ML20106E913
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Catawba Nuclear Station UFSAR Appendix 5A. Tables Appendix 5A. Tables
Catawba Nuclear Station UFSAR Table 5-1 (Page 1 of 2)
Table 5-1 System Design and Operating Parameters Unit 1 Unit 2 Plant Design Life, years 40 40 Nominal Operating Pressure, psig 2,235 2,235 Total System Volume Including Pressurizer and Surge 13,084 11,861 Line, ft3 System Liquid Volume, Including Pressurizer Water 12,429 11,206 Level, ft 3 (55% full - Unit 1)
(55 % full - Unit 2)
Pressurizer Spray Rate, maximum gpm 900 900 Pressurizer Heater Capacity, kw 1,662 1,662 Thermal Design Parameters NSSS Power, % 100 100 NSSS Power, MWt 3,488 3,430 NSSS Power, 106 BTU/hr 11,956 11,714 Analyzed Power, MWt 3,479 3,479 6
Analyzed Power, 10 BTU/hr 11,881 11,881 Licensed Reactor Power, MWt 3,469 3,411 Licensed Reactor Power, 106 BTU/hr 11,837 11,639 Thermal Design Flow, Loop gpm 95,500 94,250 Reactor 106 lb/hr 145.3 143.4 Reactor Coolant Pressure, psia 2,250 2,250 Reactor Coolant Temperature, °F Unit 1 Unit 2 Core outlet 617.4 620.8 Vessel outlet 614.9 616.7 Core average 585.3 589.6 Vessel average 585.1 587.5 Vessel/core inlet 555.3 558.3 Steam Generator outlet 555.3 558 Steam Generator Steam Temperature °F 549 541 Steam Pressure, psia 1021 970 6
Steam Flow, 10 lb/hr total 15.5 15.12 (09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-1 (Page 2 of 2)
Unit 1 Unit 2 Feed Temperature, °F 442 440 Moisture, % max 0.25 Zero Load Temperature °F 557 Hydraulic Design Parameters Pump Design Point, Flow (gpm) 101,000 101,000 Head (ft) 279 286 Mechanical Design Flow, gpm 105,000 105,000 System Pressure Drops @ Best Estimate Flow Unit 1 Unit 2 Reactor Vessel P, psi 45.1 45.1 Steam Generator P, psi 33.0 38.3 Hot Let Piping P, psi 1.3 1.3 Pump Suction Piping P, psi 3.3 3.3 Cold Leg Piping P, psi 3.3(1) 3.3(1)
Pump Head, feet 279 286 Note:
- 1. Includes pump weir P of 2.o psi (09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-2 (Page 1 of 1)
Table 5-2. Applicable Code Addenda for RCS Components Reactor Vessel ASME III, 1971 Edition through Winter '71 - Unit 1 ASME III, 1971 Edition through Winter '72 - Unit 2 Steam Generator ASME III, 1986 Edition no addenda - Unit 1 ASME III, 1971 Edition through Winter '72 - Unit 2 Pressurizer ASME III, 1971 Edition through Winter '72 CRDM Housing ASME III, 1974 Edition through Summer '74 CRDM Head Adapter ASME III, 1971 Edition through Winter '72 Reactor Coolant Pump ASME III, 1971 Edition through Summer '73 Reactor Coolant Pipe ASME III, 1974 Edition Surge Lines ASME III, 1974 Edition Valves Pressurizer Safety Dresser ASME III, 1974 Edition through Summer '74 Gate, Globe and Check Westinghouse Valve ASME III, 1974 Edition through Summer '74 Division Borg-Warner, NVD ASME III, 1971 Edition through Summer '73 Packless Globe and Check Kerotest ASME III, 1974 Edition through Summer '73 CC1 ASME III, 1974 Edition through Winter '74 Rockwell ASME III, 1977 Edition through Summer '78 Walworth ASME III, 1971 Edition through Summer '73 Fisher ASME III, 1974 Edition through Summer '74 Addendum (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-3 (Page 1 of 1)
HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED Table 5-3. Code Cases Applicable for Operation, Maintenance, and Testing Activities Number Description N-92(1698) Waiver of Ultrasonic Transfer Method N-98(1705-1) Calibration Block Tolerance N-112(1730) Acceptance Standards for Components N-113(1731) Basic Calibration Blocks for Ultrasonic Examination of Welds 10 to 14 In. Thick N-118(1738) Acceptance Standards - Surface Indications Cladding N-198-1 Exemption from Examination for ASME Class 2 Piping Located at Containment Penetrations N-209 Conditional acceptance of Identifiable, Isolated, or Random Rounded Indications N-210 Exemptions to Hydrostatic Test Repairs N-211 Recalibration of Ultrasonic Equipment Upon Change of Personnel N-242 Steam Generator Power Operated Relief Valves 1/4" Plugs 1/4" 3000# THR'D Half Couplings 1/4" 3000# S. W. 90° Elbows 1/4" 3000# S. W. Full Couplings 3/4" 3000# THR'D Caps 1" 6000# S. W. Half Couplings 1" 9000# S. W. Full Couplings 2" 6000# S. W. Special Weld Boss 2"x3/4" 6000# S. W. Special Reducer N-416-1 Alternatative pressure test requirements for welded repairs (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-4 (Page 1 of 1)
Table 5-4. ASME Code Cases Used For Catawba Units 1 & 2 Class 1 Components Equipment Unit 1 Unit 2 Steam Generators N-411-1 1355 N-20-3 1493 N-71-15 1484 N-474-1 1528 2142 2143 1498 Pressurizer 1528 ----
RC Pipe/Fitting Fab. 1423-2 1423-2 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-5 (Page 1 of 1)
Table 5-5. Typical Plant Thermal-Hydraulic Parameters. HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED Units 2-Loop 3-Loop 4-Loop Catawba Heat Output, Core MWt 1,780 2,652 3,411 3,411 System Pressure psia 2,250 2,250 2,250 2,250 Coolant Flow gpm 178,000 265,500 354,000 397,200 (Unit 1) 387,600 (Unit 2)
Average Core Mass Velocity 106 lb/hr-ft2 2.42 2.33 2.50 2.62 Inlet Temperature °F 545 544 552.5 556.3 (558.3, Unit 2)
Core Average Tmod °F 581 580 588 586.8 (589.6, Unit 2)
Core Length Ft 12 12 12 12 Average Power Density kw/l 102 100 104 104 Maximum Fuel Temperature °F <4100 <4200 <4200 <4200 Fuel Loading kg/l 2.7 2.6 2.6 2.6 Pressurizer Volume Ft3 1000 1400 1800 1800 Pressurizer Volume Ratioed to Primary System Volume 0.157 0.148 0.148 0.144 Peak Surge Rate for Pressurizer Safety Valve Sizing Ft3/sec 21.8 33.2 41.0 34.74 Transient Pressurizer Safety Valve Flow at 2500 psia - +3% Ft3/sec 26.1 36.1 43.3 43.225 Accumulation Ratio of Safety Valve Flow to Peak Surge Rate 1.197 1.087 1.056 1.244 Full Power Steam Flow per Loop lb/sec 1078 1076 1038 1056.5 Nominal Shell-side Steam Generator Water Mass per Loop lb 100,300 106,000 106,000 122,600 (Unit 1) 103,370 (Unit 2)
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-6 (Page 1 of 3)
Table 5-6. Class 1 Primary Components Material Specifications Reactor Vessel Components (Unit 1)
Head Plates SA-533, Gr. B, Class 1 (vacuum treated)
Shell, Flange and Nozzle Forgings, Nozzle Safe SA-508, Class 2, SA-182, Type 308L (Weld Ends Metal Buildup)
CRDM and/or ECCS Appurtenances, Upper Head SB-166 or 167 and SA-182, Type F304 Instrumentation Tube Appurtenances, Lower SB-166 Head Closure Studs, Nuts, Washers Inserts and SA-540, Gr. B-24 Adaptors Core Support Pads SB-166 with Carbon less than 0.10%
Monitor Tubes and Vent Pipe SA-312, Type 316 or SB-166 Vessel Supports, Seal Ledge and Head Lifting SA-516, Gr. 70 Quenched and Tempered or SA-Lugs 533, Gr. B, Class 1 or SA-509 Class 2 (vessel supports may be of weld metal build up of equivalent strength)
Cladding and Buttering Stainless steel weld metal analysis A-7 and Ni-Cr-Fe weld metal F-Number Reactor Vessel Components (Unit 2)
Shell and Head Plates (other than core region) SA-533, Gr. A, B or C, Class 1 or 2 (vacuum treated)
Shell Plates (core region) SA-533, Gr. A or B, Class 1 (vacuum treated)
Shell, Flange and Nozzle Forgings, Nozzle Safe SA-508, Class 2 or 3, SA-182, Type F304 or F316 Ends CRDM and/or ECCS Appurtenances, Upper Head SB-167 and SA-182, Type F304 Instrumentation Tube Appurtenances, Lower SB-166 or 167 and SA-182, Type F304, F304L or Head F316 Closure Studs SA-540, Class 3, Gr. B-24 Nuts and Washers SA-540, Class 3, Gr. B-23 Core Support Pads SB-166 with carbon less than 0.15% (with an aim of less than 0.10%)
Monitor Tubes and Vent Pipe SA-312 or 376, Type 316 or SB-167 Vessel Supports, Seal Ledge and Head Lifting SA-516, Gr. 70 quenched and tempered or SA-Lugs 533, Gr. A, B or C, Class 1 or 2 (vessel supports may be of weld metal build up of equivalent strength)
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-6 (Page 2 of 3)
Cladding and Buttering Stainless steel weld metal with a corrosion resistance equal to or better than Type 304 Acceptable Values:
Cr - 18% minimum Ni - 8% minimum C -.08% maximum1 Ferrite Content 15%
Ni-Cr-Fe alloy Fe content shall be 15%maximum.
