ML20189A081

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8 to Updated Final Safety Analysis Report, Chapter 5, Appendix 5A, Tables
ML20189A081
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 06/29/2020
From:
Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
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ML20189A066 List:
References
RA-20-0136
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Oconee Nuclear Station UFSAR Appendix 5A. Tables Appendix 5A. Tables

Oconee Nuclear Station UFSAR Table 5-1 (Page 1 of 1)

Table 5-1. Reactor Coolant System Pressure Settings Pressure, Capacity, psig lb/hr, total Design Pressure 2500 Pressurizer Code Safety Valves 2500 667,000 High Pressure Trip 2355(1)

Pressurizer Electromatic Relief Valve Open 2450(1) 107,000 Close 2400(1)

High Pressure Alarm 2255(1)

Pressurizer Spray Valve Open 2205(1)

Close 2155(1)

Operating Pressure 1 2155 Low Pressure Alarm 2055 Low Low Pressure Alarm 1920(1)

Low Pressure Trip 1800(1)

Hydrotest Pressure 3125 Note:

1. At sensing nozzle on reactor outlet pipe.

(31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-2 (Page 1 of 3)

Table 5-2. Transient Cycles for RCS Components Except Pressurizer Surge Line Component Transient Design Exceptions Number Transient Description (ASME Category) Cycles (See notes) 1A Heatup from 70°F to 8% Full Power (Normal)

Total 360 (1) 1B First 25% of Plant Life Cooldown from 8% Full Power (Normal) 90 Last 75% of Plant Life Cooldown from 8% Full Power (Normal) 270 Total 360 (1) 2 Power Change 0 to 15% and 15 to 0% (Normal) 1440 3 Power Loading 8 to 100% Power (Normal) 18,000 (7) 4 Power Unloading 100 to 8% Power (Normal) 18,000 (7) 5 10% Step Load Increase (Normal) 8,000 6 10% Step Load Decrease (Normal) 8,000 7 Step Load Reduction (100 to 8% Power) (Upset)

Resulting from turbine trip 160 Resulting from electrical load rejection 150 Total 310 8 Reactor Trip (Upset)

Type A 40 Type B 160 Type C 90 Trip included in transient numbers 11, 15, 16, 17 122 and 21 Total3 412 Manual Actuation of High Presure Injection System after Reactor Trip 70 (2) (3) 9 Rapid Depressurization (Upset) 40 (2) 10 Change of Reactor Coolant Flow (Upset) 412 11 Rod Withdrawal Accident (Upset) 40 12 Hydrotests (Test)

All RCS components 10 (7) 13 Steady-state Power Variations (Normal) 14 Control Rod Drop (Upset) 60 15 Loss of Station Power (Upset) 40 (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-2 (Page 2 of 3)

Component Transient Design Exceptions Number Transient Description (ASME Category) Cycles (See notes) 16 Steam Line Failure (Faulted) 1 17A Loss of Feedwater to One Steam Generator (Upset) 30 17B Stuck Open Turbine Bypass Valve (Emergency) 10 18 Loss of Feedwater Heater (Upset) 620 19 Feed and Bleed Operations (Normal) 4,000 (3) 20 Miscellaneous (Normal)

Miscellaneous A 30,000 (3)

Miscellaneous B 20,000 Miscellaneous C 4x106 21 Loss of Coolant (Faulted) 1 (4) 22 Test Transients (Test)

High Pressure Injection System 40 (3)

Core flooding check valve 240 Component and General Flaw Limiting Transient Total Allowable Location Transient Cycles Deleted row(s) per 2004 update Deleted row(s) per 2003 update (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-2 (Page 3 of 3)

Component Transient Design Exceptions Number Transient Description (ASME Category) Cycles (See notes)

Note:

1. Certain components have flaw tolerance evaluations as allowed by ASME Section XI (Refer to Section 5.2.2 and 5.2.3.12.4) that assume a reduced number of heatup and cooldown cycles. The lowest of the reduced number of cycles is used as the limit for the number of unit heatups and cooldowns. These evaluations will be updated, the flaws will be reexamined, or the flaws will be removed if the reduced number of transient cycles becomes limiting. A tabulation of the evaluations is presented below.
2. In order to analytically demonstrate a usage factor of less than 1.0, certain welds associated with the Emergency HPI nozzles have been qualified for fewer than the design number of cycles of the two, noted transients. The analysis uses a total of 29 cycles for the combined number of occurrences (i.e. the sum of the number of occurrences of Manual Actuation of HPI System after Reactor Trip Transient 8) and the number of occurrences of Rapid Depressurizations (Transient 9) can not exceed 29.

The AOTC (Allowable Operating Transient Cycle) monitoring program keeps track of the number of occurrences on each unit. The number of allowed transients has been reduced in the AOTC log books to limit the sum of these two transients to 29 for each unit.

3. Not applicable to replacement Steam Generators.
4. Deleted Per 2003 Update
5. Deleted Per 2004 Update
6. Deleted Per 2004 Update
7. The Reactor Vessel closure head assemblies are limited to 5000 power loading and unloading cycles and 15 hydrotests as discussed in Supplement B of OSS-0279.00 0003.

(31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-3 (Page 1 of 1)

Table 5-3. Stress Limits for Seismic, Pipe Rupture, and Combined Loads Case Loading Combination Stress Limits I Design loads + operating basis earthquake loads Pm 1.0 Sm (PL + Pb) 1.5 Sm II Design loads + safe shutdown earthquake loads Pm 1.2 Sm (PL + Pb) 1.2 (1.5 Sm)

III Design loads + pipe rupture loads Pm 1.2 Sm (PL + Pb) 1.2 (1.5 Sm)

IV Design loads + safe shutdown earthquake loads + Pm 2/3 Su pipe rupture loads (PL + Pb) 2/3 Su Where PL = Primary local membrane stress intensity 1

Pm = Primary general membrane stress intensity Pb = Primary bending stress intensity Sm = Allowable membrane stress intensity Su = Ultimate stress for unirradiated material at operating temperature Note:

1. All symbols have the same definition or connotation as those in ASME B&PV Code Section III, Nuclear Vessels.
2. All components will be designed to insure against structural instabilities regardless of stress levels.

(31 DEC 2010)

Oconee Nuclear Station UFSAR Table 5-4 (Page 1 of 1)

Table 5-4. Reactor Coolant System Component Codes Component Codes Addendum Reactor Vessel ASME III Class A Summer 19671 Replacement Reactor Vessel Head ASME III Class 1 1989, No addendum3,4 Pressurizer ASME III Class A Summer 19671 Reactor Coolant System Piping USAS B31.7 Errata through June5 1968 Feedwater Header USAS B31.1 1967 R. C. Pump Casings ASME III Class A Summer 1967 (not code stamped)

Safety and Relief Valves ASME III Art. 9 Summer 1967 Welding Qualifications ASME III and IX Summer 1967 Replacement Steam Generator (primary and ASME III Class 1 1989 No addendum secondary sides)

Note:

1. Welded joints tested in accordance with requirements of Article 7, Summer 1966 Addenda.
2. This table reflects original design/construction code information. Refer to UFSAR Section 5.2.2 for additional information on Reactor Coolant System Codes and Classifications.
3. Input Document for Replacement RVCHA Licensing and Safety Evaluation, Babcock

& Wilcox Canada, BWC Report No. 068S-LR-01 Rev 2; OM 201.R-0141.001.

4. History Docket for Closure Heads, Customer Spec.# OSS-0279.00-00-003, Babcock

& Wilcox Canada, BWC-Cont. 068S, 068S-01.

