NL-05-1159, 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.
1 9 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 iii Revision 2 iii Revision 2
Vogtle Unit 1 - Pressure Temperature Limits Report l
Southern Nuclear Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear 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 1000F/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 1 00F in any 1 -hour period.
- b. A maximum cooldown rate of 1 00F 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 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.
1 Revision 2 1
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 500F 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 E23151. 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.
2 Revision 2 2
Revision 2
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 Nuclear 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, 110 0F 3/4T, 950F 2500 2250 2000 1750 i
0 0
U 1500 1250 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 1000F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Errors) (Plotted Data provided 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, 950F 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 1000F/hr) Applicable for the First 36 EFPY (Without Margins for Instrumentation Error) (Plotted Data provided on Table 2-2) 5 Revision 2 5
Revision 2
Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear 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 1000F/hr Heatup Criticality Limit Leak Test Limit T- - l P
T I
P I
P T
I P
T l
P 60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 0
747 760 760 760 760 763 770 782 796 815 836 862 891 925 962 1005 1052 1105 1163 1228 1300 1380 1468 1566 1674 1793 1925 2070 2231 2408 170 170 170 170 170 170 170 170 170 170 170 170 170 170 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 0
760 760 760 760 763 770 782 796 815 836 862 891 925 962 1005 1052 1105 1163 1228 1300 1380 1468 1566 1674 1793 1925 2070 2231 2408 60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 0
730 730 730 730 730 730 730 730 733 739 747 759 774 791 812 837 865 897 933 974 1020 1071 1128 1191 1261 1339 1426 1521 1627 1743 1872 2014 2171 2344 170 170 170 170 170 170 170 170 170 170 170 170 170 170 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 0
730 730 730 730 730 730 730 733 739 747 759 774 791 812 837 865 897 933 974 1020 1071 1128 1191 1261 1339 1426 1521 1627 1743 1872 2014 2171 2344 153 170 2000 2485 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 Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear 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 Voqtle Unit 1 -
Pressure Temperature Limits Report Southern Nuclear 900 a-
-O z5 Ujcc 0a-w
-Jm 0-J
-j
-J z
0z 850 800 750 700 650 600 550 500 450 400 (201,76( )
(70,6 12
=
~I a
0 50 100 150 200 250 300 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (DEG F) 350 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/cm2, 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 B8805-3(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.
10 Revision 2 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.
1 1 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 200F (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 200F (Heat # C-0613-2)
Intermediate Shell Plate B8805-3 0.062 0.598 300F (Heat # C-0623-1)
Lower Shell Plate B8606-1 0.053 0.593 200F (Heat # C-2146-1)
Lower Shell Plate B8606-2 0.057 0.60 200F (Heat # C-2146-2)
Lower Shell Plate B8606-3 0.067 0.623 10OF (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
-800F 101-142A, B & C(b)
Circumferential Weld 101-171w'I 0.042 0.102
-800F 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.
12 Revision 2 12 Revision 2
Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 1 - 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 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)
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 (0O Azimuth)
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-'
I 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
- 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 -1 24A and lower shell longitudinal weld 101 -1 42B 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.
- 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 -1 42B which used the fluence at O0 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."
15 Revision 2 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 t
Southern Nuclear Vogtle Unit 1 -
Pressure Temperature Limits Report 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.
- 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 Vogtle Unit 1 - Pressure Temperature Limits Report Southern Nuclear Table 5-9 Unit 1 Beltline Region Materials at 54 EFPY(Q RTpTs Calculations for Vogtle 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)
Position 2.1 20.8 1.29
-80 26.8 26.8
-26 I
Notes:
- a. Initial RTNDT values are measured values.
- b.
ARTpTs = CF
- 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 Southern Nuclear VoteUi
-Pesr eprtreLmt eotSuhr ula 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/cm2, 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 1 000F/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 1 00°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
iv Revision 2 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 1 00F in any 1 -hour period.
- b. A maximum cooldown rate of 1 00F 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 500F 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 H13] 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 E23151. 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.
2 Revision 2 2
Revision 2
Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear VoteUi
-Pesr eprtreLmt eotSuhr ula 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 Nuclear 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, 1070F 2500 2250 2000 1750 ina X 1500 3a, 2 1250 0.