Steam Generator Components (Units 1 & 2)
Pressure Plates SA533 GR A, B or C, Class 1 or 2 Pressure Forgings (including nozzles and SA508 Class 2, 2a or 3 tubesheet)
Nozzle Safe Ends Unit 1 SA 336-F316N/316LN Unit 2 N/A Channel Heads Unit 1 SA 508 Class 3 Unit 2 SA216 Grade WCC Tubes Unit 1 SB163 Alloy 690, Code Case N-20-3 Unit 2 SB163 Ni-Cr-Fe, Annealed Cladding and Buttering Unit 1 SFA 5.9 ER 309L/ER 308L Unit 2 SFA 5.9 ER 309L - SS Closure Bolting SA193 Gr B-7 Pressurizer Components (Units 1 & 2)
Pressure Plates SA533 Gr A, Class 2 Pressure Forgings SA508 Class 2 Nozzle Safe Ends SA182 Type 316L Cladding and Buttering Stainless Steel Weld Metal Analysis A-8 and Ni-Cr-Fe Weld Metal F-Number 43 Closure Bolting SA193 Gr B-7/SA 194 Gr 7 Reactor Coolant Pump (Units 1
& 2)
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-6 (Page 3 of 3)
Pressure Forgings SA182 F 304, F 316, F-347 or F 348 Pressure Casting SA351 Gr CF8, CF8A or CF8M Tube & Pipe SA213, SA376 or SA312 -
Seamless Type 304 or 316 Pressure Plates SA240 Type 304 or 316 Bar Material SA479 Type 304 or 316 Closure Bolting SA193, SA320, SA540, SA453, Gr 660 Flywheel SA533 Gr B, Class 1 Reactor Coolant Piping (Units 1
& 2)
Reactor Coolant Pipe SA351 Gr CF8A centrifugal casting Reactor Coolant Fittings SA351 Gr CF8A Branch Nozzles SA182 Code Case 1423-2 Gr 304N Surge Line SA 376 Gr 304 Auxiliary Piping 1/2" through ANSI B36.19 12" and wall schedules 40S through 80S (ahead of second isolation valve)
All other Auxiliary piping ANSI B36.10 (ahead of second isolation valve)
Socket weld Fittings ANSI B16.11 Piping Flanges ANSI B16.5 Control Rod Drive Mechanism (Units 1 & 2)
Latch Housing SA182 Gr F304 or SA351 Gr CF8 Rod Travel Housing SA182 Gr F304 or SA336 Gr F8 Cap SA479 Type 304 Welding Materials Stainless Steel Weld Metal Analysis A-8 Note:
- 1. For multilayer cladding where the first layer is Type 309 material, the carbon content of the first layer shall be 0.1% maximum.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-7 (Page 1 of 2)
Table 5-7. Class 1 and 2 Auxiliary Components Material Specifications Valves Bodies SA182 Type F316 or SA351 Gr CF8 or CF8M Bonnets SA182 Type F316 or SA351 Gr CF8 or CF8M Discs SA182 Type F316 or SA564 Gr 630 Pressure Retaining Bolting SA453 Gr 660 Pressure Retaining Nuts SA453 Gr 660 or SA194 Gr 6 Auxiliary Heat Exchangers Tube Sheets SA240 Type 304, SA 182 Gr F304, SA516 Gr 70 with Stainless Steel Cladding (Analysis A-8)
Tubes SA213 TP 304, SA 249 TP 304 Tube Side Shell Side Heads SA240, Type 304, 304L, SA182 SA285 Gr C, SA516 Gr 70 Gr F304 SA403 Type 304 Nozzle Necks SA240 Type 304 SA312 Type SA106 Gr B 304, SA479 Type 304 Shells SA240 Type 304 SA106 Gr B, SA285 Gr C, SA315, Gr CF8A SA516 Gr 70 Flanges SA182 Gr F304 SA105 Auxiliary Pressure Vessels, Tanks, Filters, etc.
Shells & Heads SA-351 Gr CF8A, SA-312 TP304, SA-182 Gr F304, SA240 Type 304 or SA264 consisting of SA537 Gr B with Stainless Steel Weld Metal Analysis A-8 Cladding Flanges & Nozzles SA182 Gr F304 and SA105 or SA350 Gr LF2, LF3 with Stainless Steel Weld Metal Analysis A-8 Cladding Piping SA312 and SA240 TP304 or TP316 Seamless Pipe Fittings SA403 WP394 Seamless Closure Bolting & Nuts SA193 Gr B7 and SA194 Gr 2H Auxiliary Pumps Pump Casing & Heads SA351 Gr CF8 or CF8M, SA182 Gr F304 or F315 Flanges & Nozzles SA182 Gr F304 or F316, SA403 Gr WP316L Seamless Piping SA312 TP304 or TP316 Seamless Stuffing or Packing Box SA351 Gr CF8 or CF8M, SA240 TP304 or TP316, SA182 Gr F304 Cover Pipe Fittings SA403 Gr WP316L Seamless, SA213 TP304 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-7 (Page 2 of 2)
Closure Bolting & Nuts SA193 Gr B6, B7 or B8M and SA194 Gr2H or Gr 8M, SA193 Gr B6, B7 or B8M; SA453 Gr 660; and Nuts, SA194 Gr 2H, Gr 8M, Gr 6, and Gr 7 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-8 (Page 1 of 1)
Table 5-8. Reactor Vessels Internals for Emergency Core Cooling Forgings SA182 Type F304 Plates SA240 Type 304 Pipes SA312 Type 304 Seamless or SA376 Type 304 Tubes SA213 Type 304 Bars SA479 Type 304 & 410 Castings SA351 Gr CF8 or CF8A Bolting SA193 GrB8M (65 MYS/90MTS) Code Case 1613 Inconel 750 SA637 Gr688 Type 2 Nuts SA193 Gr B-8 Locking Devices SA479 Type 304 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-9 (Page 1 of 1)
Table 5-9. Deleted Per 1998 Update (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-10 (Page 1 of 1)
Table 5-10. Leakage Detection Sensitivity Readout Detection Device Parameter Monitored Location Leak Rate Sensitivity Response Time Containment Atmosphere Deleted Per 2007 Deleted Deleted Deleted Per 2007 Deleted Per 2007 Update Particulate Radioactivity Update Per 2007 Per 2007 Update Monitor Update Update Radioactivity Control 1 gpm The monitor Assume leakage activity containing accumulated on filters Room sensitivities are only current realistic coolant activity, from samples of given in Table 11- then leakage will be detected in 10 containment air. 20 hours or less during Mode 1.
Containment Floor and Water level in sump Control 1 gpm 1 gpm within one Leak detection within one hour.
Equipment Sump Level Room hour of leakage Indicator reaching the sump.
Ventilation Unit Water level in tank Control 1 gpm 1 gpm within one Leak detection within one hour.
Condensate Drain Tank Room hour of leakage Level Indicator reaching the tank.
Incore Instrumentation Water level in sump Control 1 gpm 1 gpm within four Leak detection within four hours.
Sump Room hours of leakage reaching the sump.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-11 (Page 1 of 2)
Table 5-11. Reactor Vessel Quality Assurance Program HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED RT1 UT1 PT1 MT1 Forgings
- 1. Flanges yes yes yes
- 2. Studs, Nuts yes yes
- 3. Head Adaptors yes yes
- 4. Head Adaptor Tube yes yes
- 5. Instrumentation Tube yes yes
- 6. Main Nozzles yes yes
- 7. Nozzle Safe Ends (Unit 2) yes yes
- 8. Shells (Unit 1) yes yes Plates yes yes Weldments
- 1. Main Seam yes yes yes
- 2. CRD Head Adaptor Connection yes
- 3. Instrumentation Tube Connection yes
- 4. Main Nozzle yes yes yes
- 5. Cladding yes yes
- 6. Nozzle Safe Ends (Forging - Unit 2) yes yes yes
- 7. Nozzle Safe Ends 8. 9. 10. 11.
(Weld deposit - Unit 1) yes yes yes
- 8. Head Adaptor Forging to Head 9. 10. 11. 12.
Adaptor Tube yes yes
- 9. All Full Penetration Ferritic Pressure Boundary Welds Accessible After Hydrotest yes yes
- 10. All Full Penetration Nonferritic Pressure Boundary Welds Accessible After Hydrotest yes
- 11. Seal Ledge yes
- 12. Head Lift Lugs yes (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-11 (Page 2 of 2)
RT1 UT1 PT1 MT1
- 13. Core Pad Welds yes Note:
Catawba Nuclear Station UFSAR Table 5-12 (Page 1 of 3)
Table 5-12. Initial (Unirradiated) Toughness Properties for the Catawba Unit 1 Reactor Vessel3 Tcv1 Shelf (50 Ft-Lb) Energy Cu P Ni TNDT (35 Mil) RTNDT NPWD1 Component Heat No. Mat'l Spec. No. (%) (%) (%) °F °F °F Ft-Lb Closure Head Dome 55888-1 A533B,CL.1 - .011 0.61 -4 70 10 86 Closure Head Ring 007055 A508,CL.2 - .006 0.86 16 23 16 101 Closure Head Flange 527038 .05 .013 0.83 -4 -2 -4 104 Vessel Flange 411212 - .004 0.86 -31 -47 -31 153 Inlet Nozzle 526827 .05 .010 0.75 -13 7 -13 87 526829 .07 .010 0.74 -4 43 -4 86 526859 .04 .013 0.77 -4 23 -4 81 526857 .05 .012 0.75 -13 23 -13 77 Outlet Nozzle 526827 .05 .011 0.75 -22 27 -22 84 526829 .07 .010 0.75 -4 2 -4 87 526859 .04 .011 0.77 -13 38 -13 81 526857 .05 .013 0.80 -4 38 -4 60 Nozzle Shell Forging 06 411077 - .007 0.85 -40 34 -26 101 6 6 Inter. Shell Forging 05 411343 .086 .004 0.858 -40 52 -8 1344 Lower Shell Forging 04 527708 .04 .008 0.83 -13 16 -13 1344 Bottom Head Ring 527428 .06 .013 0.77 -4 74 14 68 Bottom Head Segment 55292-1 A533B,CL.1 - .006 0.59 -22 5 -22 79
- .006 0.59 -13 34 -13 79 55163-2 - .011 0.61 -4 38 -4 80
- .011 0.61 -13 74 14 70 (09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-12 (Page 2 of 3)
Tcv1 Shelf (50 Ft-Lb) Energy Cu P Ni TNDT (35 Mil) RTNDT NPWD1 Component Heat No. Mat'l Spec. No. (%) (%) (%) °F °F °F Ft-Lb Dome 55178-1 - .010 0.64 -31 84 24 64 Nozzle Shell to Inter.
Shell Weld (P710) .03 - 0.75 - - 102 928 Inter. Shell to Lower Shell Weld Root (P710) .03 .009 0.757 -02 - 02 Lower Shell to Bot.
Head Ring Weld (P710) .03 - 0.75 - - 102 928 Inter. Shell to Lower Shell Weld (R747) .0396 .010 .7246 -76 -9 -51 1304 Weld Wire Flux Weld Control No. Type Heat No. Type Lot. No.
