ML20309A720
Text
McGuire Nuclear Station UFSAR Appendix 5A. Tables Appendix 5A. Tables
McGuire Nuclear Station UFSAR Table 5-1 (Page 1 of 2)
(09 OCT 2015)
Table 5-1. System Design and Operating Parameters1 Unit Design Life, Years 40 Nominal Operating Pressure, psig 2235 Total System Volume, including pressurizer and surge line, ft3 13,050 System Liquid Volume, including pressurizer water level (60% full), ft3 11,155 Deleted Per 2014 Update System Thermal and Hydraulic Data Temperatures (Based on Thermal Design Flow)
NSSS Power, MWt 3,483 NSSS Power, 106 BTU/hr 11,892 Licensed Reactor Power, MWt 3,469 Licensed Reactor Power, 106 BTU/hr 11,837 Thermal Design Flow, GPM/Loop 100,200 Total Reactor Coolant Flow, lb/hr 148 x 106 Reactor Vessel Inlet Temperature, °F 555.6 Reactor Vessel Outlet Temperature, °F 614.6 Steam Generator Primary Outlet Temperature, °F 555.1 Steam Generator at Full Power, psia 1020.7 Steam Generator Steam Temperature, °F 548.7 Steam Flow at Full Power, lb/hr (total) 15.5 x 106 Feedwater Inlet Temperature, °F 442.0 Pressurizer Spray Rate, max., gpm 900 Pressurizer Heat Capacity, kw 1800 Pressurizer Relief Tank Volume, ft3 1800 Flows and Pressure Drops (Based on Best Estimate Flow)
Best Estimate Flow, GPM/Loop 99,295 Pump Head at B.E. Flow, ft 276 Reactor Vessel P, psi 46.7 Steam Generator P, psi 33 Piping P, psi 8.0
McGuire Nuclear Station UFSAR Table 5-1 (Page 2 of 2)
(09 OCT 2015)
Note:
- 1. These values are approximate. Exact design values may be obtained from the NSSS Vendor (Westinghouse) or to the steam generator vendor (BWI)
McGuire Nuclear Station UFSAR Table 5-2 (Page 1 of 1)
(05 APR 2011)
Table 5-2. has been incorporated into Table 5-49
McGuire Nuclear Station UFSAR Table 5-3 (Page 1 of 2)
(14 OCT 2000)
Table 5-3. Load Combinations and Operating Conditions Load Combination (Except for RSGs - See Below)
Operating Condition
- 1. Normal (deadweight, thermal and pressure)
Normal Condition
- 2. Normal and Operating Basis Earthquake Upset Condition
- 3. Normal and Safe Shutdown Earthquake Faulted Condition
- 4. Normal and Design Basis Accident Faulted Condition
- 5. Normal and Safe Shutdown Earthquake and Design Basis Accident Faulted Condition The following load combinations shall be used for the structural evaluation of the RSG (pressure boundary)
Loading Conditions Service Loads/Combinations ASME SECTION III Service Stress SUBSECTION NB Limit Level Design Deadweight Operating Basis Earthquake (OBE)
Design Pressure Design Temperature Design Flow Thermal Internal Design Mechanical Loads Design Normal Deadweight Thermal Internal Normal Mechanical Loads Normal Condition Transients A
Upset Deadweight Thermal OBE Internal Upset Mechanical Loads Upset Condition Transients B
Emergency (See Note 1.)
Deadweight Thermal Internal Emergency Mechanical Loads Emergency Condition Transients C
Faulted Deadweight Thermal Safe Shutdown Earthquake (SSE)
Internal Faulted Mechanical Loads Pipe Rupture Loads Faulted Condition Transients D
McGuire Nuclear Station UFSAR Table 5-3 (Page 2 of 2)
(14 OCT 2000)
Load Combination (Except for RSGs - See Below)
Operating Condition Note:
- 1. This loading condition is not part of McGuire Nuclear Station Design Bases, but is included here for new RSG only.
McGuire Nuclear Station UFSAR Table 5-4 (Page 1 of 2)
(14 OCT 2000)
Table 5-4. Faulted Condition Stress Limits for Class 1 Components(4,5)
System (or Subsystem) Analysis Components Analysis Stress Limits for Components Test Pm Pm + Pb ELASTIC Smaller of 2.4 Sm and 0.70 Su Smaller of 3.6 Sm and 1.05 Su Note 2 ELASTIC Plastic Larger of 0.70 Su or Sy
+ 1/3(Su - Sy)
Larger of 0.70 Sut or Sy + 1/3 (Sut - Sy)
Limit Analysis 0.9 L1 Note 1 PLASTIC Plastic Larger of 0.70 Su or Larger of 0.70 Sut or 0.8 L T Elastic Sy + 1/3 (Su - Sy)
Sy + 1/3 (Sut - Sy)
Note 3 Note:
- 1. L1 = Lower bound limit load with an assumed yield point equal to 2.3 Sm.
- 2. These limits are based on a bending shape of 1.5 for simple bending cases with different shape factors, the limits will be changed proportionally.
- 3. LT = The limits established for the analysis need not be satisfied if it can be shown from the test of a prototype or model that the specified loads (dynamic or static equivalent) do not exceed 80 percent of LT, where LT is the ultimate load or load combination used in the test. In using this method, account shall be taken of the size effect and dimensional tolerances (similitude relationships) which may exist between the actual component and the tested models to assure that the loads obtained from the test are a conservative representation of the load carrying capability of the actual component under postulated loading for faulted conditions.
- 4. The use of these stress limits on Westinghouse-supplied components will be shown to be no less conservative than the limits of Appendix F to the ASME Code.
If plastic component analysis is used with elastic system analysis or with plastic system analysis, the deformations and displacement of the individual system members will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis.
Sy = Yield stress at temperature Su = Ultimate stress from engineering stress-strain curve at temperature Sut = Ultimate stress from the true stress-strain curve at temperature Sm = Stress intensity from ASME Section III at temperatue
McGuire Nuclear Station UFSAR Table 5-4 (Page 2 of 2)
(14 OCT 2000)
System (or Subsystem) Analysis Components Analysis Stress Limits for Components Test
- 5. Babcock and Wilcox replacement steam generators are designed to the requirements of Section III of the ASME Code for faulted conditions.
McGuire Nuclear Station UFSAR Table 5-5 (Page 1 of 3)
(27 MAR 2002)
Table 5-5. Active and Inactive Valves in the Reactor Coolant System Boundary Line Valve Line Size Type A-Active 2 I - Inactive Normal Position Post-LOCA Position RHR Suction 1)Motor gate 14" I
Closed (interlocked) with RCS pressure)
Closed 2)Motor gate 14" I
Closed (interlocked with second RCS pressure instrument)
Closed Loop Drains (each loop)
Charging 1)Manual globe 2"
I Closed Closed 2)Manual globe 2"
I Closed Closed 1)Check 3"
A Open Closed 2)Check 3
A Open Closed RHR Return or Cold Leg Connections (each loop) 1)Check 10" A
Closed Open - for lowhead injection and accumulator injection 2)Check 6"
A Closed Open - for lowhead injection Accumulator 1)Check 10" A
Closed Open - for accumulator and low head injection 2)Check 10" A
Closed Open - for accumulator injection SIS - Boron Injection Tank (each loop) 1)Check 11/2" A
Closed Open - for boron injection 2)Manual Throttle 11/2" I
Open Open
McGuire Nuclear Station UFSAR Table 5-5 (Page 2 of 3)
(27 MAR 2002)
Line Valve Line Size Type A-Active 2 I - Inactive Normal Position Post-LOCA Position 3)Check 11/2" A
Closed Open - for boron injection High Head Hot Leg Connections (each loop) 1)Check 6"
A Closed Open - for highhead recirculation 2)Check 2"
A Closed Open - for highhead recirculation Low Head Hot Leg Connections (each loop) 1)Check 6"
A Closed Open - for lowhead recirculation 2)Check 8"
A Closed Open - for lowhead recirculation High Head Cold Leg Connections 1)Check 10" A
Closed Open - for high head injection and recirculation 2)Check 2"
A Closed Open - For high head injection and recirculation Excess letdown 1)Manual globe 1"
I Open (Locked)
Open 2)Air-op globe 1"
I(1)(5)
Closed (fail close)
Closed - on safety injection signal 3)Air-op globe 1"
I(1)(5)
Closed (fail close)
Closed - on safety injection signal Letdown 1)Manual globe 3"
I Open Open 2)Air-op gate 3"
I(3)
Open (fail close)
Closed - on low pressurizer level signal 3)Air-op gate 3"
I(3)
Open (fail close)
Closed - on low pressurizer level signal Alternate Charging 1)Check 3"
A1 Closed Closed
McGuire Nuclear Station UFSAR Table 5-5 (Page 3 of 3)
(27 MAR 2002)
Line Valve Line Size Type A-Active 2 I - Inactive Normal Position Post-LOCA Position 2)Check 3"
A1 Closed Closed Pressurizer relief Valves (to PRT) 1)Motor gate 3"
I Open Open 2)Air-op globe 3"
A1 Closed (fail close)
Closed - low pressurizer pressure signal Pressurizer Safety Valves (to PRT) 1)Safety Valve 6"
A Closed Closed Auxiliary Spray (from CVCS) 1)Check 2"
A Closed Closed 2)Air-op globe 2"
I (4)
Closed (fail close)
Closed - operator action Note:
- 1. There is a possibility for these valves to be open when the accident occurs.
- 2. Active here means the valve must change position to perform a safety or shutdown function and is not intended to designate that the valve is postulated to fail.
- 3. LOCA isolation assured via St closure of the downstream letdown line containment isolation valves.
- 4. LOCA isolation assured via Ss closure of the upstream changing line outboard series isolation valves, and inboard containment isolation check valve.
- 5. LOCA isolation assured by St closure of downstream containment isolation valves.
McGuire Nuclear Station UFSAR Table 5-6 (Page 1 of 1)
(27 MAR 2002)
Table 5-6. Deleted Per 1998 Update
McGuire Nuclear Station UFSAR Table 5-7 (Page 1 of 1)
(14 OCT 2000)
Table 5-7. Applicable Code Addenda For Class 1 Equipment Within RCPB
- 1. Reactor Vessel (Unit 1)
ASME III, 1971 Edition thru Summer 1971 (Unit 2)
ASME III, 1971 Edition thru Winter 1971 Steam Generator ASME III, 1986 Edition No Addendum Pressurizer ASME III, 1971 Edition CRDM Housing Full Length ASME III, 1971 Edition thru Summer 1971 Part Length ASME III, 1971 Edition CRDM Head Adapter ASME III, 1971 Edition Reactor Coolant Pump ASME III, 1971 Edition thru Summer 1972 Reactor Coolant Pipe ASME III, 1971 Edition thru Winter 1971 Surge Lines ASME III, 1971 Edition thru Winter 1971
- 2. Many of these plants are purchased via multi-plant contracts. Therefore, the purchase order dates are not necessarily indicative of the applicable code addenda and it is not meaningful to supply them. As indicated in one (1) above, the code addenda required by the 10CFR 50.55a are met based on a CP date of 2/28/73.
