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| issue date = 04/30/2009
| issue date = 04/30/2009
| title = WCAP-17009-NP, Revision 1, Analysis of Capsule W from the Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program.
| title = WCAP-17009-NP, Revision 1, Analysis of Capsule W from the Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program.
| author name = Burgos B N, Conermann J M, Hunter M A, Rosier B A
| author name = Burgos B, Conermann J, Hunter M, Rosier B
| author affiliation = Westinghouse Electric Co
| author affiliation = Westinghouse Electric Co
| addressee name =  
| addressee name =  
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=Text=
=Text=
{{#Wiki_filter:Westinghouse Non-Proprietary Class 3 WCAP-17009-NP Revision 1 April 2009 Analysis of Capsule W from the Vogtle Unit No. I Reactor Vessel Radiation urveillance Program/
{{#Wiki_filter:Westinghouse Non-Proprietary Class 3 WCAP-17009-NP                                      April 2009 Revision 1 Analysis of Capsule W from                                    /
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17009-NP Revision 1 Analysis of Capsule W from the Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program B. A. Rosier*B. N. Burgos* for J. M. Conermann M. A. Hunter*April 2009 Reviewer:
the Vogtle
N. R. Jurcevich*Primary Component Asset Management Approved:
P. C. Paesano*, Manager Primary Component Asset Management
*Electronically


==SUMMARY==
==SUMMARY==
The purpose of this report is to document the testing results of the surveillance Capsule W from Vogtle Unit 1. Capsule W was removed at 18.41 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed.
 
A fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database.
The purpose of this report is to document the testing results of the surveillance Capsule W from Vogtle Unit 1. Capsule W was removed at 18.41 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database. Capsule W received a fluence of 4.36 x 10' 9 n/cm 2 (E > 1.0 MeV) after irradiation to 18.41 EFPY. The peak clad/base metal interface vessel fluence after 18.41 EFPY of plant operation was 1.05 x 1019 n/cm 2 (E > 1.0 MeV).
Capsule W received a fluence of 4.36 x 10'9 n/cm 2 (E > 1.0 MeV) after irradiation to 18.41 EFPY. The peak clad/base metal interface vessel fluence after 18.41 EFPY of plant operation was 1.05 x 1019 n/cm 2 (E > 1.0 MeV).This evaluation led to the following conclusions:
This evaluation led to the following conclusions: 1) The measured percent decrease in upper shelf energy for all the surveillance materials contained in the Vogtle Unit 1 Capsule W are less than the Regulatory Guide 1.99, Revision 2 [1] predictions. 2) The Vogtle Unit 1 surveillance plate data is judged to be not credible however the weld data is judged to be credible. This credibility evaluation can be found in Appendix D. 3) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2]. The upper shelf energy evaluation is presented in Appendix E.
: 1) The measured percent decrease in upper shelf energy for all the surveillance materials contained in the Vogtle Unit 1 Capsule W are less than the Regulatory Guide 1.99, Revision 2 [1] predictions.  
Lastly, a brief summary of the Charpy V-notch testing can be found in Section 1. All Charpy V-notch data was plotted using a symmetric hyperbolic tangent curve fitting program.
: 2) The Vogtle Unit 1 surveillance plate data is judged to be not credible however the weld data is judged to be credible.
WCAP-17009-NP                                                                                   April 2009 Revision I
This credibility evaluation can be found in Appendix D. 3) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2]. The upper shelf energy evaluation is presented in Appendix E.Lastly, a brief summary of the Charpy V-notch testing can be found in Section 1. All Charpy V-notchdata was plotted using a symmetric hyperbolic tangent curve fitting program.WCAP-17009-NP April 2009 Revision I 1-1 1  
 
1-1 1      


==SUMMARY==
==SUMMARY==
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule W, the fifth capsule removed and tested from the Vogtle Unit 1 reactor pressure vessel, led to the following conclusions:
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule W, the fifth capsule removed and tested from the Vogtle Unit 1 reactor pressure vessel, led to the following conclusions:
* Charpy V-notch test data were plotted using a symmetric hyperbolic tangent curve-fitting program.Appendix C presents the CVGRAPH, Version 5.3, Charpy V-notch plots for Capsule W and previous capsules, along with the program input data." Capsule W received an average fast neutron fluence (E > 1.0 MeV) of 4.36 xl01 9 n/cm 2 after 18.41 effective full power years (EFPY) of plant operation." Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented withthe longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 83.2°F and an irradiated 50 ft-lb transition temperature of 134.9°F. This results in a 30 ft-lb transition temperature increase of 98.1'F and a 50 ft-lb transition temperature increase of I13'F for the longitudinally oriented specimens.