5. Reactor Coolant piping was requalified to the 1983 ASME code during the Steam Generator Replacement project.

(31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-5 (Page 1 of 2)

Table 5-5. Materials of Construction Component Section Materials Reactor Vessel Pressure Plate SA-533, Grade B, Class 1 1 Pressure Forgings A-508-64, Class 2 (Code Case 1332-3)

Cladding 18-8 Stainless Steel or Ni-Cr-Fe Studs, Nuts and Washers A-540, Grade B23 or B24 (Code Case 1335-2)

Thermal Shield and Internals SA-240, Type 304 Guide Lugs Ni-Cr-Fe, SB-168 (Code Case 1336)

Replacement Pressure Forging SA-508, Class 3 2,3 Reactor Vessel Cladding 308L/309L 2,3 Head CRDM Flange SA-182, Grade F316LN 2,3 CRDM Guide Tube SB-167 UNS N06690 (ASME Section III Code Case N474-2) 2,3 Steam Generator Deleted per 2004 update Pressure Forgings SA-508 CL.3A Cladding for Heads 308L/309L Stainless Steel Cladding for Tube Sheets UNS N06052 (Code Case 2142)

Tubes SB-163 UNS N06690 (Code Case N-20-4)

Studs - Reactor Coolant Side SA-193, Grade B7 Nuts - Reactor Coolant Side SA-194, Grade 7 Studs - Secondary Side SA-193, Grade B7 Nuts - Secondary Side SA-194, Grade 7 Pressurizer Shell, Heads, and External Plate SA-212, Grade B Forgings A-508-64, Class 1 (Code Case 1332-3)

Cladding 18-8 Stainless Steel Studs and Nuts SA-320, Grade L43 Internal Plate SA-240, Type 304 Internal Piping SA-312, Type 304 Sampling and Level Indication Piping SA-479, Type 316 Safe Ends Reactor Coolant 28 in. and 36 in. SA-516, Grade 70 (Elbows) A-106, Grade C Piping (Straights)

Cladding 18-8 Stainless Steel (31 DEC 2019)

Oconee Nuclear Station UFSAR Table 5-5 (Page 2 of 2)

Component Section Materials 10 in. Surge Line and 2-1/2 in. Spray A-403, Grade WP 316 (Elbows) A-376, Type Line 316 (Straights)

Piping Safe Ends SA-479, Type 316; A-376, Type 316 and Ni-Cr-Fe, SB-166 Reactor Coolant Pumps Oconee 1 Forging Stainless Steel SA182, 304 Static Casting Stainless Steel SA-351, Gr. CF8 Seal Housing SA-351, Gr. CF3 or SA-182, F316 Tubing and Pipe Stainless Steel SA-213, Type 316 or 304 and SA-376 or 312 (Seamless) Type 304 or 316 Bolting Material SA-193, SA-540 Welding Filler Metals SA-298 or SA-371 Plate, Sheet and Strip SA-240 Oconee 2 & 3 Castings Casing A-351, Grade CF8M Stuffing Box A-351, Grade CF8M Forgings Shaft A-473, Type 316 Bolting Casing Studs A-193, Grade B7 Casing Nuts A-194, Type 2H Valves Valve Bodies A-351, Grade CF8M A-182, F316 and F347; SA-479, Type 316 Note:

1. This material is metallurgically identical to SA-302, Grade B, as modified by Code Case 1339.
2. Input Document for Replacement RVCHA Licensing and Safety Evaluation. Babcock & Wilcox Canada, BWC Report No.068S-LR-01 Rev. 2, OM 201.R-0141.001.
3. History Docket for Closure Heads, Customer Spec# OSS-0279.00-00-0003, Babcock & Wilcox Canada, BWC-CONT. 068S, 068S-01 (Vol. 1 of 4).

(31 DEC 2019)

Oconee Nuclear Station UFSAR Table 5-6 (Page 1 of 1)

Table 5-6. Summary of Primary Plus Secondary Stress Intensity for Components of the Reactor Vessel Allowable Stress 3 Sm, psi Area Stress Intensity psi (Operating Temperature)

Control Rod Housing 24,800 69,900 Head Flange 58,000 80,000 Vessel Flange 43,000 80,000 Closure Studs 89,400 107,400 Primary Nozzles -Inlet 24,000 80,000 Outlet 24,000 80,000 Bottom Head to Shell 23,300 80,000 Bottom Instrumentation 10,100 69,900 Nozzle Belt to Shell 32,300 80,000 Core Flooding Nozzle 23,660 80,000 Support Skirt 88,000 93,700 Note:

1. Locations or points of stress analysis are illustrated on Figure 5-10.
2. "The values shown in this table are historical. See calculation OSC-1815 for current values."

(31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-7 (Page 1 of 1)

Table 5-7. Summary of Cumulative Fatigue Usage Factors for Components of the Reactor Vessel Item Usage Factor1 Control Rod Housing 0.0 Head Flange 0.10 Vessel Flange 0.05 Stud bolts 0.38 Primary Nozzles - Inlet 0.06 Outlet 0.06 Bottom Head to Shell 0.0 Bottom Instrumentation 0.0 Nozzle Belt to Shell 0.0 Core Flooding Nozzle 0.02 Support Skirt 0.14 Note:

1. As defined in Section III of the ASME Boiler and Pressure Vessel Code, Nuclear Vessels.

The values shown in this table are historical. See calculation OSC-1815 for current values.

Points of stress analysis are illustrated on Figure 5-10.

(31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-8 (Page 1 of 1)

Table 5-8. Stresses Due to a Maximum Design Steam Generator Tube Sheet Pressure Differential of 2,500 psi at 650°F Stress Computed Value Allowable Value Original Steam Generator Deleted row(s) per 2004 update Replacement Steam Generator Primary Membrane 16.5 Ksi 30.0 Ksi (Sm)

Primary Membrane plus 30.1 Ksi 45.0 Ksi (1.5 Sm)

Primary Bending (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-9 (Page 1 of 1)

Table 5-9. Ratio of Allowable Stresses to Computed Stresses for a Steam Generator Tube Sheet Pressure Differential of 2,500 psi Component Part Stress Ratio Original Steam Generator Deleted row(s) per 2004 update Replacement Steam Generator Primary Head 2.21 Primary Head Tube Sheet Joint 1.53 Tubes 1.20 Tube Sheet Average Membrane SI ratio 1.82 Membrane plus bending SI ratio 1.50 (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-10 (Page 1 of 5)

Table 5-10. Fabrication Inspections Component RT UT PT MT ET

1. Reactor Vessel 1.1 Forgings 1.1.1 Flanges X1 X 1.1.2 Studs, Bar X 1.1.3 Studs After Final X Machining 1.1.4 Skirt Adaptor X1 1.1.5 Nozzle Shell Forgings X X 1.1.6 Main Nozzle Forgings X X 1

1.1.7 Dutchman Forging X X Deleted row(s) per 2004 update 1.1.10 Replacement RVH CRDM X 7,8 X 7,8 Nozzle Flange 1.1.11 Replacement RVH CRDM X 7,8 X 7,8 Nozzle Guide Tube 1.1.12 Replacement Reactor X 7,8 X 7,8 X 7,8 Vessel Closure Head 1.2 Plates 1.2.1 Head and Shell Plate X1 X6 1.2.2 Support Skirt X1 X6 1.3 Instrumentation Tubes X X 1.4 Closure O-Rings X X 1.5 Weldments 1.5.1 Longitudinal and X X Circumferential Main Seams 1.5.2 CRD Mechanism Adaptor X X to Shell 1.5.3 CRD Mechanism Adaptor X X to Flange 1.5.4 Main Nozzles X X 1.5.5 Instrumentation Nozzle X Connection 1.5.6 Nozzle Safe-Ends, Weld X X Deposit (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-10 (Page 2 of 5)

Component RT UT PT MT ET 1.5.7 Temporary Attachment X After Removal 1.5.8 All Accessible Welds After X Hydrotest 1.5.9 O-Ring Closure Weld X X 1.5.10 Cladding, Sealing X6,2 X Surfaces 1.5.11 Cladding, All Other X6,3 X 1.5.12 Insulation Support Lugs X 1.5.13 Replacement RVH CRDM Nozzle Flange to X7,8 X7,8 guide Tube Weld CRDM Nozzle to X7,8 Replacement RVCH Weld

