaks in the canopy seal weld. Inlet and outlet
nozzles are located symmetrically around the vessel. Outlet nozzles are arranged on the vessel
to facilitate optimum layout of the reactor c oolant system equipment. The inlet nozzles are tapered from the coolant loop vessel interfaces to the vessel inside wall to reduce loop pressure
drop. The bottom head of the vessel contains penetration nozzles for connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular member made of either an
Inconel or an Inconel-stainless steel composite tube. Each tube is attached to the inside of the
bottom head by a partial penetration weld.
Internal surfaces of the vessel which are in contact with primary coolant are weld overlay with 0.125-in. minimum of stainless steel or Inconel.
The reactor vessel is designed and fabricated in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Code,Section III. Principal design
parameters of the reactor vessel are given in table 5.3.3-1. The reactor vessel is shown in
figure 5.3.3-1.
There are no special design features which would prohibit the in situ annealing of the vessel. If the unlikely need for an annealing operation was required to restore the properties of the vessel material opposite the reactor core because of neutron irradiation damage, a metal temperature greater than 650
°F for a maximum period of 168 h would be applied.
(4) Various modes of heating may be used, depending on the temperature required.
The reactor vessel materials surveillance program is adequate to accommodate the annealing of the reactor vessel. Sufficient specimens are available to evaluate the effects of the annealing
treatment.
Cyclic loads are introduced by normal power changes, reactor trips, and startup and shutdown operations. These design base cycles are selected for fatigue evaluation and constitute a
conservative design envelope for the projected plant life
- a. Vessel analysis results in a usage factor that is less than 1.
The design specifications require analysis to prove that the vessel is in compliance with the fatigue and stress limits of the ASME Code,Section III. The loadings and transients specified
for the analysis are based on the most severe conditions expected during service. The heatup a Metal fatigue is evaluated as a TLAA fo r license renewal (see subsection 19.4.2).
VEGP-FSAR-5
5.3-16 REV 19 4/15 and cooldown rates are 100
°F/h for normal operations and under abnormal or emergency conditions. This rate is reflected in the vessel design specifications. 5.3.3.2 Materials of Construction The materials used in the fabrication of the reactor vessel are discussed in subsection 5.2.3. 5.3.3.3 Fabrication Methods The VEGP reactor vessel manufacturer is Combustion Engineering Corporation.
The fabrication methods used in the construction of the reactor vessel are discussed in paragraph 5.3.1.2. 5.3.3.4 Inspection Requirements The nondestructive examinations performed on the reactor vessel are described in
paragraph 5.3.1.3. 5.3.3.5 Shipment and Installation The reactor vessel is shipped in a horizontal position on a shipping sled with a vessel-lifting
truss assembly. All vessel openings are sealed to prevent the entrance of moisture, and an
adequate quantity of desiccant bags is placed inside the vessel. These are usually placed in a
wire mesh basket attached to the vessel cover. All carbon steel surfaces, except for the vessel
support surfaces and the top surface of the external seal ring, are painted with a heat-resistant
paint before shipment.
The closure head is also shipped with a shipping cover and skid. An enclosure attached to the ventilation shroud support ring protects the control rod mechanism housings. All head openings
are sealed to prevent the entrance of moisture, and an adequate quantity of desiccant bags is
placed inside the head. These are placed in a wire mesh basket attached to the head cover. All
carbon steel surfaces are painted with heat- resistant paint before shipment. A lifting frame is
provided for handling the vessel head. 5.3.3.6 Operating Conditions Operating limitations for the reactor vessel are presented in the Pressure Temperature Limit
Report (PTLR).
In addition to the analysis of primary components discussed in paragraph 3.9.1.4, the reactor vessel is further qualified to ensure against unstable crack growth under faulted conditions.
Actuation of the emergency core cooling system (ECCS) following a loss-of-coolant accident
produces relatively high thermal stresses in regions of the reactor vessel which come into
contact with ECCS water. Primary considerati on is given to these areas, including the reactor vessel beltline region and the reactor vessel primary coolant nozzle, to ensure the integrity of
the reactor vessel under this severe postulated transient. The Westinghouse Owners Group
evaluated TMI Action Item II.K.2.13, and the item is satisfied upon submittal of RT NDT values which are below the pressurized thermal shock (PTS) rule screening values. Additionally, new VEGP-FSAR-5
5.3-17 REV 19 4/15 calculations were performed based on the requirements of Generic Letter 92-01 and the results
are given in tables 5.3.3-2 and 5.3.3-3.
For the beltline region, significant developments have recently occurred in order to address PTS events. On the basis of recent deterministic and probabilistic studies, taking U.S. pressurized
water reactor operating experience into account, the Nuclear Regulatory Commission staff concluded that conservatively calculated screening criterion values of RT NDT less than 270
° for plate material and axial welds, and less than 300
° for circumferential welds, present an acceptably low risk of vessel failure from PTS events. These values were chosen as the screening criterion in the PTS rule for 10 CFR 50.34 (new plants) and 10 CFR 50.61 (operating
plants).(2) The conservative methods chosen by the NRC staff for the calculation of RT PTS for the purpose of comparison with the screening criterion is presented in paragraph (b)(2) of
10 CFR 50.61. Details of the analysis method and the basis for the PTS rule can be found in
(3) The reactor vessel beltline materials are specified in subsection 5.3.1. The fluence of 4.76 x
10 19 n/cm 2 which is the design basis fluence at the vessel inner radius, at 48 EFPY, at the peak location, was used for calculating all of the RT PTS values. Based on the latest capsule data for each unit, the expected fluence at the vessel inner radius after 56.3 EFPY will be significantly
less than the 4.76 x 10 19 n/cm 2 design basis fluence
- a. RT PTS is RT NDT, the reference nilductility transition temperature as calculated by the method chosen by the NRC staff as presented in
paragraph (b)(2) of 10 CFR 50.61, and the PTS rule. The PTS rule states that this method of
calculating RT should be used in reporting values used to be compared to the above screening
criterion set in the PTS rule. The screening criteria will not be exceeded using the method of
calculation prescribed by the PTS rule for the vessel design lifetime. The material properties, initial RT NDT , and end-of-life RT PTS values are in tables 5.3.3-2 and 5.3.3-3. The materials identified in tables 5.3.3-2 and 5.3.3-3 are those materials that are exposed to high fluence
levels at the beltline region of the reactor vessel and are, therefore, the subject of the PTS rule.
These materials, therefore, are a subset of the materials identified in subsection 5.3.1.
The principles and procedures of linear elastic fracture mechanics (LEFM) are used to evaluate thermal effects in the regions of interest. The LEFM approach to the design against failure is
basically a stress intensity consideration in which criteria are established for fracture instability
in the presence of a crack. Consequently, a bas ic assumption employed in LEFM is that a crack or crack-like defect exists in the structure. The essence of the approach is to relate the
stress field developed in the vicinity of the crack tip to the applied stress on the structure, the
material properties, and the size of defect necessary to cause failure.
The elastic stress field at the crack tip in any cracked body can be described by a single parameter designated as the stress intensity factor, K. The magnitude of the stress intensity
factor K is a function of the geometry of the body containing the crack, the size and location of
the crack, and the magnitude and distribution of the stress.
The criterion for failure in the presence of a crack is that failure will occur whenever the stress intensity factor exceeds some critical value. For the opening mode of loading (stresses
perpendicular to the major plane of the crack), the stress intensity factor is designated as K I and the critical stress intensity factor is designated K IC. Commonly called the fracture toughness, K IC is an inherent material property which is a function of temperature. Any combination of applied load, structural configuration, crack geometry, and size which yields a stress intensity
factor greater than K IC for the material will result in crack instability. The criterion of the applicability of LEFM is based on plasticity considerations at the postulated
crack tip. Strict applicability (as defined by American Society of Testing Materials (ASTM)) of a Reactor vessel embrittlement is evaluated as a TLAA for license renewal (see subsection 19.4.1).
VEGP-FSAR-5
5.3-18 REV 19 4/15 LEFM to large structures where plane strain conditions prevail requires that the plastic zone
developed at the tip of the crack does not exceed approximately 2 percent of the crack depth.
In the present analysis, the plastic zone at the tip of the postulated crack can reach 20 percent
of the crack depth. However, LEFM has been successfully used to provide conservative brittle
fracture prevention evaluations, even in cases where strict applicability of the theory is not
permitted due to excessive plasticity. Recently , experimental results from the Heavy Section Steel Technology program intermediate pressure vessel tests have shown that LEFM can be applied conservatively as long as the pressure component of the stress does not exceed the
yield strength of the material. The addition of the elastically calculated thermal stresses, which
results in total stresses in excess of the yield strength, does not affect the conservatism of the
results, provided that these thermal stresses are included in the evaluation of the stress
intensity factors. Therefore, for faulted conditions analyses, LEFM is considered applicable for
the evaluation of the vessel inlet nozzle and beltline region.
In addition, it has been well established that the crack propagation of existing flaws in a structure subjected to cyclic loading can be defined in terms of fracture mechanics parameters.
Thus, the principles of LEFM are also applicable to fatigue growth of a postulated flaw at the
vessel inlet nozzle and beltline region.
Additional details on this method of analysis of reactor vessels under severe thermal transients are given in reference 1. 5.3.3.7 Inservice Surveillance The internal and external surfaces of the reactor vessel are accessible for periodic inspection.
Visual and/or nondestructive techniques are used. During refueling, the vessel cladding is
capable of being inspected in certain areas between the closure flange and the primary coolant
inlet nozzles, and, if deemed necessary, the core barrel is capable of being removed, making
the entire inside vessel surface accessible.
The closure head is examined visually during each refueling. Optical devices permit a selective inspection of the cladding, CRDM nozzles, and the gasket seating surface. The knuckle transition piece, which is the area of highest stress of the closure head, is accessible on the
outer surface for visual inspection, dye penetrant or magnetic particle testing, and ultrasonic
testing. The closure studs and nuts can be inspected periodically using visual, magnetic
particle, and ultrasonic techniques.
The closure studs, nuts, washers, and the vessel flange seal surface, as well as the full-penetration welds in the following areas of the installed reactor vessel, are available for
nondestructive examination: A. Vessel shell, from the inside and outside surfaces. B. Primary coolant nozzles, from the inside and outside surfaces.(a) C. Closure head, from the inside and outside surfaces; bottom head, from the inside and outside surfaces. D. Field welds between the reactor vessel nozzle safe ends and the main coolant piping, from the inside and outside surfaces.
The design considerations which have been incorporated into the system design to permit the above inspection are as follows: (a) Only partial outside diameter coverage is provided.
VEGP-FSAR-5
5.3-19 REV 19 4/15 A. All reactor internals are completely removable. The tools and storage space required to permit these inspections are provided. B. The closure head is stored dry on the reactor operating deck during refueling to facilitate direct visual inspection. C. All reactor vessel studs, nuts, and washers can be removed to dry storage during refueling. D. Access is provided to the reactor vessel nozzle safe ends. The insulation covering the nozzle-to-pipe welds may be removed.
The reactor vessel presents access problems because of the radiation levels and remote
underwater accessibility to this component. Because of these limitations on access to the
reactor vessel, several steps have been incorporated into the design and manufacturing
procedures in preparation for the periodic nondestructive tests which are required by the ASME
inservice inspection code. These are as follows: A. Shop ultrasonic examinations are perform ed on all internally clad surfaces to an acceptance and repair standard to ensure an adequate cladding bond to allow
later ultrasonic testing of the base metal from inside surface. The size of
cladding bond defect allowed is 1/4 in. by 3/4 in. with the greater direction parallel
to the weld in the region bounded by 2T (T = wall thickness) on both sides of
each full-penetration pressure boundary weld. Unbounded areas exceeding
0.442 in.2 (3/4-in. diameter) in all other regions are rejected. B. The design of the reactor vessel shell is an uncluttered cylindrical surface to permit future positioning of the test equipment without obstruction. C. The weld-deposited clad surface on both sides of the welds to be inspected is specifically prepared to ensure meaningful ultrasonic examinations. D. During fabrication, all full-penetration ferritic pressure boundary welds are ultrasonically examined in additi on to code examinations. E. After the shop hydrostatic testing, all full-penetration ferritic pressure boundary welds (with the exception of the closure head welds), as well as the nozzles to
safe end welds, are ultrasonically examined from both the inside and outside
diameters in addition to ASME Code,Section III requirements. The closure head
ferritic pressure boundary welds are examined from the outside diameter only.
The vessel design and construction enables inspection in accordance with the ASME Code,Section XI. The reactor vessel inservice inspection requirements are detailed in the VEGP
inservice inspection program.
The Reactor Vessel Closure Head Stud Program is credited as a license renewal aging management program (see subsection 19.2.23). 5.3.3.8 References 1. Buchalet, C., Bamford, W. H., and Chirigos, J. N., "Method for Fracture Mechanics Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients," WCAP-8510, December 1975. 2. PTS Rule, Federal Register Vol. 50, No. 141, July 23, 1985, 10 CFR 50.34. 3. NRC Policy Issue, "Pressurized Thermal Shock," SECY-82-465, November 23, 1982.
VEGP-FSAR-5
5.3-20 REV 19 4/15 4. EPRI NP 2712, "Feasibility of and Methodology for Thermal Annealing an Embrittled Reactor Vessel," November 1982.
VEGP-FSAR-5 REV 13 4/06 TABLE 5.3.1-1 (SHEET 1 OF 2) REACTOR VESSEL QUALITY ASSURANCE PROGRAM
RT (a) UT (a) PT (a) MT (a) Forgings Flanges Yes Yes Studs and nuts Yes Yes CRDM head adapter flange Yes Yes CRDM head adapter tube Yes Yes Instrumentation tube Yes Yes Main nozzles Yes Yes Nozzle safe ends Yes Yes Plates Yes Yes Weldments CRDM head adapter to closure head connection Yes Instrumentation tube to bottom head connection Yes Main nozzle Yes Yes Yes Cladding Yes Yes Nozzle to safe ends Yes Yes Yes CRDM head adapter flange to CRDM head adapter tube Yes Yes All full-penetration ferritic pressure boundary welds accessible after hydrotest Yes Yes Full-penetration nonferritic pressure boundary welds accessible after hydrotest
- a. Nozzle to safe ends Yes Yes b. CRDM head adapter flange to CRDM head adapter tube Yes Seal ledge Yes Head lift lugs Yes Core pad welds Yes
VEGP-FSAR-5 REV 13 4/06 TABLE 5.3.1-2 (a)
VEGP UNIT 1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES Component Code No. Material Spec. No.
Cu
(%) Ni
(%)
P
(%)
(°F)
RT NDT (°F) Use NMWD(b) (ft-lb) Closure head dome B8807-1 A533B Cl. 1 0.16 0.67 0.008 -50 15 88 Closure head torus(c) B8808-1 A533B Cl. 1 0.14 0.56 0.010 -30 8 85 Closure head flange(c) B8801-1 A508 Cl. 2 - 0.70 0.011 20 20 132 Vessel flange(c) B8802-1 A508 Cl. 2 - 0.71 0.014 0 0 119 Inlet nozzle B8809-1 A508 Cl. 2 - 0.86 0.011 20 107 Inlet nozzle B8809-2 A508 Cl. 2 - 0.84 0.014 10 95 Inlet nozzle B8809-3 A508 Cl. 2 - 0.82 0.013 10 117 Inlet nozzle B8809-4 A508 Cl. 2 - 0.87 0.014 20 105 Outlet nozzle B8810-1 A508 Cl. 2 - 0.82 0.006 10 >124 Outlet nozzle B8810-2 A508 Cl. 2 - 0.79 0.006 10 >100 Outlet nozzle B8810-3 A508 Cl. 2 - 0.77 0.006 10 >102 Outlet nozzle B8810-4 A508 Cl. 2 - 0.80 0.006 10 > 75 Nozzle shell B8804-1 A533B Cl. 1 0.14 0.62 0.011 -10 28 94 Nozzle shell B8804-2 A533B Cl. 1 0.10 0.58 0.006 -40 15 104 Nozzle shell B8804-3 A533B Cl. 1 0.14 0.69 0.013 -30 40 92 Intermediate shell(c) B8805-1 A533B Cl. 1 0.08 0.59 0.004 0 0 90 Intermediate shell(c) B8805-2 A533B Cl. 1 0.08 0.59 0.004 -10 20 100 Intermediate shell(c) B8805-3 A533B Cl. 1 0.06 0.60 0.003 -20 30 107 Lower shell(c) B8606-1 A533B Cl. 1 0.05 0.59 0.005 -50 20 116 Lower shell(c) B8606-2 A533B Cl. 1 0.05 0.58 0.009 -10 20 113 Lower shell(c) B8606-3 A533B Cl. 1 0.06 0.64 0.007 -20 10 118 Bottom head torus B8813-1 A533B Cl. 1 0.13 0.50 0.009 10 88 Bottom head dome B8812-1 A533B Cl. 1 0.10 0.53 0.009 28 122 Intermediate and lower(c) G1.43 SAW 0.03 0.10 0.007 80 129 shell vertical weld seams and girth
- a. This table is based on initial testing of the vessel materials at the time the surveillance capsule program was developed.
- b. Normal to major working direction.
- c. Denotes materials in reactor vessel closure flange and beltline region.
VEGP-FSAR-5
REV 13 4/06 TABLE 5.3.1-3(a) VEGP UNIT 2 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES Component Code No. Material Spec. No. Cu
(%) Ni
(%) P
(%) NDT
(°F) RT NDT (°F) Use NMWD(b) (ft-lb) Closure head dome R9-1 A533B CI. 1 0.07 0.61 0.008 30 123 Closure head torus(c) R10-1 A533B Cl. 1 0.07 0.64 0.010 0 84 Closure head flange(c) R7-1 A508 Cl. 2 - 0.72 0.011 10 10 130 Vessel flange(c) R1-1 A508 Cl. 2 - 0.87 0.011 60 115 Inlet nozzle B9806-1 A508 Cl. 2 0.07 0.84 0.010 50 119 Inlet nozzle B9806-2 A508 Cl. 2 0.06 0.83 0.009 40 128 Inlet nozzle R5-1 A508 Cl. 2 0.09 0.87 0.008 20 147 Inlet nozzle R5-2 A508 Cl. 2 0.08 0.85 0.009 20 134 Outlet nozzle R6-3 A508 Cl. 2 - 0.69 0.011 10 122 Outlet nozzle R6-4 A508 Cl. 2 - 0.66 0.010 10 140 Outlet nozzle B9807-3 A508 Cl. 2 - 0.66 0.005 30 116 Outlet nozzle B9807-4 A508 Cl. 2 - 0.64 0.010 -10 10 132 Nozzle shell R3-1 A533B Cl. 1 0.20 0.67 0.015 0 -20 79 Nozzle shell R3-2 A533B Cl. 1 0.20 0.67 0.015 0 -40 79 Nozzle shell R3-3 A533B Cl. 1 0.15 0.62 0.010 60 84 Intermediate shell(c) R4-1 A533B Cl. 1 0.06 0.64 0.009 -20 10 95 Intermediate shell(c) R4-2 A533B Cl. 1 0.05 0.62 0.009 -10 10 104 Intermediate shell(c) R4-3 A533B Cl. 1 0.05 0.59 0.009 0 30 84 Lower shell(c) B8825-1 A533B Cl. 1 0.05 0.59 0.006 -20 40 83 Lower shell(c) R8-1 A533B Cl. 1 0.06 0.62 0.007 -20 40 87 Lower shell(c) B8628-1 A533B Cl. 1 0.05 0.59 0.007 -20 50 85 Bottom head torus R12-1 A533B CI. 1 0.17 0.64 0.012 20 89 Bottom head dome R11-1 A533B Cl. 1 0.10 0.62 0.008 30 115 Intermediate and lower(c) G1.60 SAW 0.07 0.13 0.007 10 147 shell vertical weld seams and girth Intermediate and lower(c) E3.23 SAW 0.06 0.12 0.007 30 90 shell vertical weld seams and girth
- a. This table is based on initial testing of the vessel materials at the time the surveillance capsule program was developed.
- b. Normal to major working direction.
- c. Denotes materials in reactor vessel closure flange and beltline region.
VEGP-FSAR-5 REV 17 4/12 TABLE 5.3.1-7
[HISTORICAL]
REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM VEGP Unit 1 Capsules U, V, W, X, Y, and Z Material Charpy Tensile 1/2T-CT Plate B8805-3 (long.)
15 3 4 Plate B8805-3 (trans.)
15 3 4 Weld metal (G-1.43) 15 3 4 HAZ 15 - - VEGP Unit 2 Capsules U, V, W, X, Y, and Z Material Charpy Tensile 1/2T-CT Plate B8628-1 (long.)
15 3 4 Plate B8628-1 (trans.)
15 3 4 Weld metal (E-3.23) 15 3 4 HAZ 15 - -
VEGP-FSAR-5
REV 17 4/12 TABLE 5.3.1-9 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE (UNIT 2)
Capsule Number Vessel Location Lead Factor (a) Withdrawal Time EFPY (b) Approximate Capsule Fluence (n/cm 2 , E > 1.0 MeV) (a) U 58.5° 4.10 1.20 3.56 x 10 18 (c) Y 241° 3.95 4.98 1.12 x 10 19 (c) X 238.5° 4.25 7.78 1.78 x 10 19 (c) W 121.5° 4.14 13.29 2.98 x 10 19 (c) (d) Z 301.5° 4.15 18.48 4.16 x 10 19 (c) V 61° 3.84 18.48 (e)-------
- a. Updated in Capsule W dosimetry analysis.
- b. Effective Full Power Years (EFPY) from plant startup.
- c. Plant-specific evaluation.
- d. This capsule was withdrawn at a fluence not less than once nor greater than twice the peak EOL fluence for a standard license term of 40 years (36 EFPY). In addition, this
capsule was withdrawn at a fluence not less than once nor greater than twice the
peak EOL fluence for an additional 20-year license renewal term to 60 years (54
EFPY). e. Capsule V has been removed from the reactor vessel and placed in the spent fuel pool. No testing or analysis has been performed on this capsule. Reinsertion of this capsule may be considered in the future.