* Charpy V-notch test data were plotted using a symmetric hyperbolic tangent curve-fitting program.
* Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 111.2°F and an irradiated 50 ft-lb transition temperature of 171.7°F. This results in a 30 ft-lb transition temperature increase of 94.1°F and a 50 ft-lb transition temperature increase of 109.2°F for the transversely oriented specimens." Irradiation of the Surveillance Program Weld Metal (Heat #83653) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -22.4°F and an irradiated 50 ft-lb transition temperature of 1.1°F. This results in a 30 ft-lb transition temperature increase of 34.8°F and a 50 ft-lb transition temperature increase of 31.4°F.* Irradiation of the Heat-Affected-Zone (HAZ) Material Charpy specimens resulted in an irradiated30 ft-lb transition temperature of -41.0°F and an irradiated 50 ft-lb transition temperature of 3.5°F.This results in a 30 ft-lb transition temperature increase of 45. ITF and a 50 ft-lb transition temperature increase of 58.9°F." The average upper shelf energy of the Intermediate Shell Plate B8805-3 (longitudinal orientation) resulted in an average energy decrease of 22 ft-lb after irradiation.
Appendix C presents the CVGRAPH, Version 5.3, Charpy V-notch plots for Capsule W and previous capsules, along with the program input data.
This results in an irradiated average upper shelf energy of 100 ft-lb for the longitudinally oriented specimens.
" Capsule W received an average fast neutron fluence (E > 1.0 MeV) of 4.36 xl019 n/cm 2 after 18.41 effective full power years (EFPY) of plant operation.
* The average upper shelf energy of the Intermediate Shell Plate B8805-3 (transverse orientation) resulted in an average energy decrease of 14 ft-lb after irradiation.
" Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 83.2°F and an irradiated 50 ft-lb transition temperature of 134.9°F. This results in a 30 ft-lb transition temperature increase of 98.1'F and a 50 ft-lb transition temperature increase of I13'F for the longitudinally oriented specimens.
This results in an irradiatedaverage upper shelf energy of 82 ft-lb for the transversely oriented specimens.
* Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 111.2°F and an irradiated 50 ft-lb transition temperature of 171.7°F. This results in a 30 ft-lb transition temperature increase of 94.1°F and a 50 ft-lb transition temperature increase of 109.2°F for the transversely oriented specimens.
WCAP- 17009-NP April 2009 Revision 1 1-2" The average upper shelf energy of the Surveillance Program Weld Metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 125 ft-lb for the weld metal specimens.
" Irradiation of the Surveillance Program Weld Metal (Heat #83653) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -22.4°F and an irradiated 50 ft-lb transition temperature of 1.1°F. This results in a 30 ft-lb transition temperature increase of 34.8°F and a 50 ft-lb transition temperature increase of 31.4°F.
* The average upper shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation.
* Irradiation of the Heat-Affected-Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -41.0°F and an irradiated 50 ft-lb transition temperature of 3.5°F.
This results in an irradiated average upper shelf energy of 114 ft-lb for the HAZ Material.* A comparison of the measured 30 ft-lb shift in transition temperature values for the Vogtle Unit 1reactor vessel surveillance materials is presented in Table 5-10." Based on the credibility evaluation presented in Appendix D, the Vogtle Unit 1 surveillance plate data is not credible but the surveillance weld data is credible.* Based on the upper shelf energy evaluation in Appendix E, all beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end of the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2]." The calculated 36 EFPY (end-of-license) and 54 EFPY (end-of-license renewal) neutron fluence (E > 1.0 MeV) at the core mid-plane for the Vogtle Unit 1 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows: Calculated (36 EFPY): Calculated (54 EFPY): Vessel inner radius*  
This results in a 30 ft-lb transition temperature increase of 45. ITF and a 50 ft-lb transition temperature increase of 58.9°F.
= 2.155 x 101 9 n/cm 2 (Interpolated from Table 6-2)Vessel 1/4 thickness  
" The average upper shelf energy of the Intermediate Shell Plate B8805-3 (longitudinal orientation) resulted in an average energy decrease of 22 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 100 ft-lb for the longitudinally oriented specimens.