2. Replacement Steam Generator 2.1 Tube Sheet 2.1.1 Forging X X 2.1.2 Cladding X X 2.2 Heads 2.2.1 Forging X X 2.2.2 Cladding X X 2.3 Shell 2.3.1 Forging X X 2.4 Tubes X X X 2.5 Nozzles (Forgings) X X or X 2.6 Studs, Bar 2.7 Studs After Final Machining X 2.8 Weldments 2.8.1 N/A 2.8.2 N/A 2.8.3 Shell, Circumferential X X X 2.8.4 Cladding, Sealing Surfaces X X 2.8.5 Cladding, all other X X (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-10 (Page 3 of 5)

Component RT UT PT MT ET 2.8.6 Nozzle to Shell (Steam X X X Noz) 2.8.7 Level Sensing/Drain X X Connection 2.8.8 Instrument Connection X 2.8.9 Conical Support X X 2.8.10 Tube to Tubesheet X 2.8.11 Temporary Attachment X or X after Removal 2.8.12 Hydrostatic Test (All X or X Accessible Welds) 2.8.13 Lifting Lugs X or X 2.8.14 N/A

3. Pressurizer 3.1 Heads 3.1.1 Plate X1 X 3.1.2 Cladding X6,3 X 3.2 Shell 3.2.1 Forging X1 X 3.2.2 Plate X1 X6 3.2.3 Cladding X6,3 X 3.3 Heater Bundles 3.3.1 Cover Plate X X 3.3.2 Diaphragm and Spacer X X Plate 3.3.3 Studs, Bar X 3.3.4 Studs and Nuts After Final X Machining 3.3.5 Heaters 3.3.5.1 Tubing X X6 3.3.5.2 Positioning of X Heater Element in Tube 3.4 Nozzle (Forgings) X X 3.5 Weldments (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-10 (Page 4 of 5)

Component RT UT PT MT ET 3.5.1 Shell, Longitudinal as X X Deposited by Submerged Arc 3.5.2 Shell, Longitudinal as X X X Deposited by Electroslag 3.5.3 Shell, Circumferential X X 3.5.4 Cladding, Sealing X6,2 X Surfaces 3.5.5 Cladding, All Other X6,2 X 3.5.6 Nozzle to Shell X X 3.5.7 Nozzle Safe-Ends (If X X Weld Deposit) 3.5.8 Nozzle Safe-End (If X X Forging or Bar) 3.5.9 Instrumentation and Vent X Connections 3.5.10 Support Brackets X 3.5.11 Heater Guide Tube Pad X X 3.5.12 Temporary Attachment X After Removal 3.5.13 All Accessible Welds X After Hydrotest 3.5.14 Insulation Support Pads X

4. Piping 4.1 Pipe 4.1.1 Forgings X1 X 6,3 4.1.2 Cladding X X 4.2 Bends 4.2.1 Plate X1 X6 4.2.2 Cladding X6,3 X 4.3 Nozzle Forgings X X 4.4 Weldments 4.4.1 Longitudinal X X 4.4.2 Circumferential X X 6,3 4.4.3 Cladding, Elbows X X 6,3 4.4.4 Cladding, Straight X X (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-10 (Page 5 of 5)

Component RT UT PT MT ET 4.4.5 Nozzles to Run Pipe X X 4.4.6 Thermowell Connections X 4.4.7 Insulation Support Lug X Pads

5. Reactor Coolant Pumps 5.1 Castings X X 5.2 Forgings X X 5.3 Weldments 5.3.1 Circumferential X X 5.3.2 Piping Connections X
6. Valves 6.1 Castings X X 6.2 Forgings X X Note:
1. 100% scanning for longitudinal wave technique and 100% shear wave technique.
2. UT of clad defects and bond to base metal.
3. UT of clad bond to base metal (spot check).
4. Also gas leak test--B&W requirement.
5. Over 12-inch length on each end.
6. Additional B&W requirement.

RT: Radiographic UT: Ultrasonic PT: Dye Penetrant Mt: Magnetic Particle ET: Eddy Current

7. Input Document for Replacement RVCHA Licensing and Safety Evaluation. BWC Report No. 068S-LR-01 Rev. 2, OM 201.R-0141.001.
8. History Docket for Closure Heads, Customer Spec. No. OSS-0279.00-00-003, Babcock

& Wilcox Canada, BWC-CONT. 068S, 068S-01 (Vol. 1-2 of 4).

(31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-11 (Page 1 of 2)

Table 5-11. Reactor Vessel Design Data Design Pressure, psig 2500 Design Temperature, °F 650 Coolant Operating Temperature, Inlet/Outlet, °F 554/604 Hydrotest Pressure, psig 3125 3

Coolant Volume (Hot, Core and Internals in Place), ft 4058 Reactor Coolant Flow, lb/hr 131.32 x 106 Number of Reactor Closure Head Studs 60 Diameter of Reactor Closure Head Studs, in. 6-1/2 Vessel Dimensions Overall Height of Vessel and Closure Head, ft-in.1 40-8-3/4 Shell i.d., in. 171 Flange i.d., in. 165 Straight Shell Minimum Thickness, in. 8-7/16 Shell Cladding Minimum Thickness, in. 1/8 Shell Cladding Nominal Thickness, in. 3/16 Insulation Thickness, in. 3 Replacement Closure Head Insulation Thickness, in. 3 1/46 Replacement Closure Head Nominal Thickness, in. 74,5 Lower Head Minimum Thickness, in. 5 Vessel Nozzles Function No. ID, in. Material Coolant Inlet 4 28 Carbon Steel - SS Clad Coolant Outlet 2 36 Carbon Steel - SS Clad Core Flooding - LP Injection 2 12 Carbon Steel2 - SS Clad Control Rod Drive 61 2.76 Inconel3 Axial Power Shaping Rod Drive 8 2.76 Inconel3 Row(s) Deleted Per 2000 Update In-Core Instrumentation 52 3/4 Sch 160 Inconel Dry Weight, lbs Vessel 646,000 Replacement Closure Head 155,2004,7 Studs, Nuts, and Washers 39,500 (31 DEC 2003)

Oconee Nuclear Station UFSAR Table 5-11 (Page 2 of 2)

Note:

1. Instrument nozzle to CRD flange.
2. With stainless steel safe end added after stress relief.
3. With stainless steel flanges.
4. Input Document for Replacement RVCHA Licensing and Safety Evaluation. Babcock &

Wilcox Canada, BWC Report No. 068S-LR-01 Rev. 2, OM 201.R-0141.001.

5. History Docket for Closure Heads, Customer Spec. #OSS-0279.00-00-003, Babcock &

Wilcox Canada, BWC-CONT. 068S, 068S-01 (Vol. 1 of 4).

6. Transco Drawing RT-48783-DR1, RPV Top Head Service Structure Insulation Drip Panels Key Layout and Details (Layout D1), OM 241.R0005,001.
7. BWC Drawing 068SE001, RPV Closure Head General Arrangement, OM 201.R 0001.001.

