NL-05-1159, Revision to the Pressure Temperature Limits Report

Revision to the Pressure Temperature Limits Report

ML051870322
Person / Time
Site:	Vogtle
Issue date:	07/01/2005
From:	Grissette D Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-05-1159
Download: ML051870322 (49)

Text

Don E.Grissette Southern Nuclear Vice President Operating Company. Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.6474 Fax 205.992.0341 SOUTHERNAN July 1, 2005 COMPANY Energy to Serve Your World Docket Nos.: 50-424 NL-05-1159 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant Revision to the Pressure Temperature Limits Report Ladies and Gentlemen:

In accordance with the requirements of Vogtle Electric Generating Plant (VEGP)

Technical Specification 5.6.6(d), Southern Nuclear Operating Company (SNC) submits Revision 2 to the Pressure and Temperature Limits Report (PTLR). The following pages of the PTLR have been revised based on results of the VEGP Unit I fall 2003 reactor vessel surveillance capsule X withdrawal and the VEGP Unit 2 spring 2004 reactor vessel surveillance capsule W withdrawal reported in SNC's March 21, 2005 letter to the NRC (ML050840317).

Unit 1, Table 5-5, page 14 Unit 1, Table 5-6, page 15 Unit 1, Table 5-8, page 17 Unit 1, Table 5-9, page 18 Unit 2, Table 5-5, page 14 Unit 2, Table 5-6, page 15 Unit 2, Table 5-8, page 17 This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, Don E. Grissette DEG/DRG/daj

U. S. Nuclear Regulatory Commission NL-OS-1 159 Page 2

Enclosure:

Vogtle Electric Generating Plant Pressure Temperature Limits Report, Revision 2 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. T. E. Tynan, General Manager - Plant Vogtle RType: CVC7000 U. S. Nuclear Regulatorv Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Vogtle Mr. G. J. McCoy, Senior Resident Inspector - Vogtle

Vogtle Electric Generating Plant Pressure Temperature Limits Report Revision 2

Southern Nuclear Company Vogtle Unit 1 Pressure Temperature Limits Report Revision 2, April 2005

Vogtle Unit 1- Pressure Temperature Limits Report Southern Nuclear Table of Contents List of Tables ......... iii List of Figures ......... iv 1.0 RCS Pressure Temperature Limits Report (PTLR) .1 2.0 Operating Limits .1 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) .1 3.0 Cold Overpressure Protection Systems (LCO 3.4.12) .1 3.1 Pressurizer PORV Setpoints .2 3.2 Arming Temperature .2 4.0 Reactor Vessel Material Surveillance Program .2 5.0 Supplemental Data Tables ................... ,,... . . . . 3 6.0 References. 19 ii Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Votl Uni 1 rsueTmeaueLmt eotSuhr ula List of Tables Table 2-1 Vogtle Unit 1 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) ....................................... 6 Table 2-2 Vogtle Unit 1 Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) .7 Table 3-1 Vogtle Unit 1 Data Points for COPS PORV Setpoints .8 Table 5-1 Comparison of the Vogtle Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions ........................ 10 Table 5-2 Calculation of Chemistry Factors using Vogtle Unit 1 Surveillance Capsule Data ........ 11 Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 1....... 12 Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit 1 (1019 n/cm2, E > 1.0 MeV) ................. 13 Table 5-5 Vogtle Unit 1 Calculation of the Adjusted Reference Temperature (ART)

Values for the 1/4T Location a 36 EFPY ............................. 14 Table 5-6 Vogtle Unit 1 Calculation of the ART Values for the 3/4T Location @ 36 EFPY .............. 15 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Vogtle Unit 1 Heatup/Cooldown Curves ............................ 16 Table 5-8 RTPTS Calculations for Vogtle Unit 1 Beltline Region Materials at 36 EFPY .... 17 Table 5-9 RTPTS Calculations for Vogtle Unit 1 Beltline Region Materials at 54 EFPY .... 18 Revision 2 iii iii Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report l Southern Nuclear Southern Nuclear Vogtle Unit 1 Pressure Temperature Limits Report List of Figures Figure 2-1 Vogtle Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) .................................................... 4 Figure 2-2 Vogtle Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1000 F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) .5 Figure 3-1 Vogtle Unit 1 Maximum Allowable Nominal PORV Setpoints for COPS . 9 iv Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogtfe Unit 1 Pressure Temperature Limits Report Southern Nuclear 1.0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for Vogtle Unit 1 has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Protection Systems (COPS)

Revisions to the PTLR shall be provided to the NRC after issuance.

2.0 RCS Pressure and Temperature (PMT) Limits The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodology in WCAP-14040, Revision 4t1 with exception of WCAP-16142-P, Revision 1121 (Elimination of the Flange Requirement). The operability requirements associated with COPS are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of a cold overpressure transient in accordance with the methodology specified in TS 5.6.6.

2.1 RCS P/T Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 600F.

2.1.2 The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100F in any 1-hour period.

b. A maximum cooldown rate of 100F in any 1-hour period.

c. A maximum temperature change of less than or equal to 100 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2.

3.0 Cold Overpressure Protection Systems (LCO 3.4.12)

The setpoints for the pressurizer Power Operated Relief Valves (PORVs) and arming temperature are presented in the subsections which follow. These setpoints and arming temperature have been developed using the NRC-approved methodology specified in TS 5.6.6.

11 Revision 2 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 1 Pressure Temperature Limits Report Southern Nuclear 3.1 Pressurizer PORV Setpoints The pressurizer PORV setpoints are specified in Figure 3-1 and Table 3-1.

The limits for the COPS setpoints are contained in the 36 EFPY steady-state curves (Table 2-2), which are beltline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 74 psi.

Note: These setpoints include an allowance for the 500 F thermal transport effect for heat injection transients. A calculation has been performed to confirm that the setpoints will maintain the system pressure within the established limits when the pressure difference between the pressure transmitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.

3.2 Arming Temperature COPS shall be armed when any RCS cold leg temperature is < 2200 F.

4.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedule is provided in UFSAR Table 5.3.1-8. The results of these examinations shall be used to update Figures 2-1, 2-2, and 3-1.

The pressure vessel steel surveillance program (WCAP-1101114 ,) is in compliance with Appendix H131 to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23 15 1. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640156 of Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G,'Fracture Toughness Criteria for Protection Against Failure"r7'. The surveillance capsule removal schedule meets the requirements of ASTM E185-823 1". The removal schedule is provided in UFSAR Table 5.3.1-8.

Revision 22 2 Revision

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2E91, predictions.

Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsule data. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure from Regulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 1.02.

Table 5-3 provides the required Vogtle Unit 1 reactor vessel toughness data.

Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and the PTS evaluation.

Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 36 EFPY for each beltline material in the Vogtle Unit 1 reactor vessel.

The limiting beltline material was the intermediate shell plate B8805-2.

Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Vogtle Unit 1 reactor vessel beltline materials at the 1/4T and 3/4T locations for 36 EFPY.

Table 5-8 provides RTPTS values for Vogtle Unit 1 at 36 EFPY.

Table 5-9 provides RTPTS values for Vogtle Unit 1 at 54 EFPY.