P710 NIMO 899680 Grau Lo P23 R7475 895075 Grau Lo P46 (09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-12 (Page 3 of 3)
Tcv1 Shelf (50 Ft-Lb) Energy Cu P Ni TNDT (35 Mil) RTNDT NPWD1 Component Heat No. Mat'l Spec. No. (%) (%) (%) °F °F °F Ft-Lb Notes:
- 1. Estimated per NRC Standard Review Plan Section 5.3.2 from data obtained in the principle working direction
- 2. Estimated per NRC Standard Review Plan Section 5.3.2 from charpy tests performed at 10°F
- 3. Source (except where noted otherwise): S. E. Yanichko, Catawba Unit No. 1 Reactor Material Toughness Properties, Westinghouse Internal Calc-Note dated June 29, 1978 (located in reactor vessel materials archives with MCTRs).
- 4. Source: WCAP-15609, Rev. 1, MOX Fuel Effects on Reactor Vessel Integrity at Catawba Units 1 and 2 and McGuire Units 1 and 2, dated March 2003, and Westinghouse Owners Group Calcnote 92-016, WOG USE Program - Onset of Upper Shelf Energy Calculations, J. M. Chicots, 1/19/93 [MUHP-5080].
- 5. Used for Surveillance Program Weldment.
- 6. ATI-94-012-T003, Rev. 2, A Review of Materials Data for the Catawba 1 Reactor Pressure Vessel, dated March 1999.
- 7. Source: Check analysis reported weld deposit analysis reported in De Rotterdame Drodgdak Mattschappu N.V. (The Rotterdam Dockyard Company or RDM) Welding Material Test Report.
- 8. Source: Catawba License Renewal Application (09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-13 (Page 1 of 2)
Table 5-13. Initial (Unirradiated) Toughness Properties for the Catawba Unit 2 Reactor Vessel3 Tcv1 Shelf Ft-Lb1 Energy
(
Mat'l Spec P Ni TNDT 35 RTNDT NPWD1 Component Heat No. No. Cu (%) (%) (%) °F Mil) °F °F Ft-Lb Closure Head Dome B8607-1 A533B,CL.1 .13 .007 0.64 - 40 50 - 10 106 Closure Head Torus B8608- 1 .07 .007 0.62 - 20 57 -3 118 Closure Head Flange B8601- 1 A508 CL.2 - .010 0.70 10 <10 10 152 Vessel Flange B8602- 1 - .010 0.71 10 <10 10 175 Inlet Nozzle B8609- 1 - .010 0.81 - 20 <10 - 20 119 B8609- 2 - .010 0.78 - 20 <10 - 20 124 B8609- 3 - .008 0.85 - 20 <40 - 20 109 B8609- 4 - .006 0.85 - 20 97 37 97 Outlet Nozzle B8610- 1 - .008 0.73 - 10 <50 - 10 141 B8610- 2 - .006 0.78 - 10 <50 - 10 144 B8610- 3 - .004 0.80 - 20 <40 - 20 140 B8610- 4 - .006 0.80 - 10 <50 - 10 150 Nozzle Shell B8604- 1 A533B,CL.1 .11 .007 0.61 - 10 84 24 96 B8604- 2 .11 .007 0.61 - 10 86 26 89 B8604- 3 .07 .009 0.53 - 20 110 50 70 Intermediate Shell B8605- 1 .0821 .011 0.6184 - 10 75 15 89 4 5 4 B8605- 2 .08 .012 0.613 - 20 93 33 82 B8616- 1 .0454 .010 0.5954 0 72 12 92 Lower Shell B8806- 1 .0574 .0115 0.564 - 60 66 6 83 B8806- 2 .0574 .0115 0.5934 - 40 50 - 10 102 (24 OCT 2004)
Catawba Nuclear Station UFSAR Table 5-13 (Page 2 of 2)
Tcv1 Shelf Ft-Lb1 Energy
(
Mat'l Spec P Ni TNDT 35 RTNDT NPWD1 Component Heat No. No. Cu (%) (%) (%) °F Mil) °F °F Ft-Lb B8806- 3 .0574 .0115 0.5934 - 40 68 8 105 Bottom Head Torus B8613- 1 .14 .010 0.48 - 40 52 -8 113 Bottom Head Dome B8612- 1 .14 .010 0.48 - 40 65 5 124 Nozzle Shell Vert. Weld Seams (G1.36) .1566 .0126 0.0596 - 50 <10 -50 >112 6 6 6 Nozzle Shell to Inter. Shell Weld Seam (G1.50) .153 .016 0.077 - 40 <20 -40 >102 Inter. & Lower Vert. Weld Seams (G1.45) .04 .005 .12 - 80 <-20 -80 >130 Inter. & Lower Shell Weld Seam (G1.45)
Weld Wire Flux Weld Control No. Type Heat No. Type Lot No.
G1.36 B- 4 51912 Linde 0091 3490 G1.50 B- 4 5P5622 1122 G1.452 B- 4 83648 3536 Notes:
- 1. Estimated per NRC Standard Review Plan Section 5.3.2.
- 2. Used for Surveillance Program Weldment.
- 3. Source (except where noted otherwise): WCAP-11941, Analysis of Capsule Z from the Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Program, S.E. Yanichko et al, table A-3, September 1988.
- 4. Source: ATI-94-012-T002, Rev. 2, A Review of Materials Data for the Catawba 2 Reactor Pressure Vessel, dated December 1999.
- 5. Source: Combustion Engineerings (C-Es)
- 6. Source: Combustion Engineering Report NPSD-1039, Rev. 2.
(24 OCT 2004)
Catawba Nuclear Station UFSAR Table 5-14 (Page 1 of 1)
Table 5-14. Comparison of Initial (Unirradiated) and Projected EOLE (54 EFPY) Fracture Toughness Properties Of The Catawba Unit 1 Reactor Vessel Beltline Region Material Initial Values End-of-Life Extension at 54 EFPY Inner Wall EOLE 2 USE TNDT RTNDT USE Fluence2 RTPTS2 Margin RTPTS2 USE2 Drop2 Component Code No. (°F) (°F) (ft-lb) 1019n/cm2 (°F) 2
(°F) (ft-lb)
(%)
Inter. Shell Forging 05 4113431 -40 -8 134 2.60 35.8 17 45 10 121 Lower Shell Forging 04 527708 -13 -13 134 2.60 32.7 32.7 52 21 106 Weld (W05) R7471 -76 -51 130 2.60 35.8 28 13 8 120 Notes:
- 1. Surveillance program materials.
- 2. Fluence, RTPTS and RTPTS, and USE values taken from Westinghouse Report WCAP-17669-NP, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations, "dated June 2013. Also see UFSAR Tables 5-42 and 5-44 for additional information.
(09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-15 (Page 1 of 1)
Table 5-15. Comparison of Initial (Unirradiated) and Projected EOLE (54 EFPY) Toughness Properties for the Catawba Unit 2 Reactor Vessel Initial Values End-of-Life Extension at 54 EFPY Peak Inner Wall USE EOLE Code TNDT RTNDT USE3 Fluence2 RTPTS2 RTPTS2 (Drop3 USE3 Component No. (°F) (°F) (ft-lb) 1019n/cm2 (°F) Margin (°F) (%)
(ft-lb)
Inter. Shell B8605-11 -10 15 89 3.16 57.2 17 89 6.6 90 Inter. Shell B8605-2 -20 33 82 3.16 66.3 34 1334 22 64 Inter. Shell B8616-1 0 12 92 3.16 40.3 34 86 22 72 Lower Shell B8806-1 -60 6 83 3.16 48.1 34 88 22 65 Lower Shell B8806-2 -40 -10 102 3.16 48.1 34 72 22 80 Lower Shell B8806-3 -40 8 105 3.16 48.1 34 90 22 82 Inter. to Lower Shell Circ.
Weld Seam (101-171);
Axial Weld Seams (101-142A, 101-124B), G1.451 -80 -80 146 3.16 43.4 28 -9 11 130 and (101-142B/C, 101-124A/C)
(21 OCT 2010)
Catawba Nuclear Station UFSAR Table 5-16 (Page 1 of 2)
Table 5-16. Catawba Unit 1 Closure Head Bolting Material Properties Closure Head Studs Energy Lateral Mat'l Spec. Bar 0.2 YS UNITS ELONG RA At 10°F Expansion Heat No. No. No.1 KSi KSi % % BHN Ft-Lbs Mils 35674 A540,B24 25K 142.5 160.4 18.8 57 341 42,40,40 12,16,28 35674 A540,B24 25T 136.9 157.1 18.0 57 331 38,38,39 12,8,12 35674 A540,B24 26K 143.9 161.8 19.0 57 331 46,46,46 32,28,28 35674 A540,B24 26T 138.2 162.7 19.4 56 341 44,44,43 32,24,28 35674 A540,B24 27K 141.5 161.8 18.2 56 331 40,39,39 12,28,12 35674 A540,B24 27T 141.5 160.5 18.4 56 331 44,43,44 16,24,24 35666 A540,B24 28K 143.6 162.8 18.4 52 321 48,48,46.5 28,24,24 35666 A540,B24 28T 145.0 164.9 18.0 52 331 45.5,46.5,49.5 28,20,24 35666 A540,B24 29K 145.7 163.8 18.2 55 341 42,40.5,40.5 24,16,20 35666 A540,B24 29T 145.7 163.8 18.4 55 331 38,38,40.5 20,16,24 35847 A540,B24 297K 143.6 160.4 18.0 59 321 49,52,52.5 32,24,24 35847 A540,B24 297T 145.7 161.8 18.0 59 341 49.5,53.5,53.5 32,32,35 Closure Head Nuts (Original supply. See Section 5.3.1.7.1 for discussion of alternative nuts) 36627 A540,B24 328K 133.5 153.7 20.0 61 331 60,62,58 35,45,47 36627 A540,B24 328T 135.9 154.7 20.0 60 331 60.5,67,62 39,47,43 Closure Head Washers (Original supply. See Section 5.3.1.7.1 for discussion of alternative washers) 36512 A540,B24 306K 131.2 153.7 19.6 60 331 48.5,49.5,49 35,35,32 36512 A540,B24 306T 132.5 153.7 20.2 60 321 52,52.5,50.5 43,43,43 (17 OCT 2013)
Catawba Nuclear Station UFSAR Table 5-16 (Page 2 of 2)
Closure Head Studs Energy Lateral Mat'l Spec. Bar 0.2 YS UNITS ELONG RA At 10°F Expansion Heat No. No. No.1 KSi KSi % % BHN Ft-Lbs Mils Note:
- 1. K & T denote top and bottom of bar respectively.