(31 DEC 2003)

Oconee Nuclear Station UFSAR Table 5-12 (Page 1 of 2)

Table 5-12. Reactor Vessel -- Physical Properties (Oconee 1)

Ultimate Yield Strength (103 Strength (103 Elong. in 2 in. Impact Test Impact Item Heat No. psi) psi) (%) Temp. (°F) Values Deleted row(s) per 2003 update Bottom Head A 0973-2 87.2 65.0 24.5 +10 35-30-47 Intermediate Shell Plate C 2197-2 91.5 70.0 25.0 +10 39-45-26 Upper Shell Plate (1) C 3265-1 87.0 66.2 28.1 +10 34-64-27 Upper Shell Plate (2) C 3278-1 84.5 63.5 28.1 +10 35-29-53 Lower Shell Plate (1) C 2800-1 85.0 60.5 29.0 +10 36-39-39 Lower Shell Plate (2) C 2800-2 90.5 69.0 25.0 +20 32-33-49 Core Flooding Nozzle 94894 98.0 74.0 21.5 +10 45-53-40 Core Flooding Nozzle 94894 92.5 71.0 24.0 +10 37-50-45 Inlet Nozzle 123S346VA1 90.0 67.5 25.0 +10 104-94-142 Inlet Nozzle 123S346VA2 92.7 72.5 26.0 +10 104-121-106 Inlet Nozzle 124S502VA1 97.2 76.0 25.0 +10 120-106-101 Inlet Nozzle 124S502VA1 94.0 73.5 23.5 +10 110-85-77 Outlet Nozzle 122S316VA2 90.0 67.0 26.0 +10 131-110-94 Outlet Nozzle 122S316VA1 90.0 68.5 25.0 +10 92-86-82 4P16373P156 Upper Shell Flange 6 82.5 57.4 29.0 +10 49-41-71 Dutchman Forging 122S347VA1 94.5 74.5 24.0 +10 92-70-70 Deleted row(s) per 2003 update Upper Nozzle Belt Forging ZV-2888 82.0 57.0 30.5 +34 avg 30 avg Lower Nozzle Belt Forging ZV-2861 85.0 63.5 29.0 +26 avg 30 avg (31 DEC 2003)

Oconee Nuclear Station UFSAR Table 5-12 (Page 2 of 2)

Ultimate Yield Strength (103 Strength (103 Elong. in 2 in. Impact Test Impact Item Heat No. psi) psi) (%) Temp. (°F) Values Replacement RVH Forging O1W60-1-1 87.7 avg 66.4 avg 29.4 avg Note:

1. From History Docket for Closure Heads, Customer Spec. #OSS-0279.00-00-003, Babcock & Wilcox Canada, BWC-CONT.

068S, 068S-01, Vol. 1 of 4.

(31 DEC 2003)

Oconee Nuclear Station UFSAR Table 5-13 (Page 1 of 1)

Table 5-13. Reactor Vessel - Chemical Properties (Oconee 1). (References 34, 60)

Element Heat Number C Mn P S Si Ni Mo Co V Cr Cu A1 Deleted row(s) per 2003 update A 0973-2 .21 1.34 .011 .016 .18 .46 .47 .010 -- -- -- --

C 2197-2 .21 1.28 .008 .010 .17 .50 .46 .021 -- -- -- --

C 3265-1 .21 1.42 .015 .015 .23 .50 .49 .016 -- -- -- --

C 3278-1 .19 1.26 .010 .016 .23 .60 .47 .016 -- -- -- --

C 2800-1 .20 1.40 .012 .017 .20 .63 .50 .014 -- -- -- --

C 2800-2 .20 1.40 .012 .017 .20 .63 .50 .014 -- -- -- --

94894 .22 0.62 .006 .009 .23 .87 .60 .016 -- 0.33 -- --

123S346VA1 .22 .61 .010 .010 .20 .69 .56 .01 0.01 .27 -- --

123S346VA2 .21 .62 .010 .008 .20 .69 .57 .01 .01 .28 -- --

124S502VA1 .22 .65 .010 .010 .22 .75 .59 .02 .01 .35 -- --

124S502VA2 .23 .68 .010 .014 .22 .78 .60 .02 .01 .31 -- --

122S316VA2 .20 .62 .010 .009 .28 .73 .57 .013 .01 .33 -- --

122S316VA1 .18 .58 .010 .014 .28 .68 .61 .015 .01 .32 -- --

4P16373P1566 .20 .72 .010 .012 .28 .74 .55 .011 .03 .34 -- --

122S347VA1 .20 .62 .010 .008 .25 .66 .55 .021 .02 .32 -- --

Deleted row(s) per 2003 update ZV-2888 .22 .70 .010 .008 .32 .62 .59 .007 .02 .36 -- --

ZV-2861 .18 0.64 0.006 0.010 0.29 0.65 0.57 0.01 0.01 0.31 -- --

O1W60-1-1 .18 1.46 .005 <0.000 .17 .89 .51 -- <0.003 .15 0.05 0.021 Note: 1

1. From History Docket for Closure Heads, Customer Spec. #055-0279.00-00-003, Babcock & Wilcox Canada, BWC-CONT.

068S, 068S-01, Vol. 1 of 4 (31 DEC 2003)

Oconee Nuclear Station UFSAR Table 5-14 (Page 1 of 2)

Table 5-14. Reactor Vessel - Mechanical Properties (Oconee 2 & 3). (Reference 33)

Approximate Specimen Drop Weight Cv Energy at Upper Shelf Cv Description NDT (F) +10°F (ft - lb) Energy (ft-lb)

Oconee 2 Top Shell Forging C.1/5 T1 +20 86, 46, 79 127 C.1/4 T +10 100, 89, 72 140 C.1/2 T +20 62, 77, 40 141 1

Bottom Shell C.1/5 T +20 116, 93, 104 140 Forging C.1/4 T +20 82, 83, 90 139 C.1/2 T +20 101, 89, 92 149 Top Weld Deposit 1/4T Not Available 41, 37, 43 Not Available (WF 154)

Center Weld 1/4T Not Available 38, 28, 49 Not Available Deposit (WF 25)

Bottom Weld 1/4T Not Available 35, 40, 30 Not Available Deposit (WF 112)

Oconee 3 Top Shell Forging C 1/5 T1 +40 76, 82, 46 116 C 1/4 T +30 85, 77, 78 136 C 1/2 T +30 82, 55, 91 119 Bottom Shell C 1/5 T1 +20 49, 83, 43 155 Forging C 1/4 T +40 39, 50, 66 152 C 1/2 T +20 24, 34, 14 154 Oconee 3 Top Weld Deposit 1/4 T Not Available 36, 35, 26 Not Available (WF 200)

Outer Weld 1/4 T Not Available 29, 35, 30 Not Available Deposit (WF 67)

Bottom Weld 1/4 T Not Available 42, 29, 46 Not Available Deposit (WF 169-1)

Note:

1. Circumferential, 2 inches from surface.
2. In addition to the impact tests required by the ASME Code, the Nil-Ductility Temperature and Charpy V-notch energy levels at several temperatures were obtained for the two (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-14 (Page 2 of 2)

Approximate Specimen Drop Weight Cv Energy at Upper Shelf Cv Description NDT (F) +10°F (ft - lb) Energy (ft-lb) forgings that comprise the core region of the reactor vessels. The forging material is ASTM A508-64 Class 2 as modified by Code Case 1332-4. The impact tests were taken at 2 inches from surface, 1/4 and 1/2 of the forging thickness, and oriented in the circumferential direction with the length of the notch of the Charpy V-notch perpendicular to the surface of the material. The weld deposits of the core region (circumferential welds) were impact tested at plus 10°F using Charpy V-notch specimens oriented perpendicular to the direction of welding with the notch normal to the surface. No upper shelf fracture energy levels were obtained for the weld deposits.

(31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-15 (Page 1 of 1)

Table 5-15. Reactor Coolant Flow Distribution with Less than Four Pumps Operating Oconee 1 Oconee 2 or 3 (106 lb/hr)

(106 lb/hr)

HISTORICAL INFORMATION NOT REQUIRED TO BE REVISED 3 Pumps Flow in loop with two pumps 68.92 71.1 Flow in loop with one pump 29.95 29.5 Flow of pump in one pump loop 43.50 43.6 Idle pump reverse flow 13.55 14.1 2 Pumps - 2 Loops Pump flow each loop 44.49 44.5 Steam generator flow each loop 32.67 32.6 Reverse flow each idle pump 11.82 11.9 2 Pumps - 1 Loop Operating loop flow 71.22 73.6 Idle loop reverse flow 10.82 11.9 1 Pump Operating pump flow 45.06 45.0 Operating loop idle pump reverse flow 10.65 10.6 Idle loop reverse flow 5.23 5.5 Note:

1. For the configurations with both loops in operation the temperature in the cold legs will be the core inlet temperature (about 554°F). The hot leg fluid will be at about 604°F.
2. The reactor will not be operated at power in the 2 pump - single loop configuration.