3~ ~ Reiso 2 3 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report t Southern Vogtle Unit 1 Pressure Temperature Limits Report Southern Nuclear Nuclear Material Property Basis Limiting Material: Intermediate Shell Plate B8805-2 Limiting ART Values at 36 EFPY: 1/4T, 110 0 F 3/4T, 95 0F 2500 2250 2000 1750 1500 i

0

- 1250 0

U 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-1 Vogtle Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000 F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) (Plotted Dataprovided on Table 2-1) 4 Revision 2

Vogtle Unit 1- Pressure Temperature Limits Report Southern Nuclear Material Property Basis Limiting Material: Intermediate Shell Plate B8805-2 Limiting ART Values at 36 EFPY: 1/4T, 110F 3/4T, 950 F 2500 1455366597 2250 Unacceptal e_

2000 Operation 1750 IL 1500 Cooldown CO Rates F/Hr E 1250

0. steady-state

-20

-40

-60 3 1000 -100 750---

500-2olttup _

250 Tmn 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-2 Vogtle Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 0F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Error) (Plotted Data provided on Table 2-2)

Revision 2 5 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Southern Nuclear Vogtle Unit 1 - Pressure Temperature Limits Report Table 2-1 Vogtle Unit 1 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) 60OF/hr Heatup IOOF/hr Heatup 60°F/hr Heatup Criticality Limit 1000 F/hr Heatup Criticality Limit Leak Test Limit T- -l P T I P I P T I P T l P 60 0 170 0 60 0 170 0 153 2000 60 747 170 760 60 730 170 730 170 2485 65 760 170 760 65 730 170 730 70 760 170 760 70 730 170 730 75 760 170 760 75 730 170 730 80 760 170 763 80 730 170 730 85 763 170 770 85 730 170 730 90 770 170 782 90 730 170 730 95 782 170 796 95 730 170 733 100 796 170 815 100 733 170 739 105 815 170 836 105 739 170 747 110 836 170 862 110 747 170 759 115 862 170 891 115 759 170 774 120 891 170 925 120 774 170 791 125 925 170 962 125 791 170 812 130 962 175 1005 130 812 175 837 135 1005 180 1052 135 837 180 865 140 1052 185 1105 140 865 185 897 145 1105 190 1163 145 897 190 933 150 1163 195 1228 150 933 195 974 155 1228 200 1300 155 974 200 1020 160 1300 205 1380 160 1020 205 1071 165 1380 210 1468 165 1071 210 1128 170 1468 215 1566 170 1128 215 1191 175 1566 220 1674 175 1191 220 1261 180 1674 225 1793 180 1261 225 1339 185 1793 230 1925 185 1339 230 1426 190 1925 235 2070 190 1426 235 1521 195 2070 240 2231 195 1521 240 1627 200 2231 245 2408 200 1627 245 1743 205 2408 205 1743 250 1872 210 1872 255 2014 215 2014 260 2171 220 2171 265 2344 225 2344 6 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Voqtle Unit 1 Pressure Temperature Limits Report Southern Nuclear Table 2-2 Vogtle Unit 1 Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors)

Steady State 200F1hr 4 0 °F1hr 6 0 °F1hr 100°F1hr T P T P T P TP T P 60 0 60 0 60 0 60 0 60 0 60 747 60 709 60 670 60 633 60 559 65 762 65 725 65 688 65 652 65 582 70 778 70 742 70 707 70 673 70 608 75 796 75 762 75 728 75 696 75 637 80 816 80 783 80 752 80 722 80 668 85 838 85 807 85 778 85 751 85 704 90 862 90 834 90 807 90 783 90 743 95 889 95 863 95 840 95 819 95 787 100 918 100 895 100 875 100 858 100 835 105 951 105 931 105 915 105 902 105 889 110 987 110 971 110 959 110 950 110 948 115 1027 115 1015 115 1007 115 1004 120 1071 120 1063 120 1061 125 1120 125 1117 130 1173 135 1233 140 1299 145 1371 150 1452 155 1541 160 1639 165 1747 170 1867 175 2000 180 2146 185 2308 . .-

7 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Nuclear Vogtle Unit 1 Pressure Temperature Limits Report Southern Table 3-1 Vogtle Unit 1 Data Points for the Maximum Allowable Nominal COPS PORV Setpoints Temperature PORV Setpoint (Deg.F) (psig) 70 612 90 612 140 642 201 760 202 760 350 760 8 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Nuclear Voqtle Unit 1 Pressure Temperature Limits Report Southern 900 a- 850

-O z 800 5 (201,76()

750 cc Uj 700 0

a-w 650

-J m (70,6 12 600 0

-J

-j 550 500

-J z 450

= ~I a 0

z 400 0 50 100 150 200 250 300 350 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (DEG F)

Figure 3-1: Vogtle Unit 1 Maximum Allowable Nominal PORV Setpoints for COPS 9 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogue Unit 1 Pressure Temperature Limits Report Southern Nuclear Table 5-1 Comparison of the Vogtle Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Fluence(d) Temperature Shift Decrease (x 1019 n/cm 2 , Predicted Measured Predicted Measured Material Capsule E > 1.0 MeV) (OF) (a) ('F) (b) (%) Ia) (%/0)( )

Intermediate Shell U 0.334 26.80 13.56 14.5 0 Plate B8 805- 3(h) Y 1.16 39.97 31.94 19.5 0 (Longitudinal) V 1.97 45.50 42.66 22 3 X 3.53 51.03 96.50 26 11 Intermediate Shell U 0.334 26.80 0.00e) 14.5 0 Plate 18805- 3(h) Y 1.16 39.97 15.19 19.5 0 (Transverse) V 1.97 45.50 33.79 22 2 X 3.53 51.03 60.80 26 3 Surveillance U 0.334 23.52 24.98 14.5 0 Program Y 1.16 35.08 7.70 19.5 0 Weld Metali' V 1.97 39.93 0.00 " 22 2

. X 3.53 44.79 53.40 26 3 Heat Affected U 0.334 --- 0.0- 5 Zone Y 1.16 20.78 9 Material V 1.97 --- 42.08 --- 11 X 3.53 --- 10.60 7 Notes:

a. Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

b. Calculated using measured Charpy data plotted using CVGRAPH, Version 5.0.2"].

c. Values are based on the definition of upper shelf energy given in ASTM El 85-82181.

d. The fluence values presented here are the calculated values, not the best estimate values.

e. The actual value is -9.28. This physically should not occur, therefore 0.00 will be conservatively assumed.

f. The actual value is -1.34. This physically should not occur, therefore 0.00 will be conservatively assumed.

g. The actual value is -19.35. This physically should not occur, therefore 0.00 will be conservatively assumed.

h. The heat number for lower shell plate B8805-3 is C-0623-1.

i. The Surveillance weld was fabricated Irom Wire Heat No. 83653, Flux Type Linde 0091, Flux Lot No. 3536.

Revision 2 10 10 Revision 2

Vocitle Unit I - Pressure Temperature Limits Report Southern Nuclear Voitle Unit 1 Pressure Temperature Limits Report Southern Nuclear Table 5-2 Calculation of Chemistry Factors Using Vogtle Unit 1 Surveillance Capsule Data Material Capsule Capsule al FPO) ARTNDT l FF*ARTNDT FF l Intermediate Shell U 0.334 0.698 13.56 9.5 0.487 Plate B8805-3(f) Y 1.16 1.041 31.94 33.3 1.084 (Longitudinal) V 1.97 1.185 42.66 50.6 1.404 X 3.53 1.329 96.50 128.2 1.766 Intermediate Shell U 0.334 0.698 O(e) 0.0 0.487 Plate B8805-3(t) Y 1.16 1.041 15.19 15.8 1.084 (Transverse) V 1.97 1.185 33.79 40.0 1.404 X 3.53 1.329 60.80 80.8 1.766 SUM: 358.2 9.482 CFe88 os3= (FF RTNDT) + 2( FF 2) = (358.2) + (9.482) = 37.8F Surveillance Weld U 0.334 0.698 25.48 (24.98) 17.8 0.487 Material) Y 1.16 1.041 7.85 (7.70) 8.2 1.084 V 1.97 1.185 0 0.0 1.404 X 3.53 1.329 54.47 (53.40) k' 72.4 1.766 SUM: 98.4 4.741 CF Suv. Weld = X(FF

• RTNDT) + Y( FF2) = (98.4) + (4.741) = 20.80F Notes:

a. I = Calculated fluence from capsule X dosimetry analysis results (12), (x 109n/cm2 , E> 1.0 MeV).

b. FF = fluence factor = e(.28-0.1-1ogt)

c. ARTNDT values are the measured 30 ft-lb shift values taken from App. C of Ref. 12.

d. The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 1.02.

e. Actual values for ARTNDT are -9.28 (Plate) and -1.34 (Weld). This physically should not occur, therefore for conservatism a value of zero will be used for this calculation.

f. The heat number for lower shell plate 88805-3 is C-0623-1.

g. Surveillance Weld was fabricated from Wire Heat No.83653, Flux Type Linde 0091, Flux Lot No. 3536.