VEGP-FSAR-5 REV 13 4/06 TABLE 5.3.2-4 (SHEET 1 OF 2)
VEGP UNIT 2 REACTOR VESSEL CORE BELTLINE REGION TOUGHNESS PROPERTIES Intermediate Shell Course
Plate R4-1 Plate R4-2 Plate R4-3 Temp. Energy Lat. Exp Shear Temp. Energy Lat. Exp. Shear Temp. Energy Lat. Exp. Shear (F) (ft lb) (mils) (%) (F) (ft lb) (mils) (%) (F) (ft lb) (mils) (%) -40 11 6 0 -40 6 4 0 -40 9 6 0 -40 10 5 0 -40 7 4 0 -40 7 5 0 -40 12 7 0 -40 8 5 0 -40 7 4 0 10 30 24 15 10 23 17 10 10 19 16 5 10 27 19 10 10 33 27 15 10 18 15 5 10 33 24 15 10 38 31 20 10 19 18 5 40 43 33 25 50 36 28 15 60 37 30 15 40 45 34 25 50 40 34 20 60 40 36 20 40 40 31 20 50 43 35 20 60 55 43 35 60 52 42 35 60 63 49 40 80 42 35 30 60 51 43 35 60 66 50 40 80 44 38 30 60 46 36 30 60 48 34 25 80 48 41 30 70 61 47 40 70 66 53 40 90 59 51 40 70 72 54 50 70 59 47 35 90 52 46 35 70 54 43 40 70 51 39 30 90 50 41 35 100 80 65 80 100 88 67 80 100 70 51 70 100 86 67 90 100 85 64 80 100 67 50 70 100 91 70 90 100 96 65 95 100 74 59 80 160 97 72 100 160 103 69 100 160 86 68 100 160 95 69 100 160 99 70 100 160 89 66 100 160 93 67 100 160 110 74 100 160 78 61 100 T NDT = -20 F T NDT = -10 F T NDT = 0 F RT NDT = 10 F RT NDT = 10 F RT NDT = 30 F VEGP-FSAR-5 REV 13 4/06 TABLE 5.3.2-4 (SHEET 2 OF 2)
Lower Shell Course Plate B8825-1 Plate R8-1 Plate B8628-1 Temp. Energy Lat. Exp Shear Temp. Energy Lat. Exp. Shear Temp. Energy Lat. Exp. Shear (F) (ft lb) (mils) (%) (F) (ft lb) (mils) (%) (F) (ft lb) (mils) (%)
-40 11 7 0 -40 14 7 5 -40 11 4 0
-40 11 8 0 -40 12 5 0 -40 9 3 0
-40 10 5 0 -40 11 4 0 -40 11 4 0 0 16 10 10 0 22 18 20 0 15 12 10 0 14 9 10 0 22 16 20 0 15 11 10 0 14 9 10 0 20 16 20 0 15 11 10 40 37 25 30 40 33 22 25 40 26 20 30 40 28 20 30 40 34 23 25 40 25 18 30 40 32 23 30 40 35 25 25 40 22 17 30 90 51 38 50 90 49 39 40 100 55 45 40 90 49 35 50 90 62 46 60 100 43 35 30 90 56 42 50 90 62 46 60 100 48 36 35 100 55 41 50 100 67 50 65 110 61 50 70 100 60 47 50 100 68 51 65 110 61 49 70 100 61 49 50 100 69 52 65 110 58 45 70 160 76 60 90 160 85 64 95 160 76 60 95 160 79 64 90 160 86 65 95 160 70 67 95 160 77 60 90 160 90 66 95 160 68 58 95 212 86 66 100 212 83 61 100 212 76 60 100 212 82 65 100 212 86 64 100 212 70 58 100 212 80 64 100 212 93 68 100 212 70 57 100 275 76 60 100 275 82 65 100 275 96 68 100
T NDT = -20 F T NDT = -20 F T NDT = -20 F RT NDT = 40 F RT NDT = 40 F RT NDT = 50 F
VEGP-FSAR-5
REV 16 10/10 TABLE 5.3.3-2 UNIT 1 REACTOR VESSEL VALUES FOR ANALYSIS OF POTENTIAL PRESSURIZED THERMAL SHOCK EVENTS(a)
Initial 10 CFR 50.61 Predicted RTPTS (°F)
Regulatory Guide 1.99 Predicted USE (ft-lb)
Material Cu wt-% Ni wt-% RT NDT (°F) 36 EFPY 57 EFPY Initial USE (ft-lb) 36 EFPY 57 EFPY Intermediate Shell Plate, B8805-1 0.083 0.597 0 98 104 90 72 69 Intermediate Shell Plate, B8805-2 0.083 0.61 20 118 124 100 80 77 Intermediate Shell Plate, B8805-3(b) 0.062 0.598 30 110 115 107 86 (e) 82 (e) 121(d) 126(d) Lower Shell Plate, B8606-1 0.053 0.593 20 94 97 116 93 89 Lower Shell Plate, B8606-2 0.057 0.60 20 97 101 113 90 87 Lower Shell Plate, B8606-3 0.067 0.623 10 95 100 118 94 91 Intermediate Shell Longitudinal Weld Seams 101-124 A, B, & C (c) 0.042 0.102 -80 3 -24(d) 11 -21(d) 134 107(e) 103(e) Lower Shell Longitudinal Weld Seams 101-142 A, B, & C (c) 0.042 0.102 -80 3 -24(d) 11 -21(d) 134 107 (e) 103 (e) Intermediate to Lower Shell Girth Weld 101-171 (c) 0.042 0.102 -80 3 -24(d) 11 -21(d) 134 107(e) 103(e) NOTES: a. RTPTS values are based on the peak fluences at the vessel inner radius of 2.155 E19 (for 36 EFPY) and 3.485 E19 (for 57 EFPY). USE was predicted using the 1/4T fluence values based on the peak fluence at the vessel inner radius. The vessel wa ll thickness is 8.625 inches at the beltline region. Copper and nickel values for all materials are the latest Best Estimate values as of April 2005. b. Limiting vessel material for Pressurized Thermal Shock event. c. All of the core region welds were fabricated from wire heat 83653, linde 0091 flux, lot # 3536. d. Determined using 10 CFR 50.61 with credible surveillance capsule data for the welds and noncredible surv eillance capsule dat a for the plate. e. Conservatively determined using Position 1.2 (without surveillance capsule data) of Regulatory Guide 1.99, Revision 2; however, surveillance data were available.
VEGP-FSAR-5 REV 16 10/10 TABLE 5.3.3-3 UNIT 2 REACTOR VESSEL VALUES FOR ANALYSIS OF POTENTIAL PRESSURIZED THERMAL SHOCK EVENTS(a)
Initial 10 CFR 50.61 Predicted RTPTS (°F)
Initial Regulatory Guide 1.99 Predicted USE (ft-lb)
Material Cu wt-% Ni wt-% RT NDT (°F) 36 EFPY 57 EFPY USE (ft-lb) 36 EFPY 57 EFPY Intermediate Shell Plate, R4-1 0.07 0.63 10 95.9 101.0 95 76 74 Intermediate Shell Plate, R4-2 0.06 0.61 10 87.7 91.9 104 83 81 Intermediate Shell Plate, R4-3 0.05 0.60 30 100.6 104.1 84 67 66 Lower Shell Plate, B8825-1 0.06 0.62 40 117.7 121.9 83 66 65 Lower Shell Plate, R8-1 (b) 0.07 0.63 40 125.9 131.0 87 70 68 Lower Shell Plate, B8628-1 0.05 0.59 50 120.6 124.1 85 79 (e) 79 (e) 91.1 (d) 93.4 (d) Intermediate Shell Longitudinal Weld Seams 101-124 A, B, & C (c) 0.05 0.15 -10 92.2 102.1 152 140(e) 140(e) 40.8 (d) 45.6 (d) Lower Shell Longitudinal Weld Seams 101-142 A, B, & C (c) 0.05 0.15 -10 92.2 102.1 152 140(e) 140(e) 40.8 (d) 45.6 (d) Intermediate to Lower Shell Girth Weld (c) 0.05 0.15 -30 72.2 20.8(d) 82.1 25.6(d) 90 83(e) 83(e) NOTES: a. RTPTS values are based on the peak fluence at the vessel inner radius of 1.93 E19 (for 36 EFPY) and 3.06 E19 (for 57 EFPY). USE was predicted using the 1/4T fluence values based on the peak fluence at the vessel inner radius. The vessel wall thickness is 8.625 inches at the beltline region. Copper and nickel values for all materials are the latest Best Estimate values as of April 2005. b. Limiting vessel material for Pressurized Thermal Shock event. c. The longitudinal welds were fabricated from wire heat 87005 linde 0091 flux, lot 0145. The girth weld was fabricated from weld wire heat 87005, linde 124 flux, lot 1061. d. Determined using 10 CFR 50.61 with credible surveillance capsule data. e. Determined using Position 2.2 (with credible surveillance capsule data) of Regulatory Guide 1.99, Revision 2.
REV 13 4/06 REACTOR VESSEL FIGURE 5.3.3-1
VEGP-FSAR-5 REV 13 4/06 TABLE 5.4.1-1 (SHEET 1 OF 2)
REACTOR COOLANT PUMP DESIGN PARAMETERS Unit desi g n p ressure (p si g) 2485 Unit desi gn temperature (F) 650 (a) Unit overall hei g ht (ft) 27.4 Seal water in j ection (g al/min) 8 Seal water return (g al/min) 3 Com p onent coolin g waterflow (g al/min)596 Maximum continuous com p onent coolin g water inlet temperature (F) 105 Total wei ght, dr y (lb) 201,300 Pum p Desi g n flow (g al/min) 100,600 Develo p ed head (ft) 288 NPSH re quired (ft) Fi gure 5.4.1-2 Suction temperature, thermal desi g n (F)558.2 Pum p dischar ge nozzle, inside diameter (in.)27-1/2 Pum p suction nozzle, inside diameter (in.)31 Speed (rpm) 1187 Water volume (ft 3) 80 (b)
Moto r T ype Drip-proo f s q uirrel ca g e induction, with water/ai r coolers Power (h p) 7000 Volta g e (V) 13,200 Phase 3 Fre q uenc y (Hz) 60Insulation class Class F thermalastic e p ox y insulation
Current (am p) Startin g 1750 at 13,200 V Nominal input, hot reactor coolant 253 Nominal in p ut, cold reactor coolant 336 VEGP-FSAR-5 REV 13 4/06 TABLE 5.4.1-1 (SHEET 2 OF 2)
Pump moment of inertia, maximum (lb/ft 2) Fl y wheel 70,000 Moto r 22,500 Shaft 520 Im p elle r 1980
- a. Design temperature of pressure-retaining parts of the pumpassembly exposed to the reactor coolant and injection water on the high-pressure side of the controlled leakage seal is that temperature determined for the parts for a reactor coolant loop
temperature of 650F. b. Composed of reactor coolant in the casing and of seal injection and cooling water in the thermal barrier.
VEGP-FSAR-5 REV 13 4/06 TABLE 5.4.2-2 (SHEET 1 OF 2)
STEAM GENERATOR QUALITY ASSURANCE PROGRAM RT (a) UT (a) PT (a) MT(a) ET (a) Tube Sheet Forging Yes Yes Cladding Yes (b) Yes Channel Head (if fabricated)
Fabrication Yes(c) Yes (d) Yes Cladding Yes Secondary Shell and Head Plates Yes Tubes Yes Yes Nozzles (Forgings)
Yes Yes Weldments Shell, longitudinal Yes Yes Shell, circumferential Yes Yes Cladding (channel Yes head-tube sheet joint cladding restoration) Primary nozzles to Yes Yes fab head Manways to fab head Yes Yes Steam and feedwater Yes Yes nozzles to shell Support brackets Yes Tube to tube sheet Yes Instrument connections Yes (primary and secondary)
Temporary attachments Yes after removal
VEGP-FSAR-5 REV 13 4/06 TABLE 5.4.2-2 (SHEET 2 OF 2)
RT (a) UT (a) PT (a) MT(a) ET (a) After hydrostatic test Yes (all major pressure boundary welds and complete cast channel head - where accessible)
Yes Nozzle safe ends (if Yes Yes weld deposit)
- a. RT - Radiographic.
UT - Ultrasonic.
PT - Dye penetrant. MT - Magnetic particle. ET - Eddy current.
- b. Flat surfaces only.
- c. Weld deposit.
- d. Base material only.
VEGP-FSAR-5 REV 13 4/06 TABLE 5.4.7-4 (SHEET 1 OF 5)
RESIDUAL HEAT REMOVAL SYSTEM - SAFETY GRADE COLD SHUTDOWN OPERATIONS - FAILURE MODES AND EFFECTS ANALYSIS
Effect on Failure Detection Component(a) Failure Mode Function(b) System Operation Methods(c) Remarks
- 1. Motor-operated a. Fails to open Provides a. Failure blocks reactor a. Valve open/close Valve is electrically gate valve on demand. isolation of coolant flow from hot position indication interlocked with HV-8701A fluid flow leg of RC loop 1 at CB; RC a containment (HV-8701B from the RCS through train A loop 1 or 4 hot suction valve HV-8811A analogous). to the suction of RHRS. Failure leg pressure RWST to RHR suction of RHR reduces redundancy indication at CB; line isolation pump 1. of RHR coolant RHR train A discharge valve HV-8812A, trains provided. No flow indication, with RHR to charging effect on safety for and RHR pump pump suction system operation. 1 discharge line isolation Plant cooldown pressure indication valve HV-8804A requirements are at CB. and with met by reactor coolant a "prevent-open" flow from hot leg of RC pressure interlock loop 4 flowing PT-438 (PT-408). through train B of The valve can not RHRS, however time be opened remotely required to reduce from the CB if RCS temperature is one of the indicated extended. isolation valves is open or if RC loop pressure exceeds 365 psig. The valve can be manually opened.
- b. Once the b. Failure reduces b. Valve is valves are redundancy of interlocked open the main main control with pressure control board board annunciator interlock annunciator alarm at 420 psig. PT-438 (PT-408) alarm fails No effect on safety to alarm on and RCS system operation. the main control pressure Plant operating board annunciator exceeds procedures require panel if one or 420 psig. that operators both of the close both valves valves is not prior to an RCS fully closed pressure of 420 and RCS pressure psig. Alternate exceeds 420 psig. alarm is provided by valve HV-8701B (HV-8701A).
VEGP-FSAR-5 TABLE 5.4.7-4 (SHEET 2 OF 5)
REV 13 4/06 Effect on Failure Detection Component Failure Mode Function System Operation Methods Remarks
- 2. Motor-operated Same as item 1. Same as item 1. Same as item 1. Same as item 1. Same as item 1, gate valve except for pressure HV-8702A interlock PT418 (HV-8702B (PT428) control. analogous).
- 3. RHR pump 1 Fails to deliver Provides fluid Failure results in loss Open pump switchgear The RHRS shares (RHR pump 2 working fluid. flow of reactor of reactor coolant flow circuit breaker indication components with analogous). coolant through from hot leg of RC at CB; circuit breaker the ECCS. Pumps RHR heat loop 1 through train A close position are tested as part exchanger 1 to of RHRS. Failure reduces monitor light for of the ECCS reduce RCS redundancy of RHR group monitoring of testing program. temperature coolant trains provided. components at CB; (See subsection 6.3.4.) during No effect on safety for common breaker trip cooldown operation. system operation. Plant alarm at CB; RC loop cooldown requirements 1 hot leg pressure are met by reactor indication at CB; RHR coolant flow from hot train A discharge flow leg of RC loop 4 flowing indication and low through train B of RHRS, flow alarm at CB; and however, time required pump discharge pressure to reduce RCS indication at CB. temperature is extended.
- 4. Motor-operated a. Fails closed. Provides a. Failure blocks mini- a. Valve open/close Valve is automatically globe valve regulation of flow line to suction position indication controlled to open when FV-610 fluid flow of RHR pump 1 during at CB; and RHRS pump discharge is less (FV-611 through mini- cooldown operation. train A discharge than the open setpoint analogous). flow bypass line No effect on safety flow indication (824 gpm at 350 F, to suction of for system operation. at CB. 780 gpm at 100F) RHR pump 1 to Plant cooldown and close when the protect against requirements are met discharge exceeds the overheating of by reactor coolant closed setpoint (1944 gpm the pump and flow from hot leg of at 350F, 1841 gpm at loss of discharge RC loop 4 flowing 100F). flow from the through train B of pump. RHRS. However, time required to reduce RCS temperature is extended.
VEGP-FSAR-5 TABLE 5.4.7-4 (SHEET 3 OF 5)
REV 13 4/06
Effect on Failure Detection Component Failure Mode Function System Operation Methods Remarks
- b. Fails open. b. Failure allows for a Same as item 4.a. portion of RHR heat exchanger 1 discharge flow to be bypassed to suction of RHR pump 1. RHRS train A is degraded for the regulation of coolant temperature by RHR heat exchanger 1. No effect on safety for system operation. Cool- down of RCS remains within established specification cooldown rate.
- 5. Air - a. Fails to open Controls rate a. Failure prevents cool a. RHR pump 1 discharge Valve is designed diaphragm- on demand for of fluid flow ant discharged from flow temperature to fail closed operated - flow increase bypassed RHR pump 1 from and RHRS train A and is electrically butterfly valve ("Auto" mode around RHR bypassing RHR discharge to wired so that FCV-618 CB switch heat exchanger heat exchanger 1 RCS cold leg electrical solenoid (FCV-619 selection). 1 during resulting in mixed flow temperature of the air diaphragm analogous). cooldown operation. mean temperature of recorded on the operator is energized to coolant flow to RCS plant computer; open the valve. being low. RHRS and RHRS train A Valve is normally train A is degraded discharge to RCS closed to align for the regulation cold leg flow RHRS for ECCS of controlling indication at CB. operation during plant temperature of coolant. power operation and No effect on safety load follow. for system operation. Cooldown of RCS Valve is designed within established for normal plant specification rate cooldown operation. may be accomplished It is not required for through operator safety grade cold action of throttling shutdown operations. flow control valve HCV-606 and controlling cooldown with redundant RHRS train B.
VEGP-FSAR-5 TABLE 5.4.7-4 (SHEET 4 OF 5)
REV 13 4/06
Effect on Failure Detection Component Failure Mode Function System Operation Methods Remarks
- b. Fails to close b. Failure allows coolant b. Same as item 5.a. on demand for discharged from RHR flow reduction pump 1 to bypass (Auto mode CB RHR heat exchanger 1 switch selection). resulting in mixed mean temperature of coolant flow to RCS being high. RHRS train A is degraded for the regulation controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be accomplished through operator action of throttling flow control valve HCV-606 and controlling cooldown with redundant RHRS train B; however, cooldown time is extended.
- 6. Air diaphragm- a. Fails to close Controls rate a. Failure prevents control a. Same methods of Valve is operated on demand for of fluid flow of coolant discharge detections as designed to butterfly flow reduction. through RHR flow from RHR those stated for fail open. valve HCV-606 heat exchanger heat exchanger 1 item 5.a. Valve is normally (HCV-607 1 during cool- resulting in loss of addition, monitor open to analogous). down operation. mixed mean temperature light and align RHRS for coolant flow alarm (valve ECCS operation adjustment to RCS. closed) for during plant No effect on safety group monitoring power operation for system operation. of components at and load follow. Cooldown of RCS within CB. established specification rate may be accomplished by operator action of controlling cooldown with redundant RHS train B.
- b. Fails to open b. Same as item 6.a. b. Same as item 6.a. on demand for flow increase.
VEGP-FSAR-5 TABLE 5.4.7-4 (SHEET 5 OF 5)
REV 13 4/06
Effect on Failure Detection Component Failure Mode Function System Operation Methods Remarks
- 7. Motor-operated Fails to close Provides isolation No effect on safety for Valve open/closed Valve is normally gate valve on demand. of fluid system operation. Plant position indication open to align HV-8812A from the RWST cooldown requirements at CB and RHRS for ECCS (HV-8812B to suction are met by reactor valve (closed) operation during analogous). of RHR pump 1 coolant flow from hot monitor light plant power operation during cooldown leg loop 4 flowing and alarm at CB. and load operation. through train B of follow. Valve RHRS; however, time must be closed required to reduce RCS during plant temperature is extended. cooldown to satisfy electrical interlock to permit valves HV-8701A and B (HV-8702A, B) to be opened.
- 8. Motor-operated Fails to close Provides separation Failure reduces the Same as item 7. gate valve on demand. between the redundancy for isolating HV-8716A two RHR trains RHR trains during (HV-8716B during cooldown cooldown. Negligible analogous). operation. effect on system operation. Isolation valve HV-8716B (HV-8716A) provides backup isolation between the two RHR trains.
- a. Component 7 is a component of the ECCS that performs a safety-grade cold shutdown function.
- b. List of acronyms and abbreviations.
Auto - Automatic.
BAT - Boric acid tank.
BIT - Boron injection tank (Unit 1 only).
CB - Main control board.
CVCS - Chemical and volume control system.
ECCS - Emergency core cooling system. HELB - High-energy line break.
MELB - Moderate-energy line break.
PRT - Pressurizer relief tank.
RC - Reactor coolant.
RHRS - Residual heat removal system. RWST - Refueling water storage tank.
RV - Reactor vessel.
SI - Safety injection.
VCT - Volume control tank. c. As part of plant operation, periodic tests, surveillance inspections, and instrument calibrations are made to monitor equipment and performance. Failures may be detected during such monitoring of equipment in addition to detection methods noted.
VEGP-FSAR-5 REV 19 4/15 TABLE 5.4.12-3 (SHEET 1 OF 2)
REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER VALVE SIZE (in.)
FUNCTION MAXIMUM ALLOWABLE LEAKAGE(gpm)
- 1. HV-8701A 12 RHR Suction (gate valve) 5.0 2. HV-8701B 12 RHR Suction (gate valve) 5.0 3. HV-8702A 12 RHR Suction (gate valve) 5.0 4. HV-8702B 12 RHR Suction (gate valve) 5.0 5. 1204-U4-120 2 SI-Hot Leg 2nd Isolation Valve 1.0 6. 1204-U4-121 2 SI-Hot Leg 2nd Isolation Valve 1.0 7. 1204-U4-122 2 SI-Hot Leg 2nd Isolation Valve 1.0 8. 1204-U4-123 2 SI-Hot Leg 2nd Isolation Valve 1.0 9. 1204-U6-079 10 Accumulator 2nd Isolation Valve 5.0 10. 1204-U6-080 10 Accumulator 2nd Isolation Valve 5.0 11. 1204-U6-081 10 Accumulator 2nd Isolation Valve 5.0 12. 1204-U6-082 10 Accumulator 2nd Isolation Valve 5.0 13. 1204-U6-083 10 Injection Line 1st Isolation Valve 5.0 14. 1204-U6-084 10 Injection Line 1st Isolation Valve 5.0 15. 1204-U6-085 10 Injection Line 1st Isolation Valve 5.0 VEGP-FSAR-5 REV 19 4/15 TABLE 5.4.12-3 (SHEET 2 OF 2)
REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER VALVE SIZE (in.)
FUNCTION MAXIMUM ALLOWABLE LEAKAGE(gpm)
- 16. 1204-U6-086 10 Injection Line 1st Isolation Valve 5.0 17. 1204-U6-124 6 SI-Hot Leg 1st Isolation Valve 3.0 18. 1204-U6-125 6 SI-Hot Leg 1st Isolation Valve 3.0 19. 1204-U6-126 6 SI-Hot Leg 1st Isolation Valve 3.0 20. 1204-U6-127 6 SI-Hot Leg 1st Isolation Valve 3.0 21. 1204-U6-128 8 RHR-Hot Leg 2nd Isolation Valve 4.0 22. 1204-U6-129 8 RHR-Hot Leg 2nd Isolation Valve 4.0 23. 1204-U4-143 2 SI-Cold Leg 2nd Isolation Valve 1.0 24. 1204-U4-144 2 SI-Cold Leg 2nd Isolation Valve 1.0 25. 1204-U4-145 2 SI-Cold Leg 2nd Isolation Valve 1.0 26. 1204-U4-146 2 SI-Cold Leg 2nd Isolation Valve 1.0 27. 1204-U6-147 6 RHR Cold Leg 2nd Isolation Valve 3.0 28. 1204-U6-148 6 RHR Cold Leg 2nd Isolation Valve 3.0 29. 1204-U6-149 6 RHR Cold Leg 2nd Isolation Valve 3.0 30. 1204-U6-150 6 RHR Cold Leg 2nd Isolation Valve 3.0
REV 13 4/06 REACTOR COOLANT PUMP FIGURE 5.4.1-1
REV 13 4/06 REACTOR COOLANT PUMP ESTIMATED PERFORMANCE CHARACTERISTIC FIGURE 5.4.1-2
REV 13 4/06 MODEL F STEAM GENERATOR FIGURE 5.4.2-1
REV 13 4/06 QUATREFOIL TUBE SUPPORT PLATES FIGURE 5.4.2-2
REV 13 4/06 RHRS PROCESS FLOW DIAGRAM FIGURE 5.4.7-1 (SHEET 1 OF 5)
REV 13 4/06 RHRS PROCESS FLOW DIAGRAM FIGURE 5.4.7-1 (SHEET 2 OF 5)
MODES OF OPERATION MODE A INITIATION OF RHR OPERATION When the reactor coolant temperature and pressure are reduced to 350
°F and 365 psig, approximately 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown, the second phase of plant cooldown starts with the RHRS being placed in operation. Before starting the pumps, the inlet isolation valves are opened, the heat exchanger flow control valves are set at minimum flow, and the outlet valves are verified open. The automatic miniflow valves are open and remain so until the pump flow exceeds the closed setpoint (1944 gpm at 350°F, 1841 gpm at 100
°F), at which time they trip closed. Should the pump flow drop below the open setpoint (824 gpm at 350
°F, 780 gpm at 100
°F), the miniflow valves open automatically.
Startup of the RHRS includes a warmup period, during which reactor coolant flow through the heat exchangers is limited to minimize thermal shock on the RCS. The rate of heat removal from the reactor coolant is controlled manually by regulating the reactor coolant flow through the residual heat exchangers. The total flow is regulated
automatically by control valves in the heat exchanger bypass line to maintain a constant total flow. The cooldown rate is limited to 100
°F/h based on equipment stress limits and a 120°F maximum component cooling water temperature.
MODE B END CONDITIONS OF A NORMAL COOLDOWN
This situation characterizes most of the RHRS operation. As the reactor coolant temperature decreases, the flow through the residual heat exchanger is increased until all of the flow is directed through the heat exchanger to obtain maximum cooling.
NOTE For the safeguards functions performed by the RHRS, refer to section 6.3.