(31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-16 (Page 1 of 1)

Table 5-16. Reactor Coolant Pump - Design Data (Oconee 1)

Design Pressure/Operating Pressure, psig 2500/2185 Hydrostatic Test Pressure (cold), psig 4100 Design Temperature (casing), °F 650 Operating Speed, rpm 1190 Suction Temperature, °F 554 Developed Head, ft 350 Capacity, gal/min 88,000 Seal Water Injection, gal/min 8 Seal Return, gal/min 2.2 (1A1, 1B2), 2.0 (1A2, 1B1)

Pump Discharge Nozzle i.d., in. 28 Pump Suction Nozzle i.d., in. 31 Overall Unit Height, ft-in. (Pump - Motor) 29' 9" Weight (dry), lb (without motor) 99,600 3

Coolant Volume, ft 56 Pump-motor moment of inertia, lb-ft2 70,000 Injection Water Temperature, °F 125 Cooling Water Temperature, °F 105 Motor Data Type Squirrel Cage Induction Single Speed, Water Cooled Voltage 6600 Phase 3 Frequency, Hz 60 Insulation Class F Starting Current, amp 4350 Power, HP (Nameplate) 9000 (31 DEC 2011)

Oconee Nuclear Station UFSAR Table 5-17 (Page 1 of 1)

Table 5-17. Reactor Coolant Pump - Design Data (Oconee 2, 3) (Data per Pump)

Design Pressure/Temperature, psig/°F 2500/650 Hydrotest Pressure, psig 3750 RPM at Nameplate Rating 1190 Developed Head, ft 362 Capacity, gal/min 92,200 Seal Water Injection, gal/min 10 Seal Water Return, gal/min 1.5 Injection Water Temperature, °F 120°F +/- 10°F Cooling Water Temperature, °F 105 Pump Discharge Nozzle i.d., in. 28 Pump Suction Nozzle i.d., in. 28 Overall Height, (Pump-Motor), ft-in. 29-4 Dry Weight Without Motor, lb 100,000 3

Coolant Volume, ft 98 Pump-motor Moment of Inertia, lb-ft2 70,000 Motor Data Type Squirrel Cage Induction Single Speed, Water Cooled Voltage 6600 Phase 3 Frequency, Hz 60 Insulation Class F Starting Current, amp 4350 Power, HP (Nameplate) 9000 (31 DEC 2001)

Oconee Nuclear Station UFSAR Table 5-18 (Page 1 of 3)

Table 5-18. Reactor Coolant Pump Casings - Code Allowables (Applies to Oconee 2 and 3)

Component Material Area Governing Condition Allowable Stress or Maximum Stress or Code III (See Note 1) Stress Intensity (psi) Stress Intensity (psi)

Para.

Discharge/ ASTM Extreme N414.3 D 1.5 Sm = 28,050 9,176 Suction Nozzle A351 Fibers (Outside CF8M Reinforcement)

Extreme N414.3 A + [(B + C)/2] + P 1.5 Sm = 28,050 18,164 Fibers Extreme N414.4 A + [(B + C)/2] + D + P 3.0 Sm = 56,100 18,908 Fibers Extreme N414.3 P 1.5 Sm = 28,050 18,426 Fibers Extreme Note 1 A+B+C+P 1.2 x 1.5 Sm = 33,660 18,214 Fibers Extreme Note 1 A+B+C+D+P 1.2 x 3.0 Sm = 67,320 23,072 Fibers Centerline N417.9 D 1.5 Sm = 28,050 8,420 Fibers Centerline N417.9 A + [(B + C)/2]+ P 1.0 Sm = 18,700 16,202 Fibers Centerline N414.4 A + [(B + C)/2] + D + P 3.0 Sm = 56,100 18,791 Fibers Centerline N414.1 P 1.0 Sm = 18,700 16,380 Fibers Centerline Note 1 A+B+C+P 1.2 Sm = 22,440 16,490 Fibers (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-18 (Page 2 of 3)

Component Material Area Governing Condition Allowable Stress or Maximum Stress or Code III (See Note 1) Stress Intensity (psi) Stress Intensity (psi)

Para.

Centerline Note 1 A+B+C+D+P 1.2 x 3.0 Sm = 67,320 22,621 Fibers Discharge/ ASTM Extreme N414.3 A + [(B + C)/2] + D + P 1.5 Sm = 28,050 18,440 Suction Nozzle A351 Fibers (Inside CF8M Reinforcement)

Extreme See Note 2 A+B+C+D+P 1.2 x 1.5 Sm = 33,660 18,362 Fibers Centerline N417.8 A + [(B + C)/2] + D + P 1.0 Sm = 18,700 15,970 Fibers Centerline See Note 2 A+B+C+D+P 1.2 Sm = 22,440 16,139 Fibers Bowl Section ASTM Extreme N414.3 P 1.5 Sm = 28,050 14,750 A351 Fibers CF8M Centerline N414.1 P 1.0 Sm = 18,700 10,000 Fibers Cover ASTM Extreme N712.1 Hydrostatic Test Pressure 0.9 Sy = 17,160 See Note 3 A182 Fibers Grade F316 Extreme N414.3 Operating Pressure 1.5 Sm = 25,780 20,807 Note 4 Fibers Centerline N712.1 Hydrostatic Test Pressure 1.35 Sy = 25,740 See Note 3 Fibers Centerline N414.1 Operating Pressure 1.0 Sm = 17,190 14,527 Fibers (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-18 (Page 3 of 3)

Component Material Area Governing Condition Allowable Stress or Maximum Stress or Code III (See Note 1) Stress Intensity (psi) Stress Intensity (psi)

Para.

All Fibers N412(m)(1) Operating Thermal 2.0 Sy = 38,140 62,469 Note 5 All Fibers N414.4 Operating Pressure & 3.0 Sm = 51,570 48,440 Note 6 Thermal Notes:

1. A = Dead Load Reactions B = Vertical Seismic Reactions (SSE)

C = Horizontal Seismic Reactions (SSE)

D = Thermal Expansion Reactions P = 2500 psia (operating design pressure)

2. Allowable Stress specified by B & W for Reactor Coolant Piping reactions on Pump.
3. No maximum stress associated with this case, this requirement determines the hydrostatic test pressure necessary to produce the allowable stress.
4. Element 18 has maximum stress Pm + Pb = 29,102, which is acceptable versus allowable 1.5Sm = 30,000 psi @ 225 F.
5. All thermal stresses for wet and dry cooling jackets meet the provisions of paragraph N414(m)(1) vs. the allowable of 2.0 x Y.S. At several locations at the lower inside radius the localized thermal stresses exceed 2.0 x Y.S. These stresses are deemed to be peak stresses and as such are required only to be considered from a fatigue standpoint, per paragraph N412(m)(2), N412(k) and N4i4.5.
6. Three elements have maximum stress Pm + Pb + Q qualified versus allowable at lower temperature, 3.0Sm = 60,000. One element has maximum stress Pm + Pb + Q that exceeds allowable at lower temperature, 3.0Sm = 60,000 but this stress is considered a peaking stress, and thus the requirements of N414.4 dont apply.

(31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-19 (Page 1 of 1)

Table 5-19. Deleted Per 2000 Update.

(31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-20 (Page 1 of 3)

Table 5-20. Steam Generator Design Data (Data per Steam Generator)

Original Steam Generator Deleted row(s) per 2004 update Replacement Steam Generator Steam Conditions at Full Load, Outlet Nozzles Steam Flow, lbm/hr 5.4*106 Steam Temperature, °F 597 (BOL) / 591 (EOL), 570 (Design)

Steam Pressure, psig 910 Feedwater Temperature, °F 460 Reactor Coolant Flow, lbm/hr 65.66

  • 106 (Thermal Design Flow)

Reactor Coolant Side Design Pressure, psig 2500 Design Temperature, °F 650 Hydrotest Pressure, psig 3125 Coolant Volume (Hot), ft3 2001.45 (Note 1)

Full Load Temperature, °F 604 (inlet) / 554 (outlet) @ TDF Secondary Side Design Pressure, psig 1150 Secondary Side 1200 Feedwater Design Temperature, °F 630 Hydrotest Pressure, psig 1500 Net Volume, ft3 3486.3 Mass of Steam and Water at Full Load, lbm 54,696 Energy Content of Steam and Water at Full Load, (BTU) 29.8 (BOL) / 34.0 (EOL)

  • 106 Dimensions Tube, Nom. OD/Min Wall, in. 0.625/0.034 Overall Height (Including stool), ft-in. 73-3 1/16 Shell OD, in. 148-1/8 Shell Minimum Thickness, in. 3 (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-20 (Page 2 of 3)

Shell Minimum Thickness at Tubesheets and F.W. 5 Connections, in.