(17 OCT 2013)
Catawba Nuclear Station UFSAR Table 5-17 (Page 1 of 2)
Table 5-17. Catawba Unit 2 Closure Head Bolting Material Properties Closure Head Studs Energy Lateral Mat'l Bar 0.2%YS ELONG RA At 10°F Expansion Heat No. Spec. No. No.1 KSi UTS KSi % % BHN Ft-Lbs Mils 81874 A540,B24 144 149.5 163.0 16 51.4 331 47,47,47 27,27,26 144-1 146.5 160.0 17 54.7 341 50,49,49 31,28,29 146 153.5 167.0 16 52.7 341 44,44,45 25,26,26 146-1 145.0 160.0 17 54.4 331 51,49,49 30,28,28 149 146.5 160.0 17 54.8 352 50,49,49 27,29,26 149-1 147.0 161.0 17 52.7 341 49,49,49 28,26,27 153 152.0 164.0 16 53.8 341 53,52,52 31,30,32 153-1 140.5 155.0 18 54.4 331 46,47,47 26,26,26 157 142.5 158.0 17.5 54.4 341 51,51,51 30,28,31 157-1 148.0 161.5 17 54.0 341 47,48,47 27,30,27 163 149.8 163.0 16.5 55.1 331 49,49,49 28,26,29 163-1 145.0 159.0 16.5 54.3 341 51,51,53 30,31,33 82552 197 142.5 156.0 18 53.8 341 51,51,52 28,29,30 197-1 141.5 155.0 17 53.8 341 50,51,50 28,31,28 201 146.0 158.5 15.5 51.7 341 48,48,49 27,27,27 201-1 145.5 159.5 15.5 50.6 341 47,49,47 27,30,27 207 138.0 153.0 17.0 52.5 341 51,52,51 31,32,31 207-1 138.5 153.0 16.5 51.4 341 51,50,49 28,28,27 212 141.0 155.0 17.0 53.0 341 52,52,51 32,31,30 (17 OCT 2013)
Catawba Nuclear Station UFSAR Table 5-17 (Page 2 of 2)
Closure Head Studs Energy Lateral Mat'l Bar 0.2%YS ELONG RA At 10°F Expansion Heat No. Spec. No. No.1 KSi UTS KSi % % BHN Ft-Lbs Mils 212-1 144.0 157.0 16.0 49.8 352 51,50,49 31,29,29 Closure Head Nuts and Washers (Original supply. See Section 5.3.1.7.1 for discussion of alternative nuts) 19632 A540,B23 69 141.5 155.0 17 56.4 321 55,55,55 33,30,29 69-1 145.5 158.0 17 55.0 321 49,47,47 30,28,25 73 145.5 158.0 16 51.9 - 55,56,55 34,33,31 73-1 142.5 156.0 16.5 54.5 - 48,47,45 28,26,25 (17 OCT 2013)
Catawba Nuclear Station UFSAR Table 5-18 (Page 1 of 1)
Table 5-18. Reactor Vessel Design Parameters Design Pressure, psig 2485 Design Temperature, °F 650 Overall Height of Vessel and Closure Head, ft (Bottom Head 43.833 Outside Diameter to top of Control Rod Mechanism Adaptor)
Thickness of Insulation, minimum, in 3 Number of Reactor Closure Head Studs 54 Diameter of Reactor Closure Head/Studs, in (minimum shank) 6.75 (Unit 1) 6.8125 (Unit 2)
Inside Diameter of Flange, in l67 Outside Diameter of Flange, in 205 Inside Diameter at Shell, in l73 Inlet Nozzle Inside Diameter, in 27.5 Outlet Nozzle Inside Diameter, in 29 Clad Thickness, minimum, in 0.125 Lower Head Thickness, minimum, in 5.236 (Unit 1) 5.375 (Unit 2)
Vessel Belt-Line Thickness, minimum, in 8.464 (Unit 1) 8.625 (Unit 2)
Closure Head Thickness, in 6.496 (Unit 1) 7.0 (Unit 2)
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-19 (Page 1 of 1)
Table 5-19. Chemical Composition Of The Catawba Unit 1 Reactor Vessel Beltline Region Material3 Inter. Shell Lower Shell Weld Weld Forging Forging Control No. Control No.
Element 411343 527708 R7471 P7102 C .21 .21 .069 .052 Mn .76 .72 1.97 1.97 P .004 .008 .010 .009 S .006 .007 .010 .015 Si .28 .33 .22 .25 Ni .8584 .83 .7244 .75 Cr .38 .33 .05 .04 Mo .60 .55 .56 .46 Cu .0864 .04 .0394 .03 V <.01 <.01 - .01 Co .013 .01 - -
Al .04 .01 - .014 Notes:
- 1. Submerged arc weld - NiMo wire (Heat No. 895075) and Grau Lo Flux (Lot No. P46) used to fabricate beltline region girth seam WO5 and surveillance weldment.
- 2. Submerged arc weld - NiMo wire (Heat No. 899650) and Grau Lo Flux (Lot No. P23) used to fabricate root of beltline region girth seam WO5.
- 3. Unless noted otherwise, the source of all composition data is the check analysis reported on DeRotterdame Drodgdak Mattschappu N.V. (The Rotterdam Dockyard Company or RDM) MCTRs or weld deposit analysis reported in RDMs Welding Material Test Reports.
- 4. ATI-94-012-T003, Rev. 2, A Review of Materials Data for the Catawba 1 Reactor Pressure Vessel, dated March 1999.
(09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-20 (Page 1 of 1)
Table 5-20. Chemical Composition of the Catawba Unit 2 Reactor Vessel Beltline Region Material Intermediate Shell Plate Lower Shell Plate Weld Control No.
Code. No. (Heat No.) Code No. (Heat No.) (Heat No.)
B8605-1 B8605-2 B8616-1 B8806-1 B8806-2 B8806-3 G1.452 Element (C-0543-1) C-0543-2) (A-0617-1) (C-2288-1) (C-2272-1) (C-2272-2) (83648)
C .25 .25 .24 .24 .22 .22 .13 Mn 1.40 1.40 1.39 1.44 1.36 1.36 1.23 P .011 .012 .010 .011 .011 .011 .005 S .013 .012 .021 .014 .016 .016 .009 Si .28 .28 .27 .23 .25 .25 .13 Ni4 .618 .613 .595 .56 .593 .593 .136 Cr - - - - - - -
Mo .57 .56 .54 .55 .54 .54 .59 4
Cu .082 .08 .045 .057 .057 .057 .042 V - - ND1 ND1 ND1 ND1 .006 Notes:
- 1. Not detected.
- 2. Submerged arc weld -. type B4 wire (Heat No. 83648) and Linde 0091 flux (Lot No. 3536) used to fabricate all beltline region weld seams including the surveillance weldment: chemistry data taken rom WCAP-11941, Analysis of Capsule Z from the Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Program, dated September 1988
- 3. Unless noted otherwise, the source of all composition data is the check analysis reported on Combustion Engineerings (C-Es) MCTRs or weld deposit analysis reported in C-Es Welding Material Test Reports.
- 4. All Cu and Ni values were taken from ATI-94-012-T002, Rev. 2, A Review of Materials Data for the Catawba 2 Reactor Pressure Vessel, dated December 1999.
(24 OCT 2004)
Catawba Nuclear Station UFSAR Table 5-21 (Page 1 of 4)
Table 5-21. Catawba Unit 1 Reactor Vessel Beltline Region Toughness Properties Inter. To Lower Shell Weld Weld Metal For Pressure Inter. To Lower Shell Weld Weld Code No. R747 Vessel Core Region Root Weld Core No. P710 Lat. Lat.
Temp. Energy Exp. Shear Temp. Energy Lat. Exp. Shear Temp Energy Exp. Shear
(°F) (ft/lb) (Mils) (%) (°F) (ft/lb) (Mils) (%) (°F) (ft/lb) (Mils) (%)
-148 3.5 4 0 -100 12 6 13 10 57.5 43 47
-112 8.5 12 16 -60 13 9.5 28 10 43.0 39 47
-76 21.5 23 33 -60 15 11.5 33 10 39.0 55 55
-40 45.0 39 43 -40 37 26 33
-22 57.5 47 55 -40 26 18 28
-4 54.5 43 47 -16 40 31 42 32 92.5 71 76 -16 60 45.5 37 68 104.5 81 92 -16 54 39 50 86 113.0 91 89 5 44 38 54 122 123.5 83 93 25 91 66 72 140 144.0 99 100 25 98.5 74 87 158 129.0 94 98 75 119 86 87 176 130.0 87 98 75 110 77 95 212 126.5 87 100 120 132 90.5 100 120 119 87 100 210 133 90 100 210 130 90 100 210 124 89 100 TNDT = -76°F RTNDT = -51°F (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-21 (Page 2 of 4)
Lower Shell Forging (04) Lower Shell Forging (04)
Heat No. 527708 (Tang) Heat No. 527708 (Axial)
Temp. Energy Lat. Exp Shear Temp. Energy Lat. Exp. Shear
(°F) (ft/lb) (Mils) (%) (°F) (ft/lb) (Mils) (%)
-148 2.5 2 0 -148 5 4 0
-148 3.5 4 0 -148 3 4 0
-148 4.5 4 0 -148 3 4 0
-76 11.5 12 4 -76 10 8 13
-76 11.5 8 3 -76 31 16 10
-76 11.5 8 3 -76 20 24 11
-4 57.5 47 29 -4 67 59 23
-4 57.5 51 23 -4 82 67 35
-4 53.5 47 23 -4 69 59 29 40 101.5 79 67 60 94 71 60 40 126.5 94 80 60 104 75 60 40 109.5 83 68 60 98 79 61 40 149.0 94 80 104 130 91 66 40 156.5 91 75 104 123 83 60 40 113.0 83 65 104 112 83 72 60 125.5 83 70 176 133 91 90 60 117.0 71 62 176 135 94 89 60 106.5 71 55 176 135 87 85 113 149.5 94 100 113 135.5 91 85 113 149.0 94 100 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-21 (Page 3 of 4)
Lower Shell Forging (04) Lower Shell Forging (04)
Heat No. 527708 (Tang) Heat No. 527708 (Axial)
Temp. Energy Lat. Exp Shear Temp. Energy Lat. Exp. Shear
(°F) (ft/lb) (Mils) (%) (°F) (ft/lb) (Mils) (%)
176 155.0 99 98 176 158.5 91 100 176 151.0 91 100 TNDT = -13°F RTNDT = -13°F Inter. Shell forging (05) Inter. Shell forging (05) Inter. Shell Equip. (05)
Heat No. 411343 Heat No. 411343 (Axial) Heat No. 411343 (Tang)
Temp. Energy Lat. Exp. Shear Temp. Energy Lat. Exp. Shear Temp. (°F) Energy Lat. Exp. Shear
(°F) (ft/lb) (Mils) (%) (°F) (ft/lb) (Mils) (%) (ft/lb) (Mils) (%)
-148 5 4 0 -100 2 0 0 -148 2.0 6 0
-148 5 4 0 -40 8 5.5 0 -148 3.0 2 0
-148 5 4 0 -40 16 11 0 -148 4.5 4 0
-76 10 4 8 -15 48 34 25 -76 8.5 8 3
-76 17 16 10 0 27 19 20 -76 3.5 4 0
-76 8 4 4 0 35 25 30 -76 8.5 4 0
-4 43 59 29 0 51 38 30 -4 45.5 39 27
-4 58 28 11 20 60 44 34 -4 16.5 20 16
-4 54 39 16 20 56 44 40 -4 42.0 35 17 60 86 67 50 20 69 52 45 40 101.5 79 52 60 83 63 44 75 81 60 52 40 102.5 79 49 60 80 63 49 75 89 67 61 40 100.0 71 49 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-21 (Page 4 of 4)
Inter. Shell forging (05) Inter. Shell forging (05) Inter. Shell Equip. (05)
Heat No. 411343 Heat No. 411343 (Axial) Heat No. 411343 (Tang)
Temp. Energy Lat. Exp. Shear Temp. Energy Lat. Exp. Shear Temp. (°F) Energy Lat. Exp. Shear
(°F) (ft/lb) (Mils) (%) (°F) (ft/lb) (Mils) (%) (ft/lb) (Mils) (%)
104 115 83 76 120 111 77 82 40 87.0 71 47 104 117 75 80 120 119 79.5 81 40 86.5 71 38 104 112 83 74 150 135 90 100 40 57.5 51 38 176 138 99 80 210 137 88 100 60 118.0 75 65 176 137 91 80 210 132 91 100 60 122.5 79 57 176 136 91 85 210 133 87 100 60 92.0 63 52 113 140.0 87 85 113 129.5 91 76 113 155.0 99 90 176 152.5 87 100 176 156.5 87 100 176 152.5 94 100 TNDT = -40°F RTNDT = -8°F (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-22 (Page 1 of 5)
Table 5-22. Catawba Unit 2 Reactor Vessel Beltline Region Toughness Properties Intermediate Shell Course (Transverse Data)
Plate B8605 - 1 Plate B8605 - 2 Plate B8616-1 Lat. Lat.