REV 13 4/06 RHRS PROCESS FLOW DIAGRAM FIGURE 5.4.7-1 (SHEET 3 OF 5)
REV 13 4/06 RHRS PROCESS FLOW DIAGRAM FIGURE 5.4.7-1 (SHEET 4 OF 5)
REV 13 4/06 RHRS PROCESS FLOW DIAGRAM FIGURE 5.4.7-1 (SHEET 5 OF 5)
REV 13 4/06 RHRS PUMP PERFORMANCE CURVE FIGURE 5.4.7-2
REV 13 4/06 RHRS COOLDOWN CURVE (ONE TRAIN)
FIGURE 5.4.7-3 (SHEET 1 OF 2)
REV 13 4/06 RHRS COOLDOWN CURVE (TWO TRAIN)
FIGURE 5.4.7-3 (SHEET 2 OF 2)
REV 13 4/06 PRESSURIZER FIGURE 5.4.10-1
REV 13 4/06 PRESSURIZER RELIEF TANK FIGURE 5.4.11-1
REV 13 4/06 PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORT ARRANGEMENT FIGURE 5.4.11-2
REV 13 4/06 PRESSURIZER SAFETY VALVES RELIEF RATE FIGURE 5.4.13-1
REV 13 4/06 REACTOR VESSEL SUPPORTS FIGURE 5.4.14-1
REV 13 4/06 STEAM GENERATOR SUPPORTS FIGURE 5.4.14-2
REV 13 4/06 REACTOR COOLANT PUMPS SUPPORTS FIGURE 5.4.14-3
REV 13 4/06 PRESSURIZER SUPPORTS FIGURE 5.4.14-4
VEGP-FSAR-6 REV 14 10/07 TABLE 6.1.1-1 (SHEET 1 OF 2)
PRINCIPAL ESF PRESSURE-RETAINING MATERIALS Component Material Piping/tubing SA-53 Gr. B SA-106 Gr. B and C SA-155 Gr. 70 Class 1 and Gr. KC70 Class 1 SA-213, TP 304, 304L and 316 SA-249, TP 304L SA-312, TP 304 and 304L SA-333 Gr. 1 and 6 SA-335 Gr. P11 and P22 SA-376, TP 304 and 316 SB-111 Gr. CDA 706 SB-466 Gr. CDA 706 Fittings/flanges SA-105 N SA-181 Gr. I and II SA-182, TP F304, F304L, F316, F316L SA-234 Gr. WPB, WPC, WPBW, and WPCW SA-403 WP 304, 304L, 304W, and 304LW SA-420 Gr. WPL6 SA-479, TP 304, 304L and 316 Plate SA-240, TP 304, 304L and 316L SA-283 Gr. C SA-285 Gr. A and C SA-515 Gr. 70 SA-516 Gr. 70 SA-537 Class 1 SB-171 Gr. CDA 706
Bolting/nuts/studs SA-193 Gr. B6, B7, B8, and B8M SA-194, Gr 2H, 4, 6, 7, 8H, 8M, and B8 SA-307 Gr. B SA-320 Gr. L7 SA-453 Gr. 660A and 660B SA-564 Gr. 630 VEGP-FSAR-6 REV 14 10/07 TABLE 6.1.1-1 (SHEET 2 OF 2)
Component Material Castings SA-216 Gr. WCB and WCC SA-217 Gr. WC9 SA-351 Gr. CF8M and CF3M SA-487 Gr. CA6NM SB-61 SB-62 Gr. CDA 836 SB-148 Gr. CA 952 ASTM-A276 TP 410 Forgings SA-105 SA-182, TP F304, F304L, F316, and F316L; Gr. F11 and F22 SA-240 TP 304 and 316 SA-350 Gr. LF1 and LF2 SA-479, TP 304, 304L and 316
Bars SA-479, TP 304, 316 and 410 Gr 316L and F316 SA-564 Gr. 630 Weld rod SFA 5.1, E 6010 and E 7018 SFA 5.4, E 308-16, E 308L-16 and E 309 SFA 5.9, ER 308, ER 308L, and ER 309 SFA 5.17, EM 12K SFA 5.18, E 70S-2, E 70S-3, E 70S-4, E70S-6, and E70S-1B SFA 5.20, E 70T-1 and 70T-5
VEGP-FSAR-6 REV 14 10/07 TABLE 6.1.1-2 (SHEET 1 OF 2)
PRINCIPAL ESF MATERIALS EXPOSED TO REACTOR COOLANT OR CONTAINMENT SPRAY
Component Material Piping/tubing SA-106 Gr. B and C SA-155 Gr. KC70 Class 1 and 70 Class 1 SA-213, TP 304, 304L, and 316 SA-249, TP 304L SA-312, TP 304 and 304L SA-333 Gr. 1 and 6 SA-376, TP 304 and 316 SB-111 Gr. CDA 706 SB-466 Gr. CDA 706
Fittings/flanges SA-105 N SA-181 Gr. I and II SA-182, TP F304, F304L, F316, and F316L SA-234 Gr. WPB, WPBW, WPCW, and WPC SA-403, WP 304, 304L, 304W, and 304LW SA-420 Gr. WPL6 SA-479, TP 304, 304L, and 316
Plate SA-240, TP 304, 304L, and 316L SA-285 Gr. A and C SA-515 Gr. 70 SA-516 Gr. 70 SA-537 Class 1 SB-171 Gr. CDA 706 ASTM A 515 Gr. 70
Shapes SA-36
ASTM A-36 ASTM A-500 Gr. B (Code Case N-71-10)
Bolts/nuts/studs/pins SA-193 Gr. B6, B7, B8 and B8M SA-194 Gr. 2H, 8H, 8M, 7, 4, 6 and B8 SA-307 Gr. B SA-320 Gr. L7 SA-325 Type 1 SA-453 Gr. 660A and 660B VEGP-FSAR-6 REV 14 10/07 TABLE 6.1.1-2 (SHEET 2 OF 2)
Component Material SA-564 Gr. 630 ASTM A 193 Gr. B7 ASTM A 194 Gr. 7 ASTM A 307 ASTM A 354 ASTM A 490 Bars SA-479, TP 304, 316, 410, 316L, and F316 SA-564 Gr. 630 ASTM A 108 Gr. 1018 CW (Code Case N-71-5)
Forgings SA-105 SA-182, TP F304, F304L, F316 and F316L; Gr F11 and F22 SA-240, TP 304 and 316 SA-350 Gr. LF1 and LF2 SA-479, TP 304, 304L and 316 ASTM A 668 Class C (Code Case N-71-5)
Castings SA-216 Gr. WCB and WCC SA-217 Gr. WC9 SA-351 Gr. CF8M and CF3M SA-487 Gr. CA6NM SB-61 SB-62 Gr. CDA 836 SB-148 Gr. CA 952 ASTM A 276 TP 410 ASTM A-216 Gr. WCB Cooling coil fins SB-152 Gr. CDA 122
VEGP-FSAR-6 REV 14 10/07 TABLE 6.1.2-2 (SHEET 1 OF 10)
CONTAINMENT COMPONENTS - COATING SCHEDULE Item Material(a) Thickness (in.) Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
Containment Liner Plate System (a Dome CS 0.25 30,800(1)
+/-5 Inorganic
zinc 2.5 Epoxy- polyamide 3.0 Cylinder shell CS 0.25 68,612(1)
+/-5 Inorganic
zinc 2.5 Sumps SS 0.19 220(1) +/-5 NC
Refueling canal walls
and bottom SS 0.25 8,115(1) +/-5 NC Basemat CS 0.25 14,423(2) +/-5 Inorganic zinc 2.5 Reactor cavity walls CS 0.25 3,700(1) +/-5 Inorganic zinc 2.5 Reactor cavity
bottom CS 0.25 787(2) +/-5 Inorganic zinc 2.5 Primary Shield(a)
Cylinder wall CS 0.25 805(1) +/-5 Inorganic zinc 2.5 Cylinder wall and
nozzles CS 1.00 1,287(1) +/-5 Inorganic zinc 2.5 Nozzles CS 0.50 54(1) +/-5 Inorganic zinc 2.5 Upper cone and
nozzles CS 0.38 71(1) +/-5 Inorganic zinc 2.5
Containment Locks and Hatch(a)
Equipment Hatch Spherical head CS 1.50 174(3) +15 -5 Inorganic
zinc 2.5 Exposed hatch ring CS 3.00 172(3) +15 -5 Inorganic
zinc 2.5
Hatch structure (CBI
details) CS 0.50 3,250(4) +15
-5 Inorganic
zinc 2.5
VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 2 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
Personnel Lock
Barrel CS 0.5 25(4) +15 -5 Inorganic
zinc 2.5 Sleeve CS 2.0 52(1) +15 -5 Inorganic
zinc 2.5 Escape Lock
Barrel CS 0.5 25(4) +15 -5 Inorganic
zinc 2.5 Sleeve CS 2.0 33(1) +15 -5 Inorganic
zinc 2.5 Containment
Concrete(a)
Dome
Buttresses to 50° above horizontal Concrete 72.0 2278(5) +/-10 NC Buttresses from 50° to vertical Concrete 48.0 1822(5) +/-10 NC Dome area less buttress area Concrete 45.0 26,700(5) +/-10 NC Shell
Buttresses Concrete 72.0 5920(5) +/-10 NC
Shell Concrete 45.0 62,692(5) +/-10 NC
Internal Structural
Concrete(a)
Basemat Concrete 126.0 14,423(5) +/-10 NC Reactor cavity walls Concrete 96.0 3818(6) +/-10 NC Reactor cavity
slab Concrete 96.0 787(6) +/-10 NC
Exterior refueling canal walls VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 3 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
5-ft 0-in. thick walls Concrete 60.0 258(6) +/-10 Epoxy(a) (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0 7-ft 0-in. thick walls Concrete 84.0 584(6) +/-10 Epoxy(a) (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0 4-ft 0-in. thick walls Concrete 48.0 6067(6) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0 Refueling canal slab Concrete 48.0 1394(6) +/-10 NC Filler slab at 171 ft 9 in. Concrete 33.0 13,715(7) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 15.0 Epoxy (finish) 3.0 Reactor cavity filler slab Concrete 12.0 787(1) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 15.0 Epoxy (finish) 3.0 10.0 Primary shield Concrete 108.0 2014(8) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) Epoxy (finish) 3.0 Cavity access walls Concrete 24.0 830(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0
North/south beams at 199 ft
0 in. Concrete 72.0 1849(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0 Pressurizer walls Concrete 30.0 4394(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0 Secondary shield walls Concrete 36.0 29,722(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) 10.0 Epoxy (finish) 3.0
Air Shaft Concrete Walls Walls, Nos. 1, 2, and 3 Below 181 ft 0 in. Concrete 30.0 1114(9) +/-10 Epoxy (sealer) 0.5 Above 185 ft 0 in. Concrete 36.0 3045(9) +/-10 Epoxy (sealer) 0.5 No. 4 airshaft Concrete 36.0 1055(9) +/-10 Epoxy (sealer) 0.5 No. 1 and 2 ceiling slab Concrete 48.0 229(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes
Epoxy (surfacer) 15.0 Epoxy (finish) 3.0 VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 4 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
No. 3 ceiling slab Concrete 48.0 216(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes
Epoxy (surfacer) 15.0 Epoxy (finish) 3.0 Not used
Operating deck
slab - 220 ft 0 in.
15.0 2-ft 0-in. slab, 180° to 360° Concrete 24.0 1610(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surfacer) Epoxy (finish) 3.0
2-ft 9-in. slab at 90° Concrete 33.0 625(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 15.0 Epoxy (finish) 3.0 3-ft 0-in. slab at 90° and
260° Concrete 36.0 210(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 15.0 Epxoy (finish) 3.0 5-ft 0-in. slab at 0° and 180° Concrete 60.0 940(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 15.0 Epoxy (finish) 3.0 Miscellaneous Walls 1-ft 6-in. wall at 90° Concrete 18.0 580(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 3-ft 0-in. wall Concrete 36.0 1089(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0
3-ft 0-in. w all - below R.F.
canal Concrete 36.0 229(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 Mass concrete Concrete 176.0 678(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 Walls under R.F. canal Concrete 36.0 315(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 5 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
Instrument walls Concrete 30.0 1307(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 Stairs and Elevator Shaft
Walls Stair No. 1
Walls Concrete 12.0 2206(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 Ceiling Concrete 18.0 137(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 Stair No. 2
1-ft 0-in. wall Concrete 12.0 1400(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 1-ft 6-in. wall and ceiling Concrete 18.0 578(9) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed
to fill holes Epoxy (surface r) 10.0 Epoxy (finish) 3.0 Elevator Walls Concrete 12.0 1391(9) +/-10 Epoxy (sealer) 0.5 Ceiling Concrete 18.0 63(9) +/-10 Epoxy (sealer) 0.5 Miscellaneous pads Concrete 24.0 200(4) +/-10 Epoxy (sealer) 0.5 Epoxy (filler) As needed to fill
holes Epoxy (surface r) 15.0 Epoxy (finish) 3.0 Structural Steel Platform
grating CS 0.19 86,630(4) +/-15 Galvanized Ladders and stairways CS 0.22 1328(4) +/-15 Inorganic zinc 2.5 VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 6 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
Pipe whip restraints Light and medium capacity CS 1.5 340(4) +/-20 Inorganic zinc 2.5 Heavy and extra heavy capacity CS 3.0 15,670(4) +/-20 Inorganic zinc 2.5 Pressurizer
support steel Pressurizer supports CS 1.0 914(4) +/-10 Inorganic zinc 2.5 Grating CS 0.19 5060(4) +/-10 Galvanized Platform beams CS 0.5 2588(4) +/-10 Inorganic zinc 2.5 Crossover leg
support CS 2.5 840(4) +/-10 Inorganic zinc 2.5 Platform at steam generator
Grating CS 0.19 3250(4) +100
-20 Galvanized Support beams CS 0.5 30,000(4) +100
-20 Inorganic
zinc 2.5 Polar crane runway Brackets - 37 CS 1.25 2714(4) +/-10 Inorganic zinc 2.5 Rail CS 1.5 980(4) +/-10 Inorganic zinc 2.5 Girders - 37 CS 1.25 14,950(4) +/-10 Inorganic zinc 2.5 Internal platform structure Columns CS 1.2 14,625(4) +/-10 Inorganic zinc 2.5 Beams CS 0.38 6194(4) +/-15 Inorganic 2.5
VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 7 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
zinc CS 0.5 13,096(4) +/-15 Inorganic zinc 2.5 CS 0.62 37,991(4) +/-15 Inorganic zinc 2.5 CS 0.75 13,012(4) +/-15 Inorganic zinc 2.5 CS 1.0 12,741(4) +/-15 Inorganic zinc 2.5 CS 1.3 3141(4) +/-15 Inorganic zinc 2.5 Cable trays CS 0.0613 20,697 - Galvanized Cable tray support CS 0.38 7900(4) +/-20 Galvanized CS 0.19 3500(4) +/-20 Galvanized Conduit CS 0.15 12,779 - Galvanized Conduit Boxes CS 0.105 6141 - Galvanized HVAC ducting SS 0.1 12,000 - NC
HVAC ducting
bracing and
hangers CS 0.38 2000(4) +/-20 Inorganic zinc 2.5
Pipe supports (steel members)
CS 0.25 51,787 +/- 5 Inorganic zinc 2.5
Pipe racks CS 1.25 13,750 +/- 2 Inorganic zinc 2.5 Snubbers CS 0.5 2970 +/- 5 Inorganic zinc 2.5 Spring hangers CS 0.5 386 +/- 5 Inorganic zinc 2.5 Uninsulated
piping
24 in. SS 0.375 182 +/- 5 NC
14 in. SS 0.312 205 +/- 5 NC
12 in. SS 1.312 194 +/- 5 NC
12 in. SS 1.125 53 +/- 5
12 in. SS 0.688 387 +/- 5 NC VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 8 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
12 in. SS 0.375 367 +/- 5 NC
10 in. CS 0.365 206 +/- 5 Inorganic zinc 2.5 10 in. SS 0.365 1509 +/- 5 NC
8 in. CS 0.322 832 +/- 5 Inorganic zinc 2.5 8 in. SS 0.322 4769 +/- 5 NC
8 in. SS 0.906 286 +/- 5 NC
6 in. SS 0.280 5752 +/- 5 NC
6 in. CS 0.280 1139 +/- 5 Inorganic zinc 2.5 4 in. CS 0.237 2530 +/- 5 Inorganic zinc 2.5
4 in. SS 0.337 9 +/- 5 Inorganic
zinc 2.5 4 in. SS 0.12 2806 +/- 5 NC
4 in. SS 0.237 1289 +/- 5 NC
4 in. SS 0.531 210 +/- 5 NC
3 in. CS 0.216 119 +/- 5 Inorganic zinc 2.5 3 in. CS 0.438 203 +/- 5 Inorganic zinc 2.5 3 in. SS 0.12 58 +/- 5 NC
3 in. SS 0.216 1146 +/- 5 NC
3 in. SS 0.438 388 +/- 5 NC
2 1/2 in.
CS 0.203 451 +/- 5 Inorganic zinc 2.5 2 in. CS 0.218 267 +/- 5 Inorganic zinc 2.5 2 in. CS 0.344 71 +/- 5 Inorganic zinc 2.5
2 in. CS 0.154 12 +/- 5 Inorganic zinc 2.5
2 in. SS 0.154 733 +/- 5 NC
VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 9 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
2 in. SS 0.344 512 +/- 5 NC
2 in. SS 0.218 25 +/- 5 NC
2 in. SS 0.109 21 +/- 5 NC
1 1/2 in.
CS 0.145 36 +/- 5 Inorganic zinc 2.5 1 1/2 in.
CS 0.2 421 +/- 5 Inorganic zinc 2.5
1 1/2 in.
SS 0.2 48 +/- 5 NC
1 1/2 in.
SS 0.281 463 +/- 5 NC
1 1/2 in.
SS 0.145 627 +/- 5 NC
1 in. CS 0.133 35 +/- 5 Inorganic zinc 2.5 1 in. CS 0.179 783 +/- 5 Inorganic zinc 2.5 1 in. CS 0.25 14 +/- 5 Inorganic zinc 2.5 1 in. SS 0.133 443 +/- 5 NC
1 in. SS 0.179 72 +/- 5 NC
1 in. SS 0.25 190 +/- 5 NC
3/4 in. SS 0.154 303 +/- 5 NC
3/4 in. SS 0.218 407 +/- 5 NC
3/4 in. SS 0.113 956 +/- 5
1/2 in. CS 0.188 1 +/- 5 Inorganic zinc 2.5 1/2 in. SS 0.147 140 +/- 5 NC
1/2 in. SS 0.109 15 +/- 5 NC
1/2 in. SS 0.065 278 +/- 5 NC
1/2 in. SS 0.049 123 +/- 5 NC
VEGP-FSAR-6 TABLE 6.1.2-2 (SHEET 10 OF 10)
REV 14 10/07 Item Material(a) Thickness
(in.)
Estimated
Surface Area (ft 2)(b) Tolerance
Surface Area(%) 1st Coat (c) Minimum Thickness (mils of)
1st Coat 2nd Coat Minimum Thickness (mils) of
2nd Coat 3rd Coat Minimum Thickness (mils) of
3rd Coat 4th Coat Minimum Thickness (mils) of
4th Coat
3/8 in. SS 0.065 38 +/- 5 NC
Containment
coolers CS 0.25 20,500 +10 Inorganic zinc 2.5 Epoxy 3.0 Containment auxiliary coolers CS 0.25 4400 +10 Inorganic zinc 2.5 Epoxy 3.0 Cavity cooling
coils CS 0.25 2200 +10 Inorganic zinc 2.5 Epoxy 3.0
ESF fans CS 0.25 200 +10 Inorganic zinc 2.5 Epoxy 3.0
- b. (1) Interior surface (liner plate) backed by concrete. Only one side directly exposed to containment environment. Surface area of exposed side given. (2) Interior surface (liner plate) backed by concrete and covered with filler slab (concrete). Not directly exposed to contai nment environment. Surface area of one side given. (3) Direct boundary between containment interior and ex terior environment. Interior surface area given. (4) Specified member (beams, walls, slabs, etc.) completely exposed (all sides) to the contai nment environment. Total exposed surface area given (e.g., for walls both sides are given; for wide flange shapes both sides of flanges and web are given). (5) Exterior surface. One side directly exposed to exterior environment. Surfac e area of exposed side given. (6) Interior surface covered with steel liner plate. Not directly exposed to containment env ironment. Interior surface area given. (7) Interior surface. One side directly exposed to containment environment with basemat steel liner plate on opposite side.
Exposed surface area given. (8) Interior surface. One side directly exposed to containment environment with steel liner on opposite side (liner in turn e xposed to containment environment). Surface area of directly exposed side given.
(9) Outer surface directly exposed to general containment environment with inner area almost entirely self-enclosed. Directly exposed outer surface area given.
- c. NC denotes no coating.
- d. This epoxy is applied as a wainscot which covers containment interior walls from elevations 171 ft 9 in. to 181 ft 9 in. and elevations 220 ft to 228 ft. The remainder of concrete walls are coated with 0.5 mils of epoxy only.
- e. Although actual areas for individual categories may differ slightly from the values shown, the total surface areas are considered to be within the indicated tolerances.
VEGP-FSAR-6 TABLE 6.1.2-4 OTHER ORGANIC MATERIALS REV 14 10/07 Quantity Quantity Enclosed Exposed Item Material (lb) (lb) Cable insulation Ethylene propylene 16,100 24,800 rubber and chlorosul-phonated polyethylene Heat shrink tubing Raychem WCSF-N 650 - Lug insulation AMP special indus-200 - tries type PVF
Cable ties Thomas and Bets Tefzel 100 500 Terminal blocks Diallyl-phthalate long 400 - glass fiber fill
REV 14 10/07 CONTAINMENT TEMPERATURE PROFILE FOR POST-LOCA COATING TEST FIGURE 6.1.2-1
VEGP-FSAR-6 REV 14 10/07 TABLE 6.2.1-1 CONTAINMENT DESIGN LIMITS AND CALCULATED CONTAINMENT PEAK PRESSURE AND TEMPERATURE
Peak Available Peak Pressure Margin Temperature Break (psig) (psi)
(°F) Primary Side Ruptures Double-ended pump suction, 35.9 (b) 16.1 (b) 247 (c) maximum safety injection Double-ended pump suction, 35.9 (b) 16.1 (b) 247 (c) minimum safety injection Double-ended hot leg 36.5 15.5 250 Containment Design Pressure 52 psig -3 psig Containment Atmosphere Design Temperature 381°F (a)
- a. A peak containment atmosphere temperature of 381
°F was used in calculating the thermal gradients across the containment wall.
- b. Per LDCR 2005051, the pressure is being increased by 1.3 psig from 34.6 psig, and the margin is correspondingly reduced temporarily until re-evaluated during a future analysis requiring re-evaluation of the containment mass and energy releases.
- c. Per LDCR 2005051, there will be an insignificant increase in peak temperature. However, the peak temperature remains well below the design temperature of 381 °F.
VEGP-FSAR-6 TABLE 6.2.1-2 ASSUMPTIONS FOR CONTAINMENT ANALYSIS - PART 1
Service water tem p erature (°F) 95 Refuelin g water tem p erature (°F) 130 Refuelin g water stora g e tank volume (g al)710,700 (deliverable volume
) Initial Containment Tem p erature (°F) 120 Initial p ressure (p sia) 17.7 Initial relative humidit y (%) 20 Net free volume (ft 3)2.75 x 10 6
REV 17 4/12
VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.1-60 SPECTRUM OF SECONDARY SYSTEM PIPE RUPTURES ANALYZED (a) Case Size Type Of Rupture
% Power (b) 1 Full Double-ended 102 2 Full Double-ended 70 3 Full Double-ended 30 4 Full Double-ended 0 5 0.60 ft 2 Double-ended with entrainment 102 6 0.53 ft 2 Double-ended with entrainment 70 7 0.36 ft 2 Double-ended with entrainment 30 8 0.20 ft 2 Double-ended with entrainment 0 9 0.33 ft 2 Double-ended without entrainment 102 10 0.32 ft 2 Double-ended without entrainment 70 11 0.22 ft 2 Double-ended without entrainment 30 12 0.10 ft 2 Double-ended without entrainment 0 13 0.86 ft 2 Split rupture 102 14 0.908 ft 2 Split rupture 70 15 0.944 ft 2 Split rupture 30 16 0.40 ft 2 Split rupture 0
- a. Case 1 through case 16 have assumed the loss of a diesel.
- b. % power of 3579 MWt.
VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 1 OF 9)
REV 18 9/13 MASS AND ENERGY RELEASE DATA FOR CASE 16 - PEAK CALCULATED CONTAINMENT PRESSURE FOR MSLB Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 0.00 0.00 0.000 9.40 843.43 1.005 18.60 799.79 0.954 0.20 906.37 1.077 9.60 842.21 1.003 18.80 798.43 0.953 0.40 904.86 1.075 9.80 841.85 1.003 19.00 797.53 0.952 0.60 903.06 1.073 10.00 841.43 1.002 19.20 797.05 0.951 0.80 901.31 1.071 10.20 839.56 1.000 19.40 795.86 0.950 1.00 899.95 1.070 10.40 838.11 0.999 19.60 795.70 0.950 1.20 898.72 1.068 10.60 836.90 0.997 19.80 794.60 0.948 1.40 897.06 1.066 10.80 836.71 0.997 20.00 793.65 0.947 1.60 894.80 1.064 11.00 836.34 0.997 20.20 792.95 0.946 1.80 892.96 1.062 11.20 834.53 0.994 20.40 792.33 0.946 2.00 892.40 1.061 11.40 833.05 0.993 20.60 791.61 0.945 2.20 890.84 1.059 11.60 831.83 0.991 20.80 791.36 0.945 2.40 888.62 1.057 11.80 831.34 0.991 21.00 791.16 0.944 2.60 886.71 1.054 12.00 829.79 0.989 21.20 790.40 0.943 2.80 885.95 1.054 12.20 828.86 0.988 21.40 789.33 0.942 3.00 884.39 1.052 12.40 827.84 0.987 21.60 788.93 0.942 3.20 882.89 1.050 13.00 824.87 0.983 21.80 788.56 0.941 3.40 880.80 1.048 13.20 823.95 0.982 22.00 787.66 0.940 3.60 879.46 1.046 13.40 822.96 0.981 22.20 787.82 0.940 3.80 878.70 1.045 13.60 822.04 0.980 22.40 786.59 0.939 4.00 877.29 1.044 13.80 821.04 0.979 22.60 786.22 0.939 4.20 875.24 1.041 14.00 820.14 0.978 22.80 786.48 0.939 4.40 873.61 1.039 14.20 819.17 0.977 23.00 785.80 0.938 4.80 870.88 1.036 14.40 818.88 0.976 23.20 785.75 0.938 5.00 869.46 1.035 14.60 817.85 0.975 23.40 784.42 0.936 5.20 869.10 1.034 14.80 816.40 0.973 23.60 783.92 0.936 5.40 867.83 1.033 15.00 815.61 0.973 24.00 783.12 0.935 5.60 865.86 1.030 15.20 815.16 0.972 24.20 782.74 0.934 5.80 864.34 1.029 15.40 814.13 0.971 24.80 781.65 0.933 6.20 861.82 1.026 15.60 812.69 0.969 25.40 780.51 0.932 6.40 860.80 1.025 15.80 811.88 0.968 26.00 779.34 0.930 6.60 860.07 1.024 16.00 811.45 0.968 26.40 778.53 0.930 6.80 858.87 1.022 16.20 810.44 0.967 27.00 777.29 0.928 7.00 857.05 1.020 16.40 809.04 0.965 27.60 776.01 0.927 7.20 855.98 1.019 16.60 808.20 0.964 28.20 774.70 0.925 7.40 855.36 1.018 16.80 807.86 0.964 28.80 773.36 0.924 7.60 854.30 1.017 17.00 806.87 0.962 29.80 771.07 0.921 7.80 853.01 1.016 17.20 805.45 0.961 30.40 769.67 0.919 8.00 851.60 1.014 17.40 804.57 0.960 31.40 767.32 0.917 8.20 850.38 1.013 17.60 804.24 0.959 32.60 764.48 0.913 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 2 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 8.40 848.89 1.011 17.80 803.29 0.958 34.80 759.23 0.907 8.60 847.83 1.010 18.00 801.91 0.957 39.00 749.16 0.895 8.80 847.27 1.009 18.20 801.00 0.956 41.20 743.90 0.889 9.00 845.58 1.007 18.40 800.77 0.955 43.40 738.67 0.883 45.60 733.49 0.877 116.20 578.27 0.695 128.40 506.00 0.609 47.80 728.34 0.871 116.40 576.70 0.693 128.80 504.21 0.607 50.00 723.23 0.865 116.60 575.15 0.691 129.20 502.44 0.604 52.00 718.64 0.860 116.80 573.61 0.689 129.60 500.70 0.602 54.20 713.63 0.854 117.00 572.09 0.687 129.80 499.84 0.601 56.40 708.68 0.848 117.20 570.59 0.685 130.20 498.13 0.599 58.60 703.79 0.842 117.40 569.10 0.684 130.60 496.45 0.597 60.80 698.94 0.837 117.60 567.63 0.682 130.80 495.62 0.596 62.60 695.01 0.832 117.80 566.17 0.680 131.20 493.98 0.594 62.80 694.74 0.832 118.00 564.73 0.679 131.60 492.37 0.592 65.00 689.99 0.826 118.20 563.31 0.677 131.80 491.57 0.591 67.20 685.30 0.821 118.40 561.90 0.675 132.20 489.99 0.590 69.40 680.65 0.815 118.60 560.50 0.674 132.60 488.43 0.588 71.60 676.05 0.810 118.80 559.12 0.672 132.80 487.66 0.587 73.80 671.49 0.805 119.00 557.76 0.670 133.20 486.14 0.585 74.80 669.43 0.802 119.20 556.40 0.669 133.40 485.39 0.584 76.00 666.98 0.799 119.40 555.07 0.667 133.80 483.90 0.582 78.20 662.52 0.794 119.60 553.74 0.666 134.20 482.44 0.581 80.40 658.10 0.789 119.80 552.43 0.664 134.40 481.71 0.580 82.40 654.13 0.784 120.00 551.14 0.662 134.80 480.28 0.578 84.60 649.79 0.779 120.20 549.85 0.661 135.20 478.87 0.576 86.80 645.51 0.774 120.40 548.58 0.659 135.60 477.48 0.575 89.00 641.26 0.769 120.60 547.33 0.658 136.00 476.11 0.573 91.20 637.06 0.764 120.80 546.08 0.656 136.40 474.76 0.571 93.40 632.91 0.759 121.00 544.85 0.655 136.80 473.43 0.570 95.40 629.16 0.755 121.20 543.63 0.653 137.20 472.12 0.568 97.60 625.09 0.750 121.40 542.42 0.652 137.40 471.47 0.567 99.80 621.06 0.745 121.60 541.22 0.651 137.80 470.19 0.566 102.00 617.06 0.740 121.80 540.04 0.649 138.40 468.31 0.564 104.20 613.12 0.736 122.20 537.70 0.646 138.80 467.07 0.562 106.40 609.21 0.731 122.60 535.40 0.644 139.40 465.25 0.560 108.40 605.70 0.727 123.00 533.14 0.641 139.80 464.06 0.559 110.60 601.89 0.723 123.40 530.92 0.638 140.40 462.31 0.557 111.60 600.17 0.720 123.60 529.83 0.637 141.00 460.60 0.554 112.80 598.12 0.718 124.00 527.67 0.635 141.60 458.92 0.552 114.00 596.09 0.716 124.40 525.54 0.632 142.20 457.28 0.550 114.20 595.62 0.715 124.80 523.45 0.629 142.60 456.20 0.549 114.40 593.24 0.712 125.20 521.39 0.627 143.20 454.62 0.547 114.60 591.49 0.710 125.40 520.37 0.626 143.60 453.59 0.546 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 3 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 114.80 589.76 0.708 125.80 518.36 0.623 144.20 452.06 0.544 115.00 588.06 0.706 126.20 516.38 0.621 144.60 451.06 0.543 115.20 586.38 0.704 126.40 515.40 0.620 145.20 449.59 0.541 115.40 584.72 0.702 126.80 513.46 0.618 145.60 448.63 0.540 115.60 583.08 0.700 127.20 511.55 0.615 146.20 447.21 0.538 115.80 581.46 0.698 127.60 509.68 0.613 146.80 445.82 0.537 116.00 579.85 0.696 128.00 507.83 0.611 147.40 444.46 0.535 172.20 407.41 0.491 218.80 387.01 0.466 147.80 443.57 0.534 172.80 406.86 0.490 221.00 386.69 0.466 148.40 442.26 0.532 173.40 406.32 0.489 223.20 386.41 0.465 149.00 440.97 0.531 174.40 405.44 0.488 225.40 386.15 0.465 149.60 439.71 0.529 175.40 404.60 0.487 227.40 385.93 0.465 150.20 438.48 0.528 176.40 403.79 0.486 229.60 385.72 0.465 150.80 437.28 0.527 177.00 403.32 0.486 231.80 385.52 0.464 151.40 436.10 0.525 177.60 402.85 0.485 234.00 385.34 0.464 151.80 435.33 0.524 178.20 402.40 0.485 236.20 385.19 0.464 152.40 434.19 0.523 178.80 401.96 0.484 238.40 385.04 0.464 152.80 433.45 0.522 179.80 401.25 0.483 240.40 384.93 0.464 153.40 432.35 0.521 180.80 400.56 0.482 242.60 384.81 0.463 154.00 431.27 0.519 181.40 400.16 0.482 244.80 384.71 0.463 154.60 430.22 0.518 182.00 399.77 0.481 247.00 384.62 0.463 155.20 429.19 0.517 182.60 399.38 0.481 249.20 384.54 0.463 155.60 428.52 0.516 183.20 399.01 0.481 251.40 384.47 0.463 156.20 427.52 0.515 184.20 398.40 0.480 255.60 384.36 0.463 156.60 426.87 0.514 185.20 397.82 0.479 260.00 384.28 0.463 157.20 425.91 0.513 186.40 397.15 0.478 264.20 384.22 0.463 157.60 425.28 0.512 187.40 396.62 0.478 268.60 384.18 0.463 158.20 424.35 0.511 188.40 396.10 0.477 273.00 384.16 0.463 158.60 423.75 0.510 189.40 395.61 0.476 277.60 384.15 0.463 159.20 422.85 0.509 190.60 395.03 0.476 283.80 384.16 0.463 159.80 421.98 0.508 191.80 394.49 0.475 290.00 384.19 0.463 160.40 421.12 0.507 192.80 394.05 0.475 296.00 384.23 0.463 160.80 420.55 0.506 193.80 393.64 0.474 302.20 384.29 0.463 161.40 419.72 0.505 195.00 393.15 0.474 314.60 384.41 0.463 162.00 418.90 0.504 196.20 392.69 0.473 376.40 385.10 0.464 162.60 418.11 0.504 197.20 392.33 0.473 401.00 385.36 0.464 163.20 417.32 0.503 198.60 391.84 0.472 425.80 385.61 0.464 163.80 416.56 0.502 198.80 391.65 0.472 592.20 387.24 0.466 164.40 415.81 0.501 200.40 391.13 0.471 611.40 387.41 0.467 164.80 415.32 0.500 201.40 390.82 0.471 623.80 387.51 0.467 165.40 414.60 0.499 202.40 390.52 0.470 636.00 387.59 0.467 165.80 414.13 0.499 203.60 390.18 0.470 648.40 387.66 0.467 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 4 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 166.40 413.43 0.498 204.80 389.85 0.470 654.60 387.69 0.467 167.00 412.75 0.497 205.80 389.59 0.469 667.00 387.72 0.467 167.60 412.08 0.496 206.80 389.34 0.469 686.80 387.73 0.467 168.20 411.42 0.495 208.00 389.05 0.469 692.40 387.71 0.467 168.60 410.99 0.495 209.20 388.78 0.468 698.00 387.69 0.467 169.20 410.36 0.494 210.20 388.56 0.468 703.80 387.66 0.467 169.60 409.95 0.494 211.20 388.35 0.468 716.80 387.56 0.467 170.20 409.34 0.493 212.40 388.11 0.467 720.80 387.52 0.467 170.60 408.94 0.493 213.40 387.92 0.467 732.20 387.40 0.467 171.20 408.36 0.492 215.40 387.56 0.467 743.60 387.25 0.466 171.60 407.97 0.491 216.60 387.35 0.467 755.00 387.07 0.466 766.20 386.88 0.466 868.00 370.26 0.446 878.80 333.01 0.401 777.60 386.66 0.466 868.40 369.25 0.445 879.00 332.11 0.400 789.00 386.42 0.465 868.80 368.22 0.444 879.20 331.20 0.399 800.20 386.18 0.465 869.00 367.70 0.443 879.40 330.28 0.398 811.60 385.91 0.465 869.40 366.62 0.442 879.60 329.36 0.397 823.00 385.63 0.464 869.80 365.53 0.440 879.80 328.43 0.396 834.40 385.34 0.464 870.00 364.97 0.440 880.00 327.50 0.394 858.80 384.69 0.463 870.40 363.83 0.438 880.20 326.55 0.393 859.00 384.68 0.463 870.80 362.67 0.437 880.40 325.60 0.392 859.20 384.62 0.463 871.20 361.47 0.435 880.80 323.68 0.390 859.40 384.52 0.463 871.40 360.87 0.435 881.20 321.74 0.387 859.60 384.41 0.463 871.80 359.63 0.433 881.60 319.77 0.385 859.80 384.27 0.463 872.00 359.00 0.432 882.00 317.77 0.383 860.00 384.11 0.463 872.20 358.36 0.432 882.20 316.77 0.381 860.20 383.95 0.462 872.40 357.72 0.431 882.60 314.74 0.379 860.40 383.77 0.462 872.60 357.07 0.430 883.00 312.69 0.377 860.60 383.57 0.462 872.80 356.41 0.429 883.20 311.66 0.375 860.80 383.37 0.462 873.00 355.74 0.428 883.60 309.59 0.373 861.00 383.16 0.461 873.20 355.07 0.428 884.00 307.50 0.370 861.20 382.94 0.461 873.40 354.39 0.427 884.40 305.39 0.368 861.40 382.71 0.461 873.60 353.70 0.426 884.80 303.26 0.365 861.60 382.46 0.461 873.80 353.00 0.425 885.00 302.20 0.364 861.80 382.21 0.460 874.00 352.30 0.424 885.40 300.06 0.361 862.00 381.95 0.460 874.20 351.58 0.423 886.00 296.83 0.357 862.20 381.68 0.460 874.40 350.86 0.423 886.40 294.66 0.355 862.40 381.40 0.459 874.60 350.14 0.422 886.80 292.49 0.352 862.60 381.11 0.459 874.80 349.40 0.421 887.40 289.23 0.348 862.80 380.81 0.459 875.00 348.65 0.420 889.00 280.51 0.338 863.00 380.51 0.458 875.20 347.90 0.419 889.60 277.25 0.334 863.20 380.19 0.458 875.40 347.14 0.418 890.20 274.01 0.330 863.40 379.87 0.458 875.60 346.38 0.417 890.60 271.86 0.327 863.60 379.53 0.457 875.80 345.60 0.416 891.00 269.72 0.325 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 5 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 863.80 379.19 0.457 876.00 344.82 0.415 891.40 267.59 0.322 864.00 378.84 0.456 876.20 344.02 0.414 891.80 265.47 0.319 864.20 378.48 0.456 876.40 343.22 0.413 892.00 264.42 0.318 864.40 378.11 0.455 876.60 342.42 0.412 892.40 262.33 0.316 864.60 377.74 0.455 876.80 341.60 0.411 892.80 260.25 0.313 864.80 377.36 0.455 877.00 340.78 0.410 893.20 258.19 0.311 865.20 376.57 0.454 877.20 339.94 0.409 893.60 256.15 0.308 865.60 375.75 0.453 877.40 339.10 0.408 893.80 255.13 0.307 866.00 374.90 0.452 877.60 338.26 0.407 894.00 254.12 0.306 866.20 374.47 0.451 877.80 337.40 0.406 894.20 253.12 0.305 866.60 373.58 0.450 878.00 336.54 0.405 894.40 252.12 0.303 867.00 372.66 0.449 878.20 335.67 0.404 894.60 251.13 0.302 867.40 371.72 0.448 878.40 334.79 0.403 894.80 250.14 0.301 867.80 370.75 0.447 878.60 333.90 0.402 895.00 249.16 0.300 895.20 248.18 0.299 904.40 210.60 0.253 913.80 185.77 0.223 895.40 247.22 0.297 904.60 209.93 0.252 914.00 185.37 0.222 895.60 246.25 0.296 904.80 209.28 0.251 914.20 184.97 0.222 895.80 245.30 0.295 905.00 208.63 0.251 914.40 184.57 0.221 896.00 244.35 0.294 905.20 207.99 0.250 914.60 184.18 0.221 896.20 243.41 0.293 905.40 207.35 0.249 914.80 183.79 0.221 896.40 242.47 0.292 905.60 206.72 0.248 915.00 183.41 0.220 896.60 241.54 0.291 905.80 206.10 0.248 915.20 183.03 0.220 896.80 240.62 0.289 906.00 205.48 0.247 915.40 182.66 0.219 897.00 239.71 0.288 906.20 204.87 0.246 915.60 182.29 0.219 897.20 238.80 0.287 906.40 204.27 0.245 915.80 181.92 0.218 897.40 237.90 0.286 906.60 203.67 0.245 916.00 181.56 0.218 897.60 237.00 0.285 906.80 203.08 0.244 916.20 181.20 0.217 897.80 236.11 0.284 907.00 202.49 0.243 916.40 180.85 0.217 898.00 235.23 0.283 907.20 201.91 0.242 916.60 180.50 0.217 898.20 234.36 0.282 907.40 201.34 0.242 916.80 180.16 0.216 898.40 233.49 0.281 907.60 200.77 0.241 917.00 179.81 0.216 898.60 232.63 0.280 907.80 200.20 0.240 917.20 179.48 0.215 898.80 231.78 0.279 908.20 199.09 0.239 917.40 179.14 0.215 899.00 230.93 0.278 908.40 198.54 0.238 917.60 178.81 0.214 899.20 230.09 0.277 908.60 198.00 0.238 917.80 178.49 0.214 899.40 229.26 0.276 908.80 197.46 0.237 918.00 178.16 0.214 899.60 228.43 0.275 909.00 196.93 0.236 918.20 177.85 0.213 899.80 227.62 0.274 909.20 196.40 0.236 918.40 177.53 0.213 900.00 226.80 0.273 909.40 195.88 0.235 918.60 177.22 0.213 900.20 226.00 0.272 909.60 195.37 0.235 918.80 176.91 0.212 900.40 225.20 0.271 909.80 194.86 0.234 919.20 176.31 0.211 900.60 224.41 0.270 910.00 194.35 0.233 919.40 176.01 0.211 900.80 223.62 0.269 910.20 193.85 0.233 919.80 175.43 0.210 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 6 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 901.00 222.84 0.268 910.40 193.36 0.232 920.00 175.14 0.210 901.20 222.07 0.267 910.60 192.87 0.232 920.40 174.57 0.209 901.40 221.30 0.266 910.80 192.39 0.231 920.80 174.02 0.209 901.60 220.54 0.265 911.00 191.91 0.230 921.00 173.75 0.208 901.80 219.79 0.264 911.20 191.44 0.230 921.40 173.22 0.208 902.00 219.04 0.263 911.40 190.97 0.229 921.80 172.70 0.207 902.20 218.30 0.262 911.60 190.51 0.229 922.00 172.45 0.207 902.40 217.57 0.261 911.80 190.05 0.228 922.20 172.19 0.206 902.60 216.84 0.261 912.00 189.60 0.228 922.60 171.70 0.206 902.80 216.12 0.260 912.20 189.15 0.227 923.00 171.21 0.205 903.00 215.41 0.259 912.40 188.71 0.226 923.20 170.98 0.205 903.20 214.70 0.258 912.60 188.27 0.226 923.60 170.51 0.204 903.40 214.00 0.257 912.80 187.84 0.225 924.00 170.05 0.204 903.60 213.31 0.256 913.00 187.41 0.225 924.20 169.83 0.204 903.80 212.62 0.255 913.20 187.02 0.224 924.60 169.39 0.203 904.00 211.94 0.255 913.40 186.60 0.224 925.00 168.96 0.203 904.20 211.26 0.254 913.60 186.18 0.223 925.40 168.54 0.202 925.80 168.13 0.202 945.00 156.66 0.188 1001.60 151.86 0.182 926.00 167.93 0.201 945.80 156.42 0.187 1004.40 151.83 0.182 926.40 167.53 0.201 946.40 156.24 0.187 1007.40 151.82 0.182 926.80 167.15 0.200 947.00 156.07 0.187 1010.00 151.80 0.182 927.20 166.77 0.200 947.80 155.86 0.187 1015.80 151.78 0.182 927.60 166.40 0.199 948.40 155.70 0.186 1021.40 151.76 0.182 928.00 166.04 0.199 949.20 155.51 0.186 1027.20 151.75 0.182 928.40 165.69 0.199 950.00 155.32 0.186 1038.40 151.74 0.182 928.80 165.35 0.198 950.60 155.18 0.186 1070.80 151.73 0.182 929.00 165.18 0.198 951.20 155.05 0.186 1800.00 151.74 0.182 929.40 164.85 0.198 952.00 154.89 0.185 1800.20 151.78 0.182 929.80 164.53 0.197 952.80 154.73 0.185 1800.40 152.30 0.182 930.00 164.37 0.197 953.40 154.62 0.185 1800.60 152.41 0.182 930.40 164.06 0.197 954.20 154.48 0.185 1800.80 152.36 0.182 930.80 163.76 0.196 955.00 154.34 0.185 1801.00 152.17 0.182 931.00 163.61 0.196 955.80 154.21 0.185 1801.20 151.84 0.182 931.40 163.32 0.196 956.40 154.12 0.185 1801.40 151.40 0.181 931.80 163.03 0.195 957.00 154.03 0.184 1801.60 150.84 0.181 932.20 162.75 0.195 957.80 153.91 0.184 1801.80 150.19 0.180 932.60 162.48 0.195 958.40 153.83 0.184 1802.00 149.44 0.179 933.00 162.22 0.194 959.20 153.72 0.184 1802.20 148.61 0.178 933.60 161.83 0.194 960.00 153.62 0.184 1802.40 147.70 0.177 934.20 161.46 0.193 961.20 153.48 0.184 1802.60 146.72 0.176 934.60 161.22 0.193 962.00 153.40 0.184 1802.80 145.69 0.174 935.00 160.98 0.193 962.80 153.31 0.184 1803.00 144.58 0.173 935.40 160.75 0.193 963.40 153.25 0.184 1803.20 143.43 0.172 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 7 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 935.80 160.53 0.192 964.80 153.12 0.183 1803.40 142.22 0.170 936.40 160.21 0.192 966.20 153.00 0.183 1803.60 140.93 0.169 937.00 159.89 0.192 967.60 152.89 0.183 1803.80 139.04 0.166 937.40 159.69 0.191 969.00 152.79 0.183 1804.00 137.19 0.164 937.80 159.49 0.191 970.40 152.70 0.183 1804.20 135.43 0.162 938.40 159.21 0.191 971.80 152.62 0.183 1804.40 133.68 0.160 938.80 159.02 0.190 973.20 152.54 0.183 1804.60 131.94 0.158 939.20 158.84 0.190 974.60 152.47 0.183 1804.80 130.22 0.156 939.60 158.66 0.190 976.00 152.40 0.182 1805.00 128.51 0.154 940.00 158.49 0.190 977.40 152.35 0.182 1805.20 126.82 0.151 940.60 158.24 0.190 979.00 152.29 0.182 1805.40 125.14 0.149 941.20 158.00 0.189 980.40 152.24 0.182 1805.60 123.48 0.147 941.60 157.84 0.189 981.80 152.19 0.182 1805.80 121.84 0.145 942.00 157.69 0.189 983.20 152.15 0.182 1806.00 120.22 0.143 942.40 157.54 0.189 984.60 152.12 0.182 1806.20 118.61 0.142 942.80 157.39 0.189 987.40 152.05 0.182 1806.40 117.03 0.140 943.40 157.18 0.188 990.20 152.00 0.182 1806.60 115.46 0.138 943.80 157.05 0.188 993.00 151.95 0.182 1806.80 113.92 0.136 944.20 156.92 0.188 996.00 151.91 0.182 1807.00 112.39 0.134 944.60 156.79 0.188 998.80 151.88 0.182 1807.20 110.89 0.132 1807.40 109.42 0.130 1816.60 60.30 0.071 1826.80 31.41 0.037 1807.60 107.97 0.129 1816.80 59.52 0.070 1827.00 30.99 0.036 1807.80 106.55 0.127 1817.00 58.75 0.069 1827.20 30.57 0.036 1808.00 105.14 0.125 1817.20 58.00 0.068 1827.40 30.15 0.035 1808.20 103.76 0.124 1817.40 57.25 0.068 1827.60 29.75 0.035 1808.40 102.39 0.122 1817.60 56.52 0.067 1827.80 29.35 0.034 1808.60 101.05 0.120 1817.80 55.80 0.066 1828.00 28.95 0.034 1808.80 99.73 0.119 1818.00 