11 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 1 Pressure Temperature Limits Report Southern Nuclear Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 1 Material Description Cu(%) Ni(%) Initial RTNDTIa)

Closure Head Flange B8801-1 0.70 200 F (Heat# 123J173VA1)

Vessel Flange B8802-1 0.71 0F (Heat# 123H402VA1) _

Intermediate Shell Plate B8805-1 0.083 0.597 0oF (Heat # C-0613-1)

Intermediate Shell Plate B8805-2 0.083 0.61 200 F (Heat # C-0613-2)

Intermediate Shell Plate B8805-3 0.062 0.598 300 F (Heat # C-0623-1)

Lower Shell Plate B8606-1 0.053 0.593 200 F (Heat # C-2146-1)

Lower Shell Plate B8606-2 0.057 0.60 20 0F (Heat # C-2146-2) .

Lower Shell Plate B8606-3 0.067 0.623 10 OF (Heat # C-2085-2)

Intermediate Shell Longitudinal Welds, 0.042 0.102 -80F 101-124A, B & C(b)

Lower Shell Longitudinal Welds, 0.042 0.102 -80 0 F 101-142A, B & C(b)

Circumferential Weld 101-171w'I 0.042 0.102 -800 F Surveillance Program Weld Metalo'J 0.040 0.102 Notes:

a. The initial RTNDT values for the plates and welds are based on measured data.

b. All welds, including the surveillance weld, were fabricated with weld wire heat number 83653, Linde 0091 Flux, Lot No. 3536. Per Regulatory Guide 1.99, Revision 2, 'weight percent copper" and

'weight percent nickel' are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld.

Revision 2 12 12 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report , Southern Vogtle Unit 1 Pressure Temperature Limits Report Southern Nuclear Nuclear Table 5-4 Peak Calculated Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit 1 (n/cm2 , E > 1.0 MeV)

Azimuthal Location EFPY 00 150 300 450 14.33(a) 4.35E+18 6.55E+18 8.11E+18 8.38E+18 20.00 6.13E+18 9.16E+18 1.12E+19 1.15E+19 24.00 7.38E+18 1.10E+19 1.33E+19 1.37E+19 32.00 9.89E+18 1.47E+19 1.77E+19 1.81E+19 40.00 1.24E+19 1.83E+19 2.20E+19 2.25E+19 48.00 1.49E+19 2.20E+19 2.63E+19 2.70E+19 54.00 1.68E+19 2.48E+19 2.95E+19 3.03E+19 (a) The end of cycle 11 (Actual) when Capsule X was withdrawn.

13 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Table 5-5 Vogtle Unit 1 Calculation of the ART Values for the 1/4T Location @ 36 EFPY(a1f9) I Material RG 1.99 R2 CF FF IRTNDT ARTNDT() Margin" ART""

Method (OF)

Intermediate Shell Plate Position 1.1 53.1 1.06 0 56.3 34 90 B8805-1 _

Intermediate Shell Plate Position 1.1 53.1 1.06 20 56.3 34 110 B8805-2 Intermediate Shell Plate Position 1.1 38.4 1.06 30 40.7 34 105 B8805-3

. Position 2.1 24.4 1.06 30 26.0 17t _ 73 Lower Shell Plate Position 1.1 32.8 1.06 20 34.8 34 89 B8606-1 Lower Shell Plate Position 1.1 35.2 1.06 20 37.3 34 91 B8606-2 _

Lower Shell Plate Position 1.1 41.9 1.06 10 44.4 34 88 B8606-3 Inter. Shell Longitudinal Weld Position 1.1 34.5 0.899 -80 31.0 31 -18 Seam 101 -124A Position 2.1 8.6 0.899 -80 7.7 7.7"' -65 (0OAzimuth)

Inter. Shell Long. Weld Seams Position 1.1 34.5 1.06 -80 36.6 36.6 -7 101-124B,C Position 2.1 8.6 1.06 -80 9.1 9.1T -62 (120°, 240°Azimuth)

Intermediate to Lower Shell Position 1.1 34.5 1.06 -80 36.6 36.6 -7 Girth Weld Seam 101-171 Position 2.1 8.6 1.06 -80 9.1 9.1° -62 Lower Shell Long. Weld Seams Position 1.1 34.5 1.06 -80 36.6 36.6 -7 101-142AC (60°, 300° Position 2.1 8.6 1.06 -80 9.1 9.1__ -62 A (imuth) .

Lower Shell Long. Weld Seam Position 1.1 34.5 0.899 -80 31.0 31 -18 101 -142B (180" Azimuth) Position 2.1 8.6 0.899 -80 7.7 7.7-'

Notes:

a. Calculation of ART values was based on calculated fluence projections and measured data from the previous capsule analysis results (Capsule V, Ref. 10). It should be noted that the fluence projections from Capsule V (Ref. 10) were higher than those from Capsule X (Ref. 12). However, the chemistry factors determined from surveillance data (Table 5-2) are now higher but still lower than the corresponding Position 1.1 chemistry factor.

Therefore, the limiting ART values from Plate B8805-2 could not be exceeded by an updated plate B8805-3 ART value or updated weld ART values. Thus, the original limiting ART value is still valid. Note, the surveillance plate data is now deemed "Non-Credible," but this has no impact based on the preceding discussion.

b. Initial RTNDT values are measured values.

c. ARTNDT = CF

• FF

d. M =2 '(a, + ,2)')

e. ART = Initial RTNDT + ARTNDT + Margin ("F); (Rounded per ASTM E29, using the 'Rounding Method").

f. Weld data deemed credible per References 10 and 12. Plate data is now deemed "Non-Credible." See Note "a.r

g. Neutron Fluence value used for all material is the highest value from Reference 10 for 36 EFPY with exception to intermediate shell longitudinal weld 101 -124A and lower shell longitudinal weld 101 -142B which used the fluence at 0° from Table 5-4 for 36 EFPY. It should be noted that the fluence projections from Reference 10 are higher than the updated fluence analysis in Reference 12. See Note "a."