McGuire Nuclear Station UFSAR Table 5-8 (Page 1 of 1)
(14 OCT 2000)
Table 5-8. Reactor Coolant Loop Piping Stress Analysis Summary1 Hot Leg Crossover Leg Cold Leg Evaluation Max.
Location Allowable Max.
Location Allowable Max.
Location Allowable Eq. 9 design stress intensity (ksi) 21.80 Hot Leg End of 50° Elbow 26.7 22.72 Crossover Leg Middle of 40° Elbow 26.7 25.41 Cold Leg Middle of 22° Elbow 26.7 Eq. 9 faulted stress intensity (ksi) 48.33 Hot Leg End of 50° Elbow 53.4 40.56 Crossover Leg Middle of 40° Elbow 53.4 51.44 Cold Leg End of 22° Elbow 53.4 Eq. 12 stress (ksi) 37.28 Hot Leg Middle of 50° Elbow 53.4 10.45 Middle of Crossover Leg 90° Elbow RCP Side 53.4 10.78 Cold Leg Middle of 22° Elbow 53.4 Eq. 13 stress intensity range (ksi) 39.8 RPV outlet nozzle 53.4 37.8 SG Outlet Nozzle 53.4 38.1 RPV Inlet Nozzle 53.4 Fatigue Usage Factor 0.6576 RPV outlet Nozzle 1.0 0.0511 SG Outlet Nozzle 1.0 0.106 8
RPV Inlet Nozzle 1.0 Note :
- 1. The stress intensities tabulated above from the Westinghouse analyses remain the bounding design basis for the reactor coolant loop piping. Stress intensities calculated for steam generator replacement are less than those shown.
McGuire Nuclear Station UFSAR Table 5-9 (Page 1 of 1)
(14 OCT 2000)
Table 5-9. Relief Valve Discharge to the Pressurizer Relief Tank Reactor Coolant System 3 Pressurizer Safety Valves 3 Pressurizer Power Operated Relief Valves Figure 5-1 Figure 5-1 Safety Injection System 2 SIS Discharge to Hot Leg 1 SIS Discharge to Cold Legs 1 SIS Pump Suction Figure 6-177 Figure 6-177 Figure 6-177 Residual Heat Removal System 1 RHR Pump Suction from RCS Hot Leg 2 RHR Pump Discharge to Cold Legs 1 RHR Pump Discharge to Hot Legs Figure 5-28 Figure 5-28 Figure 5-28 Chemical and Volume Control System 1 Charging Pump Suction 1 Seal Water Return Line 1 Letdown Line Figure 9-96 Figure 9-96 Figure 9-98 Containment Spray System 2 CS Pump Suction from Containment Sump Figure 6-194
McGuire Nuclear Station UFSAR Table 5-10 (Page 1 of 1)
(14 OCT 2000)
Table 5-10. Reactor Coolant System Pressure Settings Hydrostatic Test Pressure (Cold) 3107 psig Design Pressure 2485 Nominal Operating Pressure 2235 High Pressure Reactor Trip 2385 Low Pressure Reactor Trip 1945 Safety Valves Lift Setpoint 2485 Power Operated Relief Valves NC-32B & NC-36B Lift Setpoint3 2335 Power Operated Relief Valve NC-34A Lift Setpoint1,2,3
+100 psi Pressurizer Spray Valve Begin to Open1
+25 Proportional Spray Full On1
+75 Proportional Heaters Off1
+15 Proportional Heaters Full On1
-15 Backup Heaters On1
-25 High Pressure Alarm uncompensated pressure signal 2310 psig compensated pressure signal
+75 psi Low Pressure Alarm1
-25 PORV NC-32B and NC-34A Open (RCS Cold Leg 300°F)3 380 Note:
- 1. Input to this function is a compensated pressure signal
- 2. NC-34A is subjected to a 1 second lag in the initiating lift signal.
- 3. Each of the PORVs have a 2 second valve stroke time.
McGuire Nuclear Station UFSAR Table 5-11 (Page 1 of 3)
(14 OCT 2000)
Table 5-11. Reactor Coolant Pressure Boundary Materials Class 1 Primary Components Reactor Vessel Components Shell & Head Plates (other than core region)
SA533 Gr A, B or C, Class 1 or 2 (Vacuum treated)
Shell, Flange & Nozzle Forgings SA508 Class 2 or 3 Nozzle Safe Ends SA182 Type F304 or F316 CRDM & ECCS Appurtenances - Upper Head SB166 or 167 and SA182 Type F304 Instrumentation tube Appurtenances-SB166 or 167 and SA182 type 304, Lower Head F304L or F316 Closure Studs SA540 Class 3 Gr B23 or B24 Closure Nuts SA540 Class 3 or Gr B23 or B24 Closure Washers SA540 Class 3 Gr B23 or B24 Core Support Pads SB166 with Carbon less than 0.10%
Monitor Tubes & Vent Pipe SA312 or 376 Type 304 or 316 or SB167 Vessel Supports, Seal Ledge & Head Lifting Lugs SA516 Gr 70 Quenched & Tempered or SA533 Gr A, B, or C, Class 1 or 2. (Vessel Supports may be of weld metal buildup of equivalent strength)
Cladding Stainless steel weld Metal Analysis A-7 and Ni-Cr-Fe Weld Metal F-Number 43 Steam Generator Components Pressure Plates SA533 Gr B, Class 1 Pressure Forgings SA508 Class 3 Nozzle Safe Ends SA336-F316N/316LN Channel Heads SA-508 C13 Primary Divider Plate SB-168 NO6690 Tubes SB-163 Code Case N-20-3 Alloy 690 Cladding SFA 5.9 ER 309L/ER 308L (equivalent to 304L)
SFA 5.14 ER Ni-Cr3 (equivalent to Alloy 600, Inco82)
Closure Bolting SA 193 Gr B-7 Closure Nuts SA 194 Gr 7 Pressurizer Components Pressure Plates SA533 Gr A, B, or C, Class 1 or 2 Pressure Forgings SA508 Class 2 or 3
McGuire Nuclear Station UFSAR Table 5-11 (Page 2 of 3)
(14 OCT 2000)
Nozzle Safe Ends SA182 or 376 Type 316 or 316L and Ni-Cr-Fe Weld Metal F-Number 43 Cladding Stainless Steel Weld Metal Analysis A-7 and Ni-Cr-Fe Weld Metal F-Number 43 Closure Bolting SA540 Pressurizer Safety Valve Forgings SA182 Type F316 Reactor Coolant Pump Pressure Forgings SA182 Type 304, 316, or 348 Pressure Castings 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 Gr B7 or B8 SA540 Gr B23 or B24, SA453 Gr 660 Flywheel SA533 Gr B, Class 1 Part Length Mechanism Pressure Housing SA182 or SA312 Seamless Gr 304 and Code Case 1337-3 Bar Material SA479 Type 304 Welding Materials SFA 5.4 and 5.9 Type 308 and 308L and Ni-Cr-Fe Weld Metal F-Number 43 Reactor Coolant Piping Reactor Coolant Pipe Code Case 1423-a Gr F302N or 316N, or SA351 Gr CF8A or CF8M centrifugal castings Reactor Coolant Fittings SA351 Gr CF8A or CF8M Branch Nozzles SA182 Gr F304 or 316 or Code Case 1423-1 Gr F304N or 316N Surge Line & Loop Bypass SA376 Type 304 or 316 or Code Case 1423-1 Gr F304N or 316N Auxiliary Piping 1/2" through 12" and wall schedules 40S through 80S (ahead of second isolation valve)
ANSI B36.19 All other Auxiliary piping (ahead of second isolation valve)
ANSI B36.10 Socket weld fittings ANSI B16.11 Piping Flanges ANSI B16.5
McGuire Nuclear Station UFSAR Table 5-11 (Page 3 of 3)
(14 OCT 2000)
Auxiliary Piping Valves (Class 1)
SA182 Type 304 or 316 or SA351 GR CF8, CF8A or CF8M Welding Materials SFA 5.4 and 5.9 Type 304 or 308L Control Rod Drive Mechanism Pressure Housing SA182 Gr F304 or SA351 Gr CF8 Pressure Forgings SA182 Gr F304 or SA336 Gr F8 Bar Material SA479 Type 304 Welding Materials SFA 5.4 and 5.9 Type 308 or 308L Note:
- 1. UPDATING INFORMATION CONTAINED IN THIS TABLE IS NOT REQUIRED. The material information contained in this Table is "Historical" in nature and is intended to be used for reference only. Actual materials are procured in accordance with the applicable equipment specification and ASME code requirements and are documented on the approved manufacturers drawings.
McGuire Nuclear Station UFSAR Table 5-12 (Page 1 of 2)
(10 OCT 2009)
Table 5-12. Reactor Coolant Pressure Boundary Materials Class 1 and 2 Auxiliary Components Valves Motor and Manual Operated Gate and Check Valves Bodys SA182 Gr F316 Bonnets SA182 Gr F316 or SA351 Gr CF8 or CF8M Discs SA182 Gr 316 or SA479 Type XM-19 Cond A Stems SA564 Type 630 Cond 1100°F Heat Treatment and SB637 Type 07718 Closure Bolting & Nuts SA453 Gr 660 and SA194 Gr B6 Bonnet Retainer SA479 Type S21800 Air Operated Valves Bodys SA182 Gr F316 or SA351 Gr CF8 or CF8M Bonnets SA182 Gr F316 or SA351 Gr CF8 or CF8M Discs SA182 Gr F316 or SA546 Gr 630 Cond 1100°F Heat Treatment Stems SA182 Gr F318 or SA564 Gr 630 Cond 1100°F Heat Treatment or SA637 Gr 688 Type 2 Closure Bolting & Nuts SA453 Gr 660 and SA194 Gr B6 Bonnet Retainer SA182 Gr F316 Auxiliary Relief Valves Forgings SA182 Gr F316 Disc SA479 Gr 316 Miscellaneous Valves (2 inches and less)
Bodys SA479 Type 316 or SA351 Gr CF8 SA182 Type F316 Bonnets SA479 Type 316 or SA351 Gr CF8 Disc SA479 Type 316 Stems SA479 Type 410 or Type 304 A276 Type 410 Cond. H Closure Bolting & Nuts SA453 Gr 660 and SA193 Gr B6 Auxiliary Heat Exchangers Heads SA182 Gr F304 or SA240 Type 304 or 316 Flanges SA182 Gr F304 or F316
McGuire Nuclear Station UFSAR Table 5-12 (Page 2 of 2)
(10 OCT 2009)
Flange Necks SA182 Gr F304 or SA240 Type 316 or SA312 Type 304 Seamless Tubes SA213 Type 304 Tube Sheets SA240 Type 304 or 316 or SA182 Gr F304 or SA515 Gr 70 with Stainless Steel Weld Metal Analysis A-7 Cladding Shells SA351 Gr CF9 Pipe SA312 Type 304 Seamless Auxiliary Pressure Vessels Tanks, Filters, etc.