55.10 0.065 1828.20 28.57 0.033 1809.00 98.45 0.117 1818.20 54.40 0.064 1828.40 28.19 0.033 1809.20 97.17 0.116 1818.40 53.72 0.063 1828.60 27.83 0.032 1809.40 95.92 0.114 1818.60 53.04 0.062 1828.80 27.47 0.032 1809.60 94.69 0.113 1818.80 52.38 0.062 1829.00 27.11 0.032 1809.80 93.48 0.111 1819.00 51.73 0.061 1829.20 26.77 0.031 1810.00 92.28 0.110 1819.20 51.09 0.060 1829.40 26.44 0.031 1810.20 91.11 0.108 1819.40 50.46 0.059 1829.60 26.11 0.030 1810.40 89.96 0.107 1819.60 49.85 0.059 1829.80 25.79 0.030 1810.60 88.83 0.106 1819.80 49.24 0.058 1830.00 25.49 0.030 1810.80 87.72 0.104 1820.00 48.66 0.057 1830.20 25.19 0.029 1811.00 86.64 0.103 1820.20 48.08 0.057 1830.40 24.89 0.029 1811.20 85.57 0.102 1820.40 47.49 0.056 1830.60 24.59 0.029 1811.40 84.53 0.100 1821.20 45.15 0.053 1830.80 24.30 0.028 1811.60 83.46 0.099 1821.60 43.97 0.052 1831.00 24.00 0.028 1811.80 82.42 0.098 1821.80 43.39 0.051 1831.20 23.71 0.028 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 8 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 1812.00 81.38 0.097 1822.00 42.81 0.050 1831.40 23.41 0.027 1812.20 80.35 0.095 1822.20 42.23 0.050 1831.60 23.13 0.027 1812.40 79.34 0.094 1822.40 41.66 0.049 1831.80 22.84 0.027 1812.60 78.32 0.093 1822.60 41.11 0.048 1832.00 22.55 0.026 1812.80 77.32 0.092 1822.80 40.55 0.048 1832.20 22.30 0.026 1813.00 76.33 0.090 1823.00 40.00 0.047 1832.40 22.06 0.026 1813.20 75.34 0.089 1823.20 39.46 0.046 1832.60 21.81 0.025 1813.40 74.37 0.088 1823.40 38.92 0.046 1832.80 21.47 0.025 1813.60 73.40 0.087 1823.60 38.39 0.045 1833.00 21.17 0.025 1813.80 72.45 0.086 1823.80 37.87 0.044 1833.20 20.89 0.024 1814.00 71.51 0.085 1824.00 37.54 0.044 1833.40 20.62 0.024 1814.20 70.58 0.084 1824.20 37.13 0.043 1833.60 20.35 0.024 1814.40 69.66 0.082 1824.40 36.71 0.043 1833.80 20.08 0.023 1814.60 68.75 0.081 1824.60 36.28 0.042 1834.00 19.82 0.023 1814.80 67.85 0.080 1824.80 35.81 0.042 1834.20 19.56 0.023 1815.00 66.96 0.079 1825.00 35.37 0.041 1834.60 19.04 0.022 1815.20 66.09 0.078 1825.20 34.93 0.041 1834.80 18.79 0.022 1815.40 65.22 0.077 1825.40 34.49 0.040 1835.00 18.53 0.021 1815.60 64.37 0.076 1825.80 33.60 0.039 1835.20 18.26 0.021 1815.80 63.53 0.075 1826.00 33.16 0.039 1835.40 18.00 0.021 1816.00 62.71 0.074 1826.20 32.71 0.038 1835.60 17.73 0.021 1816.20 61.89 0.073 1826.40 32.28 0.038 1835.80 17.46 0.020 1816.40 61.09 0.072 1826.60 31.84 0.037 1836.00 17.19 0.020 1836.20 16.92 0.020 1836.40 16.64 0.019 1836.60 16.35 0.019 1836.80 16.06 0.019 1837.00 15.77 0.018 1837.20 15.47 0.018 1837.40 15.17 0.018 1837.60 14.86 0.017 1837.80 14.55 0.017 1838.00 14.24 0.016 1838.20 13.91 0.016 1838.40 13.59 0.016 1838.60 13.25 0.015 1838.80 12.91 0.015 1839.00 12.57 0.014 1839.20 12.22 0.014 1839.40 11.86 0.014 1839.60 11.49 0.013 1839.80 11.12 0.013 1840.00 10.73 0.012 VEGP-FSAR-6 TABLE 6.2.1-61 (SHEET 9 OF 9)
REV 18 9/13 Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) Time (s) Mass Flowrate lbm/s Energy Flowrate Btu/s (E+06) 1840.20 10.34 0.012 1840.40 9.93 0.011 1840.60 9.51 0.011 1840.80 9.08 0.010 1841.00 8.62 0.010 1841.20 8.13 0.009 1841.40 7.59 0.009 1841.60 6.98 0.008 1841.80 6.17 0.007 1842.00 0.00 0.000 1900.00 0.00 0.000
VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.1-64 (SHEET 1 OF 3)
SPECIFIC PLANT DESIGN INPUT FOR MSLB ANALYSIS
1 2 3 4 5 6 Initial steam generator inventory (lbm) Faulted 115,000 129,000 154,700 186, 300 114,500 129,000 Intact 107,700 123,500 142,700 186, 300 107,500 123,500 Initial steam pressure (psia) 1003 1051 1112 1102 1003 1051 Mass added by feedwater pumping (lbm) 15,900 7,800 4,500 9,600 14,000 10,900 Mass added by feedwater flashing (lbm) 20,800 21,500 22,500 0 20,800 21,500 Unisolatable steam line volume (ft
- 3) 470 470 470 470 470 470 Auxiliary feedwater addition rate (lbm/h) (to faulted steam
generator) 5.48E+05 5.48E+05 5.48E+05 5.48E+05 5.48E+05 5.48E+05 Main steam line isolation time (s) 11.2 11.0 10.9 10.9 11.4 11.8 Main feedwater line isolation time (s) 8.2 8.2 7.9 7.9 8.4 8.8 Termination of auxiliary feedwater addition (s) 1800 1800 1800 1800 1800 1800
VEGP-FSAR-6 TABLE 6.2.1-64 (SHEET 2 OF 3)
REV 15 4/09 Case 7 8 9 10 11 12 Initial steam generator inventory (lbm) Faulted 154,700 186,400 115,000 129,000 154,700 186,400 Intact 142,800 154,500 107,700 123,500 142,800 154,500 Initial steam pressure (psia) 1112 1102 1003 1051 1112 1102 Mass added by feedwater pumping (lbm) 11,200 23,300 16,000 18,700 11,600 29,200 Mass added by feedwater flashing (lbm) 22,600 24,900 20,800 21,500 22,600 24,900 Unisolatable steamline volume (ft
- 3) 470 470 470 470 470 470 Auxiliary feedwater addition rate (lbm/h) 5.48E+05 5.48e+05 5.48E+05 5.48E+05 5.48E+05 5.48E+05 Main steam line isolation time (s) 61.6 280.9 14.2 63.1 92.1 188.1 Main feedwater line isolation signal
reached (s) 16.4 112.7 11.2 18.3 22.2 185.1 Termination of auxiliary feedwater addition (s) 1800 1800 1800 1800 1800 1800
VEGP-FSAR-6 TABLE 6.2.1-64 (SHEET 3 OF 3)
REV 15 4/09
Case 13 14 15 16 Initial steam generator inventory (lbm)
Faulted 116,000 130,100 155,900 186,300 Intact 107,700 123,500 142,800 186,300 Initial steam pressure (psia) 1008 1051 1112 1102 Mass added by feedwater pumping (lbm) 25,100 17,600 10,700 4800 Mass added by feedwater flashing (lbm) 20,800 21,500 22,600 0 Unisolatable steam line volume (ft
- 3) 470 470 470 470 Auxiliary feedwater addition rate (lbm/h) 5.41E+05 5.41E+05 5.41E+05 5.47E+05 Main steam line isolation time (s) 50.1 48.0 46.5 114.0 Main feedwater line isolation time (s) 17.1 15.5 14.3 23.8 Termination of auxiliary feedwater addition (s) 1800 1800 1800 1800
VEGP-FSAR-6 REV 18 9/13 TABLE 6.2.1-65
SUMMARY
OF RESULTS FOR MSLB CONTAINMENT PRESSURE - TEMPERATURE ANALYSIS
Case No. Power (a) Level
(%) Break Size (ft 2 ) Break Type Max. Press. at
- Time (psig
at Time) Max.
Vapor Temp
- at Time (°F at s) Dryout Time (s) 6.29 psig
- at Time
(s) 17.19 psig
- at Time (s) 25.79 psig
- at Time (s)
1 102 Full Double ended 27.6 at 124 232 at 125 1822 2.9 28.6 94.5 2 70 Full Double ended 27.5 at 1809 232 at 157 1822 2.4 27.5 107.6 3 30 Full Double ended 30.4 at 197 238 at 197 1834 2.0 31.3 129.0 4 0 Full Double ended 30.2 at 173 238 at 173 1808 1.6 10.5 109.0 5 102 0.60 Double ended 27.1 at 1808 271 at 107 1844 6.4 61.1 201.0 6 70 0.53 Double ended 28.3 at 1833 252 at 107 1864 6.8 80.3 409.2 7 30 0.36 Double ended 32.4 at 1820 260 at 95 1912 9.4 52.1 537.4 8 0 0.20 Double ended 22.9 at 6339 226 at 119 6283 17.6 133.1 -
9 102 0.33 Double ended 28.5 at 1831 249 at 111 1898 11.0 136.5 854.1 10 70 0.32 Double ended 30.8 at 1836 273 at 71 1920 11.2 53.6 442.9 11 30 0.22 Double ended 32.7 at 2177 262 at 97 2155 15.1 83.7 1085.7 12 0 0.10 Double ended 17.0 at 8603 220 at 135 8600 33.7 -
- 13 102 0.86 Split 31.9 at 190 303 at 109 1832 10.1 40.4 89.7 14 70 0.908 Split 31.3 at 226 301 at 109 1834 8.5 38.2 93.8 15 30 0.944 Split 31.8 at 327 299 at 107 1856 7.3 36.7 96.7 16 0 0.40 Split 36.5 at 898 260 at 117 1842 16.8 103.8 484.6
____________
- a. % Power Level of 3579 MWt.
VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.1-66 SEQUENCE OF EVENTS FOR CASE 16 -
PEAK CALCULATED CONTAINMENT PRESSURE FOR MSLB
Time (s) Event 0.0 Break occurs, blowdown from all steam generators. 16.8 Containment pressure setpoint for isolation of main feedwater lines reached (6.29 psig). 23.8 Main feedwater line isolation valves closed. 103.8 Containment pressure setpoint for isolation of main steam lines reached (17.19 psig). 114 Main steam line isolation valves closed, blowdown from broken loop steam generator and unisolated steam piping only. 116.8 Air coolers start. 117.0 Peak containment vapor temperature of 260
°F is reached. 484.6 Containment pressure setpoint for actuation of containment sprays reached (25.79 psig). 584.6 Containment sprays start. 897.6 Peak containment pressure of 36.5 psig is reached. 1800.0 Auxiliary feedwater addition is terminated. 1842 Dryout occurs, steam generator dry following auxiliary feedwater termination.
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 1 OF 17)
CONTAINMENT PENETRATION/ISOLATION VALVE INFORMATION Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow
1 57 (i) Main steam and steam to auxiliary feed-water pump driver
Secondary
coolant 29.5 29.5 8
6 6
6 6
6 4
2 4
1 0.75 4
1 1
4 1
1 1
1 Yes 1X4DB159-3 2X4DB159-3 11 HV-3006A HV-3006B PV-3000 PSV-3001 PSV-3002 PSV-3003 PSV-3004 PSV-3005 HV-3009 176 088 356 X-192 HV-13005B X-117 X-119 HV-13005A X-209 X-211 (o) X-429 X-438 (p) Out Out Out Out Out Out Out Out Out In Out Out Out Out Out Out Out Out Out In Out A 20'-9" 28'-4" 45'-10" 20'-9" 20'-9" 20'-9" 20'-9" 20'-9" 10'-1"
-
7'-10" 29'-8" 28'-0" 42'-4" 23'-3" 30'-10" 26'-6" 10'-0" 45'-0"
-
- Gate Gate Globe Relief Relief Relief Relief Relief Gate Globe Gate Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe E/H E/H E/H Self Self Self Self Self Elec. motor Manual Manual Manual Manual Air Manual Manual Air Manual Manual Manual Manual N N
E N
N N
N N
E N
N N
N N
N N
N N
N N
N Auto Auto Auto Auto Auto Auto Auto Auto Remote man.
Manual Manual Manual Manual Auto Manual Manual Auto Manual Manual Manual Manual Remote man.
Remote man.
Remote man.
None None None None None None None None None None Remote man.
None None Remote man.
None None None None O O C
C C
C C
C O
C C
C C
O C
C O
C C
C C C C
C C
C C
C C
O C
C C
C C
C C
C C
C C
C C C
C C
C C
C C
O C
C C
C C
C C
C C
C C
C FC FC FC NA NA NA NA NA FAI NA NA NA NA FC NA NA FC NA NA NA NA SLI SLI Process/Signal
NA NA NA NA NA Remote man.
NA NA NA NA SLI NA NA SLI NA NA NA NA 5 5
NA NA NA NA NA NA NA NA NA NA NA 5
NA NA 5
NA NA NA NA A B
A NA NA NA NA NA B (dc)
NA NA NA NA B
NA NA A
NA NA NA NA Out
2 57 (i) Main steam and steam to auxiliary feed-water pump driver
Secondary
coolant 29.5 29.5 8
6 6
6 6
6 4
2 4
0.75 1
1 1
1 1
4 4
1 1
1 Yes 1X4DB159-3 2X4DB159-3 11 HV-3016A HV-3016B PV-3010 PSV-3011 PSV-3012 PSV-3013 PSV-3014 PSV-3015 HV-3019 178 082 X-196 X-207 X-162 (0) X-121 X-123 358 HV-13007A HV-13007B X-215 X-166 (p) X-430 Out Out Out Out Out Out Out Out Out In Out Out Out Out Out Out Out Out Out Out Out In A 28'-4 9/16" 36'-11 9/16" 16'-3" 27'-9" 27'-9" 27'-9" 27'-9" 27'-9" 5'-9"
-
4'-7" 14'-1" 27'-0" 26'-11" 31'-11" 40'-6" 10'-8" 51'-6" 57'-10" 46'-5" 26'-0"
- Gate Gate Globe Relief Relief Relief Relief Relief Gate Globe Gate Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe E/H E/H E/H Self Self Self Self Self Elec. motor Manual Manual Manual Manual Manual Manual Manual Manual Air Air Manual Manual Manual N N
E N
N N
N N
E N
N N
N N
N N
N N
N N
N N Auto Auto Auto Auto Auto Auto Auto Auto Remote man.
Manual Manual Manual Manual Manual Manual Manual Manual Auto Auto Manual Manual Manual Remote man.
Remote man.
Remote man.
None None None None None None None None None None None None None None Remote man.
Remote man.
None None None O O C
C C
C C
C O
C C
C C
C C
C C
O O
C C
C C C
C C
C C
C C
O C
C C
C C
C C
C C
C C
C C C C
C C
C C
C C
O C
C C
C C
C C
C C
C C
C C FC FC FC NA NA NA NA NA FAI NA NA NA NA NA NA NA NA FC FC NA NA NA SLI SLI Process/Signal
NA NA NA NA NA Remote man.
NA NA NA NA NA NA NA NA SLI SLI NA NA NA 5 5
NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA 5
5 NA NA NA A B
B NA NA NA NA NA A (dc)
NA NA NA NA NA NA NA NA A
B NA NA NA Out
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 2 OF 17)
Valve Location Relative to Con-
Len g th Valve
Actuation Mode
ValvePosition
Valve Power Source Penetra-tion Numbe r GDC o r RG
S ystem Name
Fluid Line Size (in.) ESF o r Support S y stems Drawing Numbe r A rrange-ment Fig.
6.2.4-1Valve Numbe r tainment Inside/
OutsideType Tests of Pipe (f t-in.)
T y pe Operato r Essential or Nonessential
Primar y
Secondar y Normal ShutdownPost- A cciden tPower Failure Actuation Si gnal Closure Time (s) 1E Bus A o r B Normal Direction ofFlow 3 57 (i) Main steam line Secondar y coolant 29.5 29.5 8
6 6
6 6
6 2
4 1
0.75 1
1 4
4 1
1 Yes 1X4DB159-1 2X4DB159-1 11 HV-3026 AHV-3026B PV-3020 PSV-3021 PSV-3022 PSV-3023 PSV-3024 PSV-3025 218 093 352 X-198 X-125 X-127 HV-13008A HV-13008B X-217 X-431 Ou tOut Out Out Out Out Out Out In Out Out Out Out Out Out Out Out In A27'-4 3/4" 35'-1 3/4" 15'-3" 26'-9" 26'-9" 26'-9" 26'-9" 26'-9"
-
3'-7" 13'-11" 26'-9" 30'-11" 39'-6" 50'-6" 56'-10" 57-'1"
-Gate Gate Globe Relief Relief Relief Relief Relief Globe Gate Globe Globe Globe Globe Globe Globe Globe Globe E/H E/H E/H Self Self Self Self Self Manual Manual Manual Manual Manual Manual Air Air Manual manual N N
E N
N N
N N
N N
N N
N N
N N
N N A uto Auto Auto Auto Auto Auto Auto Auto Manual Manual Manual Manual Manual Manual Auto Auto Manual Manual Remote man.
Remote man.
Remote man.
None None None None None None None None None None None Remote man.
Remote man.
None None O O
C C
C C
C C
C C
C C
C C
O O
C C C C
C C
C C
C C
C C
C C
C C
C C
C C C C
C C
C C
C C
C C
C C
C C
C C
C C FC FC FC NA NA NA NA NA NA NA NA NA NA NA FC FC NA N A SLI SLI Process/Signal
NA NA NA NA NA NA NA NA NA NA NA SLI SLI NA N A 5 5
NA NA NA NA NA NA NA NA NA NA NA NA 5
5 NA N A A B B
NA NA NA NA NA NA NA NA NA NA NA A
B NA N A Ou t 4 57 (i) Main steam line
Secondar y coolant 29.5 29.5 8
6 6
6 6
6 2
4 1
0.75 1
1 1
4 4
1 1 Yes 1X4DB159-1 2X4DB159-1 11 HV-3036 AHV-3036B PV-3030 PSV-3031 PSV-3032 PSV-3033 PSV-3034 PSV-3035 220 152 X-213 (o) X-194 354 X-129 X-131 HV-13006A HV-13006B X-432 X-440 (p) Ou tOut Out Out Out Out Out Out In Out Out Out Out Out Out Out Out In Out A20'-0" 27'-4" 49'-3" 20'-9 3/4" 20'-9 3/4" 20'-9 3/4" 20'-9 3/4" 20'-9 3/4"
-
6'-10" 28'-5" 27'-10" 27'-0" 22'-6" 29'-10" 26'-6" 42'-4"
-
- Gate Gate Globe Relief Relief Relief Relief Relief Globe Gate Globe Globe Globe Globe Globe Globe Globe Globe Globe E/H E/H E/H Self Self Self Self Self Manual Manual Manual Manual Manual Manual Manual Air Air Manual Manual N N
E N
N N
N N
N N
N N
N N
N N
N N
N A uto Auto Auto Auto Auto Auto Auto Auto Manual Manual Manual Manual Manual Manual Manual Auto Auto Manual Manual Remote man.
Remote man.
Remote man.
None None None None None None None None None None None None Remote man.
Remote man.
None None O O
C C
C C
C C
C C
C C
C C
C O
O C
C C C
C C
C C
C C
C C
C C
C C
C C
C C
C C C
C C
C C
C C
C C
C C
C C
C C
C C
C FC FC FC NA NA NA NA NA NA NA NA NA NA NA NA FC FC NA NA SLI SLI Process/Signal
NA NA NA NA NA NA NA NA NA NA NA NA SLI SLI NA NA 5 5
NA NA NA NA NA NA NA NA NA NA NA NA NA 5
5 NA NA A B A
NA NA NA NA NA NA NA NA NA NA NA NA A
B NA NA Ou t 5 N/A Eddy current/sludge lancin g N/A 0.75 10 NO 1X4DB159-1 2X4DB159-1 56(k) X-017Out In B--Globe Flan g eManual N A N NManual N A None None C C C C C C N A N A N A N A N A N A N A N A-
- 7 57 (i) Steam generato r blowdown Secondar y coolant 3 1.5 0.75 1 No 1X4DB159-3 2X4DB159-3 34 HV-7603 A 126 335 409 Ou t In In In A 10'
-
-
-Globe Globe Globe Globe A i rManual Manual manual N N
N N A uto Manual Manual Manual Remote man.
None None None O C
C C C C
C C C C
C C FC NA NA N A A FS NA NA N A 15 NA NA N A A ,B NA NA N A Ou t 8 57 (i) Steam generato r blowdown Secondar y coolant 3 1.5 0.75 1
1 1 No 1X4DB159-3 2X4DB159-3 34 HV-7603B 129 336 X-157 X-164 410 Ou t In In In In In A1'-0"
-
-
-
-
-Globe Globe Globe Globe Globe Globe A i rManual Manual Manual Manual Manual N N
N N
N N A uto Manual Manual Manual Manual Manual Remote man.
C C
C C C C
C C
C C C C
C C
C C FC NA NA NA NA N A A FS NA NA NA NA N A 15 NA NA NA NA N A A ,B NA NA NA NA N A Ou t
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 3 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow
9 57 (i) Steam generator blowdown Secondary
coolant 3 1.5 0.75 1
1 No 1X4DB159-1 2X4DB159-1 34 HV-7603C 132 337 X-155 407 Out In In In In A 1'-0" -
-
-
- Globe Globe Globe Globe Globe Air Manual Manual Manual Manual N N
N N
N Auto Manual Manual Manual Manual Remote man.
None None None None O C C
C C C C
C C
C C C
C C
C FC NA NA NA NA AFS NA NA NA NA 15 NA NA NA NA A,B NA NA NA NA Out
10 57 (i) Steam generator blowdown Secondary
coolant 3 1.5 0.75 1 No 1X4DB159-1 2X4DB159-1 34 HV-7603D 135 338 408 Out In In In A 1'-0" -
-
- Globe Globe Globe Globe Air Manual Manual Manual N N
N N Auto Manual Manual Manual Remote man.
None None None O C C
C C C
C C C C
C C FC NA NA NA AFS NA NA NA 15 NA NA NA A,B NA NA NA Out 11A 54 Chemical addition Secondary coolant chemicals 0.5 0.5 0.5 No 1X4DB159-1 2X4DB159-1 3A HV-5280 676 081 Out In In C 2'-0" -
- Globe Globe Globe Air Manual Manual N N
N Remote man.
Manual Manual None None None C LC C C LC C C LC C FC NA NA Remote man.(h) NA NA NA NA NA NA NA NA In
11B 57 (i) Steam generator sec-ondary side sample Secondary
coolant 0.5 0.5 0.5 0.5 0.375 No 1X4DB159-3 2X4DB159-3 35 HV-9451 HV-9553A HV-9553B 047 043 Out In In Out In A 2'-6" -
-
3'-3"
- Globe Globe Globe Globe Globe Solenoid Solenoid Solenoid Manual Manual N N
N N
N Auto Remote man.
Remote man.
Manual Manual Remote man.
None None None None O C O
C C C C
C C
C C C
C C
C FC FC FC NA NA AFS Remote man.
Remote man.
NA NA 15 NA NA NA NA A NA NA NA NA Out 11C 57 (i) Steam generator sec-ondary side sample Secondary
coolant 0.5 0.5 0.5 0.5 0.375 No 1X4DB159-3 2X4DB159-3 35 HV-9452 HV-9554A HV-9554B 048 044 Out In In Out In A 2'-6" -
-
3'-3"
- Globe Globe Globe Globe Globe Solenoid Solenoid Solenoid Manual Manual N N
N N
N Auto Remote man.
Remote man.
Manual Manual Remote man.
None None None None O C O
C C C C
C C
C C C
C C
C FC FC FC NA NA AFS Remote man.
Remote man.
NA NA 15 NA NA NA NA B NA NA NA NA Out 12A 54 Chemical addition Secondary coolant chemicals 0.5 0.5 0.5 No 1X4DB159-1 2X4DB159-1 3A HV-5281 677 084 Out In In C 2'-0" -
- Globe Globe Globe Air Manual Manual N N
N Remote man.