14 Revision 2 14 Revision 2

Vogtle Unit 1- Pressure Temperature Limits Report Southern Nuclear Table 5-6 Vogtle Unit 1 Calculation of the ART Values for the 3/4T Location @ 36 EFPY(a)(9) I Material RG 1.99 R2 CF FF IRTNDTra) ARTNDTlb) Margin"' ART""

Method ('IF)

Intermediate Shell Plate Position 1.1 53.1 0.773 0 41.0 34 75 B8805-1 Intermediate Shell Plate Position 1.1 53.1 0.773 20 41.0 34 95 B8805-2 Intermediate Shell Plate Position 1.1 38.4 0.773 30 29.7 29.7 89 B8805-3 Position 2.1 24.4 0.773 30 18.9 17x1) 66 Lower Shell Plate B8606-1 Position 1.1 32.8 0.773 20 25.4 25.4 71 Lower Shell Plate B8606-2 Position 1.1 35.2 0.773 20 27.2 27.2 74 Lower Shell Plate B8606-3 Position 1.1 41.9 0.773 10 32.4 32.4 75 Inter. Shell Longitudinal Weld Position 1.1 34.5 0.622 -80 21.5 21.5 -37 Seam 101-124A (0° Azimuth) Position 2.1 8.6 0.622 -80 5.3 5.3 -69 Inter. Shell Long. Weld Seams Position 1.1 34.5 0.773 -80 26.7 26.7 -27 101-124B,C (1200, 2400 Position 2.1 8.6 0.773 -80 6.6 6.6 -67 Azimuth)

Intermediate to Lower Shell Position 1.1 34.5 0.773 -80 26.7 26.7 -27 Girth Weld Seam 101-171 Position 2.1 8.6 0.773 -80 6.6 6.6"' -67 Lower Shell Long. Weld Seams Position 1.1 34.5 0.773 -80 26.7 26.7 -27 101-142A,C (60°' 300° Position 2.1 8.6 0.773 -80 6.6 6.6 -67 Azimuth)

Lower Shell Long. Weld Seam Position 1.1 34.5 0.622 -80 21.5 21.5 -37 101-142B (180° Azimuth) Position 2.1 8.6 0.622 -80 5.3 5.3T -69 Notes:

a. Calculation of ART values was based on calculated fluence projections and measured data from the previous capsule analysis results (Capsule V, Ref. 10). It should be noted that the fluence projections from Capsule V (Ref. 10) were higher than those from Capsule X (Ref. 12). However, the chemistry factors determined from surveillance data (Table 5-2) are now higher but still lower than the corresponding Position 1.1 chemistry factor.

Therefore, the limiting ART values from Plate B8805-2 could not be exceeded by an updated plate B8805-3 ART value or updated weld ART values. Thus, the original limiting ART value is still valid. Note, the surveillance plate data is now deemed 'Non-Credible," but this has no impact based on the preceding discussion.

b. Initial RTNDT values are measured values.

c. ARTNDT = CF ^ FF

d. M = 2 *(a,2 + q2)M

e. ART = Initial RTNDT + ARTNDT + Margin (0F); (Rounded per ASTM E29, using the 'Rounding Method").

f. Weld data deemed credible per References 10 and 12. Plate data is now deemed "Non-Credible." See Note a."

g. Neutron Fluence value used for all material is the highest value from Reference 10 for 36 EFPY with exception to intermediate shell longitudinal weld 101 -124A and lower shell longitudinal weld 101 -142B which used the fluence at O0from Table 5-4 for 36 EFPY. It should be noted that the fluence projections from Reference 10 are higher than the updated fluence analysis in Reference 12. See Note 'a."

Revision 2 15 15 Revision 2

Voqtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 1 Pressure Temperature Limits Report Southern Nuclear Table 5-7 Summary of the Vogtle Unit 1 Reactor Vessel Beltline Material ART Values Material RG 1.99 R2 1/4 ART 3/4 ART Method (OF) ('F)

Intermediate Shell Plate B8805-1 Position 1.1 90 75 Intermediate Shell Plate B8805-2 Position 1.1 110 95 Intermediate Shell Plate B8805-3 Position 1.1 105 89 Position 2.1 73 66 Lower Shell Plate B8606-1 Position 1.1 89 71 Lower Shell Plate B8606-2 Position 1.1 91 74 Lower Shell Plate B8606-3 Position 1.1 88 75 Inter. Shell Longitudinal Weld Position 1.1 -18 -37 Seam 101 -124A (00 Azimuth) Position 2.1 -65 -69 Inter. Shell Long. Weld Seams Position 1.1 -7 -27 101 -124B,C (120°, 240° Azimuth) Position 2.1 -62 -67 Intermediate to Lower Shell Position 1.1 -7 -27 Girth Weld Seam 101 -171 Position 2.1 -62 -67 Lower Shell Long. Weld Seams Position 1.1 -7 -27 101-142A,C (60°, 3000 Azimuth) Position 2.1 -62 -67 Lower Shell Long. Weld Seam Position 1.1 -18 -37 101 -142B (180° Azimuth) Position 2.1 -65 -69 16 Revision 2 16 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Vogtle Unit 1 Pressure Temperature Limits Report t Southern Nuclear Southern Nuclear Table 5-8 RTPT5 Calculations for Vogtle Unit 1 Beltline Region Materials at 36 EFPY( 11 Material RG 1.99 R2 CF FF IRTNOT(U) ART pTS(b) Margin(c) RTprs(d)

Method (OF)

Intermediate Shell Plate Position 1.1 53.1 1.19 0 63.2 34 97 B8805-1 Intermediate Shell Plate Position 1.1 53.1 1.19 20 63.2 34 117 B8805-2 Intermediate Shell Plate Position 1.1 38.4 1.19 30 45.7 34 110 B8805-3 Position 2.1 37.8 1.19 30 45.0 34(e) 109 Lower Shell Plate B8606-1 Position 1.1 32.8 1.19 20 39.0 34 93 Lower Shell Plate B8606-2 Position 1.1 35.2 1.19 20 41.9 34 96 Lower Shell Plate B8606-3 Position 1.1 41.9 1.19 10 49.9 34 94 Inter. Shell Long. Weld Seams Position 1.1 34.5 1.19 -80 41.1 41.1 2 101-124A, B, C Position 2.1 20.8 1.19 -80 24.8 24.8 -30 (00, 1200, 2400 Azimuth)

Intermediate to Lower Shell Position 1.1 34.5 1.19 -80 41.1 41.1 2 Girth Weld Seam 101-171 Position 2.1 20.8 1.19 -80 24.8 24.8 -30 Lower Shell Long. Weld Seams Position 1.1 34.5 1.19 -80 41.1 41.1 2 101-142A, B, C Position 2.1 20.8 1.19 -80 24.8 24.8 -30 (600,1800, 300° Azimuth) I Notes:

a. Initial RTNDT values are measured values.

b. ARTpTs = CF FF

c. M=2^(o+ c2)t"

d. RTPTS = RTNDT(u) + ARTpTs + Margin (OF)

e. Data deemed 'Non-Credible" per Reference 12.

I. Neutron Fluence value used for all material is the highest interpolated value from Table 5-4 for 36 EFPY.

I 17 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report t Southern Nuclear Southern Nuclear Vogtle Unit 1 Pressure Temperature Limits Report Table 5-9 RTpTs Calculations for Vogtle Unit 1 Beltline Region Materials at 54 EFPY(Q Material RG 1.99 R2 CF FF IRTNDT(U)(a) ARTPTs(b) Margin(c) RTpTs~d Method (OF)

Intermediate Shell Plate Position 1.1 53.1 1.29 0 34 34 103 B8805-1 Intermediate Shell Plate Position 1.1 53.1 1.29 20 34 34 123 B8805-2 Intermediate Shell Plate Position 1.1 38.4 1.29 30 34 34 114 B8805-3 Position 2.1 37.8 1.29 30 34 34(e) 113 Lower Shell Plate B8606-1 Position 1.1 32.8 1.29 20 34 34 96 Lower Shell Plate B8606-2 Position 1.1 35.2 1.29 20 34 34 99 Lower Shell Plate B8606-3 Position 1.1 41.9 1.29 10 34 34 98 Inter. Shell Longitudinal Position 1.1 34.5 1.29 -80 44.5 44.5 9 Weld Seams 101-124A,B,C (0°1200, 2400 Azimuth)

Position 2.1 20.8 1.29 -80 26.8 26.8 -26 Intermediate to Lower Shell Position 1.1 34.5 1.29 -80 44.5 44.5 9 Girth Weld Seam 101-1 71 Position 2.1 20.8 1.29 -80 26.8 26.8 -26 Lower Shell Long. Weld Seams Position 1.1 34.5 1.29 -80 44.5 44.5 9 101-142A, B. C (600, 1800, 3000 Azimuth) I Position 2.1 20.8 1.29 -80 26.8 26.8 -26 Notes:

a. Initial RTNDT values are measured values.

b. ARTpTs = CF

• FF

c. M= 2 (a,2 + %2)1"

d. RTpTs = RTNDT(u) + ARTpTs + Margin (OF)

e. Data deemed 'Non-Credible' per Reference 12.

f. Neutron Fluence value used for all material is the highest value from Table 5-4 for 54 EFPY.