Shells & Heads SA240 Type 304 or SA264 Type 304 Clad to SA516 Gr 70 or SA516 with Stainless Steel Weld Metal Analysis A-7 Cladding Flanges & Nozzles SA182 Gr F304 and SA105 or SA350 Gr LF2 with Stainless Steel Weld Metal Analysis A-7 Cladding Piping SA312 Type 304 or Type 316 Seamless Pipe Fittings SA403 WP304 Seamless Closure Bolting & Nuts SA193 Gr B7 or B8 and SA194 Gr 2D Auxiliary Pumps Pump Casings & Heads SA351 Gr CF8 or CF8M, SA182 Gr F304 or F316 Flanges & Nozzles SA182 Gr F304 or F316, SA403 Gr WP316L Seamless Piping SA312 TP304 or TP316 Seamless Stuffing or Packing Box Cover SA351 Gr CF8 or CF8M, SA240 TP304 or TP316 Pipe Fittings SA403 Gr WP316L Seamless Closure Bolting & Nuts SA193 Gr B6, B7 or B8M and SA194 Gr 2H or Gr 8M Note:
- 1. UPDATING INFORMATION CONTAINED IN THIS TABLE IS NOT REQUIRED. The material information contained in this Table is "Historical" in nature and is intended to be used for reference only. Actual materials are procured in accordance with the applicable equipment specification and ASME code requirements and are documented on the approved manufacturers drawings.
McGuire Nuclear Station UFSAR Table 5-13 (Page 1 of 1)
(13 OCT 2018)
Table 5-13. Reactor Vessel Internals Materials for Emergency Core Cooling Forgings SA182 Type F304; SA182 304L (Unit 2 only)
Plates SA240 Type 304 Pipes SA312 Type 304 Seamless or SA376 Type 304 Tubes SA213 Type 304; SA213 304L (Unit 2 only)
Bars SA479 Type 304 & 410; SA276 Type XM-19 Nitronic 50, SA276 Type 304L (Unit 2 only)
Castings SA351 Gr CF8 or CF8A Bolting SA (Pending) Westinghouse PD Spec. 70041EA Nuts SA193 Gr B-8 Locking Devices SA479 Type 304 Weld Buttering Stainless Steel Weld Metal Analysis A-7
McGuire Nuclear Station UFSAR Table 5-14 (Page 1 of 1)
(09 OCT 2015)
Table 5-14. Reactor Coolant Chemistry Specification Solution pH Determined by the concentration of boric acid and alkali present. Expected values range between 4.2 (high boric acid concentration) to 10.5 (low boric acid concentration) at 250°C Oxygen, ppm, maximum Oxygen concentration of the reactor coolant is maintained below 0.1 ppm for operation above 250°F. Hydrazine may be used to chemically scavenge oxygen during the heatup.
Chloride, ppm, maximum 0.15 Fluoride, ppm, maximum 0.15 Hydrogen, cc(STP/kg H2O)
Normal Power Operation 25 - 50 Unit Refueling Shutdown Less than 5 Total Suspended Solids, ppm, maximum 1.0 pH Control Agent (Li7OH)
Variable, as necessary for pH control program Boric Acid, ppm B Variable from 0 to approximately 3000
McGuire Nuclear Station UFSAR Table 5-15 (Page 1 of 2)
(24 APR 2014)
Table 5-15. McGuire Unit 1 Reactor Vessel Toughness Table Component Material Specification/
Weld Number Code Number Heat Number/
Flux Type (Lot)
Cu (%)
Ni (%)
P(%)
TNDT
(°F)
(°F)
USE (ft-lb)
Closure head dome A533BCL.1 B5086-1 A-8354-1 0.11 0.48 0.010 20 372 652 Closure head segments A533BCL.1 B5087 A-8149-2 0.11 0.62 0.008 10 102 892 Closure head flange A508CL.2 B5002 123X369VAI 0.75 0.010 402 402 1012 Vessel flange A508CL.2 B4701 122W201VAI 0.73 0.010 292 292 1012 Inlet nozzle A508CL.2 B5003-1 (113°)
ZV-3862-1 0.12 0.68 0.010 602 602 892 Inlet nozzle A508CL.2 B5003-2 (293°)
ZV-3862-2 0.10 0.71 0.012 602 602 882 Inlet nozzle A508CL.2 B5003-3 (247°)
ZV-3862-S1 0.10 0.69 0.009 602 602 792 Inlet nozzle A508CL.2 B5003-4 ( 67°)
ZV-3862-S2 0.10 0.69 0.010 602 602 772 Outlet nozzle A508CL.2 B5004-1 (202°)
AV2101-N8C6136 0.74 0.005 602 602 822 Outlet nozzle A508CL.2 B5004-2 ( 22°)
AV2120-N8C6137 0.74 0.007 602 602 752 Outlet nozzle A508CL.2 B5004-3 (338°)
AV2081-N8C6138 0.71 0.005 602 602 902 Outlet nozzle A508CL.2 B5004-4 (158°)
AV2234-N8C6139 0.79 0.006 602 602 812 Upper shell A533BCL.1 B5453-2 C-5168-1 0.14 0.58 0.011 10 153 72.42 Upper shell A533BCL.1 B5011-2 C-4371-1 0.10 0.54 0.011 10 273 68.32 Upper shell A533BCL.1 B5011-3 C-4387-1 0.13 0.56 0.010 0
03 94.72 Intermediate shell A533BCL.1 B5012-1 C-4387-2 0.111 0.611 0.010
-30 34 101 Intermediate shell A533BCL.1 B5012-2 C-4417-3 0.14 0.61 0.011 0
0 105
McGuire Nuclear Station UFSAR Table 5-15 (Page 2 of 2)
(24 APR 2014)
Component Material Specification/
Weld Number Code Number Heat Number/
Flux Type (Lot)
Cu (%)
Ni (%)
P(%)
TNDT
(°F)
(°F)
USE (ft-lb)
Intermediate shell A533BCL.1 B5012-3 C-4377-2 0.11 0.66 0.013
-20
-13 112 Lower shell A533BCL.1 B5013-1 C-4315-1 0.14 0.58 0.009
-10 0
95 Lower shell A533BCL.1 B5013-2 C-4374-2 0.10 0.51 0.010
-10 30 115 Lower shell A533BCL.1 B5013-3 C-4371-2 0.10 0.55 0.010 0
15 103 Bottom head segment A533BCL.1 B5458-1 C-5168-3 0.14 0.60 0.011
-70
-263 903 Bottom head segment A533BCL.1 B5458-2 C-5175-3 0.15 0.54 0.014
-30
-153 963 Bottom head segment A533BCL.1 B5458-3 C-5342-4 0.13 0.56 0.012
-20 23 823 Bottom head dome A533BCL.1 B5085-1 C-9120-2 0.13 0.53 0.010 0
103 793 Intermediate shell longitudinal weld seams 2-442 A,B,C (Tandem)
M1.224 20291/12008 Linde 1092 (3833/3854) 0.199 0.846 0.011
-60
-50 112 Intermediate shell to lower shell girth weld 9-442 (Single Wire)
G1.39 83640 Linde 0091 (3490) 0.051 0.096 0.006
-70
-70 143 Lower shell longitudinal weld seams 3-442 A,B,C (Tandem)
M1.32 21935/12008 Linde 1092 (3889) 0.213 0.867 0.015 02
-50 124 Note:
All data was obtained or derived from manufacturers original Material Certification and Test Report (MCTR) data, except where noted below.
- 1. Calculated from a combination of manufacturers MCTRs and surveillance program data (see ATI-93-012-T001, Rev. 2).
- 2. Estimated per U.S. NRC Standard Review Plan, NUREG-0800, Branch Technical Position MTEB 5-2 (see also WCAP-10786, Table A-1).
- 3. Also estimated per U.S. NRC Standard Review Plan, but re-evaluation of charpy data in 2002 yielded new RTNDT values.
- 4. Surveillance weldment.