= 1.284 x 1019 n/cm 2 Vessel 3/4 thickness  
* The average upper shelf energy of the Intermediate Shell Plate B8805-3 (transverse orientation) resulted in an average energy decrease of 14 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 82 ft-lb for the transversely oriented specimens.
= 4.56 x 101 8 n/cm 2 Vessel inner radius* = 3.30 x 1019 n/cm 2 (Taken from Table 6-2)Vessel 1/4 thickness  
WCAP- 17009-NP                                                                                       April 2009 Revision 1
= 1.967 x 1019 n/cm 2 Vessel 3/4 thickness  
 
= 6.99 x 1018 n/cm 2* Clad/base metal interface.
1-2
WCAP- 17009-NP April 2009 Revision 1 2-1 2 INTRODUCTION This report presents the results of the examination of Capsule W, the fifth capsule removed from the reactor in the continuing surveillance program, which monitors the effects of neutron irradiation on the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressure vessel materials under actual operating conditions.
" The average upper shelf energy of the Surveillance Program Weld Metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 125 ft-lb for the weld metal specimens.
The surveillance program for the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation.
* The average upper shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 114 ft-lb for the HAZ Material.
A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-1 1011, "Georgia Power Company Alvin W. Vogtle Unit No. 1 ReactorVessel Radiation Surveillance Program" [3]. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82 [4], "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Capsule W was removed from the reactor after 18.41 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
* A comparison of the measured 30 ft-lb shift in transition temperature values for the Vogtle Unit 1 reactor vessel surveillance materials is presented in Table 5-10.
This report summarizes the testing of the post-irradiation data obtained from surveillance Capsule W removed from the Southern Nuclear Operating Company Vogtle Unit 1 reactor vessel and discusses the analysis of the data.WCAP- 17009-NP April 2009 WCAP- I17009-NP April 2009 Revision I 3-1 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry.
"   Based on the credibility evaluation presented in Appendix D, the Vogtle Unit 1 surveillance plate data is not credible but the surveillance weld data is credible.
The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment.
* Based on the upper shelf energy evaluation in Appendix E, all beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end of the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2].
The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Vogtle Unit I reactor pressure vessel beltline) are well documented in the literature.
"   The calculated 36 EFPY (end-of-license) and 54 EFPY (end-of-license renewal) neutron fluence (E > 1.0 MeV) at the core mid-plane for the Vogtle Unit 1 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows:
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.
Calculated (36 EFPY):             Vessel inner radius* = 2.155 x 1019 n/cm 2 (Interpolated from Table 6-2) 2 Vessel 1/4 thickness = 1.284 x 1019 n/cm 8    2 Vessel 3/4 thickness = 4.56 x 101 n/cm Calculated (54 EFPY):              Vessel inner radius* = 3.30 x 1019 n/cm 2 (Taken from Table 6-2) 2 Vessel 1/4 thickness = 1.967 x 1019 n/cm Vessel 3/4 thickness = 6.99 x 1018 n/cm2
A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code [5]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E208 [6]) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KI, curve) which appears in Appendix G to Section XI of the ASME Code [5]. The K 1 , curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel.When a given material is indexed to the K 1 , curve, allowable stress intensity factors can be obtained for this material as a function of temperature.
* Clad/base metal interface.
Allowable operating limits can then be determined using these allowable stress intensity factors.RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties.
WCAP- 17009-NP                                                                                     April 2009 Revision 1
The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Vogtle Unit 1 reactor vessel radiation surveillance program, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement.
 
This ART (RTNDT Initial + M + ARTNDT) is used to index the material to the KIc curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.
2-1 2       INTRODUCTION This report presents the results of the examination of Capsule W, the fifth capsule removed from the reactor in the continuing surveillance program, which monitors the effects of neutron irradiation on the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressure vessel materials under actual operating conditions.
WCAP- 17009-NP April 2009 Revision 1 4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Vogtle Unit I reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.The capsules contain specimens made from the following:
The surveillance program for the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-1 1011, "Georgia Power Company Alvin W. Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program" [3]. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82 [4], "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Capsule W was removed from the reactor after 18.41 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of the post-irradiation data obtained from surveillance Capsule W removed from the Southern Nuclear Operating Company Vogtle Unit 1 reactor vessel and discusses the analysis of the data.