18 Revision 2

Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear 6.0 References

1. WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J.D. Andrachek, et. al.

2. WCAP-16142-P, Revision 1, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units 1 and 2," Warren Bamford, et. al.,

February 2004.

3. Code of Federal Regulations,1 OCFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.

4. WCAP-11011, Georgia Power Company Alvin W. Vogtle Unit No. 2 Reactor Vessel Radiation Surveillance Program, L. R. Singer, February 1986.

5. ASTM E23 Standard Test Method Notched Bar Impact Testing of Metallic Materials, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

6. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section Xl, Division 1," February 26, 1999.

7. Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness Criteria for Protection Against Failure.

8. ASTM El 85-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.

9. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988.

10. WCAP-1 5067, Analysis of Capsule V From the Southern Nuclear Vogtle Electric Generating Plant Unit 1 Reactor Vessel Radiation Surveillance Program, T.J.

Laubham, et. al., dated September 1998. [Note that the Testing/Analysis reports for surveillance capsules U and Y from Vogtle Unit 1 were documented under WCAP-12256 and WCAP-13931, Rev. 1, respectively.]

11. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 5.0.2, developed byATI Consulting, February 2003.

12. WCAP-1 6278, Analysis of Capsule X From the Southern Nuclear Operating Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program, K.G.

Knight, et. al., dated July 2004.

19 Revision 2

Southern Nuclear Company Vogtle Unit 2 Pressure Temperature Limits Report Revision 2, April 2005

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Table of Contents List of Tables......... iii List of Figures .. iv 1.0 RCS Pressure Temperature Limits Report (PTLR) ................................................. 1 2.0 Operating Limits ................................................ 1 2.1 RCS Pressure and Temperature (PiT) Limits (LCO 3.4.3) ................................... 1 3.0 Cold Overpressure Protection Systems (COPS) (LCO 3.4.12) ........................................ 1 3.1 Pressurizer PORV Setpoints ................................................. 2 3.2 Arming Temperature ................................................. 2 4.0 Reactor Vessel Material Surveillance Program ................................................. 2 5.0 Supplemental Data Tables ................................................. 3 6.0 References ............................................... 19 ii Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report eotSuhr Southern Nuclearula VoteUi -Pesr eprtreLmt List of Tables Table 2-1 Vogtle Unit 2 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) ............................................... 6 Table 2-2 Vogtle Unit 2 Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) .7 Table 3-1 Vogtle Unit 2 Data Points for COPS PORV Setpoints .8 Table 5-1 Comparison of the Vogtle Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions ......................... 10 Table 5-2 Calculation of Chemistry Factors Using Vogtle Unit 2 Surveillance Capsule Data ........ 11 Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 2 ....... 12 Table 5-4 Peak Calculated Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit 2 (10'9 n/cm 2, E > 1.0 MeV) . 13 Table 5-5 Vogtle Unit 2 Calculation of the Adjusted Reference Temperature (ART) Values for the 1/4T Location © 36 EFPY .14 Table 5-6 Vogtle Unit 2 Calculation of the ART Values for the 3/4T Location @ 36 EFPY 15 Table 5-7 Summary of the Vogtle Unit 2 Reactor Vessel Beltline Material ART Values. 16 Table 5-8 RTPTS Calculations for Vogtle Unit 2 Beltline Region Materials at 36 EFPY . 17 Table 5-9 RTPTs Calculations for Vogtle Unit 2 Beltline Region Materials at 54 EFPY . 18 iii Revision 2

Vogtle Unit 2- Pressure Temperature Limits Report Southern Nuclear List of Figures Figure 2-1 Vogtle Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 1000F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) .................................................... 4 Figure 2-2 Vogtle Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) .5 Figure 3-1 Vogtle Unit 2 Maximum Allowable Nominal PORV Setpoints for COPS . 9 Revision 2 iv iv Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear 1.0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for Vogtle Unit 2 has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.6. The TS addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Protection Systems (COPS)

Revisions to the PTLR shall be provided to the NRC after issuance.

2.0 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)

The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodology in WCAP-1 4040, Revision 4111 with exception of WCAP-1 6142-P, Revision 1121 (elimination of the flange requirement). The operability requirements associated with the COPS are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of a cold overpressure transient in accordance with the methodology specified in TS 5.6.6.

2.1 RCS P/T Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 600F.

2.1.2 The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100F in any 1-hour period.

b. A maximum cooldown rate of 100F in any 1-hour period.

c. A maximum temperature change of less than or equal to 100F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2-1 and 2-2.

3.0 Cold Overpressure Protection Systems (LCO 3.4.12)

The setpoints for the pressurizer Power Operated Relief Valves (PORVs) and arming temperature are presented in the subsections which follow. These setpoints and arming temperature have been developed using the NRC-approved methodology specified in TS 5.6.6.

1 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear 3.1 Pressurizer PORV Setpoints The pressurizer PORV setpoints are specified in Figure 3-1 and Table 3-1.

The limits for the COPS setpoints are contained in the 36 EFPY steady-state curves (Table 2-2), which are beltline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 74 psi.

Note: These setpoints include an allowance for the 500 F thermal transport effect for heat injection transients. A calculation has been performed to confirm that the setpoints will maintain the system pressure within the established limits when the pressure difference between the pressure transmitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.

3.2 Arming Temperature COPS shall be armed when any RCS cold leg temperature is < 2200 F.

4.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedule is provided in UFSAR Table 5.3.1-9. The results of these examinations shall be used to update Figures 2-1, 2-2, and 3-1.

The pressure vessel steel surveillance program (WCAP-1138114 1)is in compliance with Appendix H1 3] to 10 CFR 50, 'Reactor Vessel Material Surveillance Program Requirements." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23 151. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640[5l of Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure"m. The surveillance capsule removal schedule meets the requirements of ASTM E185-82181 . The removal schedule is provided in UFSAR Table 5.3.1-9.

Revision 2 2 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report eotSuhr Southern Nuclear ula VoteUi -Pesr eprtreLmt 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2l9], predictions.

Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsule data. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure from Regulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 1.19.

Table 5-3 provides the required Vogtle Unit 2 reactor vessel toughness data.

Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and the PTS evaluation.

Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 36 EFPY for each beltline material in the Vogtle Unit 2 reactor vessel.

The limiting beltline material was the lower shell plate R8-1.

Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Vogtle Unit 2 reactor vessel beltline materials at the 1/4T and 3/4T locations for 36 EFPY.

Table 5-8 provides RTPTS values for Vogtle Unit 2 at 36 EFPY.

Table 5-9 provides RTPTS values for Vogtle Unit 2 at 54 EFPY.

3 Revision 2 3 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Vogtle Unit 2- Pressure Temperature Limits Report Southern Nuclear Nuclear MATERIAL PROPERTY BASIS Limiting Material: Lower Shell Plate R8-1 Limiting ART Values at 36 EFPY: 1/4T, 120 0F 3/4T, 1070F 2500 2250 2000 1750 a

in X 1500 3a, 2 1250 0.