McGuire Nuclear Station UFSAR Table 5-16 (Page 1 of 1)
(14 OCT 2000)
Table 5-16. Deleted Per 1993 Update
McGuire Nuclear Station UFSAR Table 5-17 (Page 1 of 3)
(09 OCT 2015)
Table 5-17. McGuire Unit 2 Reactor Vessel Toughness Table Upper Shelf Energy Component Material Specification
/ Weld Number Code/Item/
Weld Number Heat Number/
Flux Type (Lot)
Cu
(%)
Ni
(%)
P
(%)
TNDT
(°F)
(°F)1 MWD (ft-lb)
NMWD (ft-lb)
Closure head dome A533B, Cl. 1 10 55154-1 0.63 0.011
-31 12 132 Closure head ring A508 Cl. 2 09 007055 0.86 0.006 16 16 156 Closure head flange A508 Cl. 2 08 526916 0.82 0.012
-13 1
155 Vessel flange A508 Cl. 2 07 218572 0.82 0.016
- 4
- 4 179 Inlet nozzle A508 Cl. 2 11 ( 67°)
526341-1 0.04 0.76 0.006
-13
-13 142 Inlet nozzle A508 Cl. 2 12 (113°)
526395-1 0.05 0.73 0.009
-31
-31 108 Inlet nozzle A508 Cl. 2 13 (247°)
526537 0.06 0.76 0.009
-22
-22 129 Inlet nozzle A508 Cl. 2 14 (293°)
526537 0.06 0.78 0.009
-40
-40 132 Outlet nozzle A508 Cl. 2 15 ( 22°)
526341 0.04 0.77 0.007
-13
- 7 122 Outlet nozzle A508 Cl. 2 16 (158°)
525789 0.05 0.83 0.011
-40
-24 103 Outlet nozzle A508 Cl. 2 17 (202°)
525789 0.05 0.84 0.010
-49
-16 116 Outlet nozzle A508 Cl. 2 18 (338°)
526395-2 0.05 0.74 0.010
-40
-30 1213 Upper shell A508 Cl. 2 06 411085 0.89 0.006
- 4 25 151 982 Intermediate shell A508 Cl. 2 05 526840 0.153 1
0.7931 0.012
- 4
- 42 147 94 Lower shell A508 Cl. 2 04 411337-11 0.15 0.88 0.004
-30
-302 152 141
McGuire Nuclear Station UFSAR Table 5-17 (Page 2 of 3)
(09 OCT 2015)
Upper Shelf Energy Component Material Specification
/ Weld Number Code/Item/
Weld Number Heat Number/
Flux Type (Lot)
Cu
(%)
Ni
(%)
P
(%)
TNDT
(°F)
(°F)1 MWD (ft-lb)
NMWD (ft-lb)
Bottom head ring A508 Cl. 2 03 527428 0.06 0.77 0.013
- 4 15 1093 712 Bottom head segment A533B, C1.1 02-01 (240°)
55126-2-3 0.59 0.007
-49 Bottom head segment A533B, Cl. 1 02-02 (180°)
55126-2-2 0.59 0.007
-49
- 2 136 Bottom head segment A533B, Cl. 1 02-03 (120°)
55126-2-1 0.59 0.007
-40
-40 131 Bottom head segment A533B, Cl. 1 02-04 (
60°)
55292-2-3 0.58 0.006
-13
-13 141 Bottom head segment A533B, Cl. 1 02-05 ( 0°)
55292-2-2 0.58 0.006
-13
-13 132 Bottom head segment A533B, C1.1 02-06 (300°)
55292-2-1 0.58 0.006
-13 Bottom head dome A533B, Cl. 1 01 55292-3 0.58 0.006
-40
-40 127 Intermediate to lower shell girth weld Submerged arc weld W054 895075/ Grau Lo (P46) 0.039 0.724 0.010
-76
-68 132
McGuire Nuclear Station UFSAR Table 5-17 (Page 3 of 3)
(09 OCT 2015)
Upper Shelf Energy Component Material Specification
/ Weld Number Code/Item/
Weld Number Heat Number/
Flux Type (Lot)
Cu
(%)
Ni
(%)
P
(%)
TNDT
(°F)
(°F)1 MWD (ft-lb)
NMWD (ft-lb)
Note:
All data was obtained or derived from manufacturers original Material Certification and Test Report (MCTR) data, except where noted below.
- 1. Calculated from a combination of manufacturers MCTRs and surveillance program data (see ATI-94-012-T004, Rev. 2).
- 2. Estimated per U.S. NRC Standard Review Plan, NUREG-0800, Branch Technical Position MTEB 5-2 (see also WCAP-11029, Table A-1).
- 3. 100% shear not reached, upper shelf energy is greater than listed.
- 4. Surveillance weldment.
McGuire Nuclear Station UFSAR Table 5-18 (Page 1 of 1)
(14 OCT 2000)
Table 5-18. Deleted Per 1993 Update
McGuire Nuclear Station UFSAR Table 5-19 (Page 1 of 1)
(14 OCT 2000)
Table 5-19. Unit 1 Surveillance Weld Metal (Ht. 20291 & 12008, Linde 1092, Lot 3854) Charpy V-Notch Impacts Temp.
Ft/Lbs.
% Shear Mils Lat.
Exp.
Temp. °F Energy Ft/Lbs.
Shear %
Lat. Exp.
Mils
-80 15 0
12
-100 4.5 6
1
-80 16 0
12
-100 4.5 6
1.5
-40 49 30 36
-35 11.5 23 11
-40 38 30 32
-35 30 27 26
-40 45 30 35 0
15 35 17
+10 71 40 52 0
31 40 29
+10 60 35 44 0
65 47 53
+10 73 40 49 25 75 60 61
+40 81 45 58 25 60 47 49
+40 90 50 63 50 74.5 60 63
+40 88 50 64 50 58 55 47
+110 100 90 81 100 105 93 84
+110 93 95 83 100 96 85 80
+160 111 100 82 150 105 92 86
+160 110 100 88 150 113 100 89 210 113 100 87 210 112 100 84 210 110 100 86 NDTT (°F): -60
McGuire Nuclear Station UFSAR Table 5-20 (Page 1 of 1)
(14 OCT 2000)
Table 5-20. Unit 1 Heat Affected Zone Charpy Data [3]
Temp.
Ft/Lbs.
% Shear Mils Lat.
Exp.
Temp.
(°F)
Energy Ft-lb Shear %
Lat. Exp.
Mils
-80 21 1
14
-100 13 20 7
-80 19 1
12
-100 21 18 11
-40 25 20 22
-50 24 32 14
-40 37 25 28
-50 39.5 37 27
-40 43 30 32
-25 58 50 36
+10 43 35 34
-25 35 37 26
+10 40 30 33 10 90 65 61
+10 95 60 61 10 65 48 45
+40 58 40 42 50 122 80 76
+40 61 45 45 50 74 65 54
+40 62 40 46 50 63.5 70 46
+110 94 90 67 100 126 100 79
+110 98 90 71 100 83 90 69
+160 82 100 70 110 104 94 73
+160 80 100 70 150 132 100 86 150 106 100 76 210 131 100 81 210 102 100 79 NDTT (°F): - 50
McGuire Nuclear Station UFSAR Table 5-21 (Page 1 of 1)
(14 OCT 2000)
Table 5-21. Unit 1 Intermediate Shell Plate C-4387-2 Longitudinal Data Transverse Data Temp.
°F Energy Ft-Lb Shear %
Lat. Exp.
Mils Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils
-40 17 0
12
-40 23 14 13
-40 19 0
16
-40 10 9
4.5
-40 12 0
9
-40 20 14 12 10 32 20 25 0
33 25 27 10 52 30 38 0
33 20 23 10 42 25 31 0
35 25 28 40 92 40 68 30 49 34 36 40 74 35 52 30 41 30 28 40 68 35 50 30 35 29 34 110 108 80 75 80 57 52 45 110 122 85 77 80 46 43 40.5 160 143 100 86 80 33 25 23 160 145 100 87 110 79 92 62 212 139 100 89 110 68 59 53 212 136 100 85 110 69 65 52.5 210 103 100 80.5 210 98 100 77 210 103 100 80 NDTT (°F) = -30 NDTT (°F) = -30
McGuire Nuclear Station UFSAR Table 5-22 (Page 1 of 1)
(14 OCT 2000)
Table 5-22. Unit 1 Intermediate Shell Plate C-4417-3 Longitudinal Data Transverse Data Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils
-40 12 0
13
-40 19 14 11.5
-40 17 0
15
-40 9
14 3.5
-40 11 0
10
-40 7
9 3.5 10 38 25 31 10 32 29 25 10 80 40 59 10 20 25 15 10 70 35 54 10 44 25 30.5 40 76 35 59 60 53 43 37 40 55 30 44 60 51 34 36 40 86 40 66 60 50 34 39 110 99 75 71 110 61 60 53 110 110 80 79 110 68 60 57 110 125 85 85 110 78 64 63 160 132 99 87 160 88 90 70 160 137 100 89 160 100 93 73 160 142 100 88 160 98 98 77 210 102 100 80 210 106 100 75 210 106 100 78 NDTT (°F) = 0
McGuire Nuclear Station UFSAR Table 5-23 (Page 1 of 1)
(14 OCT 2000)
Table 5-23. Unit 1 Intermediate Shell Plate C-4377-2 Longitudinal Data Transverse Data Temp. °F Energy Ft-Lb Shear %
Lat.
Exp.
Mils Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils
-40 15 0
12
-20 43 14 31
-40 35 20 26
-20 10 18 5
-40 16 0
14
-20 21 14 14.5 10 78 35 60 40 53 34 36 10 60 30 46 40 60 34 48 10 61 30 47 40 47 29 34 40 71 45 55 75 70 50 50 40 101 50 72 75 73 55 56 40 104 50 73 75 64 52 46 110 127 85 84 110 80 59 57 110 123 85 75 110 83 79 62.5 110 129 85 86 110 89 77 66 160 158 100 88 160 101 100 80 160 153 100 91 160 112 100 82 160 156 100 90 160 104 100 76 210 103 100 76 210 113 100 82 210 119 100 84 NDTT (°F) = -20
McGuire Nuclear Station UFSAR Table 5-24 (Page 1 of 1)
(14 OCT 2000)
Table 5-24. Unit 1 Lower Shell Plate C-4315-1 Longitudinal Data Transverse Data Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils Temp.
°F Energy Ft-Lb Shear %
Lat. Exp.
Mils
-40 9
0 6
-40 16 18 11
-40 28 10 22
-40 26 14 15
-40 12 0
12
-40 18 14 15 10 46 15 35 10 17 23 16 10 45 15 36 10 43 30 30 10 46 15 35 10 37 34 26 40 69 25 50 50 50 43 39 40 65 25 47 50 48 40 40 40 53 20 40 50 49 43 36 110 95 70 70 90 65 59 54 110 105 80 77 90 58 56 52 110 97 70 69 90 61 61 51 160 132 100 90 150 86 90 67 160 130 100 87 150 90 100 72 160 125 100 84 150 93 100 68 210 97 100 75 210 93 100 72 210 96 100 70 Note:
- 1. NDTT = -10°F
McGuire Nuclear Station UFSAR Table 5-25 (Page 1 of 1)
(14 OCT 2000)
Table 5-25. Unit 1 Lower Shell Plate C-4374-2 Longitudinal Data Transverse Data Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils
-40 15 0
12
-40 12 9
9
-40 22 0
19
-40 15 14 7
-40 14 0
12
-40 14 18 9
10 53 20 41 10 42 30 33 10 49 20 37 10 29 25 20 10 56 20 43 10 33 25 23.5 40 56 20 43 50 66 48 40 40 67 25 49 50 52 40 40 40 75 30 54 50 41 32 34 110 118 70 79 90 79 59 59 110 126 85 83 90 50 50 47 110 119 80 81 90 77 59 60 160 143 100 87 150 108 93 80 160 148 100 92 150 109 94 83 160 136 100 85 150 100 92 76 210 113 100 83 210 117 100 82.5 210 116 100 81 Note:
- 1. NDTT = -10°F
McGuire Nuclear Station UFSAR Table 5-26 (Page 1 of 1)
(14 OCT 2000)
Table 5-26. Unit 1 Lower Shell Plate C-4371-2 Longitudinal Data Transverse Data Temp. °F Energy Ft-Lb Shear %
Lat. Exp.
Mils Temp.
°F Energy Ft-Lb Shear %
Lat. Exp.