Tube Sheet Minimum Thickness, in. 22 1/16 Dry Weight, lbm 929600 Tube Length Between Tubesheets ft-in. 52-5 (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-20 (Page 3 of 3)

Function No. ID, in. Material Nozzles - Reactor Coolant Side Inlet 1 36 Carbon Steel - SS Clad Outlet 2 28 Carbon Steel - SS Clad Drain N/A N/A N/A Manways 2 16 Carbon Steel - SS Clad Handholes 1 6 Carbon Steel - SS Clad Nozzles - Secondary Side Steam 2 24 Carbon Steel Vent 1 1-1/2 Sch 80 E7018-A1 SFA5.5 buildup Drains 6 1-1/2 Sch 80 E7018-A1 SFA5.5 buildup Level Sensing 8 1 Sch 80 E7018-A1 SFA5.5 buildup Temperature Well 3 1-1/2 NPT Alloy 690 buildup Manways 2 16 Carbon Steel Feedwater Connections 32 3 Sch 80 2 1/4 Cr-Mo piping & nozzles, Alloy 600 spargers Emergency Feedwater 6 3 Sch 80 2 1/4 Cr-Mo piping & nozzles, Connections Alloy 600 sleeves Handholes 5 6 Alloy 690 Inspection Ports 30 3 Carbon Steel Note:

1. OSC-2729 Oconee Nuclear Station RETRAN Transient Analysis Model, Rev. 9, App K (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-21 (Page 1 of 2)

Table 5-21. Reactor Coolant Piping Design Data Reactor Inlet Piping Pipe i.d., in. 28 Design Pressure/Temperature, psig/°F 2500/650 Hydrotest Pressure, psig 3125 Minimum Thickness, in. 2-1/4 Coolant Volume (Hot - System Total), ft3 1085 Dry weight, System Total, lb 214,000 Reactor Outlet Piping Pipe i.d. in. 36 Design Pressure/Temperature, psig/°F 2500/650 Hydrotest Pressure, psig 3125 Minimum Thickness, in. 2-7/8 Coolant Volume (Hot - System Total), ft3 979 Dry Weight, System Total, lb 200,000 Pressurizer Surge Piping Pipe Size, in. 10, Schd 140 Design Pressure/Temperature, psig/°F 2500/670 Hydrotest Pressure, psig 3125 3

Coolant Volume, hot, ft 20 Dry Weight, lb 5000 Pressurizer Spray Piping Pipe Size, in. 2-1/2, Sch 160 Design Pressure/Temperature, psig/°F 2500/650 & 670 Hydrotest Pressure, psig 3125 3

Coolant Volume, hot, ft 2 Dry Weight, lb 650 Nozzles:

Function No. ID, in. Material On Reactor Inlet Piping 1

High Pressure Injection 4 2-1/2 Sch 160 Pressurizer Spray 1 2-1/2 Sch 160 Stainless Steel 2

Drain/Letdown 1 2-1/2 Sch 160 (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-21 (Page 2 of 2)

Nozzles:

Function No. ID, in. Material 3

Drain 3 1-1/2 Sch 160 Pressure Sensing 4 1 Sch 160 Temperature Well 4 0.375 Inconel Temperature Sensing 4 0.613 Inconel On Reactor Outlet Piping 2

Decay Heat 1 12 Sch 140 3

Vent 2 1 Sch 160 3

Conn. on Flow Meters 4 1 Sch 160 3

Pressure Sensing 4 1 Sch 160 Temperature Well 2 3/8 Inconel Temperature Sensing 6 0.613 Inconel 2

Surge Line 1 10 Sch 140 On Pressurizer Surge Piping Drain 1 1 Sch 160 Stainless Steel On Pressurizer Spray Piping Auxiliary Spray 1 1-1/2 Sch 160 Stainless Steel Spray Valve Bypass 2 1/2 Sch 160 Stainless Note:

1. Carbon Steel - SS Clad - With Stainless Steel Safe End Added after Stress Relief
2. Carbon Steel - SS Clad - with Inconel Safe End
3. SS pipe with Alloy 690 nozzles
4. Deleted per 2004 update (31 DEC 2004)

Oconee Nuclear Station UFSAR Table 5-22 (Page 1 of 1)

Table 5-22. Pressurizer Design Data Design/Operating Pressure, psig 2500/2166 Design/Operating Temperature, °F 670/648 3

Steam Volume, ft 700 3

Water Volume, ft 800 Hydrotest Pressure, psig 3 3125 Electric Heater Capacity, kW 16384 Dimensions Overall Height, ft-in. 44-11-3/4 Shell o.d., in. 96-3/8 Shell Minimum Thickness, in. 6.188 Dry Weight, lb. 291,000 Nozzles Function No. ID, in. Material Surge Line 1 10 Sch 140 Carbon Steel - SS Clad1 Spray Line 1 4 Sch 120 Carbon Steel - SS Clad2 Relief Valve 3 2-1/2 Carbon Steel - SS Clad1 Vent 1 1 Sch 160 Inconel7 Sample 1 1 Sch 160 Carbon Steel - SS Clad6 Temperature Well 1 3/8 Inconel Level Sensing 6 1 Sch 160 Carbon Steel - SS Clad6 Heater Bundle 3 19-1/8 Carbon Steel - SS Clad Manway 1 16 Carbon Steel - SS Clad Note:

1. With stainless steel safe end added after stress relief.
2. With Inconel safe end.
3. Pressure retaining part (inlet bushing) of pressurizer relief valves shop hydrotested at 3750 psig.
4. Total kw could be less depending on operational status of some heater elements.
5. Operating pressure is nominal steam space pressure based on 2155 psig at the hot leg nozzle for the pressure transmitter.
6. With Inconel or stainless steel (SA-479 Type 316) safe end.
7. With Inconel or stainless steel (SA-479 Type 316) vent nozzle.

(31 DEC 2006)

Oconee Nuclear Station UFSAR Table 5-23 (Page 1 of 3)

Table 5-23. Operating Design Transient Cycles for Pressurizer Surge Line Design Cycles Transient Oconee -

Number Transient Description - (ASME Category) 1/ -2/ -3 1A Heatup from 70°F to 8% Full Power (Normal) 1A1 11 13 10 1A2 12 14 6 1A3 64 70 33 1A4 45 43 51 1A5 228 220 260 Total Heatup Events 360 360 360 1B Cooldown from 8% Full Power (Normal) 1B1 60 60 60 1B2 300 300 300 Total Cooldown Events 360 360 360 2 Power Change 0 to 15% and 15 to 0%

(Normal) 2A 1,400 2B 1,440 3 Power Loading 8% to 100% Power 18,000 (Normal) 4 Power Unloading 100% to 8% Power 18,000 (Normal) 5 10% Step Load Increase (Normal) 8,000 6 10% Step Load Decrease (Normal) 8,000 7 Step Load Reduction (100% to 8% Power) 310 (Upset) 8 Reactor Trip (Upset)

Type A 80 Type B 172 Type C 90 Type D 70 Total Reactor Trips 412 9 Rapid Depressurization (Upset) 40 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-23 (Page 2 of 3)

Design Cycles Transient Oconee -

Number Transient Description - (ASME Category) 1/ -2/ -3 10 Change of Flow (Upset) 412 (1) 11 Rod Withdrawal Accident (Upset) 2 12 Hydrotests (Test) 13 Steady-State Power Operations 1.4E5 14 Control Rod Drop (Upset) 60 1