Temp. Lat. Exp. Shear Temp. Energy Exp. Shear Temp Energy Exp. Shear
(°F) Energy (ft/lb) Mils % °F (ft/lb) Mils (%) (°F) (ft/lb) (Mils) (%)
- 80 5 2 3 - 40 5 4 0 - 40 7 6 0
- 80 6 5 3 - 40 4 4 0 - 40 11 8 0
- 40 9 6 13 - 40 6 5 0 - 40 10 7 0
- 40 16 10 18 10 15 12 5 10 26 18 20
- 40 18 13 9 10 22 18 10 10 30 19 20 0 34 21 29 10 14 12 5 10 29 21 20 0 35 27 25 40 25 22 15 40 35 30 20 0 41 35 29 40 37 27 20 40 42 32 25 40 35 27 38 40 37 26 20 40 50 36 30 40 52 41 43 75 42 31 25 74 53 41 40 40 54 40 34 75 50 37 30 74 52 39 40 80 59 47 41 75 37 30 25 74 56 43 40 80 64 48 44 100 70 54 50 100 65 49 60 80 71 50 45 100 58 42 40 100 64 47 60 100 53 43 47 100 63 46 40 100 69 51 60 100 66 50 55 160 78 60 95 160 89 68 100 100 77 54 59 160 83 63 95 160 93 69 100 120 66 47 62 160 86 64 100 160 95 71 100 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-22 (Page 2 of 5)
Intermediate Shell Course (Transverse Data)
Plate B8605 - 1 Plate B8605 - 2 Plate B8616-1 Lat. Lat.
Temp. Lat. Exp. Shear Temp. Energy Exp. Shear Temp Energy Exp. Shear
(°F) Energy (ft/lb) Mils % °F (ft/lb) Mils (%) (°F) (ft/lb) (Mils) (%)
120 84 61 68 120 94 70 100 180 92 64 100 180 96 72 100 180 99 71 100 240 100 71 100 240 102 78 100 320 84 65 100 320 100 75 100 TNDT = -10°F TNDT = -20°F TNDT = 0°F RTNDT = 15°F RTNDT = 33°F RTNDT = 12°F Lower Shell Course (Transverse Data)
Plate B8806 - 1 Plate B8806 - 2 Plate B8806 - 3 Temp. Energy Lat. Shear Temp. Energy Lat. Shear Temp Energy Lat. Shear
(°F) (ft/lb) Exp. % °F (ft/lb) Exp. (%) . (ft/lb) Exp. (%)
Mils Mils (°F) (Mils)
- 40 16 11 0 - 40 16 12 0 - 40 14 9 0
- 40 15 9 0 - 40 16 12 0 - 40 9 6 0
- 40 13 10 0 - 40 15 10 0 - 40 18 13 5 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-22 (Page 3 of 5)
Intermediate Shell Course (Transverse Data)
Plate B8806 - 1 Plate B8806 - 2 Plate B8806-1 Energ Lat. Lat.
Temp. y Exp. Shear Temp. Energy Exp. Shear Temp Energy Lat. Exp.
(°F) (ft/lb) Mils % °F (ft/lb) Mils (%) (°F) (ft/lb) (Mils) Shear (%)
10 36 26 15 10 38 25 15 10 35 23 15 10 40 28 20 10 35 23 15 10 26 19 10 10 38 28 20 10 34 21 15 10 44 33 20 40 48 37 30 40 53 34 25 40 39 24 15 40 49 36 30 40 61 42 40 40 43 31 20 40 46 34 30 40 49 33 20 40 44 31 20 74 60 45 50 74 61 40 40 74 53 39 30 74 53 41 50 74 59 40 40 74 58 43 35 74 70 52 70 74 64 44 40 74 72 50 60 100 77 62 90 100 86 61 70 100 75 50 70 100 72 59 80 100 75 52 60 100 81 52 70 100 84 64 100 100 69 46 60 100 88 59 80 160 82 61 100 160 108 74 100 160 102 69 100 160 85 65 100 160 98 69 100 160 103 71 100 160 82 64 100 160 100 72 100 160 110 73 100 TNDT = - 60°.F TNDT = -40°F TNDT = - 40°F RTNDT = - 6°F RTNDT = -10°F RTNDT = 8°F (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-22 (Page 4 of 5)
Intermediate And Lower Shell Vertical Weld Seams And Girth Weld Seam Weld Code No G1.45 Weld Metal for Pressure Vessel Core Region Temp Energy Lat. Exp. Temp Energy Lat. Exp. Shear
(°F) (ft/lb) (mils) (°F) (ft-lb) (mils) (%)
- 20 108 74 -140 4 1 9
- 20 84 56 -140 5 2 13
- 20 79 54 -80 6 1 13 10 130 84 -80 8 3 18 10 129 76 -80 26 11 18 10 132 85 -60 11 6 28
-60 15 10 33 TNDT = -80°F -60 15 6 28 RTNDT = - 80°F -40 46 31 47
-40 58 41 40
-40 73 51 52 0 58 45 65 0 96 60 73 0 101 72 71 40 121 79 96 40 125 78 93 40 135 80 84 80 131 79 93 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-22 (Page 5 of 5)
Intermediate And Lower Shell Vertical Weld Seams And Girth Weld Seam Weld Code No G1.45 Weld Metal for Pressure Vessel Core Region Temp Energy Lat. Exp. Temp Energy Lat. Exp. Shear
(°F) (ft/lb) (mils) (°F) (ft-lb) (mils) (%)
80 138 88 96 80 147 86 94 120 142 88 100 120 146 88 100 120 151 87 100 220 139 86 100 220 148 91 100 320 152 90 100 320 164 86 100 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-23 (Page 1 of 2)
Table 5-23. Reactor Coolant Pump Design Parameters Unit Design Pressure, psig 2485 Unit Design Temperature, °F 6501 Unit Overall Height, ft 27.6 Seal Water Injection, gpm 8 Seal Water Return, gpm 3 Cooling Water Flow, gpm 436 Maximum Continuous Cooling Water Inlet Temperature, °F 105 Pump Capacity, gpm 101,000 +/- 2000 Developed Head, ft 289 +/- 12 NPSH Required, ft Figure 5-13 Suction Temperature, °F 557.8 Pump Discharge Nozzle, Inside Diameter, in 27-1/2 Pump Suction Nozzle, Inside Diameter, in 31 Speed, rpm 1185 Water Volume, ft3 78.6 2 Total Pump/Motor Weight (dry), lbs 201, 200 Motor Type Drip proof, squirrel cage induction, water/air cooled Power, Hp 7000 Voltage, volts 6600 Phase 3 Frequency, Hz 60 Insulation Class Class F Starting Current, Amps 3000 Amp @ 6600 volts Input, Hot Reactor Coolant 492 +/- 17 amps Input, Cold Reactor Coolant 654 +/- 23 amps Pump Moment of Inertia, 1b-ft2 minimum 95,000 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-23 (Page 2 of 2)
Notes:
- 1. Design temperature of pressure retaining parts of the pump assembly exposed to the reactor coolant and injection water on the high pressure side of the controlled leakage seal shall be that temperature determined for the parts for a primary loop temperature of 650°F.
- 2. Composed of reactor coolant in the casing and of injection and cooling water in the thermal barrier.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-24 (Page 1 of 1)
Table 5-24. Reactor Coolant Pump Quality Assurance Program HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED RT1 UT1 PT1 MT1 Castings yes yes Forgings
- 1. Main Shaft yes yes
- 2. Main Studs yes yes
- 3. Flywheel (Rolled Plate) yes yes yes Weldments
- 1. Circumferential yes yes
- 2. Instrument Connections yes Note:
Catawba Nuclear Station UFSAR Table 5-25 (Page 1 of 1)
Table 5-25. Steam Generator Design Data Unit 1 Unit 2 Design Pressure, reactor coolant side, psig 2485 2485 Design Pressure, steam side, psig 1185 1185 Design Temperature, reactor coolant side, °F 650 650 Design Temperature, steam side, °F 600 600 Total Heat Transfer Surface Area, ft2 79,800 48,165 Maximum Moisture Carryover, wt percent 0.25 0.25 Overall Height, ft-in 67-8 67-8 Number of U-Tubes 6633 4570 U-Tube Nominal Diameter, in. .6875 .750 Tube Wall Nominal Thickness, in. .040 .043 Number of Manways 3 4 Inside Diameter of Manways, in. 21 16 Number of Inspection Ports 2.0" Dia. 12 4(2) 2.5" Dia. 0 2 2.7" Dia. N/A 1 on S/G 2C only 6.0" Dia. 10 5 Design Fouling Factor(Btu/hr °F ft2)-1 0.00002(1) 0.00005 Preheat Section NA 0.00010 Notes:
- 1. There is no specified design fouling factor for McGuire or Catawba Unit 1. Fouling factor of 0.00002 is used for start-up, and various fouling factors are used thereafter for performance analysis under variable fouling.