Mils
-40 24 5
17
-40 16 9
10
-40 21 5
15
-40 19 14 12
-40 14 0
12
-40 12 9
8 10 40 10 29 10 36 29 24 10 51 20 36 10 32 29 22 10 38 10 25 10 30 29 20.5 40 48 20 32 60 43 38 30 40 51 20 36 60 45 42 38 40 50 20 37 60 48 38 33 110 97 70 68 100 63 59 50 110 100 70 64 100 70 59 54 110 85 70 60 100 68 54 50 160 128 100 86 160 86 92 70 160 134 100 87 160 93 96 71 160 130 100 85 160 90 96 71 210 109 100 74 210 100 100 75 210 102 100 79 NDTT (°F) = 0
McGuire Nuclear Station UFSAR Table 5-27 (Page 1 of 1)
(14 OCT 2000)
Table 5-27. Unit 2 Heat Affected Zone Charpy Data Temp.
(°F)
Ft-Lbs.
% Shear Mils Lat.
Exp.
Temp.
(°F)
Energy ft-lb Shear
(%)
Lat. Exp.
Mils
-184 8.7 5
8
-150 3
0 1.5
-166 23.1 11 12
-125 6.5 0
0
-148 12.2 11 12
-125 3.5 0
0
-130 20.3 20 16
-100 29 18 16
-112 23.7 17 20
-75 73 56 48
-94 14.5 16 24
-75 34 47 26
-76 50.9 35 39
-50 43 20 28
-58 50.9 26 39
-16 66 69 43
-40 82.7 67 59
-16 67 64 48.5
-22 104.7 70 67
-16 78 77 51
+10.4 115.7 79 75 32 82 91 56 40 121.5 85 67 32 67 87 58 68 127.9 98 94 100 107 100 69 86 119.2 100 97 100 81 100 56 104 118.0 85 75 150 101 100 70 122 129.6 90 87 210 116 100 66 140 115.7 92 79 210 121 100 67 158 133.7 96 79 250 99 100 69 NDTT (°F) = -76
McGuire Nuclear Station UFSAR Table 5-28 (Page 1 of 1)
(14 OCT 2000)
Table 5-28. Unit 2 All Weld Metal Charpy V-Notch Impacts Temp.
(°F) ft-Lbs.
% Shear Mils Lat.
Exp.
Temp.
(°F)
Energy ft-lb Shear
(%)
Lat. Exp.
Mils
-184 3.5 0
4
-150 1
0 0
-166 5.2 6
8
-150 1
0 0
-148 9.3 5
12
-75 20 30 15.5
-130 13.9 16 16
-75 16.5 29 9.5
-112 11.0 11 12
-35 34 46 31
-94 26.0 16 16
-35 52 54 40
-76 28.9 30 28
-16 50 65 40
-58 28.9 34 32
-16 57 65 46
-40 33.6 47 28
-16 59.5 65 47
-22 54.4 56 47 25 93 81 67
+10.4 70.0 66 63 25 83.5 77 64 40 92.0 76 71 71 112 75 82 68 117.5 92 87 71 110 79 80 86 106.5 98 91 125 125 98 91 104 112.3 90 94 125 124 99 92 122 123.8 95 94 210 140 100 96 140 122.1 100 91 210 132.5 99 96.5 158 134.8 100 94 275 144.5 100 98 NDTT (°F) = -76
McGuire Nuclear Station UFSAR Table 5-29 (Page 1 of 2)
(14 OCT 2000)
Table 5-29. Unit 2 Core Region Lower Shell Forging, 04 NDT = -30°F Intermediate Shell Forging, 05 NDT = -4°F Tangential Data Tangential Data Temp. (°F)
Energy (Ft-Lb)
Shear
(%)
Lat.
Exp.
(Mils)
Temp.
(°F)
Energy (Ft-Lb)
Shear
(%)
Lat. Exp.
(Mils)
-148 5.8 0
4
-148 6.4 0
4
-148 8.1 0
8
-148 7.5 0
4
-148 7.5 0
8
-148 6.4 0
6
- 76 33.6 6
20
- 76 40.5 12 35
- 76 45.1 12 35
- 76 42.8 12 32
- 76 19.7 6
16
- 76 43.4 12 35
- 4 107.0 52 83
- 4 39.9 17 35
- 4 68.9 29 55
- 4 86.8 46 71
- 4 70.0 29 55
- 4 86.2 35 67 60 124.4 78 83 60 137.7 90 91 60 136.6 73 83 60 146.4 95 91 60 133.1 75 79 60 122.1 75 79 113 147.6 98 94 113 160.9 98 79 113 152.2 100 91 113 152.2 90 94 113 153.3 100 91 113 148.1 100 87 176 153.3 100 91 176 144.7 95 83 176 146.4 96 91 176 141.2 95 91 176 147.6 94 91 176 137.7 95 87
McGuire Nuclear Station UFSAR Table 5-29 (Page 2 of 2)
(14 OCT 2000)
Axial Data Axial Data Temp.
(°F)
Energy (Ft-Lb)
Shear
(%)
Lat.
Exp.
(Mils)
Temp.
(°F)
Energy (Ft-Lb)
Shear
(%)
Lat. Exp.
(Mils)
-100 9
4 2
-100 5.5 2
1
-100 13 5
2
-100 7.5 2
1
-50 39 20 23
-50 21 15 12
-50 16 15 6
-25 29 15 20
-25 11 23 7
-25 28.5 15 18 0
74 45 47 0
34 30 26 0
71 42 48 25 49 45 40 0
38 34 21 25 53 45 39 30 85.5 55 52 56 68 54 47 30 90.5 63 56 56 63 54 46 30 98.5 65 62 56 67 54 74 98 73 66 100 71.5 90 54 74 102 72 70 100 74 92 61 75 101 68 67 100 72 94 55 125 140 100 83 140 93 100 71 125 143 100 83 140 92.5 100 64 210 140 100 85 210 95 100 67 210 140 100 77 210 97 100 97 NDTT (°F) = -4
McGuire Nuclear Station UFSAR Table 5-30 (Page 1 of 2)
(13 APR 2020)
Table 5-30. Reactor Coolant Leakage Detection Sensing Device Parameter Monitored Readout Location Sensitivity Scintillation Detector Radioactivity accumulation on filter paper from sample of containment air Control Room The instrument sensitivity for air particulate is 10-9
µCi/cc. The response times are as follows:
- 1. For a 1.0 gal/min leak into Containment in mode 1, with activity from water activation products, leakage will be detected in less than 60 minutes.1
- 2. For a 1 gal/min leak into Containment, with corrosion products only, leakage will be detected in approximately 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />s1 Containment Sump Floor and Equipment Level Monitor Containment Floor and Equipment Sump Level Control Room A 1 gal/min leak is detectable in approximately 1 hour2.
Incore Instrument Sump Level Alarm Incore Instrument Sump Level Control Room A leak of 1 gal/min is detectable5.
Containment Ventilation Unit Condensate Drain Tank Level Monitor Water level in tank Control Room 1 gal/min is detectable in approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of leakage reaching the tank.
Deleted Per 2006 Update Scintillation Detector Noble gas activity in condenser air ejector effluent Control Room A leak of < 30 gal/day is detectable3.
McGuire Nuclear Station UFSAR Table 5-30 (Page 2 of 2)
(13 APR 2020)
Sensing Device Parameter Monitored Readout Location Sensitivity Scintillation Detector Steam Generator N-16 leakage detection monitor Control Room 100 to 105 gallons per day leakage to individual steam generator indication for power levels from 40% to 100% rated power.
Volume Control Water Level Detector Volume Control Tank water level Control Room A 1 gal/min leak is detectable4.
Note:
- 1. The sensitivities indicated assume instantaneous mixing, reference MCC-1223.03-00-0044.
- 2. Sensitivity is based on Reg. Guide 1.45 criteria to detect a 1 gpm leak in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, using OAC rate of change alarm for each sump level transmitter. The following are assumed: 1 gpm leak is detectable in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after leakage has reached the sump, embedded piping is considered in the alarm setpoint, leakage is cumulative between sumps A and B.
- 3. The sensitivity indicated is based upon the detector sensitivity in Table 11-28. The count rate of the detector is proportional to leak rate and coolant activity. With a typical background of 2.5 x 102 CPM, any combination of defects and leaks resulting in a count rate approximately 3 sigma above detector background is detectable.
- 4. The sensitivity indicated is based upon the change in water level in the volume control tank, pressurizer level, charging rate, letdown rate, reactor coolant pump seal water injection rate, and after condensate has reached the tank.
- 5. Leakage of 1 gal/min is detectable within four hours after reaching the sump.
McGuire Nuclear Station UFSAR Table 5-31 (Page 1 of 1)
(14 OCT 2000)
Table 5-31. Deleted Per 1996 Update
McGuire Nuclear Station UFSAR Table 5-32 (Page 1 of 1)
(22 APR 2017)
Table 5-32. Reactor Vessel Design Parameters Design/Operating Pressure, psig 2485/2235 Design Temperature, °F 650 Overall Height of Vessel and Closure Head, ft-in, (Bottom Head O.D. to top of control Rod Mechanism Adaptor) 43-9 5/8 Thickness of Insulation, min., in.
3 Deleted per 2017 Update Number of Reactor Closure Head Studs 54 Diameter of Reactor Closure Head Studs, in.
7 ID of Flange, in.
205 ID at Shell, in.
173 Inlet Nozzle ID, in.
27-1/2 Outlet Nozzle ID, in.
29 Clad Thickness, min., in.
5/32 Vessel Belt-Line Thickness, min., in.
8-1/2 Lower Head Thickness, min., in.
5-1/2 Closure Head Thickness, min., in.
6-1/2
McGuire Nuclear Station UFSAR Table 5-33 (Page 1 of 2)
(13 OCT 2018)
Table 5-33. Surveillance Capsule Removal Schedule [Note 1]
Unit Capsule Withdrawal End of Cycle (EOC)
Lead Factor
[Note 4]
Withdrawal EFPY (from plant startup Fluence (x1019 n/cm2)
[Note 4]
Reference Deleted Per 2014 Update Unit 1 U
1 4.96 1.09 0.382 WCAP-10786 Unit 1 X
5 4.83 4.30 1.40 WCAP-12354 Unit 1 V
8 4.15 7.24 1.93 WCAP-13949 Unit 1 [Note 2]
Z 8
4.75 7.24 2.21 WCAP-13949 Unit 1 Y
11 4.20 10.21 2.65 WCAP-14993 Unit 1 [Note 3]
[Note 5]
W 18 4.92 19.22 5.08 WCAP-17014-NP Deleted Row per 2017 Update Deleted per 2014 Update Unit 2 V
1 4.16 1.03 0.302 WCAP-11029 Unit 2 X
5 4.81 4.16 1.38 WCAP-12556 Unit 2 U
7 4.66 6.05 1.90 WCAP-13516 Unit 2 [Note 2]
Y 8
4.03 7.18 1.94 WCAP-14231 Unit 2 [Note 2]
Z 8
4.60 7.18 2.21 WCAP-14231 Unit 2 [Note 6]
W 10 4.64 9.44 2.82 WCAP-14799 Deleted Row per 2017 Update Deleted Per 2014 Update Note 1:
All in-vessel surveillance capsules have been withdrawn. Thus, this table is a summary of all past surveillance capsule withdrawals.