Manual Manual None None None C LC C C LC C C LC C FC NA NA Remote man.(h) NA NA NA NA NA NA NA NA In
12B 57 (i) Steam generator sec-ondary side sample Secondary
coolant 0.5 0.5 0.5 0.5 0.375 0.5 0.5 No 1X4DB159-1 2X4DB159-1 35 HV-9453 HV-9555A HV-9555B 049 045 X-185 X-423 (o) Out In In Out In In In A 2'-6" -
-
3'-3"
-
-
- Globe Globe Globe Globe Globe Globe Globe Solenoid Solenoid Solenoid Manual Manual Manual Manual N N
N N
N N
N Auto Remote man.
Remote man.
Manual Manual Manual Manual Remote man.
None None None None None None O C O
C C
C C C C
C C
C C
C C C
C C
C C
C FC FC FC NA NA NA NA AFS Remote man.
Remote man.
NA NA NA NA 15 NA NA NA NA NA NA B NA NA NA NA NA NA Out 12C 57 (i) Steam generator sec-ondary side sample Secondary
coolant 0.5 0.5 0.5 0.5 0.375 1X4DB159-1 2X4DB159-1 35 HV-9454 HV-9556A HV-9556B 050 046 Out In In Out In A 2'-6" -
-
3'-3"
- Globe Globe Globe Globe Globe Solenoid Solenoid Solenoid Manual Manual N N
N N
N Auto Remote man.
Remote man.
Manual Manual Remote man.
None None None None O C O
C C C C
C C
C C C
C C
C FC FC FC NA NA AFS Remote man.
Remote man.
NA NA 15 NA NA NA NA A NA NA NA NA Out
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 4 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 13A 56 Containment air radioactivity moni-
tor inlet Containment
atmosphere 1
1 1 No 1X4DB213-2 2X4DB213-2 36 HV-12975 HV-12976 X-001 In Out In C C -
3'-0"
- Gate Gate Globe Solenoid Solenoid Manual N N
N Auto Auto Manual Remote man.
Remote man.
None O O C C C
C C C
C FC FC NA CVI CVI NA 15 15 NA A B
NA Out 13B 56 Containment air radioactivity moni-
tor outlet Containment
atmosphere 1
1 1 No 1X4DB213-2 2X4DB213-2 45 HV-12977 HV-12978 X-003 Out In In C C 2'-0"
-
- Globe Globe Globe Solenoid Solenoid Manual N N
N Auto Auto Manual Remote man.
Remote man.
None O O C C C
C C C
C FC FC NA CVI CVI NA 15 15 NA B A
NA In 13C 1.141 Containment pressure detector DC 702 silicone oil Tubing Yes 1X4DB131 2X4DB131 48 None - A - - - - - - - - - - - - - -
14A 1.141 Reactor vessel water level
instrumentation Water Tubing Yes 1X4DB113 2X4DB113 54 None - A - - - - - - - - - - - - - -
14B 1.141 Reactor vessel water level
instrumentation W ater Tubing Yes 1X4DB113 2X4DB113 54 None - A - - - - - - - - - - - - - -
14C 1.141 Reactor vessel water level
instrumentation Water Tubing Yes 1X4DB113 2X4DB113 54 None - A - - - - - - - - - - - - - -
15 54 Purification water supply to refueling cavity Borated water 3 3 No 1X4DB130 2X4DB130 37 050 051 Out In C 1'-6" - Dia Dia Manual Manual N N Manual Manual None None LC LC LC (c) LC(c) LC LC NA NA NA NA NA NA NA NA In
18 57 (i) Feedwater Secondary coolant 16 1
No 1X4DB168-3 2X4DB168-3
12 HV-5229 X-031 Out Out A 11'-0" 4'-1" Gate Globe E/H Manual N N Auto Manual Remote man.
NA O C C C C C FC NA FI NA 5 NA A,B In
19 57 (i) Feedwater Secondary coolant 16 1
1 No 1X4DB168-3 2X4DB168-3 12 HV-5228 X-036 (p) X-037 (o) Out Out Out A 11'-0" 4'-5"
- Gate Globe Globe E/H Manual Manual N N
N Auto Manual Manual Remote man.
NA NA O C C C C
C C C
C FC NA NA FI NA NA 5 NA NA A,B NA NA In
20 57 (i) Feedwater Secondary coolant 16 1 No 1X4DB168-3 2X4DB168-3 12 HV-5230 X-073 Out Out A 8'-0" 3'-0" Gate Globe E/H Manual N N Auto Manual Remote man.
NA O C C C C C FC NA FI NA 5 NA A,B NA In
21 57 (i) Feedwater Secondary coolant 16 1
1 No 1X4DB168-3 2X4DB168-3 12 HV-5227 X-075 (o) X-076 (p) Out Out Out A 8'-9" 3'-0"
- Gate Globe Globe E/H Manual Manaul N N
N Auto Manual Manual Remote man.
NA NA O C C C C
C C C
C FC NA NA FI NA NA 5 NA NA A,B NA NA In 22 54 Demineralized water supply Demin.
water 2 2
1.5 1.0 1.0 No AX4DB190-2 38 005 038 PSV-17589 X-065 X-950 Out In In In In C 2'-2"
-
-
-
- Globe Check Relief Globe Globe Manual Self Self Manual Manual N N
N N
N Manual Auto Auto Manual Manual None None None None None LC -
-
-
C LC LC NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 5 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 23 54 Breathing air supply Compressed
air 1.5 1.5 0.75 No 1X4DB186-1 2X4DB186-1 53 211 184 226 Out In In C 2'-9" -
- Gate Check Globe Manual Self Manual N N
N Manual Auto Manual None None None LC -
C LC
-
C LC
-
C NA NA NA NA NA NA NA NA NA NA NA NA In 24 55 Hot leg sample line Primary coolant 0.5 0.5 No 1X4DB140 2X4DB140 52 HV-3502 HV-3548 Out In C 8" -
Globe Globe Air Elec. motor
N N
Auto Auto Remote man.
Remote man.
O O O
O C
C FC FAI CIA CIA 15 20 A
B
Out 28 54 ACCW supply Water with corrosion
inhibitors 10 10 No 1X4DB138-2 1X4DB138-1 40 HV-1978 HV-1979 In Out C - 1'-6" B-fly B-fly Elec. motor
Elec. motor E (g) E Remote man.
Remote man. Manual Manual O O O O C (g) C (g) FAI FAI Remote man.
Remote man.
NA NA B A In 0.75 No 1X4DB138-2 2X4DB138-2 40 PSV-1978 In 1'-6" Relief Self N Auto None C C C N/A N/A N/A N/A In 29 54 ACCW return Water with corrosion
inhibitors 10 10 0.75 No 1X4DB138-2 1X4DB138-1 24 HV-1974 HV-1975 113 In Out In C - 1'-6" B-fly B-fly Check Elec. motor
Elec. motor
Self E (g) E N Remote man.
Remote man.
Auto Manual Manual None O O
- O O
- C (g) C (g) - FAI FAI NA Remote man.
Remote man.
NA NA NA NA B A
NA Out 30 55 Safety injection to cold leg Borated water 4 2
2 2
2 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 1
1 1
1 1 Yes 1X4DB121 2X4DB121 19 HV-8835 143 144 145 146 113 HV-8823 X-119 X-120 X-121 X-122 X-123 X-124 X-125 X-126 X-181 X-239 X-241 X-346 X-348 Out In In In In In In In In In In In In In In In In In In In A 2'-0" -
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
- Gate Check Check Check Check Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor
Self Self Self Self Manual Air Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E E
E E
E N
N N
N N
N N
N N
N N
N N
N N Remote man.
Auto Auto Auto Auto Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual None None None None None Remote None None None None None None None None None None None None None O -
-
-
-
C C
C C
C C
C C
C C
C C
C C
C O
-
-
-
-
C C
C C
C C
C C
C C
C C
C C
C O
-
-
-
-
C C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA FC NA NA NA NA NA NA NA NA NA NA NA NA NA Remote man.
NA NA NA NA NA CIA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA 15 NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA B
NA NA NA NA NA NA NA NA NA NA NA NA NA In 31 55 Safety injection to hot leg Borated water 4 0.75 2
2 0.75 0.75 0.75 0.75 0.75 1
1 Yes 1X4DB121 2X4DB121 20 HV-8802B (l) HV-8824 122 123 063 X-288 X-289 X-290 X-291 X-286 X-310 Out In In In In In In In In In In A 2'-6" -
-
-
-
-
-
-
-
-
- Gate Globe Check Check Globe Globe Globe Globe Globe Globe Globe Elec. motor
Air Self Self Manual Manual Manual Manual Manual Manual Manual E N
E E
N N
N N
N N
N Remote man.
Auto Auto Auto Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None C C
-
-
C C
C C
C C
C C C
-
-
C C
C C
C C
C O (b) C
-
-
C C
C C
C C
C FAI FC NA NA NA NA NA NA NA NA NA Remote man.
CIA NA NA NA NA NA NA NA NA NA NA 15 NA NA NA NA NA NA NA NA NA B B
NA NA NA NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 6 OF 17)
Valve Location Relative to Con-
Len g th Valve
Actuation Mode
ValvePosition
Valve Power Source Penetra-tion Numbe r GDC o r RG
S ystem Name
Fluid Line Size (in.) ESF o r Support S y stems Drawing Numbe r A rrange-ment Fig.
6.2.4-1 Valve Numbe r tainment Inside/
OutsideType Tests of Pipe (f t-in.)
T y pe Operato r Essential or Nonessential
Primar y
Secondar y Normal ShutdownPost- A cciden tPower Failure Actuation Si gnal Closure Time (s) 1E Bus A o r B Normal Direction ofFlow 32 55 Boron injection line to cold leg Borated wate r 4 4
3 0.75 0.75 Yes 1X4DB119 2X4DB119 25 HV-8801 AHV-8801B
013 HV-8843 173 Ou tOut
In In In A 16'-5" 16'-3"
-
-
-Gate Gate
Check Globe Globe Elec. moto r Elec. motor
Self Air Manual E E
E N
N A uto Auto
Auto Auto Manual Remote man.
Remote man.
None Remote man.
None C C
-
C C C C
-
C C O O
-
C C FAI FAI
NA FC N A SI SI
NA CI N A N A NA
NA NA N A A B
NA NA N A In 33 55 Safety injection to hot leg Borated wate r 4 0.75 2
2 0.75 0.75 0.75 0.75 0.75 Yes 1X4DB121 2X4DB121 20 HV-8802A (l) HV-8881 120 121 290 X-292 X-293 X-294 X-295 Ou t In In In In In In In In A1'-8"
-
-
-
-
-
-
-
- Gate Globe Check Check Globe Globe Globe Globe Globe Elec. moto r Air Self Self Manual Manual Manual Manual Manual E N
E E
N N
N N
N Remote man.
Auto Auto Auto Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None C C
-
-
C C
C C
C C C
-
-
C C
C C
C O (b) C
-
-
C C
C C
C FAI FC NA NA NA NA NA NA NA Remote man.
CIA NA NA NA NA NA NA NA N A 15 NA NA NA NA NA NA NA A B NA NA NA NA NA NA NA 34 56 Containment spra y supply Borated wate r 8
8 2
0.75 Yes 1X4DB131 2X4DB131 26 HV-9001B 016 014 046 Ou t In Out Ou tC2'-0"-
2'-0" 2'-0"Gate Check Globe Globe Elec. moto r Self Manual Manual E E
N N A uto Auto Manual Manual Remote man.
None None None C
-
C C C
-
C C O
-
C C FAI NA NA N A CS NA NA N A N A NA NA N A B NA NA N A In 35 56 Containment spra y supply Borated wate r 8
8 2
0.75 Yes 1X4DB131 2X4DB131 26 HV-9001 A 015 013 047 Ou t In Out Ou tC7'-5"-
2'-1" 2'-1"Gate Check Globe Globe Elec. moto r Self Manual Manual E E
N N A uto Auto Manual Manual Remote man.
None None None C
-
C C C
-
C C O
-
C C FAI NA NA N A CS NA NA N A N A NA NA N A A NA NA N A In 36 56 RHR emergenc y sump suction Borated wate r 14 Yes 1X4DB122 2X4DB122 27 HV-8811B (l) Ou t A6'-6"GateElec. moto r E Auto Remote man.CC O (b) FAI SI (e) N ABOu t 37 56 RHR emergenc y sump suction Borated wate r 14 Yes 1X4DB122 2X4DB122 27 HV-8811A (l) Ou t A6'-6"GateElec. moto r E Auto Remote man.CC O (b) FAI SI (e) N A A Ou t 38 56 Containment spra y emergency sump
suction Borated wate r 12 Yes 1X4DB131 2X4DB131 23 HV-9002B (l) Ou t A6'-9"GateElec. moto r ERemote man. ManualCC O (b) FAIRemote man.N ABOu t
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 7 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 39 56 Containment spray emergency sump
suction Borated water
12 Yes 1X4DB131 2X4DB131 23 HV-9002A (l) Out A 6'-9" Gate Elec. motor E Remote man. Manual C C O (b) FAI Remote man. NA A Out
40 54 Fire protection water Well water 4 6
1 No 1X4DB174-4 2X4DB174-4 41 HV-27901 036 018 Out In In C 2'-6" -
- Gate Check Gate Air Self Manual N N
N Auto Auto Manual Remote man.
None None C -
C O
-
C C
-
C FC NA NA CIA NA NA 20 NA NA A,B NA NA In 41 54 Accumulator test and drain line
Borated water 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 No 1X4DB121 2X4DB121 1 HV-8871 HV-8964 HV-8888 016 X-165 X-444 (o) X-835 (o) 293 (p) 324 (p) PSV-8871 In Out Out Out Out Out Out Out Out In C 3'-7" 2'-7" 2'-9" 2'-9" 2'-6" 6'-8" 1'-6" 1'-0" 5'-6" 3'-7" Globe Globe Globe Globe Globe Globe Globe Globe Globe Relief Air Air Air Manual Manual Manual Manual Manual Manual Self N N
N N
N N
N N
N N Auto Auto Auto Manual Manual Manual Manual Manual Manual Auto Remote man.
Remote man.
Remote man.
None None None None None None None C C C
C C
C C
C C
C C C
C C
C C
C C
C C C C
C C
C C
C C
C C FC FC FC NA NA NA NA NA NA NA CIA CIA CIA NA NA NA NA NA NA NA 15 15 15 NA NA NA NA NA NA NA A B
B NA NA NA NA NA NA NA Out
Out 42 54 Nitrogen supply to accumulator N 2 1 1 0.75 No 1X4DB120 2X4DB120 3 HV-8880 017 013 Out In In C 2'-6" -
- Globe Check Globe Air Self Manual N N
N Auto Auto Manual Remote man.
None None C -
C C
-
C C
-
C FC NA NA CIA NA NA 15 NA NA A NA NA In 43 57 NSCW supply to reactor cavity
coolers Treated well water 8 0.75 1
1 0.5 0.5 1
1 1
1 1
1 1
1 1
1 No 1X4DB135-1 2X4DB135-1 17 HV-2134 PSV-11673
225 X-658 X-185 X-186 X-181 X-182 (p) 309 315 317 323 325 346 348 X-206 (o) Out In In In In In In In In In In In In In In In A 3'-0" -
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fly Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor
Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual N N
N N
N N
N N
N N
N N
N N
N N Auto Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None O C C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C C C
C C
C C
C C
C C
C C
C C
C C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA 40 NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 8 OF 17)
Valve Location Relative to Con-
Len g th Valve
Actuation Mode
ValvePosition
Valve Power Source Penetra-tion Numbe r GDC o r RG
S ystem Name
Fluid Line Size (in.) ESF o r Support S y stems Drawing Numbe r A rrange-ment Fig.
6.2.4-1Valve Numbe r tainment Inside/
OutsideType Tests of Pipe (f t-in.)
T y pe Operato r Essential or Nonessential
Primar y
Secondar y Normal ShutdownPost- A cciden tPower Failure Actuation Si gnal Closure Time (s) 1E Bus A o r B Normal Direction ofFlow 44 57 NSCW return from reactor cavity
coolers Treated well water 8 1
1 0.5 0.5 0.75 0.75 1
1 1
1 1
1 1
1 No 1X4DB135-1 2X4DB135-1 46 HV-2138X-183 X-184 (p) X-187 X-188 198 PSV-2136 313 311 321 319 327 350 352 X-222 (o) Ou t In In In In In In In In In In In In In In A2'-0"
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fl yGlobe Globe Globe Globe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Elec. moto rManual Manual Manual Manual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual N N
N N
N N
N N
N N
N N
N N
N A uto Manual Manual Manual Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None O C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA 40 NA NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA NA NA NA NA NA NA NA NA NA Ou t 45 57 NSCW supply to reactor cavity
coolers Treated well water 8 1
0.75 0.75 1
1 1
0.5 1
0.5 0.1 0.1 0.5 0.5 1
1 1
1 1
1 No 1X4DB135-2 2X4DB135-2 17 HV-2135 X-062 (o) 159 PSV-2137 022 X-654 X-565 X-389 X-490 X-388 X-405 X-431 X-391 X-392 427 443 447 455 461 429 Ou tOut Out In In In In In In In In In In In In In In In In In A1'-8" 2'-0" 2'-0"
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fl yGlobe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. moto rManual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual N N
N N
N N
N N
N N
N N
N N
N N
N N
N N A uto Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None None None None None O C
C C
C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
O C
C C C C
C C
C C
C C
C C
C C
C C
C C
C C
C C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA 40 NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA B NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA In 46 57 NSCW return from reactor cavity
coolers Treated well water 8 1
1 1
1 1
1 0.75 1
1 1
1 1
1 1
1 1
1 1
1 No 1X4DB135-2 2X4DB135-2 46 HV-2139X-040 164 X-491 (o) X-568 X-390 (o) X-387 PSV-11772
431 433 445 449 453 457 461 (p) X-816 (o) X-241 (p) X-676 (p) 451 459 (o) Ou tOut In In In In In In In In In In In In In In In In In In A1'-8" 1'-6"
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
B-fl yGlobe Globe Globe Globe Globe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. moto rManual Manual Manual Manual Manual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual N N
N N
N N
N N
N N
N N
N N
N N
N N
N N A uto Manual Manual Manual Manual Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
Manual Manual Manual Manual None None None None None None None None None None None None None None None O C
C C
C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C C C
C C
C C
C C
C C
C C
C C
C C
C C
C C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA 40 NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA B NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA Ou t VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 9 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 48 54 Normal letdown line Primary coolant 3 3
0.75 No 1X4DB114 2X4DB114 7 HV-8160 HV-8152 502 (l)(o) In Out Out C - 4'-5"
- Globe Globe Globe Air Air Manual N N
N Auto Auto Manual Remote man.
Remote man.
None O O C O O
C C C
C FC FC NA CIA CIA NA 15 15 NA A B
NA Out 49 54 Excess letdown and seal water leakoff Primary coolant 2 2
0.75 No 1X4DB114 2X4DB114 31 HV-8100 HV-8112 021 Out In In C 2'-4" -
- Globe Globe Check Elec. motor
Elec. motor
Self N N
N Auto Auto Auto Remote man.
Remote man.
None O O
- O O
- C C
- FAI FAI NA CIA CIA NA 15 15 NA B A
NA Out 50 54 Normal charging line Primary coolant 3
3 0.75 No 1X4DB114 2X4DB114 33 HV-8105
032 465 Out
In In C 3'-0"
-
- Gate
Check Globe Elec. motor
Self Manual N N N Auto
Auto M anual Remote man.
N one None O
-
C O
-
C C
-
C FAI
NA NA SI
NA NA 17 (m)
NA NA B
NA NA In 51 54 Reactor coolant pump seal water supply (pump loop No. 4) Primary coolant 1.5 1.5 0.75 Yes 1X4DB114 2X4DB114 32 HV-8103D 355 452 Out In In A 1'-3" -
- Globe Check Globe Elec. motor
Self Manual E E
N Remote man.
Auto Manual Manual None None O -
C O
-
C O
-
C FAI NA NA Remote man.
NA NA NA NA NA B NA NA In 52 54 Reactor coolant pump seal water supply (pump loop No. 3) Primary coolant 1.5 1.5 0.75 Yes 1X4DB114 2X4DB114 32 HV-8103C 354 451 Out In In A 1'-3" -
- Globe Check Globe Elec. motor
Self Manual E E
N Remote man.
Auto Manual Manual None None O -
C O
-
C O
-
C FAI NA NA Remote man.
NA NA NA NA NA B NA NA In 53 54 Reactor coolant pump seal water supply (pump loop No. 2) Primary coolant 1.5 1.5 0.75 Yes 1X4DB114 2X4DB114 32 HV-8103B 353 450 Out In In A 1'-4" -
- Globe Check G l obe Elec. motor
Self Manual E E
N Remote man.
Auto Manual Manual None None O -
C O
-
C O
-
C FAI NA NA Remote man.
NA NA NA NA NA B NA NA In 54 54 Reactor coolant pump seal water supply (pump loop No. 1) Primary coolant 1.5 1.5 0.75 Yes 1X4DB114 2X4DB114 32 HV-8103A 004 449 Out In In A 1'-4" -
- Globe Check Globe Elec. motor
Self Manual E E
N Remote man.
Auto Manual Manual None None O -
C O
-
C O
-
C FAI NA NA Remote man.
NA NA NA NA NA B NA NA In 55 NA Eddy current/sludge lancing NA 0.75 10 No 1X4DB159-1 2X4DB159-1 56(k) X-018 Out In B - - Globe (q)Flange Manual NA N N Manual None None C C C C C C NA NA NA NA NA NA NA - NA 56 55 RHR pump discharge to hot leg Borated water 12 8
8 0.75 0.75 1
0.75 Yes 1X4DB121 2X4DB121 21 HV-8840(l) 128 129 HV-8825 112 (o) X-435 296 (p) Out In In In In In In A 3'-0" -
-
-
-
-
- Gate Check Check Globe Globe Globe Globe Elec. motor
Self Self Air Manual Manual Manual E E
E N
N N
N Remote man.
Auto Auto Auto Manual Manual Manual Manual None None Remote man.
None None None C -
-
C C
C C C
-
-
C C
C C O (b) -
-
C C
C C FAI NA NA FC NA NA NA Remote man.
NA NA CIA NA NA NA NA NA NA 15 NA NA NA B NA NA B
NA NA NA In 57 55 RHR loop into cold leg Borated water 8 6
6 0.75 0.75 Yes 1X4DB121 2X4DB121 30 HV-8809A 147 148 111 HV-8890A Out In In In In A 2'-0" -
-
-
- Gate Check Check Globe Globe Elec. Motor
Self Self Manual Air E E
E N
N Remote man.
Auto Auto Manual Auto Manual None None None Remote man.
O -
-
C C O
-
-
C C O (b) -
-
C C FAI NA NA NA FC Remote man.
NA NA NA CIA NA NA NA NA 15 A NA NA NA B In 58 55 RHR loop into cold leg Borated water 8 6
6 0.75 0.75 1 Yes 1X4DB121 2X4DB121 30 HV-8809B 149 150 110 HV-8890B X-410 Out In In In In In A 1'-10" -
-
-
-
- Gate Check Check Globe Globe Globe Elec. motor
Self Self Manual Air Manual E E
E N
N N Remote man.
Auto Auto Manual Auto Manual Manual None None None Remote man.
None O -
-
C C
C O
-
-
C C
C O (b) -
-
C C
C FAI NA NA NA FC NA Remote man.
NA NA NA CIA NA NA NA NA NA 15 NA B NA NA NA B
NA In VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 10 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 59 55 RHR suction from hot leg Primary coolant 12 3 Yes 1X4DB122 2X4DB122 4 HV-8701A PSV-8708A In In A - - Gate Relief Elec. motor
Self E N Remote man.
Auto Manual None C C O C C C FAI NA Remote man.
NA NA NA A NA Out 60 55 RHR suction from hot leg Primary coolant 12 3 Yes 1X4DB122 2X4DB122 4 HV-8702A PSV-8708B In In A - - Gate Relief Elec. motor
Self E N Remote man.
Auto Manual None C O C C C FAI NA Remote man.
NA NA NA B NA Out 62 54 Pressurizer relief tank sample to waste gas compressor suction and N supply Waste gas/N 1 1 No 1X4DB112 2X4DB112 2 HV-8033 HV-8047 Out In C 1'-6" - Dia Dia Air Air N N Auto Auto Remote man.
Remote man.
C C C C C C FC FC CIA CIA 15 15 B A O u t/In 63 54 Pressurizer relief tank makeup water supply Demin.
water 3 3
1 No 1X4DB112 2X4DB112 39 HV-8028 112 020 Out In In C 1'-6" -
- Dia Check Globe Air Self Manual N N
N Auto Auto Manual Remote man.