- 2. 6 for S/G 2A (15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-26 (Page 1 of 2)
HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED Table 5-26. Steam Generator Quality Assurance Program RT1 UT1 PT1 MT1 ET2 Tubesheet
- 1. Forging yes yes
- 2. Cladding Unit 1 yes yes3 Unit 2 yes2 yes Channel Head
- 1. Casting (Unit 2) yes yes
- 2. Forging (Unit 1) yes
- 3. Cladding yes Secondary Shell & Head
- 1. Plates yes Tubes yes yes Nozzles (Forgings) yes yes Weldments
- 1. Shell, longitudinal Unit 1 yes yes yes Unit 2 yes yes
- 2. Shell, circumferential Unit 1 yes yes yes Unit 2 yes yes
- 3. Cladding (channel head- tubesheet yes joint cladding restoration)
- 4. Steam and feedwater nozzle to shell Unit 1 yes yes yes Unit 2 yes yes
- 5. Support brackets yes
- 6. Tube to tubesheet yes
- 7. Instrument connections (primary and secondary)
Unit 1 yes yes Unit 2 yes (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-26 (Page 2 of 2)
RT1 UT1 PT1 MT1 ET2
- 8. Temporary attachments after yes removal
- 9. After hydrostatic test (all welds and yes complete cast channel head - where accessible)
- 10. Nozzle safe ends (if forgings) yes yes
- 11. Nozzle safe ends (if weld deposit)
Unit 1 yes Unit 2 yes Notes:
- 2. Flat Surfaces Only
- 3. Weld Deposit Areas Only (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-27 (Page 1 of 1)
Table 5-27. Reactor Coolant Piping Design Parameters Reactor Inlet Piping, inside diameter, in 27-1/2 Reactor Inlet Piping, nominal wall thickness, in 2.32 Reactor Outlet Piping, inside diameter, in 29 Reactor Outlet Piping, nominal wall thickness, in 2.45 Coolant Pump Suction Piping, inside diameter, in 31 Coolant Pump Suction Piping, nominal wall thickness, in 2.60 Pressurizer Surge Line Piping, nominal pipe size, in 14 Pressurizer Surge Line Piping, nominal wall thickness, in 1.405 Reactor Coolant Loop Piping Design Pressure, psig 2485 Design Temperature, °F 650 Pressurizer Surge Line Design Pressure, psig 2485 Design Temperature, °F 680 Pressurizer Safety Valve Inlet Line Design Pressure, psig 2485 Design Temperature, °F 680 Pressurizer (Power-Operated) Relief Valve Inlet Line Design Pressure, psig 2485 Design Temperature, °F 680 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-28 (Page 1 of 1)
HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED Table 5-28. Reactor Coolant Piping Quality Assurance Program RT1 UT1 PT1 Fittings and Pipe (Castings) yes yes Fittings and Pipe (Forging) yes yes Weldments
- 1. Circumferential yes yes
- 2. Nozzle to runpipe (Except no RT for nozzles less than 6 inches) yes yes
- 3. Instrument connections yes Castings yes yes (after finishing)
Forgings yes yes (after finishing)
Note:
Catawba Nuclear Station UFSAR Table 5-29 (Page 1 of 1)
Table 5-29. Design Bases for Residual Heat Removal System Operation Residual Heat Removal System Start Up. ~4 hours after Reactor Shutdown Reactor Coolant System Initial Pressure, psig ~385 Reactor Coolant System Initial Temperature, °F 350 Component Cooling Water Design Temperature, °F 105 Cooldown Time, Hours After Initiation Of ~16 Residual Heat Removal System Operation Reactor Coolant System Temperature At End Of 200 Cooldown, °F Decay Heat Generation At 20 Hours After Reactor 78.2 x 106 Shutdown, BTU/hr (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-30 (Page 1 of 1)
Table 5-30. Residual Heat Removal System Component Data Residual Heat Removal Pump Number 2 Design Pressure, psig 600 Design Temperature, °F 400 Design Flow, gpm 3000 Design Head, ft 375 Maximum Calculated Runout Flow (ECCS), gpm 3800 NPSH Required at test 3980 gpm, ft 16 NPSH Available at 3980 gpm, ft. NPSH available does 33 not include the losses associated with the ECCS Sump Strainer. NPSH available will increase throughout the event as the containment sump pool temperature decreases.
Assumed "Runout Flow" per SER Supplement 2 5300 Resolution of Confirmatory Issue 22, gpm NPSH Required at 5300 gpm, ft. 22.75 NPSH Available at 5300 gpm, ft. 24.0 (Second case which includes 2 ft. of water above floor level inside containment as reviewed and approved by NRC in SER Supplement 2).
Power, HP 400 Residual Heat Exchanger Number 2 Design Heat Removal Capacity, BTU/hr 30.96 x 106 Estimated UA, BTU/hr °F 2.07 x 106 Tube-Side Shell-Side Design Pressure, psig 600 150 Design Temperature, °F 400 200 Design Flow, 1b/hr 1.48 x 106 2.48 x 106 Inlet Temperature, °F 140 105 Outlet Temperature, °F 119.0 117.5 Material Austenitic Stainless Steel Carbon Steel Fluid Reactor Coolant Component Cooling Water (18 APR 2009)
Catawba Nuclear Station UFSAR Table 5-31 (Page 1 of 10)
Table 5-31. Failure Mode and Effects Analysis-Residual Heat Removal System Active Components-Plant Cooldown Operation Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 1. Motor operated gate a. Fails to open on a. Failure blocks reactor a. Valve position 1. Valve is electrically valve ND2A demand (open coolant, flow from hot indication (closed to interlocked with the (ND37A analogous) manual mode CB leg of RC loop B open position change) containment sump switch selection). through train A of at CB; RC loop B hot isolation valve RHRS. Fault reduces leg pressure (NI185A), the RWST the redundancy of indication (NCP5120) isolation valve RHR coolant train at CB; RHR train A (FW27A), the RHR provided. No effect discharge flow to charging pump on safety for system indication suction line isolation operation. Plant (NDP5190) and low valve (ND28A), the cooldown flow alarm at CB; and residual spray valve requirements will be RHR pump discharge (NS43A) and with a met by reactor coolant pressure indication prevent open pressure flow from hot leg of (NDP5090) at CB interlock (PB-405A)
RC loop C flowing of RC loop B hot leg.
through train B of The valve can not be RHRS, however, time opened remotely from required to reduce the CB if any of the RCS temperature will indicated isolation be extended. valves is open or if RC loop pressure exceeds 385.5 psig.
- 2. Motor operated gate a. Same failure modes as a. Same effect on system a. Same methods of 1. Same remarks as valve. ND1B those stated for item operation as that detection as those those stated for item (ND36B analogous). #1. stated for item #1. stated for item #1 #1 except for pressure interlock (PB-403A) control.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 2 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 3. Residual heat a. Fails to deliver a. Failure results in loss a. Open pump 1. The RHRS shares removal pump A, working fluid. of reactor coolant switchgear circuit components with the (pump B analogous) flow from hot leg of indication at CB; ECCS. Pumps are RCS loop B through circuit breaker tested as part of the the train A of RHRS. position monitor light ECCS testing Fault reduces the for group monitoring program (see Section redundancy of RHR of components at CB; 6.3.4) Pumps failure coolant trains common breaker may also be detected provided. No effect trouble alarm at CB; during ECCS testing.
on safety for system RCS loop B hot leg operation. Plant pressure indication cooldown (NCP5120) at CB; requirements will be RHR train A met by reactor coolant discharge flow flow from hot leg of indication at CB RC loop C flowing (NDP5190); and through train B of pump discharge RHRS. However, pressure indication time required to (NDP5090) at CB.
reduce RCS temperature will be extended (15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 3 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 4. Motor operated globe a. Fails to open on a. Failure blocks a. Valve position 1. Valve is valve ND25A demand (open manual miniflow line to indication (closed to automatically (ND59B analogous) mode CB switch suction of RHR pump open position change) controlled to open selection). A during cooldown at CB. when pump discharge operation or checking is less than 533 gpm boron concentration and to close when the level of coolant in discharge exceeds train A of RHRS. No 1400 gpm. The valve effect on safety for protects the pump system operation. from dead heading Operator may during ECCS establish miniflow for operation. No auto RHR pump A position for this operation by opening valve. While the CVCS letdown pump is ON, the control valve valve may be (NV135) to allow manually positioned flow to CVCS. whenever the flowrate is between 533 gpm and 1400 gpm. While the pump is OFF, the valve automatically closes.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 4 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- b. Fails to close on b. Failure allows a b. Valve position demand portion of RHR heat indication (open to exchanger A closed position discharge flow to be change) and RHRS bypassed to suction of train A discharge flow RHR pump A. RHRS indication (NDP5190) train A is degraded. at CB.
No effect on safety for system operation.
Cooldown of RCS within the specified cooldown rate may be accomplished through operator action of throttling flow control valve ND26 and with redundant RHRS train B.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 5 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 5. Air diaphragm a. Fails to open on a. Failure prevents a. Valve position 1. Valve is designed to operated butterfly demand (auto mode coolant discharged indication closed to fail closed and is valve ND27 (ND61 CB switch selection from RHR pump A open position change) electrically wired so analogous) on manual/auto from bypassing RHR at CB; pump A that solenoid of the station) heat exchanger A discharge flow air diaphragm resulting in mixed temperature recording operator is energized mean temperature of (NDCR5060) at CB; to open the valve.
coolant flow to RCS and RHRS train A Solenoid valve being low. RHRS discharge to RCS cold receives S signal to train A is degraded leg flow indication close valve for ECCS for the regulation of (NDP5190) at CB. operation. Valve is controlling normally closed temperature of during power coolant. No effect on operations.
safety for system operation. Cooldown of RCS within established specification rate may be accomplished through operator action of throttling flow control valve ND26 and controlling cooldown with redundant RHRS train B.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 6 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- b. Fails to close on b. Failure allows coolant b. Same methods of demand (auto mode discharged from RHR detection as those CB switch selection pump A to bypass stated above except on manual/auto RHR heat exchanger open to closed valve station). A resulting in mixed position change mean temperature of indication at CB.
coolant flow to RCS being high. RHRS train A is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS with in established specification rate may be accomplished through operator action of throttling flow control valve ND26 and controlling cooldown with redundant RHRS train B, however, cooldown time will be extended.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 7 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 6. Air diaphragm a. Fails to close on a. Failure prevents a. Same methods of 1. Valve is designed to operated butterfly demand for flow control of coolant detection as those fail open and is valve ND26 (ND60 reduction through heat discharge flow from stated for item #5, electrically wired so analogous) exchanger. RHR heat exchanger except degree of that the solenoid of A resulting in loss in valve being open the air diaphragm being able to adjust position indication at operator is energized mixed mean CB. to close the valve.
temperature of The solenoid receives coolant flow to RCS. S signal to open the No effect on safety valve. The valve is for system operation. normally open Cooldown of RCS during power within established operations.
specification rate may be accomplished by operator action of controlling cooldown with redundant RHRS train B.