M IU 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-1 Vogtle Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) (Plotted Dataprovided on Table 2-1)

Revision 2 4 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear MATERIAL PROPERTY BASIS Limiting Material: Lower Shell Plate R8-1 Limiting ART Values at 36 EFPY: 1/4T, 1200F 3/4T, 107 0F 2500 2250 2000 1750

e. 1500 0

07 en 2 1250 5

LU 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-2 Vogtle Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1 0 F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Error) (PlottedData provided on Table 2-2) 5 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Nuclear Vogtle Unit 2- Pressure Temperature Limits Report Southern Table 2-1 Vogtle Unit 2 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) 60°F/hr Heatup 60°1/hr Heatup Criticality Limit j 100lF/hr Heatup 1000F/hr Heatup Criticalit Limit Leak Test Limit T l P T I P T I P T I P T I P 60 0 180 0 60 0 180 0 163 2000 60 717 180 734 60 684 180 684 180 2485 65 717 180 727 65 684 180 684 70 717 180 720 70 684 180 684 75 717 180 717 75 684 180 684 80 717 180 717 80 684 180 684 85 717 180 721 85 684 180 684 90 721 180 729 90 684 180 684 95 729 180 739 95 684 180 684 100 739 180 753 100 684 180 686 105 753 180 769 105 686 180 691 110 769 180 788 110 691 180 699 115 788 180 811 115 699 180 709 120 811 180 837 120 709 180 722 125 837 180 866 125 722 180 737 130 866 180 899 130 737 180 756 135 899 180 936 135 756 180 777 140 936 185 977 140 777 185 801 145 977 190 1023 145 801 190 829 150 1023 195 1074 150 829 195 860 155 1074 200 1130 155 860 200 896 160 1130 205 1193 160 896 205 935 165 1193 210 1262 165 935 210 979 170 1262 215 1339 170 979 215 1029 175 1339 220 1424 175 1029 220 1084 180 1424 225 1517 180 1084 225 1144 185 1517 230 1621 185 1144 230 1212 190 1621 235 1735 190 1212 235 1287 195 1735 240 1861 195 1287 240 1369 200 1861 245 2000 200 1369 245 1461 205 2000 250 2154 205 1461 250 1562 210 2154 255 2324 210 1562 255 1673 215 2324 215 1673 260 1796 220 1796 265 1932 225 1932 270 2082 230 2082 275 2248 235 2248 280 2430 240 2430 6 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Voqtle Unit 2- Pressure Temperature Limits Report Southern Nuclear Table 2-2 Vogtle Unit 2 Cooldown Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors)

Steady State I 2 0 °F1hr 4 0 °F/hr I 60°Flhr I 100°Flhr P 6T0 T I P I P T0T P T I P 60 0 60 0 60 0 60 0 60 0 60 722 60 681 60 640 60 599 60 518 65 734 65 694 65 654 65 614 65 537 70 747 70 708 70 670 70 632 70 557 75 762 75 724 75 687 75 651 75 581 80 778 80 742 80 706 80 672 80 606 85 796 85 761 85 728 85 695 85 635 90 816 90 783 90 752 90 721 90 667 95 838 95 807 95 778 95 750 95 703 100 862 100 833 100 807 100 782 100 742 105 889 105 863 105 839 105 818 105 786 110 918 110 895 110 875 110 858 110 834 115 951 115 931 115 914 115 901 115 888 120 987 120 971 120 958 120 950 120 948 125 1027 125 1015 125 1007 125 1003 130 1071 130 1063 130 1060 135 1120 135 1117 140 1173 145 1233 150 1299 155 1371 160 1452 165 1541 170 1639 175 1747 180 1867 185 2000 190 2146 195 2308 Revision 2 7

7 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 2- Pressure Temperature Limits Report Southern Nuclear Table 3-1 Vogtle Unit 2 Data Points for the Maximum Allowable Nominal COPS PORV Setpoints Temperature PORV Setpoint (Deg.F) (psig) 70 580 90 580 140 612 201 760 202 760 350 760 Revision 2 8

8 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Vocjtle Unit 2- Pressure Temperature Limits Report Southern Nuclear Nuclear 0e z

0~

a-w co a:

0 EL

-J 0

-J

-J x

-J z

0 z

400' 0 50 100 150 200 250 300 350 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (DEG F)

Figure 3-1: Vogtle Unit 2 Maximum Allowable Nominal PORV Setpoints for COPS 9 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Table 5-1 Comparison of the Vogtle Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 1019 nicm2) (oF) (a) (OF) (b) (%/")(a) (0%)(c)

Lower Shell U 0.356 22.2 2.0 15 0 Plate B8628-1(e) Y 1.12 31.9 5.8 19.5 0 (Longitudinal X 1.78 36.0 29.4 22 3 W 2.98 40.0 39.0 25 6 Lower Shell U 0.356 22.2 0.077 15 0 Plate B8628-1e) Y 1.12 31.9 1.9 19.5 0 (Transverse) X 1.78 36.0 29.8 22 7 W 2.98 40.0 45.5 25 1 Weld Metal ° U 0.356 26.0 o.oota J 15 0 Y 1.12 37.5 18.7 19.5 7 X 1.78 42.2 19.9 22 7 W 2.98 47.0 31.4 25 5 HAZ Metal U 0.356 ... O.O.0'1 ... 0 Y 1.12 - -- .077 - -- 0 X 1.78 --- o.oI--- 7

_ _ _W 2.98 3.2 --- 6 Notes:

a. Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

b. Calculated using measured Charpy data plotted using CVGRAPH, Version 5.0.2"1

c. Values are based on the definition of upper shelf energy given in ASTM El 85-82181.

d. Actual values for ARTNDT are -7.1 (Plate), -17.3 (Weld), -24.3 (HAZ Cap. U), -10.1 (HAZ Cap. Y) and

-2.5 (HAZ Cap. X). This physically should not occur, therefore for conservatism a value of zero will be reported (i.e.,No Change in T3o).

e. The heat number for lower shell plate B8628-1 is C-3500-2.

f. The Surveillance weld was fabricated from Wire Heat No. 87005, Flux Type Linde 124, Flux Lot No. 1061.

10 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Table 5-2 Calculation of Chemistry Factors using Vogtle Unit 2 Surveillance Capsule Data Material Capsule Capsule I FFBT ARTNDT (C) FF*ARTNDT FF4 Lower Shell" U 0.356 0.715 2.0 1.43 0.511 Plate B8628-1 Y 1.12 1.03 5.8 5.97 1.06 (Longitudinal) X 1.78 1.16 29.4 34.10 1.35 W 2.98 1.29 39.0 50.31 1.66 Lower Shell"' U 0.356 0.715 0O.e) 0.00 0.511 Plate B8628-1 V 1.12 1.03 1.9 1.96 1.06 (Transverse) X 1.78 1.16 29.8 34.57 1.35 W 2.98 1.29 45.5 58.70 1.66 SUM: 187.04 9.162 CFU8 626 1 = X(F* RTNDT) + V( FF2 ) = (187.04) + (9.162) = 20.4'F Surveillance Weld U 0.356 0.715 _ 0_v_ 0.00 0.511 Materiallg)

(Heat # 87005)

Y X

W 1.12 1.78 2.98 1.03 1.16 1.29 22.25( 18.7) 23.68(19.

37.37(31.4) j)V)

) 22.92 27.47 48.21 1.06 1.35 1.66 SUM: 98.60 4.581 (4.581) = 21.5 0 F 2

CFsurv. Weld = X(FF

• RTNDT) + J( FF ) = (98.60) +

Notes:

a. f = Calculated fluence from capsule W dosimetry analysis results (12), (X10'9 n/cm 2, E > 1.0 MeV).

b. FF = fluence factor= f(02 8 - 0.11og Q C. ARTNDT values are the measured 30 ft-lb shift values taken from App. C of Ref. 12.

d. The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of 1.19.

e. Actual values for ARTNDT are -7.1 (Plate) and -17.3 (Weld). This physically should not occur; therefore for conservatism a value of zero will be used.

f. The heat number for lower shell plate B8628-1 is C-3500-2.

g. Surveillance Weld was fabricated from Wire Heat No. 87005, Flux Type Linde 124, Flux Lot No. 1061.