None None O -
LC O
-
LC C
-
LC FC NA NA CIA NA NA 15 NA NA B NA NA In 64A NA Flow verification and pressure sensing
piping Containment
atmosphere 0.5 1 No 1X4DB132 2X4DB132 51 119 NA In In B - Globe Flange Manual NA N N Manual NA None None C NA C NA C NA NA NA NA NA NA NA NA NA 64B NA Flow verification and pressure sensing
piping Containment
atmosphere 0.5 1 N o 1X4DB132 2X4DB132 51 120 NA In In B - Globe Flange Manual NA N N Manual NA None None C NA C NA C NA NA NA NA NA NA NA NA NA 67A 55 Pressurizer steam sample line Primary coolant 0.5 0.5 No 1X4DB140 2X4DB140 7 HV-3514 HV-3513 Out In C 4'-0" - Globe Globe Air Air N N Auto Auto Remote man.
Remote man.
C C C C C C FC FC CIA CIA 15 15 A B Out 67B 55 Pressurizer liquid sample line Primary coolant 0.5 0.5 No 1X4DB140 2X4DB140 7 HV-3507 HV-3508 In Out C - 3'-0" Globe Globe Air Air N N Auto Auto Remote man.
Remote man.
C C C C C C FC FC CIA CIA 15 15 B A Ou t 67C 1.141 Containment pressure detector DC 702 Silicone oil Tubing Yes 1X4DB131 2X4DB131 48 - A - - - - - - - - - - - - - -
68 NA Containment leak rate test Containment
atmosphere 0.75 8
0.75 No
1X4DB132 2X4DB132 18 018 (o) NA 019 (p) In In In B - -
- Globe Flange Globe Manual NA Manual N N
N Manual NA Manual None None None C NA C C NA C C NA C NA NA NA NA NA NA NA NA NA NA NA NA
69A 54 Chemical addition Secondary coolant chemicals 0.5 0.5 0.5 No 1X4DB159-3 2X4DB159-3 3B HV-5278 678 087 Out In In C 1'-3" -
- Globe Globe Globe Air Manual Manual N N
N Remote Manual Manual None None None C LC C C LC C C LC C FC NA NA Remote man.(h) NA NA NA NA NA NA NA NA In
69B 54 Chemical addition Secondary coolant chemicals 0.5 0.5 0.5 No 1X4DB159-3 2X4DB159-3 3B HV-5279 679 090 Out In In C 1'-6" -
- Globe Globe Globe Air Manual Manual N N
N Remote Manual Manual None None None C LC C C LC C C LC C FC NA NA Remote man.(h) NA NA NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 11 OF 17)
Valve Location Relative to Con-
Len g th Valve
Actuation Mode
ValvePosition
Valve Power Source Penetra-tion Numbe r GDC o r RG
S ystem Name
Fluid Line Size (in.) ESF o r Support S y stems Drawing Numbe r A rrange-ment Fig.
6.2.4-1Valve Numbe r tainment Inside/
OutsideType Tests of Pipe (f t-in.)
T y pe Operato r Essential or Nonessential
Primar y
Secondar y Normal ShutdownPost- A cciden tPower Failure Actuation Si gnal Closure Time (s) 1E Bus A o r B Normal Direction ofFlow 69C 1.141 Containment pressure detecto r DC 702 silicone oil Tubing Yes 1X4DB131 2X4DB131 48 None-A------ -------- 70 A 56 Containment H 2 monitor suction Containmen t atmosphere 0.75 0.75 0.75 Yes 1X4DB213-2 2X4DB213-2 14 HV-2791 AHV-2790A HV-2790B Ou t In In C2'-9"-
- Globe Globe Globe Solenoid Solenoid Solenoid E E
E Remote man.
Remote man.
Remote man.
Remote man.
(n) None None C C
C C C
C O O
O FC FC FC Remote man.
Remote man.
Remote man.
N A NA NA A B B Ou t 70B 56 Containment H 2 monitor discharge Containmen t atmosphere 0.75 0.75 0.75 Yes 1X4DB213-2 2X4DB213-2 8 HV-2793 A 001 039 Ou t In In C1'-2"-
- Globe Check Globe Solenoid Self Manual E E
N Remote man.
Auto Manual Remote man.
(n) None None C
-
C C
-
C O
-
C FC NA NA Remote man.
NA NA N A NA NA A NA NA In 70C 1.141 Containment pressure detecto r DC 702 silicone oil Tubing Yes 1X4DB131 2X4DB131 48 None-A------ -------- 71 A 56 Containment H 2 monitor suction Containmen t atmosphere 0.75 0.75 0.75 Yes 1X4DB213-2 2X4DB213-2 14 HV-2792BHV-2791B HV-2792A In Out In C-1'-9"
- Globe Globe Globe Solenoid Solenoid Solenoid E E
E Remote man.
Remote man.
Remote man. None Remote man.(n) None C C C C C
C O O
O FC FC FC Remote man.
Remote man.
Remote man.
N A NA NA A B A Ou t 71B 56 Containment H 2 monitor discharge Containmen t atmosphere 0.75 0.75 0.75 Yes 1X4DB213-2 2X4DB213-2 8 HV-2793B 002 040 Ou t In In C2'-3"-
- Globe Check Globe Solenoid Self Manual E E
N Remote man.
Auto Manual Remote man.
(n) None Manual C
-
C C
-
C O
-
C FC NA NA Remote man.
NA NA N A NA NA B NA NA In 71C 1.141 Containment pressure detecto r DC 702 silicone oil Tubing Yes 1X4DB131 2X4DB131 48 None-A------ -------- 72 A 54 Accumulator sample line Borated water 0.75 0.75 0.75 No 1X4DB120 2X4DB120 28 HV-10950 159 178 In Out In C-2'-9"
-Globe Globe Globe Solenoid Manual Manual N N
N A uto Manual Manual Remote man.
None None C LC C C LC C C LC C FC NA N A CI A NA N A 15 NA N A A NA N A Ou t 72B 54 Accumulator sample line Borated water 0.75 0.75 0.75 No 1X4DB120 2X4DB120 28 HV-10952 161 183 In Out In C-2'-4"
-Globe Globe Globe Solenoid Manual Manual N N
N A uto Manual Manual Remote man.
None None C LC C C LC C C LC C FC NA N A CI A NA N A 15 NA N A A NA N A Ou t 73 A 54 Accumulator sample line Borated water 0.75 0.75 0.75 No 1X4DB120 2X4DB120 28 HV-10951 160 181 In Out In C-2'-9"
-Globe Globe Globe Solenoid Manual Manual N N
N A uto Manual Manual Remote man.
None None C LC C C LC C C LC C FC NA N A CI A NA N A 15 NA N A B NA N A Ou t 73B 54 Accumulator sample line Borated water 0.75 0.75 0.75 No 1X4DB120 2X4DB120 28 HV-10953 162 185 In Out In C-2'-4"
-Globe Globe Globe Solenoid Manual Manual N N
N A uto Manual Manual Remote man.
None None C LC C C LC C C LC C FC NA N A CI A NA N A 15 NA N A B NA N A Ou t 77 54 Reactor coolant drain tank pump discharge
Primar y coolant 3 3
1 1
1 1
1 1
1 1 No 1X4DB127 2X4DB127 29 HV-7699 HV-7136 X-186 (o) X-189 (o) X-218 (o) X-220 (o) X-153 (p) X-154 (p) X-173 (p) X-229 (p) InOut In In In In In In In In C-1'-0"
-
-
-
-
-
-
-
- Dia Dia Globe Globe Globe Globe Globe Globe Globe Globe A i r Air Manual Manual Manual Manual Manual Manual Manual Manual N N
N N
N N
N N
N N A uto Auto Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
Remote man.
None None None None None None None None O O
C C
C C
C C
C C O O
C C
C C
C C
C C C C
C C
C C
C C
C C FC FC NA NA NA NA NA NA NA NA CI A CIA NA NA NA NA NA NA NA NA 15 15 NA NA NA NA NA NA NA NA A B NA NA NA NA NA NA NA NA Ou t 0.75 PSV-7699In1'-0"ReliefSelfN Auto NoneCCCN/A N/A N/A N/A Ou t VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 12 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 78 54 Normal containment sump pumps
discharge Drains 3 3 0.75 No 1X4DB143
2X4DB143 10 HV-780 HV-781 PSV-0780 In Out In C - 3'-9" 3'-9" Gate Gate Relief Air Air Self N N
N Auto Auto Auto Remote man.
Remote man.
None O O C C C
C C C
C FC FC N/A CIA CIA N/A 15 15 N/A A B
N/A Out
Out 79 54 Reactor coolant drain tank vent
and H 2 supply Gas 0.75 0.75 No 1X4DB127 2X4DB127 9 HV-7126 HV-7150 In Out C - 1'-9" Dia Dia Air Air N N Auto Auto Remote man.
Remote man.
O O C C C C FC FC CIA CIA 15 15 A B Out 80 56 Service air and post- LOCA purge air supply Compressed
air 4 4
0.75 0.75 No 1X4DB186-1 2X4DB186-1 6 HV-9385 034 228 229 Out In Out In C 2'-0" -
1'-3"
- Gate Check Globe Globe Air Self Manual Manual N N
N N Auto Auto Manual Manual Remote man.
None None None C -
C C C
-
C C C
-
C C FC NA NA NA CIA NA NA NA 20 NA NA NA A,B NA NA NA In 81 54 Instrument air Compressed air 2 2
0.75 No 1X4DB186-4 2X4DB186-4 5 HV-9378 049 256 Out In In C 1'-6" -
- Globe Check Globe Air Self Manual N N
N Auto Auto Manual Remote man.
None None O - C O -
C C
-
C FC NA NA CIA NA NA 15 NA NA A,B NA NA In 83 56 Normal containment purge supply and equalizing Containment
atmosphere 24 24 14 14 0.75 No 1X4DB213-1 2X4DB213-1 15 HV-2626A HV-2627A HV-2626B HV-2627B 001 In Out In Out Out C - 7'-0"
-
6'-0" 2'-3" B-fly B-fly B-fly B-fly Gate Elec. motor
Elec. motor
Air Air Manual N N
N N
N Auto Auto Auto Auto Manual Remote man.
Remote man.
Remote man.
Remote man.
None LC LC C
C LC O O
C C
LC C C
C C
LC FAI FAI FC FC NA CVI CVI CVI CVI NA 10 10 5
5 NA A B
A B
NA In 84 56 Normal containment purge exhaust and equalizing Containment
atmosphere 24 24 14 14 0.75 No 1X4DB213-1 2X4DB213-1 47 HV-2628A HV-2629A HV-2628B HV-2629B 001 In Out In Out Out C - 13'-0"
-
7'-0" 2'-3" B-fly B-fly B-fly B-fly Gate Elec. motor
Elec. motor
Air Air Manual N N
N N
N Auto Auto Auto Auto Manual Remote man.
Remote man.
Remote man.
Remote man.
None LC LC C
C LC O O
C C
LC C C
C C
LC FAI FAI FC FC NA CVI CVI CVI CVI NA 10 10 5
5 NA A B
A B
NA Out 85C 1.141 Containment pres-sure detector DC 702 silicone oil Tubing Yes 1X4DB131 2X4DB131 48 None - A - - - - - - - - - - - - - -
86A 56 Post-accident sampling Containment
atmosphere 1.0 1.0 0.5 No 1X4DB110 2X4DB110 45 HV-8211 HV-8212 X-002 In Out In C - 8'-6"
- Globe Globe Globe Solenoid Solenoid Manual N N
N Auto Auto Manual Remote man.
Remote man.
None C C C C C
C C C
C FC FC NA CIA CIA NA 15 15 NA B A
NA In
87 NA Containment leak rate test Containment
atmosphere 0.75 8.0 0.75 No 1X4DB132 2X4DB132 18 019 (o) NA 018 (p) In In In B - -
- Globe Flange Globe Manual NA Manual N N
N Manual NA Manual None None None C NA C C NA C C NA C NA NA NA NA NA NA NA NA NA NA NA NA -
88A 1.141 Reactor vessel water level
instrumentation Water Tubing Yes 1X4DB113 2X4DB113 54 None - A - - - - - - - - - - - -
- -
88B 1.141 Reactor vessel water level
instrumentation Water Tubing Yes 1X4DB113 2X4DB113 54 None - A - - - - - - - - - - - - - -
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 13 OF 17)
Valve Location Relative to Con-
Len g th Valve
Actuation Mode
ValvePosition
Valve Power Source Penetra-tion Numbe r GDC o r RG
S ystem Name
Fluid Line Size (in.) ESF o r Support S y stems Drawing Numbe r A rrange-ment Fig.
6.2.4-1Valve Numbe r tainment Inside/
OutsideType Tests of Pipe (f t-in.)
T y pe Operato r Essential or Nonessential
Primar y
Secondar y Normal ShutdownPost- A cciden tPower Failure Actuation Si gnal Closure Time (s) 1E Bus A o r B Normal Direction ofFlow 88C 1.141 Reactor vessel water level
instrumentation Wate r Tubing Yes 1X4DB113 2X4DB113 54 None-A------ -------- 89 N A Transfer tube Refueling pool water 20 0.50 No 1X4DB130 2X4DB130 13 N A 090 In In B-FlangeGlobe N AManual N N N A Manual None None C C C (c) C C C N A NA N A NA N A NA N A NA - 90 N A Eddy current/sludge lancin g N A 0.75 10 No 1X4DB159-1 2X4DB159-1 56(k) X-019 Ou t In B--Globe Flan g eManual N NManual None None C N A C N A C N A N A N A N A N A N A N A N A N A- 91 57 NSCW supply to containment coolers Treated well water 8 1
0.75 1
1 1
0.5 0.5 1
0.5 0.5 1
1 0.75 1
1 1
1 1
1 1
1 Yes 1X4DB135-2 2X4DB135-2 55 HV-1809 019 PSV-1817 X-005 X-652 X-486 X-394 X-395 X-393 X-397 X-398 X-396 X-487 PSV-11774
365 369 375 377 387 381 439 441 Ou t In In Out In In In In In In In In In In In In In In In In In In A2'-6"
-
-
1'-6"
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-B-fl yGlobe Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Elec. moto rManual Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N
N N
N N
N N
N N
N N A uto Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None None None None None None None O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA N A SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA N A N A NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA N A B NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA N A In 92 57 NSCW supply to containment coolers Treated well water 8 3
1 0.75 1
1 0.5 0.5 1
1 0.5 0.5 1
1 1
1 1
1 1
1 1 Yes 1X4DB135-2 2X4DB135-2 44 HV-1807 001 018 PSV-1815 X-077 X-656 X-400 X-401 X-500 (o) X-245 X-403 X-404 393 397 403 405 409 415 435 437 X-399 (p) Ou t In In In Out In In In In In In In In In In In In In In In In A2'-6"
-
-
-
1'-6"
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fl yGate Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. moto rManual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N
N N
N N
N N
N N
N A uto Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None None None None None None O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA N A NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA B NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 14 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 93 57 NSCW supply to containment coolers Treated well water 8 3
0.75 1
0.5 0.5 1
0.5 0.5 1
0.75 1
1 1
1 1
1 1
1 1
1 Yes 1X4DB135-1 2X4DB135-1 44 HV-1806 002 PSV-1814 X-127 X-196 X-197 X-199 X-120 X-121 X-212 (o) PSV-11672 285 291 293 297 303 305 338 340 482 X-198 (p) Out In In In In In In In In In In In In In In In In In In In In A 2'-0" -
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fly Gate Relief Globe Globe Globe Globe Globe Globe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor Manual Self Manual Manual Manual Manual Manual Manual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N
N N
N N
N N
N N
N Auto Manual Auto Manual Manual Manual Manual Manual Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None None None None None None O C C
C C
C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA In 94 57 NSCW supply to containment coolers Treated well water 8 1
0.75 1
0.5 0.5 1
1 0.5 0.5 0.75 1
1 1
1 1
1 1
1 Yes 1X4DB135-1 2X4DB135-1 55 HV-1808 015 PSV-1816 X-259 X-193 X-194 X-192 X-189 X-190 X-191 PSV-11671
261 267 269 273 279 281 330 332 Out In In In In In In In In In In In In In In In In In In A 3'-0" -
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fly Globe Relief Globe Globe Globe Globe Globe Globe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor Manual Self Manual Manual Manual Manual Manual Manual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N
N N
N N
N N
N Auto Manual Auto Manual Manual Manual Manual Manual Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None None None None O C C
C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 15 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 95 57 NSCW return from containment coolers
Treated well water 8 1
1 1
1 1
1 1
1 1
1 1
1 1
1 Yes 1X4DB135-2 16 HV-1831 X-046 034 X-488 X-489 X-383 X-384 367 373 371 379 385 383 389 391 Out Out In In In In In In In In In In In In In A 2'-0" 1'-6"
-
-
-
-
-
-
-
-
-
-
-
-
- B-fly Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N
N N
N Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None O C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA B NA NA NA NA NA NA NA NA NA NA NA NA NA NA Out 96 57 NSCW return from containment coolers Treated well water 8 1
1 1
1 0.75 1
1 1
1 1
1 1
1 1
1 1 Yes 1X4DB135-2 2X4DB135-2 16 HV-1823 X-052 032 X-501 (o) X-386 PSV-11773 395 (o) 401 399 407 413 411 417 (o) 419 X-385 (p) 396 (p) X-813 (p) Out Out In In In In In In In In In In In In In In In A 2'-6" 1'-6"
-
-
-
-
-
-
-
-
-
-
-
-
-
-
- B-fly Globe Globe Globe Globe Relief Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor Manual Manual Manual Manual Self Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N
N N
N N
N Auto Manual Manual Manual Manual Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None None None None None None O C C
C C
C C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C
C C
C C
C FAI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA B NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA Out
97 57 NSCW return from containment coolers Treated well water 8 1
1 1
1 1
1 1
1 1
1 1 Yes 1X4DB135-1 2X4DB135-1 16 HV-1830 017 X-177 X-178 263 265 271 277 275 283 334 336 Out In In In In In In In In In In In A 2'-0" -
-
-
-
-
-
-
-
-
-
- B-fly Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None O C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C FAI NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA NA NA NA NA NA NA Out
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 16 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow 98 57 NSCW return from containment coolers Treated well water 8 1
1 1
1 1
1 1
1 1
1 1 Yes 1X4DB135-1 2X4DB135-1 16 HV-1822 016 X-179 289 287 295 301 299 307 342 344 X-180 Out In In In In In In In In In In In A 3'-0" -
-
-
-
-
-
-
-
-
-
- B-fly Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Globe Elec. motor Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual E N
N N
N N
N N
N N
N N Auto Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Manual Remote man.
None None None None None None None None None None None O C C
C C
C C
C C
C C
C O C
C C
C C
C C
C C
C C O C
C C
C C
C C
C C
C C FAI NA NA NA NA NA NA NA NA NA NA NA SI NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A NA NA NA NA NA NA NA NA NA NA NA Out 100 56 Post-accident air exhaust Containment
atmosphere 4
4 4
0.75 No 1X4DB213-1 2X4DB213-1 22 HV-2624A HV-2624B 012 001 In In Out Out C - -
6'-3" 6'-6" B-fly B-fly Gate Gate Elec. motor
Elec. motor Manual Manual N N
N N Auto Auto Manual Manual Remote man.
Remote man.
LC LC C C
LC LC FAI FAI NA NA CVI CVI NA NA NA (j) NA (j) NA NA A B
NA NA Out
101 57 (i) Auxiliary feedwater Secondary coolant 6 6
4 6
1 0.5 0.5 1
1 1
1 Yes 1X4DB168-3 2X4DB168-3 49 128 120 115 HV-15198 X-194 (o) 136 HV-5196 X-186 X-191 X-193 (p) X-241 In Out Out Out Out Out Out Out Out Out In A - 5'-8" 3'-3" 8'-3" 1'-8" 2'-2" 5'-2" 18'-9" 31'-4"
-
- Check Check Stop Gate Globe Check Globe Globe Globe Globe Globe Self Self Self Air Manual Self Air Manual Manual Manual Manual E N
E N
N N
N N
N N
N Auto Auto Auto Auto Manual Auto Remote man.
Manual Manual Manual Manual None None None Remote man.
None None None None None None None - -
LO O
C
-
C C
C C
C -
-
LO C
C
-
C C
C C
C -
-
LO C
C
-
C C
C C
C NA NA NA FC NA NA FC NA NA NA NA NA NA NA FI NA NA Remote man.(h) NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A,B NA NA NA NA NA NA NA In
102 57 (i) Auxiliary feedwater Secondary coolant 6 6
4 6
1 0.5 0.5 1.0 1
1 Yes 1X4DB168-3 2X4DB168-3 49 126 118 114 HV-15197 X-195 134 HV-5195 X-188 X-197 X-237 In Out Out Out Out Out Out Out Out In A - 6'-4" 6'-6" 9'-0" 4'-7" 2'-9" 8'-0" 20'-0" 30'-7"
- Check Check Stop Gate Globe Check Globe Globe Globe Globe Self Self Self Air Manual Self Air Manual Manual Manual E N
E N
N N
N N
N N Auto Auto Auto Auto Manual Auto Remote man.
Manual Manual Manual None None None Remote man.
None None None None None None - -
LO O
C
-
C C
C C -
-
LO C
C
-
C C
C C -
-
LO C
C
-
C C
C C NA NA NA FC NA NA FC NA NA NA NA NA NA FI NA NA Remote Man.(h) NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA A,B NA NA NA NA NA NA In
VEGP-FSAR-6 REV 17 4/12 TABLE 6.2.4-1 (SHEET 17 OF 17)
Valve Location Relative to Con-
Length
Valve
Actuation Mode
Valve Position
Valve
Power Source Penetra-tion Number GDC or RG
System Name
Fluid Line Size (in.) ESF or Support Systems Drawing Number Arrange-ment Fig.
6.2.4-1 Valve Number tainment Inside/
Outside Type Tests of Pipe (ft-in.)
Type
Operator Essential or Nonessential
Primary
Secondary Normal
Shutdown Post-Accident Power Failure Actuation
Signal Closure Time (s) 1E Bus A or B Normal Direction
of Flow
103 57 (i) Auxiliary feedwater Secondary coolant 6 6
4 6
1 0.5 0.5 1
1 Yes 1X4DB168-3 2X4DB168-3 49 127 119 116 HV-15199 X-180 (p) 135 HV-5197 X-215 X-181 (o) In Out Out Out Out Out Out In Out A - 13'-0" 10'-9" 15'-10" 5'-2" 7'-1" 7'-10"
-
- Check Check Stop Gate Globe Check Globe Globe Globe Self Self Self Air Manual Self Air Manual Manual E N
E N
N N
N N
N Auto Auto Auto Auto Manual Auto Remote man.
Manual Manual None None None Remote man.
LO O
C
-
C C
C -
-
LO C
C
-
C C
C -
-
LO C
C
-
C C
C NA NA NA FC NA NA FC NA NA NA NA NA FI NA NA Remote Man.(h) NA NA NA NA NA NA NA NA NA NA NA NA NA NA A,B NA NA NA NA NA In
104 57 (i) Auxiliary feedwater Secondary coolant 6 6
4 6
1 0.5 0.5 1 Yes 1X4DB168-3 2X4DB168-3 49 125 117 113 HV-15196 X-178 133 HV-5194 X-233 In Out Out Out Out Out Out In A - 3'-9" 9'-1" 17'-2" 5'-2" 7'-1" 7'-10"
- Check Check Stop Gate Globe Check Globe Globe Self Self Self Air Manual Self Air Manual E N
E N
N N
N N Auto Auto Auto Auto Manual Auto Remote man.
Manual None None None Remote man.
None None None None - -
LO O
C
-
C C -
-
LO C
C
-
C C -
-
LO C
C
-
C C NA NA NA FC NA NA FC NA NA NA NA FI NA NA Remote Man.(h) NA NA NA NA NA NA NA NA NA NA NA NA A,B NA NA NA NA In 56 Equipment hatch Containment atmosphere NA No NA 42 NA NA B NA None None N Manual Manual C C C NA NA NA NA 56 Personnel locks Containment atmosphere NA No NA 43 NA NA B NA None None N Manual Manual C C C NA NA NA NA 56 Emergency doors Containment NA No NA 43 NA NA B NA None None N Manual Manual C C C NA NA NA NA
- a. The following is a list of abbreviations:
GDC - General Design Criteria SLI - steam line isolation RG - NRC Regulatory Guides FI - feedwater isolation Dia - diaphragm CVI - containment ventilation isolation B-fly - butterfly CIA - c ontainment isolation phase A O - open CIB - containment isolation phase B C - closed SI - safety injection signal LC - locked closed CS - containment spray signal FC - failed closed E/H - electrohydraulic Self - actuated by the fluid pressure FAI - fail as is NA - not applicable AFS - auxiliary feedwater automatic start signal Auto - automatic b. These valves are required to function for long term cooldown and recirculation.
- c. These valves and/or flanges are opened for the refueling operation.