15 Loss of Station Power (Upset) 2 16 Steam Line Failure (Faulted) 1 17A Loss of Feedwater to One Steam Generator (Upset) 1 17B Stuck Open Turbine Bypass Valve (Emergency) 18 Loss of Feedwater Heater (Upset) -

19 Feed and Bleed Operations (Normal) 4,000 20 Miscellaneous (Normal)

A - Change in Makeup Flow Rate 30,000 B - Miscellaneous Spray Actuation 20,000 C - Change in Makeup Flow Rate 4.0E6 D1 - Pzr Boron Equilibration (on/off 2.55E4 valve)

D2 - Pzr Boron Equilibration 8.5E3 (modulating valve) 2 21 Loss of Coolant (Faulted) 22 Test Transients (Test)

A1 - High Pressure Injection System 5 Test B1 - High Pressure Injection System 15 Test C1 - High Pressure Injection System 10 Test D1 - High Pressure Injection System 10 Test Total Safety Injection Tests 40 A2 - HPI Check Valve Tests 8 8 8 B2 - HPI Check Valve Tests 48 48 48 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-23 (Page 3 of 3)

Design Cycles Transient Oconee -

Number Transient Description - (ASME Category) 1/ -2/ -3 C2 - HPI Check Valve Tests 15 12 9 D2 - HPI Check Valve Tests 80 84 88 Total Safety Injection Tests 151 152 153 23 Steam Generator Filling, Draining, Flushing -

and Cleaning (Normal) 3 24 Hot Functional Testing (Test)

Note:

1. Included in Transient 8, Reactor Trip.
2. Refer to the appropriate RCS Functional Specification for number of transient event cycles.
3. Included in Transient 1, Plant Heatup and Cooldown (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-24 (Page 1 of 2)

Table 5-24. Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 48 EFPY - Oconee Unit 1 Chemical Material Description Composition Reactor Fluence Vessel , n/cm2 RTPTS, Beltline Inside RTNDT, F at Screenin Region Matl. Heat Cu NI Initial Chemistr Surface F at 48 48 g Location Ident. Number Type wt% wt% RTNDT y Factor EFPY Margin EFPY Criteria 10 CFR 50.61 (Tables)

Lower Nozzle ZV- A 508 Cl. 1.11E+1 Belt Forging AHR 54 2861 2 0.16 0.65 +3 119.3 8 52.2 70.7 126.0 270 Intermediate C2197- SA-302 1.18E+1 Shell Plate C2197-2 2 Gr. BM1 0.15 0.50 +1 104.5 9 109.3 63.6 174.0 270 Upper Shell C3265- SA-302 1.31E+1 Plate C3265-1 1 Gr. BM1 0.10 0.50 +1 65.0 9 69.9 63.6 134.5 270 Upper Shell C3278- SA-302 1.31E+1 Plate C3278-1 1 Gr. BM1 0.12 0.60 +1 83.0 9 89.2 63.6 153.9 270 Lower Shell C2800- SA-302 1.31E+1 Plate C2800-1 1 Gr. BM 0.11 0.63 +1 74.5 9 80.0 63.6 144.7 270 Lower Shell C2800- SA-302 1.31E+1 Plate C2800-2 2 Gr. BM1 0.11 0.63 +1 74.5 9 80.0 63.6 144.7 270 LNB to IS Circ. Weld SA- ASA/Linde 1.11E+1 (100%) 1135 61782 80 0.23 0.52 -5 157.4 8 69.0 68.5 132.4 300 IS Longit.

Weld (Both SA- ASA/Linde 9.24E+1 230.3(2

)

100%) 1073 1P0962 80 0.21 0.64 -5 170.6 8 166.8 68.5 270 IS to US Circ. SA- ASA/Linde 1.19E+1 Weld (ID 61%) 1229 71249 80 0.23 0.59 +10 167.6 9 175.7 56.0 241.7 300 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-24 (Page 2 of 2)

Chemical Material Description Composition Reactor Fluence Vessel , n/cm2 RTPTS, Beltline Inside RTNDT, F at Screenin Region Matl. Heat Cu NI Initial Chemistr Surface F at 48 48 g Location Ident. Number Type wt% wt% RTNDT y Factor EFPY Margin EFPY Criteria US Longit.

Weld (Both SA- ASA/Linde 1.12E+1 100%) 1493 8T1762 80 0.19 0.57 -5 152.4 9 157.3 68.5 220.8 270 US to LS Circ. SA- ASA/Linde 1.27E+1 Weld (100%) 1585 72445 80 0.22 0.54 -5 158.0 9 168.5 68.5 232.0 300 LS Longit. SA- ASA/Linde 1.08E+1 Weld (100%) 1426 8T1762 80 0.19 0.57 -5 152.4 9 155.8 68.5 219.3 270 LS Longit. SA- ASA/Linde 1.08E+1 Weld (100%) 1430 8T1762 80 0.19 0.57 -5 152.4 9 155.8 68.5 219.3 270 10 CFR 50.61 (Surveillance Data)

LNB to IS Circ. Weld SA- ASA/Linde 1.11E+1 (100%) 1135 61782 80 0.23 0.52 -5 141.1 8 61.8 48.3 105.1 300 US to LS Circ. SA- ASA/Linde 1.27E+1 Weld (100%) 1585 72445 80 0.22 0.54 -5 145.2 9 155.8 48.3 199.1 300 Note:

1. SA-302 Grade B modified by ASME Code Case 1339
2. Controlling value of RTPTS reference temperature (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-25 (Page 1 of 1)

Table 5-25. Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 48 EFPY - Oconee Unit 2 Chemical Material Description Composition Fluence, 2

Reactor Vessel n/cm RTNDT, RTPTS, Beltline Region Matl. Heat Cu NI Initial Chemistry Inside F at 48 F at 48 Screening Location Ident. Number Type wt% wt% RTNDT Factor Surface EFPY Margin EFPY Criteria 10 CFR 50.61 (Tables)

Lower Nozzle Belt Forging AMX77 123T382 A 508 Cl. 2 0.13 0.76 +3 95.0 1.19E+19 99.6 70.7 173.3 270 Upper Shell Forging AAW 163 3P2359 A 508 Cl. 2 0.04 0.75 +20 26.0 1.28E+19 27.8 27.8 75.6 270 Lower Shell Forging AWG 164 4P1885 A 508 Cl. 2 0.02 0.80 +20 20.0 1.27E+19 21.3 21.3 62.7 270 LNB to US Circ. ASA/Linde Weld (100%) WF-154 406L44 80 0.27 0.59 -5 182.6 1.19E+19 191.5 68.5 255.0 300 US to LS Circ. ASA/Linde 1

Weld (100%) WF-25 299L44 80 0.34 0.68 -5 220.6 1.23E+19 233.3 68.5 296.8 300 10 CFR 50.61 (Surveillance Data)

None Note:

1. Controlling value of RTPTS refereence temperature (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-26 (Page 1 of 2)

Table 5-26. Evaluation of Reactor Vessel Pressurized Thermal Shock Toughness Properties at 48 EFPY - Oconee Unit 3 Chemical Material Description Composition Reactor Vessel Fluence Beltline , n/cm2 RTNDT, RTPTS, Region Matl. Heat Cu NI Initial Chemistr Inside F at 48 F at 48 Screenin Location Ident. Number Type wt% wt% RTNDT y Factor Surface EFPY Margin EFPY g Criteria 10 CFR 50.61 (Tables)

Lower Nozzle A 508 1.14E+1 Belt Forging 4680 4680 Cl. 2 0.13 0.91 +3 96.0 9 99.5 70.7 173.2 270 Upper Shell AWS A 508 1.26E+1 Forging 192 522314 Cl. 2 0.01 0.73 +40 20.0 9 21.3 21.3 82.6 270 Lower Shell ANK A 508 1.26E+1 Forging 191 522194 Cl. 2 0.02 0.76 +40 20.0 9 21.3 21.3 82.6 270 LNB to US ASA/

Circ. Weld WF- Linde 1.14E+1 (100%) 200 821T44 80 0.24 0.63 -5 178.0 9 184.6 68.5 248.1 300 ASA/