Revision 2 1111 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Vogtle Unit 2 Material Description Cu(°/% Ni %) l Initial RTNDT~a Closure Head Flange R7-1 ... 0.72 100 F (Heat# 125L630VA 1)

Vessel Flange R1-1 --- 0.87 -600 F Intermediate Shell Plate R4-1 0.07 0.63 100F (Heat # C-3527-1)

Intermediate Shell Plate R4-2 0.06 0.61 100 F (Heat # C-3527-2)

Intermediate Shell Plate R4-3 0.05 0.60 300F (Heat # C-3552-1)

Lower Shell Plate B8825-1 0.06 0.62 400F (Heat # C-3500- 1)

Lower Shell Plate R8-1 0.07 0.63 400 F (Heat# C-4304- 1)

Lower Shell Plate B8628-1 0.05 0.59 500 F (Heat # C-3500-2)

Intermediate Shell Longitudinal Weld 0.05 0.15 -100 F Seams 101-124A, B & C Lower Shell Longitudinal Weld Seams 0.05 0.15 -100 F 101-142A, B & C Intermediate to Lower Shell Plate 0.05 0.15 -300 F Circumferential Weld Seam 101-171 Surveillance WeldD 0.04 0.13 Notes:

a. The initial RTNDT values for the plates and welds are based on measured data.

b. The weld material in the Vogtle Unit 2 surveillance program was made of the same wire and flux as the reactor vessel intermediate to lower shell girth seam weld (101-171). These welds were fabricated using weld wire heat no. 87005, Linde 124 Flux, lot no. 1061. The intermediate shell longitudinal weld seams (101-124A,B,C) and the lower shell longitudinal weld seams (101-142A,B,C) were fabricated using weld wire heat no. 87005, Linde 0091 Flux, lot no. 0145. Hence the surveillance weld is representative of all beltline welds.

12 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 2- Pressure Temperature Limits Report Southern Nuclear Table 5-4 Peak Calculated Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Vogtle Unit 2 (n/cm 2 , E > 1.0 MeV)

Azimuthal Location EFPY 00 150 300 450 13.29(a) 4.01 E+18 5.91 E+18 7.02E+18 7.20E+18 14.59 4.42E+18 6.52E+18 7.70E+18 7.87E+18 32.00 9.87 E+ 18 1.46E+19 1.69E+19 1.70E+19 40.00 1.24E+19 1.84E+19 2.12E+19 2.11 E+19 48.00 1.49E+19 2.21 E+19 2.54E+19 2.53E+19 54.00 1.68E+19 2.49E+19 2.86E+19 2.85E+19 (a) The end of cycle 10 (Actual) when Capsule 'W" was withdrawn.

13 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Southern Nuclear Vogtle Unit 2- Pressure Temperature Limits Report Table 5-5 Vogtle Unit 2 Calculation of the ART Values for the 1/4T Location @ 36 EPpyw)(9) I Material RG 1.99 R2 CF FF IRTNDTb) ARTNDT(C Margin'd) ARTVe Method (,IF)

Intermediate Shell Plate R4-1 Position 1.1 44.0 1.051 10 46.2 34 90 Intermediate Shell Plate R4-2 Position 1.1 37.0 1.051 10 38.9 34 83 Intermediate Shell Plate R4-3 Position 1.1 31.0 1.051 30 32.6 32.6 95 Lower Shell Plate B8825-1 Position 1.1 37.0 1.051 40 38.9 34 113 Lower Shell Plate R8-1 Position 1.1 44.0 1.051 40 46.2 34 120 Lower Shell Plate B8628-1 Position 1.1 31.0 1.051 50 32.6 32.6 115 Position 2.1 12.9 1.051 50 13.6 13.6- 77 I

Intermediate Shell Longitudinal Position 1.1 43.3 1.051 -10 45.5 45.5 81 Weld Seams 101-124A, B, C Position 2.1 16.7 1.051 -10 17.6 17.6 25 81 I Lower Shell Longitudinal Position 1.1 43.3 1.051 -10 45.5 45.5 Weld Seams 101-142A, B, C Position 2.1 16.7 1.051 -10 17.6 17.6" 25 I Intermediate to Lower Shell Position 1.1 43.3 1.051 -30 45.5 45.5 61 Circ. Weld Seam 101-171 Position 2.1 16.7 1.051 -30 17.6 17.6"T5 I Notes:

a. Calculation of ART values was based on calculated fluence projections and measured data from the previous capsule analysis results (Capsule X, Ref. 10). It should be noted that the fluence projections from Capsule X (Ref. 10) were higher than those from Capsule W (Ref. 12). However, the chemistry factors determined from surveillance data (Table 5-2) are now higher, but still lower than the corresponding Position 1.1 chemistry factor.

Therefore, the limiting ART values from Plate R8-1 could not be exceeded by an updated plate B8628-1 ART value or updated weld ART values. Thus, the original limiting ART value is still valid.

b. Initial RTNDT values are measured values.

c. ARTNDT = CF

• FF

d. M = 2 *(i++2)2

e. ART = Initial RTNDT + ARTNDT + Margin (0F); (Rounded per ASTM E29, using the 'Rounding Method").

f. Data deemed credible per References 10 and 12.

I

g. Neutron Fluence value used for all material is the highest value from Reference 10 for 36 EFPY. It should be noted that the fluence projections from Reference 10 are higher than the updated fluence analysis in Reference

12. See Note "a."

Revision 2 14 14 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits R/eport Southern Nuclear Nuclear Vogtle Unit 2 Pressure Temperature Limits Report Southern Table 5-6 Vogtle Unit 2 Calculation of the ART Values for the 3/4T Location @ 36 EFpy(a)(g) I Material RG 1.99 R2 CF FF IRTNDT(D) ARTNDT(C) Margin')' ARTO) l Method (OF)

Intermediate Shell Plate R4-1 Position 1.1 44.0 0.763 10 33.6 33.6 77 Intermediate Shell Plate R4-2 Position 1.1 37.0 0.763 10 28.2 28.2 66 Intermediate Shell Plate R4-3 Position 1.1 31.0 0.763 30 23.7 23.7 77 Lower Shell Plate B8825-1 Position 1.1 37.0 0.763 40 28.2 28.2 96 Lower Shell Plate R8-1 Position 1.1 44.0 0.763 40 33.6 33.6 107 Lower Shell Plate B8628-1 Position 1.1 31.0 0.763 50 23.7 23.7 97 Position 2.1 12.9 0.763 50 9.8 9.8 ' 70 Intermediate Shell Longitudinal Position 1.1 43.3 0.763 -10 33.0 33.0 56 Weld Seams 101-124A, B, C Position 2.1 16.7 0.763 -10 12.7 12.71" 15 Lower Shell Longitudinal Position 1.1 43.3 0.763 -10 33.0 33.0 56 Weld Seams 101-142A, B, C Position 2.1 16.7 0.763 -10 12.7 12.7') 15 Intermediate to Lower Shell Position 1.1 43.3 0.763 -30 33.0 33.0 36 Circ. Weld Seam 101-171 Position 2.1 16.7 0.763 -30 12.7 T2i.7m' -5 Notes:

a. Calculation of ART values was based on calculated fluence projections and measured data from the previous capsule analysis results (Capsule X, Ref. 10). It should be noted that the fluence projections from Capsule X (Ref. 10) were higher than those from Capsule W (Ref. 12). However, the chemistry factors determined from surveillance data (Table 5-2) are now higher, but still lower than the corresponding Position 1.1 chemistry factor.