April 2009 WCAP- 17009-NP I17009-NP                                                                                 April 2009 Revision I
 
3-1 3         BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Vogtle Unit I reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.
A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code [5]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).
RTNDT   is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E208 [6]) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KI, curve) which appears in Appendix G to Section XI of the ASME Code [5]. The K1 , curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel.
When a given material is indexed to the K1 , curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
RTNDT   and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Vogtle Unit 1 reactor vessel radiation surveillance program, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RTNDT Initial + M + ARTNDT) is used to index the material to the KIc curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.
WCAP- 17009-NP                                                                                       April 2009 Revision 1
 
4-1 4       DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Vogtle Unit I reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.
The capsules contain specimens made from the following:
* Intermediate Shell Plate B8805-3 (longitudinal orientation)
* Intermediate Shell Plate B8805-3 (longitudinal orientation)
* Intermediate Shell Plate B8805-3 (transverse orientation)" Weld metal fabricated with 3/16-inch Mil B-4 weld filler wire, Heat Number 83653 Linde Type 0091 flux, Lot Number 3536, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld and all longitudinal weld seams of both the intermediate and lower shell plates of the pressure vessel." Weld heat-affected-zone (HAZ) material of
* Intermediate Shell Plate B8805-3 (transverse orientation)
    "  Weld metal fabricated with 3/16-inch Mil B-4 weld filler wire, Heat Number 83653 Linde Type 0091 flux, Lot Number 3536, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld and all longitudinal weld seams of both the intermediate and lower shell plates of the pressure vessel.
    "  Weld heat-affected-zone (HAZ) material of Intermediate Shell Plate B8805-1 Test material obtained from the intermediate shell course plate (after thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the '/4 and 3/4 thickness locations of the plate after performing a simulated post-weld stress-relieving treatment on the test material. Test specimens were also removed from weld and heat-affected-zone metal of a stress-relieved weldment joining Intermediate Shell Plate B8805-1 and adjacent Lower Shell Plate B8606-3. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of the Intermediate Shell Plate B8805-1.
Charpy V-notch impact specimens from Intermediate Shell Plate B8805-3 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major rolling direction). The core-region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular to the weld direction. The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.
Tensile specimens from Intermediate Shell Plate B8805-3 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented perpendicular to the weld direction.
Compact Test (CT) specimens from Intermediate Shell Plate B8805-3 were machined in the longitudinal and transverse orientations. Compact test specimens from the weld metal were machined with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399 [7].
All six capsules contained dosimeter wires of pure iron, copper, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Neptunium ( 23 7Np) and Uranium ( 2 38U) were placed in the capsules to measure the integrated flux at specific neutron energy levels.
WCAP- 17009-NP                                                                                    April 2009 Revision 1
 
4-2 The capsules contained


==6.1 INTRODUCTION==
A-5 For the least squares evaluation of the Vogtle Unit I surveillance capsule dosimetry, the FERRET code
[A-6] was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (ý(E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.
The application of the least squares methodology requires the following input:
: 1.      The calculated neutron energy spectrum and associated uncertainties at the measurement location.
: 2.      The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.
: 3.      The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
For the Vogtle Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [A-7]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (1iB)" [A-8].
The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" [A-9].
The following provides a summary of the uncertainties associated with the least squares evaluation of the Vogtle Unit 1 surveillance capsule sensor sets.
WCAP- 17009-NP                                                                                  April 2009 Revision 1


This section describes a discrete ordinates Sn transport analysis performed for the Vogtle Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules.In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule W, withdrawn at the end of the fourteenth plant operating cycle, is provided.
A-6 Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.
In addition, to provide an up-to-date data base applicable to the Vogtle Unit 1 reactor, the sensor sets from the previously withdrawn capsules (U, Y, V, and X) were re-analyzed using the current dosimetry evaluation methodology. These dosimetry updates are presented in Appendix A of this report. Comparisons of the results from these dosimetry evaluations with the analytical predictions served to validate the plant-specific neutron transport calculations.
After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:
These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY).The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.Because of this potential shift away from a threshold fluence toward an energy-dependent damage function for data correlation, ASTM Standard Practice E853-01, "Analysis and Interpretation of Light-Water Reactor Surveillance Results," [21] recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference.