Note 2:
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.
Note 3:
The management of Capsule W is controlled by Calvert Cliffs Unit 1. It was withdrawn at EOC 18 (19.22 EFPY) for evaluation. The dosimetry and weld specimens were analyzed. The plate specimens were stored at Westinghouse and are available for testing or additional irradiation if ever deemed necessary.
Note 4:
Capsule fluence and lead factors were updated in WCAP-17455.
Note 5:
15 specimens from Unit 1 Capsule W have been installed in the Shearon Harris Reactor Vessel as part of the EPRI PWR Supplemental Program (PSSP). PSSP Capsule CQL-P is scheduled for removal in 2028.
McGuire Nuclear Station UFSAR Table 5-33 (Page 2 of 2)
(13 OCT 2018)
Unit Capsule Withdrawal End of Cycle (EOC)
Lead Factor
[Note 4]
Withdrawal EFPY (from plant startup Fluence (x1019 n/cm2)
[Note 4]
Reference Note 6:
12 specimens from Unit 2 Capsule W have been installed in the Shearon Harris Reactor Vessel as part of the EPRI PWR Supplemental Surveillance Program (PSSP). PSSP Capsule CQL-P is scheduled for removal in 2028.
McGuire Nuclear Station UFSAR Table 5-34 (Page 1 of 1)
(14 OCT 2000)
Table 5-34. Reactor Vessel Quality Assurance Program RT1 UT1 PT1 MT1 Forgings & Tubes
- 1. Flanges yes yes
- 2. Studs yes yes
- 3. CRDM and UHI Adaptors yes yes
- 4. CRDM and UHI Adaptor Tubes yes yes
- 5. Instrumentation Tube yes yes
- 6. Main Nozzles yes yes Plates yes yes Weldments
- 1. Main Seam yes Yes3 yes
- 2. CRDM Head Adaptor to Head Connection yes
- 3. UHI Adaptor to Head Attachments yes yes yes
- 4. Instrumentation Tube Connection yes
- 5. Main Nozzles yes Yes3 yes
- 6. Cladding Yes2 yes
- 7. Nozzle-safe ends (weld deposit) yes Yes3 yes
- 8. CRDM Head Adaptor Forging to Head Adaptor Tube yes yes
- 9. UHI Adaptor Forging to Adaptor Tube yes Yes3 yes
- 10. All Ferritic Welds Accessible After Hydrotest yes
- 11. All Non-ferritic Welds Accessible After Hydrotest yes
- 12. Seal Ledge yes
- 13. Head Lift Lugs yes
- 14. Core Pad Welds yes yes Yes Note:
- 2. UT of Clad Bond-to-base Metal
- 3. UT map for Section IX
McGuire Nuclear Station UFSAR Table 5-35 (Page 1 of 1)
(14 OCT 2000)
Table 5-35. Reactor Coolant Pump Design Parameters Design pressure, psig 2485 Design temperature, °F 650 Capacity per pump, gpm 99,000 Developed head, ft.
288 NPSH required, ft.
245 Suction temperature, °F 557.8 RPM 1186 Discharge nozzle, ID, inches 27-1/2 Suction nozzle, ID, inches 31 Overall unit height, ft.-in.
27'-7.2" Water volume, ft3 80 Total rotating inertia, ft-lb.
95,000 Weight, dry, lb.
201,900 Motor Type AC induction, single speed totally enclosed water cooled Power, H.P.
7000 Voltage, volts 6600 Insulation class minimum B Frequency, Hz 60 Phase 3
Starting Current, amps 3000 Input, hot reactor coolant, kw 5250 +/- 180 Input, cold reactor coolant, kw 6850 +/- 240 Seal water injection, gpm 8
Seal water return, gpm 3
McGuire Nuclear Station UFSAR Table 5-36 (Page 1 of 1)
(14 OCT 2000)
Table 5-36. Reactor Coolant Pump Quality Assurance Program RT1 UT1 PT1 MT1 Castings yes yes Forgings
- 1. Main Shaft yes yes
- 2. Main Studs yes yes
- 1. Circumferential yes yes
- 2. Instrument Connections yes Note:
McGuire Nuclear Station UFSAR Table 5-37 (Page 1 of 1)
(14 OCT 2000)
Table 5-37. Steam Generator Design Data Design pressure, reactor coolant side, psig 2485 Design pressure, steam side, psig 1185 Design temperature, reactor coolant side, °F 650 Design temperature, steam side °F 600 Total heat transfer surface area ft2 79,800 Maximum moisture, carryover, wt percent 0.25 Overall height, ft-in.
68' - 1 3/8" Number of U-tubes 6633 Tube wall thickness, nominal, in.
0.040 U-tube outer diameter, in.
0.6875 Number of manways 3
Number of handholes 10 ID of handholes, in.
6 ID of manways, in.
21
McGuire Nuclear Station UFSAR Table 5-38 (Page 1 of 1)
(14 OCT 2000)
Table 5-38. Steam Generator Quality Assurance Program RT1 UT1 PT1 MT1 ET1 Tube Sheet
- 1. Forging yes yes
- 2. Cladding yes Yes2 Channel Head
- 1. Forging yes yes
- 2. Cladding yes Secondary Shell & Head
- 1. Plates yes Tubes yes yes Nozzles (Forgings) yes yes Weldments
- 1. Shell, longitudinal yes yes yes
- 2. Shell, circumferential yes yes yes
- 3. Cladding (Channel Head-Tube Sheet joint cladding restoration) yes
- 4. Steam and Feedwater Nozzle to shell yes yes yes
- 5. Support brackets yes
- 6. Tube to tube sheet yes
- 7. Instrument connections (primary and secondary) yes yes
- 8. Temporary attachments after removal yes
- 9. After hydrostatic test (all welds and complete channel head - where accessible) yes
- 10. Nozzle safe ends (if forgings) yes yes
- 1. RT - Radiographic RT - Radiographic UT - Ultrasonic PT - Dye Penetrant MT - Magnetic Particle ET - Eddy Current
- 2. Weld Deposit Areas Only
McGuire Nuclear Station UFSAR Table 5-39 (Page 1 of 1)
(14 OCT 2000)
Table 5-39. Reactor Coolant Piping Design Parameters Reactor inlet piping, ID, in.
27-1/2 Reactor inlet piping, nominal wall thickness, in.
2.30 Reactor outlet piping, ID, in.
29 Reactor outlet piping, nominal wall thickness, in.
2.42 Coolant pump suction piping, ID, in.
31 Coolant pump suction piping, nominal wall thickness, in.
2.58 Pressurizer surge line piping, ID, in.
11.50 Pressurizer surge line piping, nominal wall thickness, in.
1.25 Water volume, all loops and surge line, ft.3 1030 Design/operating pressure, psig 2485/2235 Design Temperature, °F 650 Design Temperature, pressurizer surge line, °F 680 Design pressure, pressurizer relief line, psig From pressurizer to safety valve 2485 psig, 680°F Design Temperature, pressurizer relief lines, °F From safety valve to pressurizer relief tank 500 psig, 500°F
McGuire Nuclear Station UFSAR Table 5-40 (Page 1 of 1)
(14 OCT 2000)
Table 5-40. Reactor Coolant Piping Quality Assurance Program RT1 UT1 PT1 Fittings and Pipe (Castings) yes yes Fittings and Pipe (Forgings) yes yes Weldments
- 1. Circumferential yes yes
- 2. Nozzle to runpipe (except no RT for nozzles less than 4 inches) yes yes yes
- 3. Instrument connections yes yes Note:
McGuire Nuclear Station UFSAR Table 5-41 (Page 1 of 1)
(10 OCT 2009)
Table 5-41. 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
<450 Reactor Coolant System initial temperature, °F
<350 Component cooling water design, temperature, °F 95 Cooldown time, hours after initiation of RHRS operation
~16 Reactor Coolant System temperature at end of cooldown, °F 140 Decay heat generation at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after Reactor shutdown, BTU/hr 77.21 x 106 Reactor Coolant System flow rate during refueling mode, gpm 1000, and as required to maintain RCS temperature 140°F
McGuire Nuclear Station UFSAR Table 5-42 (Page 1 of 1)
(14 OCT 2000)
Table 5-42. 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 Residual Heat Exchanger Number 2
Design Heat Removal Capacity, BTU/hr.
34.15 x 106 Tube-Side Shell-Side Design Pressure, psig 600 150 Design Temperature, °F 400 200 Design Flow, lb/hr.
1.48 x 106 2.48 x 106 Inlet Temperature, °F 137 95 Outlet Temperature, °F 114 108.8 Material Austenitic Stainless Steel Carbon Steel Fluid Reactor Coolant Component Cooling Water
McGuire Nuclear Station UFSAR Table 5-43 (Page 1 of 1)
(14 OCT 2000)
Table 5-43. Pressurizer Design Data Design Pressure 2485 Design Temperature, °F 680 Surge Line Nozzle Diameter, in 14 Heatup Rate of Pressurizer Using Heaters Only, °F/hr.
55 Internal Volume, cu. ft.
1800
McGuire Nuclear Station UFSAR Table 5-44 (Page 1 of 1)
(11 NOV 2006)
Table 5-44. Pressurizer Quality Assurance Program RT1 UT1 PT1 MT1 Heads
- 1. Plates yes
- 2. Cladding yes Shell
- 1. Plates yes Cladding yes Heaters
- 1. Tubing 3 yes yes
- 2. Centering of element yes Nozzle (Forgings) yes yes2 yes2 Weldments
- 1. Shell, longitudinal yes yes
- 2. Shell, circumferential yes yes
- 3. Cladding yes
- 4. Nozzle Safe End (if forging) yes yes4 yes4
- 5. Instrument Connections yes
- 6. Support Skirt yes yes
- 7. Temporary Attachments (after removal) yes
- 8. All welds and plate heads after hydrostatic tests Yes Note:
- 1. RT - Radiographic Deleted Per 2006 Update UT - Ultrasonic PT - Dye Penetrant MT - Magnetic Particle
- 4. Weld Overlay Installation, UT and PT.