Therefore, the limiting ART values from Plate R8-1 could not be exceeded by an updated plate B8628-1 ART value or updated weld ART values. Thus, the original limiting ART value is still valid.

b. Initial RTNDT values are measured values.

c. ARTNDT = CF

• FF

d. M = 2 '(, 2 + 2

e. ART = Initial RTNDT + ARTNDT + Margin (°F); (Rounded per ASTM E29, using the 'Rounding Method").

f. Data deemed credible per References 10 and 12.

g. Neutron Fluence value used for all material is the highest value from Reference 10 for 36 EFPY. It should be I

noted that the lluence projections from Reference 10 are higher than the updated fluence analysis in Reference

12. See Note "a."

15 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Nuclear Vogtle Unit 2- Pressure Temperature Limits Report Southern Table 5-7 Summary of the Vogtle Unit 2 Reactor Vessel Beltline Material ART Values Material RG 1.99 R2 114 ART 3/4 ART Method ('F) (OF)

Intermediate Shell Plate R4-1 Position 1.1 90 77 Intermediate Shell Plate R4-2 Position 1.1 83 66 Intermediate Shell Plate R4-3 Position 1.1 95 77 Lower Shell Plate B8825-1 Position 1.1 113 96 Lower Shell Plate R8-1 Position 1.1 120 107 Lower Shell Plate B8628-1 Position 1.1 115 97 Position 2.1 77 70 Intermediate Shell Longitudinal Position 1.1 81 56 Weld Seams 101-124A, B, C Position 2.1 25 15 Lower Shell Longitudinal Position 1.1 81 56 Weld Seams 101-142A, B, C Position 2.1 25 15 Intermediate to Lower Shell Position 1.1 61 36 Circ. Weld Seam 101-171 Position 2.1 5 -5 Revision 2 16 16 Revision 2

Vo tLe Unit 2 - Pressure Temperature Limits Report Southern Nuclear Southern Nuclear Voqtle Unit 2- Pressure Temperature Limits Report Table 5-8 0

RTPTS Calculations for Vogtle Unit 2 Beltline Region Materials at 36 EFPY(

Material RG 1.99 R2 CF FF IRTNDT(u)(a) ARTprTS(b) Margin(c) RTps(d)

Method (OF)

Intermediate Shell Plate R4-1 Position 1.1 44.0 1.18 10 51.9 34 96 Intermediate Shell Plate R4-2 Position 1.1 37.0 1.18 10 43.7 34 88 Intermediate Shell Plate R4-3 Position 1.1 31.0 1.18 30 36.6 34 101 Lower Shell Plate B8825-1 Position 1.1 37.0 1.18 40 43.7 34 118 Lower Shell Plate R8-I Position 1.1 44.0 1.18 40 51.9 34 126 Lower Shell Plate B8628-1 Position 1.1 31.0 1.18 50 36.6 34 121 Position 2.1 20.4 1.18 50 24.1 17(e) 91 Inter. Shell Long. Weld Seams Position 1.1 43.3 1.18 -10 51.1 51.1 92 101-124A, B, C Position 2.1 21.5 1.18 -10 25.4 25.4(e) 41 (00, 1200, 2400 Azimuth)

Intermediate to Lower Shell Position 1.1 43.3 1.18 -30 51.1 51.1 72 Girth Weld Seam 101-171 Position 2.1 21.5 1.18 -30 25.4 25.4(e) 21 Lower Shell Long. Weld Seams Position 1.1 43.3 1.18 -10 51.1 51.1 92 101-142A, B, C Position 2.1 21.5 1.18 -10 25.4 25.40) 41 (900, 210°, 330° Azimuth)

Notes:

a. Initial RTNDT values are measured values. I

b. RTpTs = CF ^ FF

c. M = 2 '(. 2 + aa 2)1t2

d. RTpTs = RTNDT(U) + ARTpTs + Margin (OF)

e. Data deemed credible per References 10 and 12. I

f. Neutron Fluence value used for all materials is the highest interpolated value from Table 5-4 for 36 EFPY.

Revision 2 17 17 Revision 2

Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Voqtle Unit 2- Pressure Temperature Limits Report Southern Nuclear Table 5-9 RTPTS Calculations for Vogtle Unit 2 Beltline Region Materials at 54 EFPY(0 Material RG 1.99 R2 CF FF IRTNDTrU)(a) ARTpTs(b) Margin(c) RTpTS(d)

Method (OF)

Intermediate Shell Plate R4-1 Position 1.1 44.0 1.28 10 56.3 34 100 Intermediate Shell Plate R4-2 Position 1.1 37.0 1.28 10 47.4 34 91 Intermediate Shell Plate R4-3 Position 1.1 31.0 1.28 30 39.7 34 104 Lower Shell Plate B8825-1 Position 1.1 37.0 1.28 40 47.4 34 121 Lower Shell Plate R8-1 Position 1.1 44.0 1.28 40 56.3 34 130 Lower Shell Plate B8628-1 Position 1.1 31.0 1.28 50 39.7 34 124 Position 2.1 20.4 1.28 50 26.1 17(e) 93 Inter. Shell Long. Weld Seams Position 1.1 43.3 1.28 -10 55.4 55.4 101 101-124A, B, C Position 2.1 21.5 1.28 -10 27.5 27.5(e) 45 (00, 1200, 2400 Azimuth)

Intermediate to Lower Shell Position 1.1 43.3 1.28 -30 55.4 55.4 81 Girth Weld Seam 101-171 Position 2.1 21.5 1.28 -30 27.5 27.5(e) 25 Lower Shell Long. Weld Seams Position 1.1 43.3 1.28 -10 55.4 55.4 101 101-142A, B, C Position 2.1 21.5 1.28 -10 27.5 27.5(e) 45 (90°, 2100, 330° Azimuth)

Notes:

a. Initial RTNDT values are measured values.

b. ARTpTs = CF FF

c. M = 2 *(i + qG2)"

d. RTpTs = RTNDT(U) + ARTPTS + Margin (OF)

e. Data deemed credible per Reference 10 and 12.

f. Neutron Fluence value used for all materials is the highest value from Table 5-4 for 54 EFPY.

Revision 2 18 18 Revision 2

Voqtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Vogue Unit 2- Pressure Temperature Limits Report Southern Nuclear 6.0 References

1. WCAP-14040-NP-A, Revision 4, 'Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al.

2. WCAP-1 6142-P, Revision 1, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Vogtle Units 1 and 2", Warren Bamford, et. al.,

February 2004.

3. Code of Federal Regulations, 10CFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.

4. WCAP-11381, Georgia Power Company Alvin W. Vogtle Unit No. 2 Reactor Vessel Radiation Surveillance Program, L. R. Singer, April 1986.

5. ASTM E23 Standard Test Method Notched Bar Impact Testing of Metallic Materials, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.

6. ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1," February 26, 1999.

7. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness Criteria for Protection Against Failure.

8. ASTM El 85-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.

9. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988.

10. WCAP-1 5159, Analysis of Capsule X From the Southern Nuclear Vogtle Electric Generating Plant Unit 2 Reactor Vessel Radiation Surveillance Program, T.J.

Laubham, et. al., dated March 1999. [Note that the Testing/Analysis reports for surveillance capsules U and Y from Vogtle Unit 2 were documented under WCAP-13007 and WCAP-14532, respectively.]

11. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 5.0.2 developed by ATI Consulting, February 2003.

12. WCAP-1 6382, Analysis of Capsule W From the Southern Nuclear Operating Company Vogtle Unit 2 Reactor Vessel Radiation Surveillance Program, T.J.

Laubham, et. al., dated January 2005.

Revision 2 19 19 Revision 2