- b. Fails to open on b. Same effect on system b. Same methods as demand for increased operation as that those stated above.
flow through heat stated above.
exchanger.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 8 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 7. Motor operated globe a. Fails closed a. Failure blocks flow a. CVCS letdown flow 1. Valve is normally valve from train A of RHRS indication (NVP5530) closed to align the (ND24A(ND58B to CVCS letdown at CB. RHRS for ECCS analogous) heat exchanger. Fault operation during plant prevents (during the power operation.
initial phase of plant cooldown) equalizing boron concentration of coolant in RHRS train A and in the RCS using the RHR cleanup line to CVCS.
No effect on safety for system operation.
Operator can balance boron concentration levels by cracking open flow control valve ND26 to permit flow to cold leg of loop B of RCS in order to balance levels using normal CVCS letdown flow. Later during cooldown, letdown flow comes from train B of RHRS.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 9 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 8. Air diaphragm a. Fails to open on a. Failure blocks flow a. Valve position 1. The valve is normally operated globe valve demand. from train A and B of indication (degree of closed and designed NV135. RHRS to CVCS opening) at CB and to Fail closed.
letdown heat CVCS letdown flow exchanger. Fault indication (NVP5530) prevents use of RHR at CB.
cleanup line to CVCS for balancing boron concentration levels of RHRS trains A and B with RCS during initial cooldown operation. Later in plant cooldown RHRS letdown flow is blocked. No effect on safety for system operation. Operator can balance boron concentration as stated above for item
- 7. Water clarity can alternately be maintained utilizing the FW pump through KF system purification loop.
- 2. Valve is a component of the CVCS that interfaces with the RHRS during plant cooldown.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-31 (Page 10 of 10)
Effect on System Failure Detection Component Failure Mode Operation1 Method2 Remarks
- 9. Motor operated gate a. Fails to close a. Failure to close results a. Valve position 1. Valve is a component valve FW27A in loss of ability to indication (open to of the ECCS that (FW55B analogous) open the associated closed position performs an RHR trains RCS loop change) at CB and function during plant suction valve. In this valve (closed) cooldown. Valve is case the alternate monitor light and normally open to RHR loop is used for alarm at CB. align the RHRS for RHR cooling. ECCS operation during plant power operation.
List of acronyms and abbreviations Auto--Automatic RC --Reactor Coolant CB --Control Board RCS --Reactor Coolant System CVCS--Chemical and Volume Control System RHR --Residual Heat Removal ECCS--Emergency Core Cooling System RHRS--Residual Heat Removal System MO --Motor Operated RWST--Refueling Water Storage Tank Notes:
- 1. See list at end of table for definition of acronyms and abbreviations used.
- 2. 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.
(15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-32 (Page 1 of 1)
Table 5-32. Pressurizer Design Data Design Pressure, psig 2485 Design Temperature, 0F 680 Surge Line Nozzle Diameter, in 14 Heatup Rate of Pressurizer Using 55 Heaters Only, 0F/hr Internal Volume ft3 1800 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-33 (Page 1 of 1)
Table 5-33. Pressurizer Quality Assurance Program
[HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED]
RT1 UT1 PT1 MT1 Heads
- 1. Plates yes
- 2. Cladding yes Shell
- 1. Plates yes
- 2. Cladding yes Heaters
- 1. Tubing 2 yes yes
- 2. Centering of element yes Nozzle (Forgings) yes yes3 yes3 Weldments
- 1. Shell, longitudinal yes yes
- 2. Shell, circumferential yes yes
- 3. Cladding yes
- 4. Nozzle Safe End yes yes4 yes4
- 5. Instrument Connection yes
- 6. Support Skirt, Longitudinal Seam yes yes
- 7. Support Skirt to Lower Head yes yes
- 8. Temporary Attachments (after removal) yes
- 9. All external pressure boundary yes welds after shop hydrostatic test Notes:
- 4. Weld Overlay Installation, UT and PT (15 NOV 2007)
Catawba Nuclear Station UFSAR Table 5-34 (Page 1 of 1)
Table 5-34. Pressurizer Relief Tank Design Data Design Pressure, psig 100 Normal Operating Pressure, psig 3 Final Operating Pressure, psig 50 Rupture Disc Release Pressure, psig Normal 91 Range 86-100 Normal Water Volume, ft3 1350 Normal Gas Volume, ft3 450 Design Temperature, 0F 340 Maximum Initial Operating Water Temperature, 0F 120 0
Maximum Final Operating Water Temperature, F 200 Total Rupture Disc Relief Capacity at 100 psig, 1b/hr 1.6 x 106 Cooling Time Required Following Maximum Discharge (Approximate), hr 1 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-35 (Page 1 of 1)
Table 5-35. Relief Valve Discharge To The Pressurizer Relief Tank Reactor Coolant System 3 Pressurizer Safety Valves Figure 5-3 3 Pressurizer Power-Operated Relief Valves Figure 5-3 Residual Heat Removal System 2 Residual Heat Removal Pump Figure 5-17 Suction Line from the Reactor Coolant System Hot Legs Chemical and Volume Control System 1 Seal Water Return Line Figure 9-89 1 Letdown Line Figure 9-89 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-36 (Page 1 of 1)
Table 5-36. Reactor Coolant System Boundary Valve Design Parameters Reactor Coolant Boundary Valves Parameter Design Pressure, psig 2485 Pre-Operational Hydrotest, psig 3107
- a. Design Temperature, 0F 650 Other than pressurizer safety and power operated relief valves
- b. Pressurizer safety and 680 power operated relief valves (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-37 (Page 1 of 1)
Table 5-37. Pressurizer Valves Design Parameters Pressurizer Spray Control Valves Parameters Number 2 Design pressure, psig 2485 Design temperature, 0F 650 Design flow for valves full open, each, gpm 450 Pressurizer Safety Valves Number 3 Maximum relieving capacity, ASME rated flow, 420,000 1b/hr (per valve)
Set pressure, psig 2485 Fluid Saturated steam Backpressure:
Normal, psig 3 to 5 Design, psig 500 Pressurizer Power Relief Valves Number 3 Design pressure, psig 2485 Design temperature, 0F 680 High pressure setpoint, psig 2335 Relieving capacity, 1b/hr (per valve) 210,000 Fluid Saturated Steam Low pressure setpoint, psig (NC-32B and NC-34A only) 400 psig Relieving capacity, gpm (per valve) 1060 Fluid Water (@600F)
(21 OCT 2010)
Catawba Nuclear Station UFSAR Table 5-38 (Page 1 of 1)
Table 5-38. Component Supports. Loading Combinations and Code Requirements Loading Combination Code or Stress Requirements Normal and Upset Conditions Preliminary Design Final Design
OL = Normal Operating Load: These loads are associated with plant operations in addition to weight of permanent equipment.
LL = Live Load, including construction loads.
OBE = Operating Basis Earthquake load.
SSE = Safe Shutdown Earthquake load.
LOCA = Accident loads including reactions due to pipe rupture and thermal loads.
AISC = Specifications for Design, Fabrication and Erection of Structural Steel Buildings, Seventh Edition, 1973.
Fs = Steel allowable stresses as specified in AISC Part 1.
Fy = Yield stress of structural steel.
ASME = 1974 ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, through Summer of 1974 addenda, including Appendix F and Appendix XVII.
Notes:
- 1. For loading combinations (3) and (4) which are ultimate loading conditions, the allowable stresses for the structural steel are as follows:
Type of Stress Allowable Stress Tension, Compression and Bending 0.9 Fy Shear 0.55 Fy Compression with Buckling 1.7 Fs
- 2. Allowable stresses are not specified for the Reactor Coolant Pump bolts as material is not defined in Section III, Appendix I of the ASME Code. Allowable stresses shown for the preliminary design are used in the final design check of these bolts.
(22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-39 (Page 1 of 1)
Table 5-39. Materials Material used in these supports include:
Plate - SA-516 Grade 70 - SA-516 Grade 60
- SA-533, Class 1 - SA-516 Grade 55
- A-588 - SA-106, Grade B
- SA-36 - SA-240, Type 304 Rod - SA-306 Grade 70
- SA-306 Grade 60 Bolts - SA-637 Grade 688, Type 2 - SA-325
- 4340 (Modified) - SA-193, GR B7 Forging - A-471, Class 9 SA 540 Grade B22
- SA-540 Grade B24 (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-40 (Page 1 of 2)
Table 5-40. Reactor Vessel Material Surveillance Program - Withdrawal Schedule Withdrawal Fluence Unit 1 Vessel Time EFPY at Lead (n/cm2 x Capsules Location (EOC) EOC Date Withdrawal Factor 1019) Reference Z 301.5° 1 8/8/86 0.79[e] 3.85 0.292 WCAP-11527 Y 241° 6 7/10/92 4.98[e] 3.73 1.31 WCAP-13720 W 121.5° 14 11/18/03 14.69[e] 4.00 3.51 Note 6 X 238.5° 10 11/28/97 9.29[e] 3.88 2.41 WCAP-15117 Note 3 U 58.5° 10 11/28/97 9.29[e] 3.88 2.41 WCAP-15117 Note 3 V 61° 10 11/28/97 9.29[e] 3.72 2.31 WCAP-15117 Ex-vessel NA 16 11/11/2006 17.35[e] NA -- WCAP-16869-Dosimetry NP, Rev 1 Unit 2 Capsules Z 301.5° 1 12/23/87 0.86 4.13 0.323 WCAP-11941 X 241° 5 1/23/93 4.52 4.14 1.23[a] WCAP-13875 W 121.5° 14 3/17/06 13 4.28 3.0[d] Note 3 U 58.5° Note 4 Note 4 Note 4 --- --- --
Y 238.5° 9 9/13/98 9.24 4.33 2.49 WCAP-15243 V 61° 9 9/13/98 9.24 4.13 2.38[b][c] WCAP-15243 Ex-vessel NA 14 4/5/2006 13 NA -- --
Dosimetry
- a. Approximate fluence at vessel 1/4 thickness location, at 32 EFPY
- b. Approximate fluence at vessel inner wall location, at 32 EFPY
- c. Approximate fluence at vessel 1/4 thickness location, at 54 EFPY
- d. Approximate fluence at vessel inner wall location at 54 EFPY
- e. The fluence evaluation supporting this effort did not consider 0.11 Effective Full Power Years (EFPY) of neutron exposure resulting from pre-commercial operation of Catawba Unit 1, between January 1985 and June 1985. The impact of this omission on the fluence evaluation results has been assessed to be negligible, and the results remain valid within the 20% uncertainty criterion for fluence calculations. Future fluence evaluations will consider this pre-commercial operation phase of Catawba Unit 1.