M 1000 IU 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 Data provided on Table 2-1) 4 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, 1070F 2500 2250 2000 1750
- e. 1500 07 0
en 2 1250 5 1000 LU 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) (Plotted Data provided on Table 2-2) 5 Revision 2
Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 2-Pressure Temperature Limits Report Southern Nuclear Table 2-1 Vogtle Unit 2 Heatup Limits at 36 EFPY (Without Uncertainties for Instrumentation Errors) 60°F/hr Heatup 60°1/hr Heatup j 100lF/hr Heatup 1000F/hr Heatup Leak Test Limit Criticality Limit Criticalit Limit T l P
T I
P T
I P
T I
P T
I P
60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 0
717 717 717 717 717 717 721 729 739 753 769 788 811 837 866 899 936 977 1023 1074 1130 1193 1262 1339 1424 1517 1621 1735 1861 2000 2154 2324 180 180 180 180 180 180 180 180 180 180 180 180 180 180 180 180 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 0
734 727 720 717 717 721 729 739 753 769 788 811 837 866 899 936 977 1023 1074 1130 1193 1262 1339 1424 1517 1621 1735 1861 2000 2154 2324 60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 0
684 684 684 684 684 684 684 684 684 686 691 699 709 722 737 756 777 801 829 860 896 935 979 1029 1084 1144 1212 1287 1369 1461 1562 1673 1796 1932 2082 2248 2430 180 180 180 180 180 180 180 180 180 180 180 180 180 180 180 180 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 0
684 684 684 684 684 684 684 684 686 691 699 709 722 737 756 777 801 829 860 896 935 979 1029 1084 1144 1212 1287 1369 1461 1562 1673 1796 1932 2082 2248 2430 163 180 2000 2485 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
20 °F1hr 4 0°F/hr I
60°Flhr I
100°Flhr 6T0 P
T I
P 0T I
P T
P T
I P
60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 0
722 734 747 762 778 796 816 838 862 889 918 951 987 1027 1071 1120 1173 1233 1299 1371 1452 1541 1639 1747 1867 2000 2146 2308 60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 0
681 694 708 724 742 761 783 807 833 863 895 931 971 1015 1063 1117 60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 0
640 654 670 687 706 728 752 778 807 839 875 914 958 1007 1060 60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 0
599 614 632 651 672 695 721 750 782 818 858 901 950 1003 60 60 65 70 75 80 85 90 95 100 105 110 115 120 0
518 537 557 581 606 635 667 703 742 786 834 888 948 7
Revision 2 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 8
Revision 2 8
Revision 2
Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Vocjtle Unit 2-Pressure Temperature Limits Report Southern Nuclear 0e z
0~
a-w co a:0 EL
-J 0
-J
-J x
-J z
0z 400' 0
50 100 150 200 250 300 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (DEG F) 350 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 0 O.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)
Y 1.12 1.03 22.25( 18.7)
)
22.92 1.06 (Heat # 87005)
X 1.78 1.16 23.68(19.
j)V) 27.47 1.35 W
2.98 1.29 37.37(31.4) 48.21 1.66 SUM:
98.60 4.581 CFsurv. Weld = X(FF
- a.
f = Calculated fluence from capsule W dosimetry analysis results (12), (X 10'9 n/cm2, E > 1.0 MeV).
- b.
FF = fluence factor= f(0 2 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.
11 Revision 2 1 1 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 100F (Heat# 125L630VA 1)
Vessel Flange R1-1 0.87
-600F Intermediate Shell Plate R4-1 0.07 0.63 100F (Heat # C-3527-1)
Intermediate Shell Plate R4-2 0.06 0.61 100F (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 400F (Heat# C-4304-1)
Lower Shell Plate B8628-1 0.05 0.59 500F (Heat # C-3500-2)
Intermediate Shell Longitudinal Weld 0.05 0.15
-100F Seams 101-124A, B & C Lower Shell Longitudinal Weld Seams 0.05 0.15
-100F 101-142A, B & C Intermediate to Lower Shell Plate 0.05 0.15
-300F 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/cm2, 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 Vogtle Unit 2-Pressure Temperature Limits Report Southern Nuclear 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 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 Lower Shell Longitudinal Position 1.1 43.3 1.051
-10 45.5 45.5 81 Weld Seams 101-142A, B, C Position 2.1 16.7 1.051
-10 17.6 17.6" 25 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
I I
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.
e.
f.
g.
M = 2 *(i+ +2)2 ART = Initial RTNDT + ARTNDT + Margin (0F); (Rounded per ASTM E29, using the 'Rounding Method").
Data deemed credible per References 10 and 12.
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."
I 14 Revision 2 14 Revision 2
Vogtle Unit 2 - Pressure Temperature Limits R/eport Southern Nuclear Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear 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
- 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 noted that the lluence projections from Reference 10 are higher than the updated fluence analysis in Reference
- 12. See Note "a."
I 15 Revision 2
Vogtle Unit 2 - Pressure Temperature Limits Report Southern Nuclear Vogtle Unit 2-Pressure Temperature Limits Report Southern Nuclear 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 16 Revision 2 16 Revision 2
Vo tLe Unit 2 - Pressure Temperature Limits Report Southern Nuclear Voqtle Unit 2-Pressure Temperature Limits Report Southern Nuclear Table 5-8 Unit 2 Beltline Region Materials at 36 EFPY(0 RTPTS Calculations for Vogtle 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.
- b.
- c.
M = 2 '(. 2 + aa 2)1t2
- d.
RTpTs = RTNDT(U) + ARTpTs + Margin (OF)
- e. Data deemed credible per References 10 and 12.
- f.
Neutron Fluence value used for all materials is the highest interpolated value from Table 5-4 for 36 EFPY.
I I
17 Revision 2 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.
- 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.
18 Revision 2 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.
19 Revision 2 19 Revision 2