- d. Valves identified in this table are shown with an asterisk (*) in figure 6.2.4-1.
- f. Valve closure time is defined in SRP 6.2.4 paragraph II.6.N.
- g. The ACCW penetrations are not part of an engineered safety features system nor are they required for safe shutdown; however, the penetrati ons are classified as essential due to the importance of maintaining cooling water to the reactor coolant pump. These valves remain open post-accident and are only closed for certain accident conditions.
- h. Power to the solenoid shall be removed during all modes of operation, except during wet layup operation and periodic testing of the valves, by moving the slide links of the states terminal blocks, located in the terminal box, to the open position.
- i. These penetrations are associated with the secondary side of the steam generators. The present design isolation provisions for secondary system lines meet the intent of GDC 57 (i.e., one isolation valve capable of remote manual operation is provided), but GDC 57 was not the design basis. The steam generator and associated secondary system piping form the primary barrier to the outside environmen t, much the same as the containment liner Plate. The valves associated with these penetrations do not receive a containment isolation signal and are not credited with affecting containment isolation in the safety analysis. The verification of the integrity of this barrier is accomplished in the type A test. j. The stroke time of these values is 20 s. This penetration is verified closed by the Technical Specifications.
- k. These penetrations are associated with the steam generator eddy current and sludge lancing operation. During modes 5 and 6, the blind flanges may be removed and replaced with a fixture that allows cables and hoses to pass through. The cables are sealed, and manually operated valves are provided on t he hoses both inside and outside containment to provide the capability for containment closure during fuel movement as required by Technical Specifications.
- l. Valve disk is provided with bonnet vent on the containment side of disk.
- m. See table 6.3.2-3 for other stroke time information.
- n. For LOCA conditions with a loss of one train of dc power, the secondary actuation mode for valves HV-2791A/B and HV-2793A/B is accomplished by providing an alternate train power connection in the QPCP panel per plant procedures.
- o. Unit 1 only.
- p. Unit 2 only.
- q. Unit 2 is a gate valve.
VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 1 OF 15)
CONTAINMENT ISOLATION VALVES (a) Valve Number Function Valve Closure Time(s) 1. Containment Isolation Phase "A" 1HV-3502 Hot leg sample line 15 1HV-3548 Hot leg sample line 20 2HV-3502 Hot leg sample line and gross failed fuel detector 15 2HV-3548 Hot leg sample line and gross failed fuel detector 20 HV-8823 Safety injection pump discharge to cold leg 15 HV-8824 Safety injection pump discharge to hot leg 15 HV-8843 Boron injection line to cold leg 15 HV-8881 Safety injection pump discharge to hot leg 15 HV-27901 Fire protection water 20 HV-8871 Accumulator test and drain line 15 HV-8964 Accumulator test and drain line 15 HV-8888 Accumulator test and fill line 15 HV-8880 Nitrogen supply to accumulator 15 HV-8160 Normal letdown line 15 VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 2 OF 15)
Valve Number Function Valve Closure Time(s) 1. Containment Isolation Phase "A" (continued)
HV-8152 Normal letdown line 15 HV-8100 Excess letdown and seal water leakoff 15 HV-8112 Excess letdown and seal water leakoff 15 HV-8825 RHR pump discharge to hot leg 15 HV-8890A RHR pump discharge to cold leg 15 HV-8890B RHR pump discharge to cold leg 15 HV-8033 Pressurizer relief tank sample to waste gas compressor suction 15 HV-8047 Pressurizer relief tank sample to waste gas compressor suction 15 HV-8028 Pressurizer relief tank makeup water supply 15 HV-3514 Pressurizer steam sample line 15 HV-3513 Pressurizer steam sample line 15 HV-3507 Pressurizer liquid sample line 15 HV-3508 Pressurizer liquid sample line 15 HV-10950 Accumulator sample line 15 HV-10952 Accumulator sample line 15 HV-10951 Accumulator sample line 15 HV-10953 Accumulator sample line 15 HV-7699 Reactor coolant drain tank pump discharge 15 VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 3 OF 15)
Valve Number Function Valve Closure Time(s) 1. Containment Isolation Phase "A" (continued)
HV-7136 Reactor coolant drain tank pump discharge 15 HV-0780 Normal containment sump pumps discharge 15 HV-0781 Normal containment sump pumps discharge 15 HV-7126 Reactor coolant drain tank vent and H 2 supply 15 HV-7150 Reactor coolant drain tank vent and H 2 supply 15 HV-9385 Service air and post-LOCA purge air supply 20 HV-9378 Instrument air 15 HV-8211 Post-accident sampling 15 HV-8212 Post-accident sampling 15 2. Containment Ventilation Isolation HV-12975 Containment air radioacitvity monitor inlet 15 HV-12976 Containment air radioactivity monitor inlet 15 HV-12977 Containment air radioactivity monitor outlet 15 HV-12978 Containment air radioactivity monitor outlet 15 HV-2626A Containment pre-access purge supply and equalizing 10 HV-2627A Containment pre-access purge supply and equalizing 10 VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 4 OF 15)
Valve Number Function Valve Closure Time(s) 2. Containment Ventilation Isolation (continued)
HV-2626B Containment mini-purge supply and equalizing 5 HV-2627B Containment mini-purge supply and equalizing 5 HV-2628A Containment pre-access purge exhaust and equalizing 10 HV-2629A Containment pre-access purge exhaust and equalizing 10 HV-2628B Containment mini-purge exhaust and equalizing 5 HV-2629B Containment mini-purge exhaust and equalizing 5 HV-2624A Post-accident air exhaust N/A HV-2624B Post-accident air exhaust N/A 3. Safety Injection HV-8811B (d) RHR emergency sump suction N/A HV-8811A (d) RHR emergency sump suction N/A HV-2134 (e) NSCW supply to reactor cavity coolers 40 HV-2138 (e) NSCW return from reactor cavity coolers 40 HV-2135 (e) NSCW supply to reactor cavity coolers 40 HV-2139 (e) NSCW return from reactor cavity coolers 40 HV-8105 Normal charging line 17 (j) HV-1809 (e) NSCW supply to containment coolers N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 5 OF 15)
Valve Number Function Valve Closure Time(s) 3. Safety Injection (continued)
HV-1807 (e) NSCW supply to containment coolers N/A HV-1806 (e) NSCW supply to containment coolers N/A HV-1808 (e) NSCW supply to containment coolers N/A HV-1831 (e) NSCW return from containment coolers N/A HV-1823 (e) NSCW return from containment coolers N/A HV-1830 (e) NSCW return from containment coolers N/A HV-1822 (e) NSCW return from containment coolers N/A HV-8801A (d) Boron injection line to cold leg N/A HV-8801B (d) Boron injection line to cold leg N/A 4. Check Valves 1418-U4-038 Demineralized water supply N/A 2401-U4-184 Breathing air supply N/A 1217-U4-113 ACCW return N/A 1204-U4-143 Safety injection to cold leg N/A 1204-U4-144 Safety injection to cold leg N/A 1204-U4-145 Safety injection to cold leg N/A 1204-U4-146 Safety injection to cold leg N/A 1204-U4-122 Safety injection to hot leg N/A 1204-U4-123 Safety injection to hot leg N/A 1204-U6-013 Boron injection to cold leg N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 6 OF 15)
Valve Number Function Valve Closure Time(s) 4. Check Valves (continued) 1204-U4-120 Safety injection to hot leg N/A 1204-U4-121 Safety injection to hot leg N/A 1206-U6-016 Containment spray supply N/A 1206-U6-015 Containment spray supply N/A 2301-U4-036 Fire protection water N/A 2402-U4-017 Nitrogen supply to accumulator N/A 1208-U4-021 Excess letdown and seal water leakoff N/A 1208-U6-032 Normal charging line N/A 1208-U4-355 Reactor coolant pump seal water supply N/A 1208-U4-354 Reactor coolant pump seal water supply N/A 1208-U4-353 Reactor coolant pump seal water supply N/A 1208-U4-004 Reactor coolant pump seal water supply N/A 1204-U6-128 RHR pump discharge to hot leg N/A 1204-U6-129 RHR pump discharge to hot leg N/A 1204-U6-147 RHR loop into cold leg N/A 1204-U6-148 RHR loop into cold leg N/A 1204-U6-149 RHR loop into cold leg N/A 1204-U6-150 RHR loop into cold leg N/A 1201-U6-112 Pressurizer relief tank makeup water supply N/A 1513-U4-001 Containment H 2 monitor discharge N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 7 OF 15)
Valve Number Function Valve Closure Time(s) 4. Check Valves (continued) 1513-U4-002 Containment H 2 monitor discharge N/A 2401-U4-034 Service air and post-LOCA purge air supply N/A 2420-U4-049 Instrument air N/A 1302-U4-126 (k) Auxiliary feedwater N/A 1302-U4-118 (k) Auxiliary feedwater N/A 1302-U4-114 (k) Auxiliary feedwater N/A 1302-U4-134 (k) Auxiliary feedwater N/A 1302-U4-128 (k) Auxiliary feedwater N/A 1302-U4-120 (k) Auxiliary feedwater N/A 1302-U4-115 (k) Auxiliary feedwater N/A 1302-U4-136 (k) Auxiliary feedwater N/A 1302-U4-127 (k) Auxiliary feedwater N/A 1302-U4-119 (k) Auxiliary feedwater N/A 1302-U4-116 (k) Auxiliary feedwater N/A 1302-U4-135 (k) Auxiliary feedwater N/A 1302-U4-125 (k) Auxiliary feedwater N/A 1302-U4-117 (k) Auxiliary feedwater N/A 1302-U4-113 (k) Auxiliary feedwater N/A 1302-U4-133 (k) Auxiliary feedwater N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 8 OF 15)
Valve Number Function Valve Closure Time(s) 5. Remote Manual HV-5280 (b) Chemical Addition N/A HV-5281 (b) Chemical Addition N/A HV-1978 ACCW Supply N/A HV-1979 ACCW Supply N/A HV-1974 ACCW Return N/A HV-1975 ACCW Return N/A HV-8835 Safety injection to cold leg N/A HV-8802B Safety injection to hot leg N/A HV-8802A Safety injection to hot leg N/A HV-9002B (f) Containment spray emergency sump suction N/A HV-9002A (f) Containment spray emergency sump suction N/A HV-8103D Reactor coolant pump seal water supply N/A HV-8103B Reactor coolant pump seal water supply N/A HV-8103C Reactor coolant pump seal water supply N/A HV-8103A Reactor coolant pump seal water supply N/A HV-8840 RHR pump discharge to hot leg N/A HV-8809A RHR loop into cold leg N/A HV-8809B RHR loop into cold leg N/A HV-8701A RHR suction from hot leg N/A HV-8702A RHR suction from hot leg N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 9 OF 15)
Valve Number Function Valve Closure Time(s) 5. Remote Manual (continued)
HV-5278 (b) Chemical addition N/A HV-5279 (b) Chemical addition N/A HV-2791A (i) Containment H 2 monitor suction N/A HV-2790A (i) Containment H 2 monitor suction N/A HV-2790B (i) Containment H 2 monitor suction N/A HV-2793A (i) Containment H 2 monitor discharge N/A HV-2792B (i) Containment H 2 monitor suction N/A HV-2791B (i) Containment H 2 monitor suction N/A HV-2792A (i) Containment H 2 monitor suction N/A HV-2793B (i) Containment H 2 monitor discharge N/A HV-5194 (b)(k) Auxiliary feedwater N/A HV-5197 (b)(k) Auxiliary feedwater N/A HV-5195 (b)(k) Auxiliary feedwater N/A HV-5196 (b)(k) Auxiliary feedwater N/A HV-9556A (k) Steam generator secondary side sample N/A HV-9556B (k) Steam generator secondary side sample N/A HV-9555A (k) Steam generator secondary side sample N/A HV-9555B (k) Steam generator secondary side sample N/A HV-9554A (k) Steam generator secondary side sample N/A HV-9554B (k) Steam generator secondary side sample N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 10 OF 15)
Valve Number Function Valve Closure Time(s) 5. Remote Manual (continued)
HV-9553A (k) Steam generator secondary side sample N/A HV-9553B (k) Steam generator secondary side sample N/A HV-3009 (h) Main steam to auxiliary feedwater pump driver N/A HV-3019 (h) Main steam to auxiliary feedwater pump driver N/A 6. Manual 1213-U6-050 (c) Purification water supply to refueling cavity N/A 1213-U6-051 (c) Purification water supply to refueling cavity N/A 1418-U4-005 (c) Demineralized water supply N/A 2401-U4-211 (c) Breathing air supply N/A 1411-U4-676 (c) Chemical addition N/A 1411-U4-677 (c) Chemical addition N/A 1411-U4-678 (c) Chemical addition N/A 1411-U4-679 (c) Chemical addition N/A 1204-U4-159 (c) Accumulator sample line N/A 1204-U4-161 (c) Accumulator sample line N/A 1204-U4-160 (c) Accumulator sample line N/A 1204-U4-162 (c) Accumulator sample line N/A 1202-U4-001 NSCW supply to containment fire protection N/A 1202-U4-002 NSCW supply to containment fire protection N/A 1508-U4-012 (c) Post-accident air exhaust N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 11 OF 15)
Valve Number Function Valve Closure Time(s) 7. Containment Spray HV-9001A Containment spray supply N/A HV-9001B Containment spray supply N/A 8. Pressure Relief Valves PSV-17589 Plant demineralized water to containment N/A PSV-11673 NSCW supply to reactor cavity coolers N/A PSV-2136 NSCW from reactor cavity coolers N/A PSV-2137 NSCW supply to reactor cavity coolers N/A PSV-11772 NSCW from reactor cavity coolers N/A PSV-8708A RHR suction from hot leg N/A PSV-8708B RHR suction from hot leg N/A PSV-1817 NSCW supply to containment coolers N/A PSV-11774 NSCW supply to containment coolers N/A PSV-1815 NSCW supply to containment coolers N/A PSV-1814 NSCW supply to containment coolers N/A PSV-11672 NSCW supply to containment coolers N/A PSV-1816 NSCW supply to containment coolers N/A PSV-11671 NSCW supply to containment coolers N/A PSV-11773 NSCW supply to containment coolers N/A PSV-3001 Main steam N/A PSV-1978 ACCW Supply N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 12 OF 15)
Valve Number Function Valve Closure Time(s) 8. Pressure Relief Valves (continued)
PSV-8871 Accumulator Test and Drain N/A PSV-7699 RCDT Pump Discharge N/A PSV-0780 Normal Containment Sump Pump N/A PSV-3002 Main steam N/A PSV-3003 Main steam N/A PSV-3004 Main steam N/A PSV-3005 Main steam N/A PSV-3011 Main steam N/A PSV-3012 Main steam N/A PSV-3013 Main steam N/A PSV-3014 Main steam N/A PSV-3015 Main steam N/A PSV-3021 Main steam N/A PSV-3022 Main steam N/A PSV-3023 Main steam N/A PSV-3024 Main steam N/A PSV-3025 Main steam N/A PSV-3031 Main steam N/A PSV-3032 Main steam N/A PSV-3033 Main steam N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 13 OF 15)
Valve Number Function Valve Closure Time(s) 8. Pressure Relief Valves (continued)
PSV-3034 Main steam N/A PSV-3035 Main steam N/A 9. Other Automatic Valves HV-3006A (g) Main steam 5 HV-3006B (g) Main steam 5 HV-3016A (g) Main steam 5 HV-3016B (g) Main steam 5 HV-13005A (g) Main steam 5 HV-13005B (g) Main steam 5 HV-13007A (g) Main steam 5 HV-13007B (g) Main steam 5 HV-3026A (g) Main steam 5 HV-3026B (g) Main steam 5 HV-13008A (g) Main steam 5 HV-13008B (g) Main steam 5 HV-3036A (g) Main steam 5 HV-3036B (g) Main steam 5 HV-13006A (g) Main steam 5 HV-13006B (g) Main steam 5 VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 14 OF 15)
Valve Number Function Valve Closure Time(s) 9. Other Automatic Valves (continued)
HV-7603A (h) Steam generator blowdown 15 HV-7603B (h) Steam generator blowdown 15 HV-7603C (h) Steam generator blowdown 15 HV-7603D (h) Steam generator blowdown 15 HV-9451 (h) Steam generator secondary side sample 15 HV-9452 (h) Steam generator secondary side sample 15 HV-9453 (h) Steam generator secondary side sample 15 HV-9454 (h) Steam generator secondary side sample 15 HV-5229 (g) Feedwater 5 HV-5228 (g) Feedwater 5 HV-5230 (g) Feedwater 5 HV-5227 (g) Feedwater 5 HV-15198 (g) Feedwater 5 HV-15197 (g) Feedwater 5 HV-15196 (g) Feedwater 5 HV-15199 (g) Feedwater 5 PV-3000 Main steam N/A PV-3010 Main steam N/A PV-3020 Main steam N/A PV-3030 Main steam N/A VEGP-FSAR-6 REV 19 4/15 TABLE 6.2.4-2 (SHEET 15 OF 15)
- a. See FSAR subsection 6.2.4 for discussion of the containment isolation system.
- b. The containment isolation valves will be main tained closed by administratively controlling the air supply valve links in an open position.
- c. Locked closed.
- d. These valves are included for table completeness. The requirements of Technical Specification 3.6.3 do not apply; instead the requirements of Technical Specification 3.5.2 apply. Valve stroke times where specified will be tested pursuant to the IST program.
- e. These valves are included for table completeness. The requirements of Technical Specification 3.6.3 do not apply; instead the requirements of Technical Specification 3.7.8 apply. Valve stroke times where specified will be tested pursuant to the IST program.
- f. These valves are included for table completeness. The requirements of Technical Specification 3.6.3 do not apply; instead the requirements of Technical Specification 3.6.6 apply. Valve stroke times where specified will be tested pursuant to the IST program.
- g. These valves are included for table completeness. The requirements of Technical Specification 3.6.3 do not apply; instead the requirements of Technical Specification 3.7.2 apply to the main steam isolation and bypass valves and Technical Specifications 3.3.2 and 3.7.3 apply to the main feedwater isolation valves.
- h. These valves are included for table completeness. The requirements of Technical Specification 3.6.3 do not apply; instead the requirements of Technical Specification 3.7.5 apply. Valve stroke times where specified will be tested pursuant to the IST program.
- i. These valves may be opened on an intermittent basis under administrative control.
- j. See table 6.3.2-3 for other stroke time information.
- k. These valves are included for table completeness. The requirements of Technical Specification 3.6.3 do not apply; instead verification of containment isolation is accomplished in the type A leakage test which credits the steam generator tubes and associated piping as the primary barrier to the outside environment. These valves are associated with the secondary side of the steam generators and are not subject to GDC 57. These valves do not receive a containment isolation signal and are not credited with effecting containment isolation in the safety analyses. Reference UFSAR paragraph 6.2.4.2.1 and table 6.2.4-1, penetrations 102, 103, and 104 and reference note "I." NOTE: The valve table as shown reflects the inservice test program as modified by the second 10-year interval update. The updat ed IST program will be fully implemented by May 31, 1998.
VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.5-6 (SHEET 1 OF 4)
PLANT PARAMETERS USED TO CALCULATE POST-ACCIDENT HYDROGEN PRODUCTION Core thermal power (MWt) (a) 3565 Containment free volume (ft
- 3) 2.75 x 10 6 Normal containment temperature (°F) 120 Weight of zirconium clad incore (lb) 45,914 Percent zirconium-water reaction (%)
1.5 Hydrogen recombiner flowrate (sf 3/min) 100 Hydrogen recombiner efficiency (%)
95 Baseline Aluminum Inventory in Containment
Weight Surface Component (lb)
(ft 2)
Flux map drive system 183 48 Nuclear instrumentation system 244 57 Digital rod position indicators 199 241 Control rod drive mechanism (CRDM) 129 68 connectors Miscellaneous valves (nuclear steam 230 86 supply system) (NSSS)
Radiation monitoring system 4 4 Containment fan cooler return 765 3000 bend assemblies Communication equipment 65 10 Miscellaneous valves (balance of 11 2 plant (BOP))
Containment lighting fixture plugs 10 6 Contingency (NSSS) 250 85 Baseline Total 2,090 3,607 VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.5-6 (SHEET 2 OF 4)
Baseline Zinc Inventory in Containment
Weight Surface Type (lb)
(ft 2) Snubbers Zn (b) 11 555 Integrated reactor ZBP (c) 17,734 180,100 vessel (RV) head/CRDM shroud
Platform grating GS (d) 12,453 99,625 Pressurizer grating GS 696 5,566 Steam generator grating GS 813 6,500 Cables and related items Zn 4,988 56,127 Cable tray supports Zn 1,710 13,680 Inorganic zinc-based paint ZBP 23,476 500,691 Miscellaneous BOP items ZBP 36 31 Baseline Total 61,917 862,875
VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.5-6 (SHEET 3 OF 4)
Regulatory Guide 1.7 Hydrogen Production Calculational Assumptions
Core Cooling Solution Radiolysis Sources:
Percent of total halogens retained 50 in the core
Percent of total noble gases retained 0 in the core
Percent of other fission products 99 retained in the core
Energy absorption by core cooling solution:
Percent of gamma energy absorbed by 10 solution
Percent of beta energy absorbed by 0 solution Hydrogen production:
Molecules hydrogen produced per 100 eV 0.50 energy absorbed by solution
VEGP-FSAR-6 REV 15 4/09 TABLE 6.2.5-6 (SHEET 4 OF 4)
Sump Solution Radiolysis
Sources:
Percent of total halogens released to 50 sump solution
Percent of noble gases released to 0 sump solution Percent of other fission products 1 released to sump solution
Energy absorption by sump solution:
Percent of total energy (beta and gamma) 100 which is absorbed by the sump solution
Hydrogen production:
Molecules of hydrogen produced per 100 eV 0.5 of energy absorbed by the sump solution
Long-Term Aluminum Corrosion Rate
Mils per year 200
- a. Hydrogen generation analysis for core thermal power of 3565 MWt is bounding for MUR power uprate of 3625.6 MWt (see figure 6.2.5-7)
- b. Zn - zinc metal.
- c. ZBP - zinc-based paint.
- d. GS - galvanized steel.
REV 13 4/06 CONTAINMENT PRESSURE TRANSIENT LOCA-DOUBLE-ENDED PUMP SUCTION-MINIMUM SAFETY INJECTION FIGURE 6.2.1-1
REV 13 4/06 CONTAINMENT PRESSURE TRANSIENT LOCA-DOUBLE-ENDED PUMP SUCTION, MAXIMUM SAFETY INJECTION FIGURE 6.2.1-2
REV 13 4/06 CONTAINMENT PRESSURE TRANSIENT LOCA-DOUBLE-ENDED HOT LEG GUILLOTINE FIGURE 6.2.1-3
REV 13 4/06 CONTAINMENT TEMPERATURE TRANSIENT LOCA-DOUBLE-ENDED PUMP SUCTION- MINIMUM SAFETY INJECTION FIGURE 6.2.1-4
REV 13 4/06 CONTAINMENT TEMPERATURE TRANSIENT LOCA-DOUBLE-ENDED PUMP SUCTION- MAXIMUM SAFETY INJECTION FIGURE 6.2.1-5
REV 13 4/06 CONTAINMENT TEMPERATURE TRANSIENT LOCA-DOUBLE-ENDED HOT LEG GUILLOTINE FIGURE 6.2.1-6
REV 13 4/06 CONTAINMENT FAN COOLER PERFORMANCE LOCA-DOUBLE-ENDED PUMP SUCTION- MAXIMUM SAFETY INJECTION FIGURE 6.2.1-7
REV 13 4/06 FIGURE 6.2.1-8 THROUGH FIGURE 6.2.1-14
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E1 FIGURE 6.2.1-15 (SHEET 1 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E2 FIGURE 6.2.1-15 (SHEET 2 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E3 FIGURE 6.2.1-15 (SHEET 3 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E4 FIGURE 6.2.1-15 (SHEET 4 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E5 FIGURE 6.2.1-15 (SHEET 5 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E6 FIGURE 6.2.1-15 (SHEET 6 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E7 FIGURE 6.2.1-15 (SHEET 7 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E8 FIGURE 6.2.1-15 (SHEET 8 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E9 FIGURE 6.2.1-15 (SHEET 9 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E10 FIGURE 6.2.1-15 (SHEET 10 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E11 FIGURE 6.2.1-15 (SHEET 11 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E12 FIGURE 6.2.1-15 (SHEET 12 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E13 FIGURE 6.2.1-15 (SHEET 13 OF 65)
REV 13 4/06 REACTOR CAVITY PRESSURE RESPONSE - NODE E14 FIGURE 6.2.1-15 (SHEET 14 OF 65)