McGuire Nuclear Station UFSAR Table 5-45 (Page 1 of 1)
(14 OCT 2000)
Table 5-45. Pressurizer Relief Tank Design Data Design Pressure, psig 100 Rupture Disc Release Pressure, psig 100+/-5%
Design Temperature, °F 340 Total Rupture Disc Relief Capacity, lb/hr at 100 psig 1.6 x 106
McGuire Nuclear Station UFSAR Table 5-46 (Page 1 of 1)
(14 OCT 2000)
Table 5-46. Reactor Coolant System Boundary Valve Design Parameters Design/Normal Operating Pressure, psig 2485/2235 Pre-Operational Hydrotest, psig 3107 Design Temperature, °F 650
McGuire Nuclear Station UFSAR Table 5-47 (Page 1 of 1)
(14 OCT 2000)
Table 5-47. Pressurizer Valves Design Parameters Parameters Pressurizer Spray Control Valves Number 2
Design pressure, psig 2485 Design temperature, °F 650 Design flow for valves full open, each, gpm 450 Pressurizer Safety Valves Number 3
Maximum relieving capacity, ASME rated flow, lb/hr (per valve) 420,000 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, °F 680 High pressure setpoint, psig 2335 Relieving capacity, lb/hr (per valve) 210,000 Fluid Saturated Steam Low pressure setpoint, psig (NC-32B and NC-34A only) 380 Relieving capacity, gpm (per valve) 1060 Fluid Water @60°F
McGuire Nuclear Station UFSAR Table 5-48 (Page 1 of 1)
(14 OCT 2000)
Table 5-48. Component Supports Loading Combinations and Code Requirements Loading Combination Code Or Stress Requirements
- 2. DL + OL + OBE AISC with Allowable Stresses of Fs. Allowable stresses in concrete and reinforcement are in accordance to Chapter 10 of the AC1-318, 1962 Code, WSD.
- 4. DL + OL + SSE + LOCA ACI-318, 1963 USD with no load factors applied to loadings. For allowable stresses in structural steel see Note 1.
DL
=
Dead Load, including own weight of the support.
=
Normal Operating Load: These loads are associated with plant operations in addition to weight of permanent equipment.
=
Live Load, including construction loads.
=
Operating Basis Earthquake load.
=
Safe Shutdown Earthquake load LOCA
=
Accident loads including reactions due to pipe rupture and thermal loads.
=
Specifications for Design, Fabrication and Erection of Structural Steel Buildings, Seventh Edition, 1969.
WSD
=
Working Stress Design USD
=
Ultimate Strength Design.
Fs
=
Steel allowable stresses as specified in AISC Part 1.
Fy
=
Yield stress of structural steel.
Type of Stress Allowable Stress Tension, Compression and Bending 0.9 Fy Shear 0.55 Fy Compression with Buckling 1.7 Fs Note:
- 1. For loading combinations 3 and 4 which are ultimate loading conditions, the allowable stresses for the structural steel are as follows:
McGuire Nuclear Station UFSAR Table 5-49 (Page 1 of 4)
(05 APR 2011)
Table 5-49. Design Transients for the Reactor Coolant System Including the BWI Replacement Steam Generators (RSGs)
Design Transients Allowable Occurrences(1)
Unit 1 Unit 2 Normal (Level A) Transients
(**) Plant Heatup 200 200
(**) Plant Cooldown 200 200 RHR Suction 200 200 RHR Injection 200 200 Refueling 80 80
(**) Plant Loading at 5%/min (15% to 100%)
13,200 13,200
(**) Plant Unloading at 5%/min (100% to 15%)
13,200 13,200 Small Step Load Increase 15 - 25%
300 300
(**) 90 - 100%
2000 2000 Small Step Load Decrease 25 - 15%
300 300
(**) 100 - 90%
2000 2000
(**) Large Step Load Decrease (100%-5%) with Steam Dump 200(2) 200(2)
Feedwater Cycling at No Load 2000 2000
(**) Steady State Fluctuations Infinite(3)
Infinite(3)
Plant Loading and Unloading between 0% and 15%
power 750 Load(2) 750 Load(2) 750 Unload(2) 750 Unload(2)
Loop Out of Service Normal Pump Shutdown 80 80 Normal Pump Startup 70 70 Boron Concentration Equalization 26,400 26,400 Reactor Coolant Pump Startup/Shutdown Cold Conditions 750 750 Hot Conditions 3750 3750 RCS Venting Affected Loops 0(7) 0(7)
Unaffected Loops 0(7) 0(7)
McGuire Nuclear Station UFSAR Table 5-49 (Page 2 of 4)
(05 APR 2011)
Design Transients Allowable Occurrences(1)
Unit 1 Unit 2 Vacuum Refill 480 480 Normal/Charging/Letdown Shutoff and Return to Service 60 60 Letdown Trip with Prompt Return to Service 200 200 Letdown Trip with Delayed Return to Service 20 20 Charging Trip with Prompt Return to Service 20 20 Charging Trip with Delayed Return to Service 20 20 Charging Flow 50% Decrease 24,000 24,000 Charging Flow 50% Increase 24,000 24,000 Letdown Flow 40% Decrease and Return to Normal 2000 2000 Letdown Flow 60% Increase 24,000 24,000 Letdown Shutoff and Momentary Excess Letdown 100 100 Switch of Charging Pump Suction to FWST and Back 180 180 Auxiliary Spray Actuation during Heatup 200 200 Auxiliary Spray Actuation during Cooldown 200 200 Pressurizer Relief Valve Operation 100 100 LTOP Pressurizer PORV Operation 200 200 Pzr Blk/Drn Vlv Operation 400 400 Pressurizer Safety Valve Operation 40 40 Upset (Level B) Transients
(**) Loss of Load without Immediate Turbine or Reactor Trip 80 80
(**) Loss of Power (Blackout with Natural Circulation) 40 40
(**) Loss of Flow in One Loop 80 80 Reactor Trip from Full Power
(**) Nominal 230 230
(*) Inadvertent Cooldown 0(7) 10
(*) Inadvertent RCS Depressurization 0(7) 20 Inadvertent Startup of an Inactive Loop 15 15 Control Rod Drop 0(7) 0(7)
Operating Basis Earthquake (OBE)
McGuire Nuclear Station UFSAR Table 5-49 (Page 3 of 4)
(05 APR 2011)
Design Transients Allowable Occurrences(1)
Unit 1 Unit 2 RSG 30 (20 cycles/occurrence)
Reactor Coolant Pump, Pressurizer 20 (20 cycles/occurrence)
Reactor Vessel 50 cycles Piping See Section 3.7.3.1 Excessive Feedwater Flow 45 45 Excessive Bypass Feedwater 0(7) 0(7)
Cold Feedwater to Dry, Pressurized RSG 2
2 Complete Loss of Flow 0(7) 0(7)
(**) Inadvertent Auxiliary Spray 10 10 Inadvertent SI Accumulator Blowdown during Plant Cooldown 4
4 High Head Safety Injection 17 17 Boron Injection 25 32 Faulted (Level D) Transients(4)
(**) Reactor Coolant Pipe Break (Large LOCA) 1(5) 1(5)
(**) Large Steam Line Break 1
1 Safe Shutdown Earthquake (SSE) 1 (10 cycles)
Steam Generator Tube Rupture 8
8 Cold Feedwater to Dry, Depressurized RSG 1
1 High Head Safety Injection 2
2 Boron Injection 2
2 Test Conditions
(**) Turbine Roll Test 10(6) 10(6)
(**) Primary Side Hydrostatic Test 5
5 Secondary Side Hydrostatic Test 10 10
(**) Primary Side Leakage Test 50 50 Secondary Side Leakage Test 80 80 Tube Leak Test Secondary Side Pressure 200 600 600 400 300 300 600 180 180
McGuire Nuclear Station UFSAR Table 5-49 (Page 4 of 4)
(05 APR 2011)
Design Transients Allowable Occurrences(1)
Unit 1 Unit 2 840 80 80 Notes:
- 1. Allowed occurrences is the minimum controlling analyzed or postulated number of occurrences for the plant design life, including period of extended operation, considering ASME Section III and XI limits.
- 2. The RSGs are designed for several cycles of swapping of main feedwater supply between the auxiliary and the main feedwater nozzles with and without tempering flow during plant loading and unloading and large step load decrease. The design is for 100 swaps without tempering flow and 750 with tempering flow.
- 3. Pressurizer surge line is analyzed for 150,000 initial fluctuations and 3,000,000 random fluctuations. Various NSSS components are analyzed for 150,000 to 30,000,000 fluctuations of various minor amplitudes.
- 4. In accordance with the ASME Code, faulted conditions are not included in fatigue evaluations.
Some components analyzed/piping segments are individually qualified for faulted events which are not listed because the unit as a whole is therefore not qualified for such.
- 5. This condition has been excused for RCS pressure boundary qualification by Leak Before Break.
Some components analyzed after LBB was approved were not analyzed for this condition.
- 6. Since Turbine Roll Test is performed only prior to fuel load, the RSGs are not designed for this event.
- 7. Some components are not designed for this event, thus it is not considered to have been part of the analysis for any component. If it occurs, it will be dispositioned, by, for example, considering it as an another, actually analyzed event.