Notes:
- 1. EFPY= Effective Full Power Year
- 2. EOC=End of Cycle
- 3. Capsule specimens have been removed and stored at Westinghouse after reading dosimetry. These specimens are available for testing or additional irradiation if ever deemed necessary.
(09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-40 (Page 2 of 2)
- 4. Capsule U is not available for irradiation and testing.
- 5. For CNS-2 Capsule X was discovered in the Y location, Capsule Y was in the X location. Values listed are actual corrected as-found locations.
- 6. CNS-1 Capsule W was placed in the spent fuel pool following removal.
(09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-41 (Page 1 of 1)
Table 5-41. Reactor Coolant System Pressure Isolation Valves Valve Number Function NI59 Accumulator Discharge NI60 Accumulator Discharge NI70 Accumulator Discharge NI71 Accumulator Discharge NI81 Accumulator Discharge NI82 Accumulator Discharge NI93 Accumulator Discharge NI94 Accumulator Discharge NI124 Safety Injection (Hot Leg)
NI125 Residual Heat Removal (Hot Leg)
NI126 Safety Injection (Hot Leg)
NI128 Safety Injection (Hot Leg)
NI129 Residual Heat Removal (Hot Leg)
NI134 Safety Injection (Hot Leg)
NI156 Safety Injection (Hot Leg)
NI157 Safety Injection (Hot Leg)
NI159 Safety Injection (Hot Leg)
NI160 Safety Injection (Hot Leg)
NI165 Safety Injection/Residual Heat Removal (Cold Leg)
NI167 Safety Injection/Residual Heat Removal (Cold Leg)
NI169 Safety Injection/Residual Heat Removal (Cold Leg)
NI171 Safety Injection/Residual Heat Removal (Cold Leg)
NI175 Safety Injection/Residual Heat Removal (Cold Leg)
NI176 Safety Injection/Residual Heat Removal (Cold Leg)
NI180 Safety Injection/Residual Heat Removal (Cold Leg)
NI181 Safety Injection/Residual Heat Removal (Cold Leg)
ND1B Residual Heat Removal ND2A Residual Heat Removal ND36B Residual Heat Removal ND37A Residual Heat Removal (22 OCT 2001)
Catawba Nuclear Station UFSAR Table 5-42 (Page 1 of 1)
Table 5-42. RT PTS Calculations for Catawba Unit 1 Beltline Region Materials at 54 EFPY Fluence @
54 EFPY (1019 RTNDT RT PTS Material CF n/cm2) FF (U) RT PTS M °F Upper Shell Forging 06 123.5 0.116 0.4472 -26 55.2 34.0 63 Intermediate Shell Forging 58 2.60 1.2559 -8 72.8 34.0 99 05 Using Surveillance 28.5 2.60 1.2559 -8 35.8 17.0 45 Capsule Data Lower Shell Forging 04 26 2.60 1.2559 -13 32.7 32.7 52 Bottom Head Ring 03 37 0.195 0.5634 14 20.8 20.8 56 Upper to Intermediate 41 0.116 0.4472 10 18.3 18.3 47 Shell Circumferential Weld W06 Intermediate to Lower 54 2.60 1.2559 -51 67.8 56.0 73 Shell Circumferential Weld W05 Using Surveillance 28.5 2.60 1.2559 -51 35.8 28.0 13 Capsule Data Lower Shell to Bottom 41 0.195 0.5634 10 23.1 23.1 56 Head Ring Weld W04 Data Sources:
All Materials: Westinghouse Report WCAP-17669-NP, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations," dated June 2013.
(09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-43 (Page 1 of 1)
Table 5-43. RT PTS Calculations for Catawba Unit 2 Beltline Region Materials at 54 EFPY Fluence @
54 EFPY RT PTS Material CF (1019 n/cm2) FF RTNDT(U) RT PTS M °F Bounding Nozzle Shell Material 77 0.11 0.43 50 32.9 32.9 115.8 Bounding Nozzle Weld Material 81 0.11 0.43 -40 34.6 34.6 29 Intermediate Shell Plate B8605-1 51 3.16 1.3 15 66.3 34 115 Using Surveillance Capsule Data 44 3.16 1.3 15 57.2 17 89 Intermediate Shell Plate B8605-2 51 3.16 1.3 33 66.3 34 133 Intermediate Shell Plate B8616-1 31 3.16 1.3 12 40.3 34 86 Lower Shell Plate B8806-1 37 3.16 1.3 6 48.1 34 88 Lower Shell Plate B8806-2 37 3.16 1.3 -10 48.1 34 72 Lower Shell Plate B8806-3 37 3.16 1.3 8 48.1 34 90 Intermediate, Lower and Intermediate to 37.3 3.16 1.3 -80 48.5 48.5 17 Lower Shell Weld Seams Using Surveillance Capsule Data 33.4 3.16 1.3 -80 43.4 28 -9 Data Sources:
Bounding Nozzle Materials: Internal calculation DPC-1201.01-00-0006, CNC-1201.01-00-0020, USE and RTPTS Values for Reactor Vessel Nozzle Region Locations, Rev. 0, dated July 2002.
All Other Beltline Materials: WCAP-15449, Rev. 1, Evaluation of Pressurized Thermal Shock for Catawba and McGuire Units 1 & 2 @
54 EFPY, dated October 2002.
(24 OCT 2004)
Catawba Nuclear Station UFSAR Table 5-44 (Page 1 of 1)
Table 5-44. Evaluation of Upper Shelf Energy for Catawba Unit 1 Beltline Region Materials at 54 EFPY 1/4 T EOL Projected Projected Weight % Fluence Unirradiated USE Decrease USE @ 54 Material of Cu (1019 n/cm2) USE (ft-lb) (%) EFPY (ft-lb)
Upper Shell Forging 06 0.16 0.070 101 14 87 Intermediate Shell Forging 05 0.09 1.565 134 21 106 Using Surveillance Capsule Data 0.09 1.565 134 10 121 Lower Shell Forging 04 0.04 1.565 134 21 106 Bottom Head Ring 03 0.06 0.117 68 12 60 Upper to Intermediate Shell 0.03 0.070 92 10 83 Circumferential Weld W06 Intermediate to Lower Shell 0.04 1.565 130 21 103 Circumferential Weld W05 Using Surveillance Capsule Data 0.04 1.565 130 8 120 Lower Shell to Bottom Head Ring 0.03 0.117 92 12 81 Weld W04 Data Sources:
All Materials: Westinghouse Report WCAP-17669-NP, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations," dated June 2013.
(09 OCT 2016)
Catawba Nuclear Station UFSAR Table 5-45 (Page 1 of 1)
Table 5-45. Evaluation of Upper Shelf Energy for Catawba Unit 2 Beltline Region Materials at 54 EFPY 1/4 T EOL Projected USE Weight % of Fluence Unirradiated Projected USE @ 54 EFPY Material Cu (1019 n/cm2) USE (ft-lb) Decrease (%) (ft-lb)
Bounding Nozzle Shell Material 0.11 0.063 65 10.0 58.5 Bounding Nozzle Weld Material 0.16 0.063 102 16.0 85.7 Intermediate Shell Plate B8605-1 0.08 1.88 89 6.6 90 Intermediate Shell Plate B8605-2 0.08 1.88 82 22 64 Intermediate Shell Plate B8616-1 0.05 1.88 92 22 72 Lower Shell Plate B8806-1 0.06 1.88 83 22 65 Lower Shell Plate B8806-2 0.06 1.88 102 22 80 Lower Shell Plate B8806-3 0.06 1.88 105 22 82 Intermediate Shell 0.04 1.13 146 10 131 Longitudinal Weld Seams 101-142A, B, C 1.88 11 130 1.88 11 130 Intermediate Shell to Lower Shell 0.04 1.88 146 11 130 Circumferential Weld Seams Lower Shell Longitudinal Weld Seams 0.04 1.88 146 11 130 101-124 A, B, C 1.13 10 131 1.88 11 130 Data Soruces:
Bounding Nozzle Materials: Internal calculation DPC-1201.01-00-0006, CNC-1201.01-00-0020, USE and RTPTS Values for Reactor Vessel Nozzle Region Locations, Rev. 0, dated July 2002.
All Other Beltline Materials: Westinghouse Letter DPC 00 069, dated October 22, 2000 (21 OCT 2010)
Catawba Nuclear Station UFSAR Table 5-46 (Page 1 of 1)
Table 5-46. Summary of Reactor Coolant System Leakage Detection Instrumentation Exceptions and Comments to Regulatory Guide (RG) 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, (Rev. 0)
RG 1.45 Regulatory Position Exception/Comment C.2 Leakage to the primary reactor containment Incore sump alarm will detect a 1 gpm input from unidentified sources should be collected and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of leakage reaching the sump.
the flow rate monitored with an accuracy of one gallon per minute or better.
C.5 The sensitivity and response time of each Exception taken for containment particulate leakage detection system in regulatory position 3 radiation monitor and incore sump level alarm.
above employed for unidentified leakage should The particulate radiation monitor sensitivity be adequate to detect a leakage rate, or its will be 10-9 uCi/cc. The particulate monitor equivalent, of one gpm in less than one hour.
alarm setting wil be as low as practicable based on background and sufficiently high enough to prevent spurious alarms.
Operability will be based on the sensitivity and surveillance testing.
The incore sump alarm will actuate within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of leakage reaching the sump.
Clarified Containment Floor and Equipment sump and Containment Ventilation Unit Condensate Drain Tank Level sensitivity of 1 gpm after leakage has reached the sump/tank.
C.6 The leakage detection systems should be Exception taken for the radioactivity monitoring capable of performing their functions following system design for a seismic event.
seismic events that do not require plant shutdown.
The airborne particulate radioactivity monitoring system should remain function when subjected to the SSE.
C.7 Indicators and alarms for each leakage Exception taken for incore sump indication in the detection system should be provided in the main control room - alarm only.
control room. Procedures for converting various The particulate radiation monitor and incore sump indications to a common leakage equivalent will alarm during the presence of a leak but are should be available to the operators. The not converted to a leakage equivalent (e.g. gpm).
calibration of the indicators should account for needed independent variables.
C.8 The leakage detection systems should be Exception taken for incore sump level alarm for equipped with provisions to readily permit testing testing and calibration during plant operation.
for operability and calibration during plant operation.
(15 NOV 2007)