Reaction                    Uncertainty 63          60 Cu(nCC)    Co                    5%
The energy-dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693-01, "Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom" [22]. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials" [I].All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [23]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 [24].WCAP- 17009-NP April 2009 Revision 1 6-2 6.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Vogtle Unit 1 reactor vessel is shown in Figure 4-1.Six irradiation capsules attached to the neutron pad are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.5', 610, 121.5', 238.50, 2410, and 301.50 as shown in Figure 4-1. These full-core positions correspond to the followingoctant symmetric locations represented in Figure 6-1: 29' from the core cardinal axes (for the 610 and 2410 dual surveillance capsule holder locations found in octants with a 22.5' neutron pad segment) and 31.5' from the core cardinal axes (for the 121.50 and 301.5' single surveillance capsule holder locations found in octants with a 20.00 neutron pad segment, and for the 58.50 and the 238.5' dual surveillance capsule holder locations found in octants with a 22.50 neutron pad segment).
54 Fe(n,p)54Mn                        5%
The stainless steel specimen containers are 1.182-inch by 1-inch and are approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.From a neutronic standpoint, the surveillance capsules and associated support structures are significant.
58 Ni(n,p) 58Co                      5%
The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pads and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.In performing the fast neutron exposure evaluations for the Vogtle Unit 1 reactor vessel and surveillance capsules, a series of fuel cycle specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:
23 8 U(n,f) 13 7Cs                    10%
p(r,0, z) = 9(r,0 (r,z) (Eqn. 6-1)where d(r,0,z) is the synthesized three-dimensional neutron flux distribution, o(rO) is the transport solution in r,0 geometry, o(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and ý(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation.
2 37 Np(n,f) 137 Cs                    10%
This synthesis procedure was carried out for each operating cycle at Vogtle Unit 1.For the Vogtle Unit I transport calculations, the rO models depicted in Figure 6-1 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric.
59          60 Co(n,7)    Co                    5%
These rO models include the core, the reactor internals, the neutron pads -including explicit representations of octants not containing surveillance capsules and octants with surveillance capsules at 290 and 31.50, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall.These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations.
These uncertainties are given at the Icy level.
In developing these analytical models, nominal design dimensions were employed for the various structural components.
Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.
Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and WCAP- 17009-NP April 2009 Revision I 6-3 miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,0 reactor models consisted of 183 radial by 99 azimuthal intervals.
For sensors included in the Vogtle Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.
Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the rO calculations was set at a value of 0.001.The rz model used for the Vogtle Unit 1 calculations is shown in Figure 6-2 and extends radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation below the lower core plate to above the upper core plate. As in the case of the rO models, nominal design dimensions and full power coolant densities were employed in the calculations.
WCAP- 17009-NP                                                                                   April 2009 Revision 1
In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of these reactor models consisted of 153 radial by 188 axial intervals.
 
As in the case of the r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis.The pointwise inner iteration flux convergence criterion utilized in the rz calculations was also set at a value of 0.001.The one-dimensional radial model used in the synthesis procedure consisted of the same 153 radial mesh intervals included in the rz model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.The core power distributions used in the plant-specific transport analysis were provided by the Nuclear Fuels Division of Westinghouse for each of the first fifteen fuel cycles at Vogtle Unit 1. Specifically, the data utilized included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions.
A-7 Reaction                      Uncertainty 63          60 Cu(n,c() Co                    4.08-4.16%
This information was used to develop spatial and energy dependent core sourcedistributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle.
54 Fe(n,p) 54Mn                    3.05-3.11%
In constructing these core source distributions, the energy distribution of the source was based on anappropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.
58 Ni(n,p) 58Co                  4.49-4.56%
All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.2 [25] and the BUGLE-96 cross-section library [26]. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications.
238 U(n,f) 137 Cs                 0.54-0.64%
In these analyses, anisotropic scattering was treated with a P 5 legendre expansion and angular discretization was modeled with an S16 order of angular quadrature.
237 Np(n,f) 137 Cs               10.32-10.97%
Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-6. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 290 dual capsule, 31.50 dual capsule, and 31.50 single capsule. These WCAP- 17009-NP 'April 2009 Revision 1 6-4 results, representative of the axial midplane of the active core, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future. Similar information is provided in Table 6-2 for the reactor vessel inner radius at four azimuthal locations.