(**,*) Transient specified for Reactor Vessel design for (Units 1&2, Unit 2 only) [Section 5.4.4.4]
McGuire Nuclear Station UFSAR Table 5-50 (Page 1 of 2)
(14 OCT 2000)
Table 5-50. Reactor Coolant System Pressure Isolation Valves Valve Number Function NI60 Accumulator Discharge NI71 Accumulator Discharge NI59 Accumulator Discharge NI70 Accumulator Discharge NI82 Accumulator Discharge NI94 Accumulator Discharge NI81 Accumulator Discharge NI93 Accumulator Discharge NI134 Safety Injection (Hot Leg)
NI159 Safety Injection (Hot Leg)
NI156 Safety Injection (Hot Leg)
NI128 Safety Injection (Hot Leg)
NI124 Safety Injection (Hot Leg)
NI160 Safety Injection (Hot Leg)
NI157 Safety Injection (Hot Leg)
NI126 Safety Injection (Hot Leg)
NI129 Safety Injection (Hot Leg)
NI125 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)
McGuire Nuclear Station UFSAR Table 5-50 (Page 2 of 2)
(14 OCT 2000)
Valve Number Function ND2A Residual Heat Removal
McGuire Nuclear Station UFSAR Table 5-51 (Page 1 of 2)
(24 APR 2014)
Table 5-51. RT PTS Calculations for McGuire Unit 1 Reactor Vessel Materials at 54 EFPY Material CF
(ºF)
Fluence @
54 EFPY (x1019 n/cm2)
(ºF)
(ºF)
M
(ºF)
(°F)
Upper Shell Plate B5453-2 99.1 0.0547 0.3072 15 30.4 30.4 76 Upper Shell Plate B5011-2 65 0.0547 0.3072 27 20.0 20.0 67 Upper Shell Plate B5011-3 89.8 0.0547 0.3072 0
27.6 27.6 55 Intermediate Shell Plate B5012-1 74.2 2.56 1.2521 34 92.9 34 161 Using Surveillance Capsule Data 63.5 2.56 1.2521 34 79.5 17 131 Intermediate Shell Plate B5012-2 100.3 2.56 1.2521 0
125.6 34 160 Intermediate Shell Plate B5012-3 74.9 2.56 1.2521
-13 93.8 34 115 Lower Shell Plate B5013-1 99.1 2.57 1.2531 0
124.2 34 158 Lower Shell Plate B5013-2 65 2.57 1.2531 30 81.4 34 145 Lower Shell Plate B5013-3 65 2.57 1.2531 15 81.4 34 130 Upper Shell Longitudinal Weld Seams 1-442A, B, C 201.3 0.0451 0.2767
-50 55.7 55.7 61 Using Surveillance Capsule Data 155.2 0.0451 0.2767
-50 43.0 28 21 Upper Shell to Intermediate Shell Circumferential Weld Seam 8-442 170.5 0.0547 0.3072
-56 52.4 62.4 59
McGuire Nuclear Station UFSAR Table 5-51 (Page 2 of 2)
(24 APR 2014)
Material CF
(ºF)
Fluence @
54 EFPY (x1019 n/cm2)
(ºF)
(ºF)
M
(ºF)
(°F)
Intermediate Shell Longitudinal Weld Seams 2-442A, B, C 201.3 2.13 1.2055
-50 242.7 56 249 Using Surveillance Capsule Data 155.2 2.13 1.2055
-50 187.1 28 165 Intermediate Shell to Lower Shell Circumferential Weld Seam 9-442 37.5 2.47 1.2432
-70 46.6 46.6 23 Lower Shell Longitudinal Weld Seams 3-442 A, B,C 208.2 2.13 1.2055
-50 251.40 56 257 Using S/C Data from Diablo Canyon 2 186.4 2.13 1.2055
-50 224.7 28 203 Deleted Per 2014 Update NOTE: The information above was taken from WCAP-17455.
McGuire Nuclear Station UFSAR Table 5-52 (Page 1 of 1)
(24 APR 2014)
Table 5-52. RT PTS Calculations for McGuire Unit 2 Reactor Vessel Materials at 54 EFPY Material CF
(ºF)
Fluence @
54 EFPY (x1019 n/cm2)
(ºF)
(ºF)
M
(ºF)
(°F)
Upper Shell Forging 06 123.9 0.0711 0.3521 25 43.6 34 103 Deleted Per 2014 Update Intermediate Shell Forging 05 117.2 2.41 1.2370
-4 145.0 34 175 Using Surveillance Capsule Data 85.5 2.41 1.2370
-4 105.7 17 119 Lower Shell Forging 04 115.8 2.48 1.2442
-30 144.1 34 148 Bottom Head Ring 03 37 0.336 0.6997 15 25.9 25.9 67 Upper Shell to Intermediate Shell Circumferential Weld Seam W06 82.9 0.0711 0.3521 10 29.2 29.2 68 Intermediate Shell to Lower Shell Circumferential Weld Seam W05 52.7 2.34 1.2296
-68 64.8 56 53 Using Surveillance Capsule Data 27.1 2.34 1.2296
-68 33.3 28
-7 Lower Shell to Bottom Head Ring Weld W04 41 0.336 0.6997 10 28.7 28.7 67 NOTE: The information above was taken from WCAP-17455.
McGuire Nuclear Station UFSAR Table 5-53 (Page 1 of 2)
(24 APR 2014)
Table 5-53. Evaluation of Upper Shelf Energy for McGuire Unit 1 Reactor Vessel Materials at 54 EFPY Material Weight % of Cu 1/4 T EOLE Fluence (x1019 n/cm2)
Unirradiated USE (ft-lb)
Projected USE Decrease (%)
Projected USE
@ 54 EFPY (ft-lb)
Upper Shell Plate B5453-2 0.14 0.033 72.4 11 64.4 Upper Shell Plate B5011-2 0.10 0.033 68.3 8.6 62.4 Upper Shell Plate B5011-3 0.13 0.033`
94.7 9.8 85.4 Intermediate Shell Plate B5012-1 0.11 1.525 101 23 77.8 Using Surveillance Capsule Data 0.11 1.525 101 11 89.9 Intermediate Shell Plate B5012-2 0.14 1.525 105 26 77.7 Intermediate Shell Plate B5012-3 0.11 1.525 112 23 86.2 Lower Shell Plate B5013-1 0.14 1.531 95 26 70.3 Lower Shell Plate B5013-2 0.10 1.531 115 21 90.9 Lower Shell Plate B5013-3 0.10 1.531 103 21 81.4 Upper Shell Longitudinal Weld Seams 1-442A, B, C 0.199 0.027 112 15 95.2 Using Surveillance Capsule Data 0.199 0.027 112 19 90.7 Upper Shell to Intermediate Shell Circumferential Weld Seam 8-442 0.183 0.033 109 15 92.7 Intermediate Shell Longitudinal Weld Seams 2-442 A, B, C 0.199 1.269 112 36 71.7 Using Surveillance Capsule Data 0.199 1.269 112 46 60.5 Intermediate Shell to Lower Shell Circumferential Weld Seam 9-442 0.051 1.472 143 21 113.0 Lower Shell Longitudinal Weld Seams 3-442 A, B, C 0.213 1.269 124 38 76.9
McGuire Nuclear Station UFSAR Table 5-53 (Page 2 of 2)
(24 APR 2014)
NOTE: The information above was taken from WCAP-17455.
McGuire Nuclear Station UFSAR Table 5-54 (Page 1 of 1)
(24 APR 2014)
Table 5-54. Evaluation of Upper Shelf Energy for McGuire Unit 2 Reactor Vessel Materials at 54 EFPY Material Weight % of Cu 1/4 T EOLE Fluence (x1019 n/cm2)
Unirradiated USE (ft-lb)
Projected USE Decrease (%)
Projected USE @ 54 EFPY (ft-lb)
Upper Shell Forging 06 0.16 0.043 98 12 86.2 Intermediate Shell Forging 05 0.153 1.450 94 27 68.6 Using Surveillance Capsule Data 0.153 1.450 94 23 72.4 Lower Shell Forging 04 0.15 1.492 141 27 102.9 Bottom Head Ring 03 0.06 0.202
>71 13
>61.8 Upper Shell to Intermediate Shell Circumferential Weld W06 0.11 0.043
>71 12
>62.5 Intermediate Shell to Lower Shell Circumferential Weld W05 0.039 1.408 132 21 104.3 Using Surveillance Capsule Data 0.039 1.408 132 3.4 127.5 Lower Shell to Bottom Head Ring Weld W04 0.03 0.202 99 13 86.1 NOTE: The information above was taken from WCAP-17455.
McGuire Nuclear Station UFSAR Table 5-55 (Page 1 of 1)
(11 NOV 2006)
Table 5-55. Summary of Reactor Coolant System Leakage Detection Instrumentation Exceptions to Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems (Rev. 0)
Regulatory Position Exception C.2 Leakage to the primary reactor containment from unidentified sources should be collected and the flow rate monitored with an accuracy of one gallon per minute or better.
Incore sump alarm will detect a 1 gpm input 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.
C.5 The sensitivity and response time of each leakage detection system in regulatory position 3 above employed for unidentified leakage should be adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour.
Exception taken for containment particulate radiation monitor and incore sump level alarm.
The particulate radiation monitor sensitivity will be 10-9 µCi/cc. The particulate monitor alarm setting will 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 CFAE and CVUCDT sensitivity of 1 gpm after leakage has reached the sump/tank.
C.7 Indicators and alarms for each leakage detection system should be provided in the main control room. Procedures for converting various indications to a common leakage equivalent should be available to the operators. The calibration of the indicators should account for needed independent variables.
Exception taken for incore sump indication in the control room - alarm only.
The particulate radiation monitor and incore sump will alarm during the presence of a leak but are not converted to a leakage equivalent (e.g. gpm).
C.8 The leakage detection systems should be equipped with provisions to readily permit testing for operability and calibration during plant operation.
Exception taken for incore sump level alarm for testing and calibration during plant operation.
McGuire Nuclear Station UFSAR Table 5-56 (Page 1 of 1)
(22 APR 2017)
Table 5-56. Ex-Vessel Dosimetry Capsule Summary Table Unit Capsule Ex-Vessel Azimuthal Location
[Note 1]
Axial Location Installed Cycle Removed Cycle Best Estimate Flux (n/cm2-s)
Reference Unit 1 A
0° Core Midplane 12 12 3.68E+08 WCAP -17799-NP Unit 1 B
15° Core Midplane 12 12 5.64E+08 WCAP -17799-NP Unit 1 C
30° Core Midplane 12 12 6.00E+08 WCAP -17799-NP Unit 1 E
45° Core Midplane 12 12 5.77E+08 WCAP -17799-NP Unit 1 D
45° Core Top 12 12 2.36E+08 WCAP -17799-NP Unit 1 F
45° Core Bottom 12 12 2.48E+08 WCAP -17799-NP Unit 1 M
0° Core Midplane 13 22 3.68E+08 WCAP -17799-NP Unit 1 N
15° Core Midplane 13 22 5.31E+08 WCAP -17799-NP Unit 1 O
30° Core Midplane 13 22 5.63E+08 WCAP -17799-NP Unit 1 Q
45° Core Midplane 13 22 5.53E+08 WCAP -17799-NP Unit 1 P
45° Core Top 13 22 1.96E+08 WCAP -17799-NP Unit 1 R
45° Core Bottom 13 22 2.40E+08 WCAP -17799-NP Unit 2 G
0° Core Midplane 12 12 4.11E+08 WCAP-17767-NP Unit 2 H
15° Core Midplane 12 12 6.35E+08 WCAP-17767-NP Unit 2 I
30° Core Midplane 12 12 6.80E+08 WCAP-17767-NP Unit 2 K
45° Core Midplane 12 12 5.46E+08 WCAP-17767-NP Unit 2 J
45° Core Top 12 12 5.04E+08 WCAP-17767-NP Unit 2 L
45° Core Bottom 12 12 8.84E+07 WCAP-17767-NP Unit 2 S
0° Core Midplane 13 21 4.02E+08 WCAP-17767-NP Unit 2 T
15° Core Midplane 13 21 5.52E+08 WCAP-17767-NP Unit 2 U
30° Core Midplane 13 21 5.86E+08 WCAP-17767-NP Unit 2 W
45° Core Midplane 13 21 5.93E+08 WCAP-17767-NP Unit 2 X
45° Core Top 13 21 4.97E+08 WCAP-17767-NP Unit 2 V
45° Core Bottom 13 21 6.22E+07 WCAP-17767-NP