US to LS Circ. Linde 1.22E+1 Weld (ID 75%) WF-67 72442 80 0.26 0.60 -5 180.0 9 190.0 68.5 253.5(1) 300 10 CFR 50.61 (Surveillance Data)

Upper Shell AWS A 508 Forging 192 522314 Cl. 2 0.01 0.73 +40 36.0 1.26E+19 38.3 34.0 75.5 270 Lower Shell ANK A 508 Forging 191 522194 Cl. 2 0.02 0.76 +40 17.4 1.26E+19 18.5 17.0 112.3 270 LNB to US ASA/

Circ. Weld WF- Linde (100%) 200 821T44 80 0.24 0.63 -5 158.3 1.14E+19 159.5 48.3 202.8 300 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-26 (Page 2 of 2)

Chemical Material Description Composition Reactor Vessel Fluence Beltline , n/cm2 RTNDT, RTPTS, Region Matl. Heat Cu NI Initial Chemistr Inside F at 48 F at 48 Screenin Location Ident. Number Type wt% wt% RTNDT y Factor Surface EFPY Margin EFPY g Criteria Note:

1. Controlling value of RTPTS reference temperature (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-27 (Page 1 of 2)

Table 5-27. Evaluation of Reactor Vessel Extended Life (48EFPY) Charpy V-Notch Upper-Shelf Energy - Oconee Unit 1 Material Description 48 EFPY Fluence Estimated 48 Copper Initial T/4 48 EFPY EFPY %

Reactor Vessel Beltline Matl. Heat Compositio CvUSE, Location CvUSE at Drop at Region Location Ident. Number Type n w/o ft-lbs n/cm2 T/4 T/4 Regulatory Guide 1.99, Revision 2, Position 1 Lower Nozzle Belt Forging AHR-54 ZV-2861 A508 Cl.2 0.16 109 6.64E+17 95 13 Intermediate Shell Plate C2197-2 C2197-2 SA-302 Gr. B M 0.15 81 7.06E+18 63 22 Upper Shell Plate C3265-1 C3265-1 SA-302 Gr. B M 0.10 81 7.84E+18 66 18 Upper Shell Plate C3278-1 C3278-1 SA-302 Gr. B M 0.12 81 7.84E+18 65 20 Lower Shell Plate C2800-1 C2800-1 SA-302 Gr. B M 0.11 81 7.84E+18 66 19 Lower Shell Plate C2800-2 C2800-2 SA-302 Gr. B M 0.11 81 7.84E+18 66 19 LNB to IS Circ. Weld (100%) SA-1135 61782 ASA/Linde 80 0.23 70 6.64E+17 56 19 IS Longit. Weld (Both 100%) SA-1073 1P0962 ASA/Linde 80 021 70 5.53E+18 49 30 IS to US Circ. Weld (61%ID) SA-1229 71249 ASA/Linde 80 0.23 70 7.12E+18 46 34 IS to US Circ. Weld (39%OD) WF-25 299L44 ASA/Linde 80 0.34 70 -- -- --

US Longit. Weld (Both 100%) SA-1493 8T1762 ASA/Linde 80 0.19 70 6.70E+18 49 30 US to LS Circ. Weld (100%) SA-1585 72445 ASA/Linde 80 0.22 70 7.60E+18 46 34 LS Longit. Weld (100%) SA-1430 8T1762 ASA/Linde 80 0.19 70 6.46E+18 49 30 LS Longit. Weld (100%) SA-1426 8T1762 ASA/Linde 80 0.19 70 6.46E+18 49 30 Regulatory Guide 1.99, Revision 2, Position 2 Upper Shell Plate C3265-1 C3265-1 SA-302 Gr. B M -- 108 7.84E+18 91 16 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-27 (Page 2 of 2)

Material Description 48 EFPY Fluence Estimated 48 Copper Initial T/4 48 EFPY EFPY %

Reactor Vessel Beltline Matl. Heat Compositio CvUSE, Location CvUSE at Drop at Region Location Ident. Number Type n w/o ft-lbs n/cm2 T/4 T/4 LNB to IS Circ. Weld (100%) SA-1135 61782 ASA/Linde 80 -- 70 6.64E+17 55 22 IS to US Circ. Weld (61%ID) SA-1229 71249 ASA/Linde 80 -- 70 7.12E+18 46 34 IS to US Circ. Weld (39%OD) WF-25 299L44 ASA/Linde 80 -- 70 -- -- --

US to LS Circ. Weld (100%) SA-1585 72445 ASA/Linde 80 -- 70 7.60E+18 48 32 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-28 (Page 1 of 1)

Table 5-28. Evaluation of Reactor Vessel Extended Life (48 EFPY) Charpy V-Notch Upper-Shelf Energy - Oconee Unit 2 Material Description Estimate 48 EFPY d 48 Reactor Vessel Copper Initial Fluence T/4 EFPY Beltline Region Matl. Heat Compositio CvUSE, Location CvUSE at 48 EFPY %

Location Ident. Number Type n w/o ft-lbs n/cm2 T/4 Drop at T/4 Regulatory Guide 1.99, Revision 2, Position 1 Lower Nozzle Belt AMX-Forging 77 123T382 A508 C1.2 0.13 109 7.12E+18 87 20 AAW-Upper Shell Forging 163 3P2359 A508 C1.2 0.04 128 7.66E+18 113 12 AWG-Lower Shell Forging 164 4P1885 A508 C1.2 0.02 140 7.60E+18 126 10 LNB to US Circ. Weld WF- ASA/Linde (100%) 154 406L44 80 0.27 70 7.12E+18 43 38 US to LS Circ. Weld ASA/Linde (100%) WF-25 299L44 80 0.34 70 7.36E+18 39 40 Regulatory Guide 1.99, Revision 2, Position 2 Upper Shell Forging AAW-163 3P2359 A508 C1.2 -- 128 7.66E+18 101 21 NB to US Circ. ASA/Linde Weld (100%) WF-154 406L44 80 -- 70 7.12E+18 45 36 US to LS Circ. ASA/Linde Weld (100%) WF-25 299L44 80 -- 70 7.60E+18 44 37 (31 DEC 2000)

Oconee Nuclear Station UFSAR Table 5-29 (Page 1 of 1)

Table 5-29. Evaluation of Reactor Vessel Extended Life (48 EFPY) Charpy V-Notch Upper-Shelf Energy - Oconee Unit 3 Material Description 48 EFPY Fluence Estimated Reactor Vessel Copper Initial T/4 48 EFPY Beltline Region Matl. Heat Composition CvUSE, ft- Location CvUSE at 48 EFPY %

Location Ident. Number Type w/o lbs n/cm2 T/4 Drop at T/4 Regulatory Guide 1.99, Revision 2, Position 1 Lower Nozzle Belt Forging 4680 4680 A508 C1.2 0.13 109 6.82E+18 87 20 AWS-Upper Shell Forging 192 522314 A508 C1.2 0.01 90 7.54E+18 82 9 ANK-Lower Shell Forging 191 522194 A508 C1.2 0.02 110 7.54E+18 99 10 LNB to US Circ. Weld ASA/Linde (100%) WF-200 821T44 80 0.24 70 6.82E+18 46 35 US to LS Circ. Weld ASA/Linde (75%ID) WF-67 72442 80 0.26 70 7.30E+18 44 37 US to LS Circ. Weld ASA/Linde (25%OD) WF-70 72105 80 0.32 70 ----- ----- -----

Regulatory Guide 1.99, Revision 2, Position 2 AWS-Upper Shell Forging 192 522314 A508 C1.2 ----- 90 7.54E+18 77 15 ANK-Lower Shell Forging 191 522194 A508 C1.2 ----- 110 7.30E+18 85 23 NB to US Circ. Weld ASA/Linde (100%) WF-200 821T44 80 ----- 70 6.82E+18 55 21 US to LS Circ. Weld WF-70 72105 ASA/Linde ----- 70 ----- -- --

(25%OD) 80 (31 DEC 2000)

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