59 Co(n,7) 6°Co                   0.79-3.59%
The vessel data given in Table 6-2 were taken at the clad/base metal interface, and thus, represent maximum calculated exposure levels on the vessel.From the data provided in Table 6-2 it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the fourteenth fuel cycle (i.e., after 18.41 EFPY of plant operation) was 1.05 x 1019 n/cm 2.Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Tables 6-1 and 6-2. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the fourteenth fuel cycle as well as future projections to 20, 23, 26, 30, 34, 38, 42, 48, and 54 EFPY. The calculations account for uprates from 3411 MWt to 3565 MWt that occurred at the onset of Cycle 5 and from 3565 MWt to 3626 MWt that occurred at the onset of Cycle 15. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 15 were representative of future plant operation.
The future projections are also based on the current reactor power level of 3626 MWt.Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-3 and 6-4, respectively. The data, based on the cumulative integrated exposures from Cycles 1 through 14, are presented on a relative basis for each exposure parameter at several azimuthal locations.
Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-3 and 6-4.The calculated fast neutron exposures for the five surveillance capsules withdrawn from the Vogtle Unit 1 reactor are provided in Table 6-5. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the Vogtle Unit 1 reactor.From the data provided in Table 6-5, Capsule W received a fluence (E > 1.0 MeV) of 4.36 x 101" n/cm 2after exposure through the end of the fourteenth fuel cycle (i.e., after 18.41 EFPY of plant operation).
Updated lead factors for the Vogtle Unit 1 surveillance capsules are provided in Table 6-6. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface.
In Table 6-6, the lead factors for capsules that have been withdrawn from the reactor (U, Y,. V, X, and W) were based on the calculated fluence values for the irradiation period correspondingto the time of withdrawal for the individual capsules.
The lead factor for Capsule Z, which was withdrawn at the end of Cycle 14 and stored in the spent fuel pool, corresponds to the calculated fluence values at the end of Cycle 14, the last completed fuel cycle for Vogtle Unit 1.WCAP-17009-NP April 2009 Revision 1 6-5 6.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures previously reported in Section 6.2 is demonstrated by adirect comparison against the measured sensor reaction rates and via a least squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serves to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least squares evaluation comparisons are documented in Appendix A.The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule W, that was withdrawn from Vogtle Unit 1 at the end of the fourteenth fuel cycle, is summarized below.Reaction Rates (rps/atom)
Reaction Measured Calculated M/C Ratio63 Cu(n, C)6 0 Co 4.51 E- 17 4.13E-17 1.0954 Fe(n,p)5 4 Mn 4.36E-15 4.51E-15 0.97 5 8 Ni(n,p)5 8 Co 6.23E- 15 6.31 E- 15 0.99238 U(n,f)137 Cs (Cd) 2.59E-14 2.40E-14 1.08237 Np(n,f)137 Cs (Cd) 2.17E-13 2.33E-13 0.93 Average: 1.01% Standard Deviation:
7.0 The measured-to-calculated (M/C) reaction rate ratios for the Capsule W threshold reactions range from 0.93 to 1.09, and the average M/C ratio is 1.01 +/- 7.0% (1c). This direct comparison falls well within the+/- 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the Vogtle Unit I reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Vogtle Unit 1.6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Vogtle Unit 1 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190.In particular, the qualification of the methodology was carried out in the following four stages: 1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.
WCAP- 17009-NP April 2009 Revision 1 6-6 3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant specific transport calculations used in the neutron exposure assessments.
: 4. Comparisons of the plant-specific calculations with all available dosimetry results from the Vogtle Unit 1 surveillance program.The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections.
This phase, however, didnot test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations.
The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters.
The overall calculational uncertainty applicable to the Vogtle Unit 1 analysis was established from results of these three phases of the methods qualification.The fourth phase of the uncertainty assessment (comparisons with Vogtle Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures previously described in Section 6.2. As such, the validation of the Vogtle Unit 1 analytical model based on the measured plant dosimetry is completely described in Appendix A.The following summarizes the uncertainties developed from the first three phases of the methodology qualification.
Additional information pertinent to these evaluations is provided in Reference 24.Capsule Vessel IRPCA Comparisons 3% 3%H. B. Robinson Comparisons 3% 3%Analytical Sensitivity Studies 10% 11%Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%Net Calculational Uncertainty 12% 13%The net calculational uncertainty was determined by combining the individual components in quadrature.
Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results.The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for Vogtle Unit 1.WCAP- 17009-NP April 2009 Revision 1 6-7 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)Cumulative Cumulative Neutron Flux (E > 1.0 MeV) [n/cm 2-sl Cycle Irradiation Irradiation Length Time Time Cycle IEFPS] IEFPSI [EFPY] Dual 290 Dual 31.50 Single 31.50 1 3.60E+07 3.60E+07 1.14 8.57E+10 9.20E+10 9.10E+10 2 3.58E+07 7.19E+07 2.28 7.44E+ 10 7.97E+10 7.90E+ 10 3 4.19E+07 1.14E+08 3.61 7.65E+10 8.44E+10 8.36E+10 4 3.92E+07 1.53E+08 4.85 6.13E+10 6.64E+10 6.57E+10 5 4.19E+07 1.95E+08 6.17 6.53E+10 7.07E+10 7.00E+10 6 4.24E+07 2.37E+08 7.52 6.62E+10 7.11E+10 7.04E+10 7 3.97E+07 2.77E+08 8.78 5.94E+10 6.43E+10 6.36E+10 8 4.22E+07 3.19E+08 10.11 6.18E+10 6.80E+10 6.74E+10 9 4.61E+07 3.65E+08 11.57 7.38E+10 8.58E+10 8.52E+10 10 4.28E+07 4.08E+08 12.93 7.63E+10 8.37E+10 8.29E+10 11 4.41E+07 4.52E+08 14.33 7.01E+10 7.79E+10 7.72E+10 12 4.28E+07 4.95E+08 15.68 6.61E+10 7.22E+10 7.15E+10 13 4.41E+07 5.39E+08 17.08 6.71E+10 7.30E+10 7.23E+10 14 4.21E+07 5.81E+08 18.41 6.56E+10 7.15E+10 7.09E+10 Future 5.01E+07 6.31E+08 20.00 7.16E+10 7.93E+10 7.86E+10 Future 9.47E+07 7.26E+08 23.00 7.16E+
10 7.93E+10 7.86E+10 Future 9.47E+07 8.20E+08 26.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 9.47E+08 30.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 1.07E+09 34.00 7.16E+10 7.93E+10 7.86E+10Future 1.26E+08 1.20E+09 38.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 1.33E+09 42.00 7.16E+10 7.93E+10 7.86E+10 Future 1.89E+08 1.51E+09 48.00 7.16E+10 7.93E+10 7.86E+10 Future 1.89E+08 1.70E+09 54.00 7.16E+10 7.93E+10 7.86E+10 Note: (a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.WCAP- 17009-NP April 2009 Revision 1 6-8 Table 6-1 (Continued)
Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)Cumulative Cumulative Neutron Fluence (E > 1.0 MeV) In/cm 2]Cycle Irradiation Irradiation Length Time Time Cycle [EFPSI [EFPSI
[EFPY] Dual 290 Dual 31.5' Single 31.50 1 3.60E+07 3.60E+07 1.14 3.09E+18 3.32E+18 3.28E+18 2 3.58E+07 7.19E+07 2.28 5.76E+18 6.17E+18 6.11E+18 3 4.19E+07 1.14E+08 3.61 8.96E+18 9.71E+18 9.62E+18 4 3.92E+07 1.53E+08 4.85 1.14E+19 1.23E+19 1.22E+19 5 4.19E+07 1.95E+08 6.17 1.41E+19 1.53E+19 1.51E+19 6 4.24E+07 2.37E+08 7.52 1.69E+19 1.83E+19 1.81E+19 7 3.97E+07 2.77E+08 8.78 1.93E+19 2.08E+19 2.06E+19 8 4.22E+07 3.19E+08 10.11 2.19E+19 2.37E+19 2.35E+19 9 4.61E+07 3.65E+08 11.57 2.53E+19 2.77E+19 2.74E+19 10 4.28E+07 4.08E+08 12.93 2.85E+19 3.12E+19 3.1OE+19 11 4.41E+07 4.52E+08 14.33 3.16E+19 3.47E+19 3.44E+19 12 4.28E+07 4.95E+08 15.68 3.45E+19 3.78E+19 3.74E+19 13 4.41E+07 5.39E+08 17.08 3.74E+19 4.10E+19 4.06E+19 14 4.21E+07 5.81E+08 18.41 4.02E+19 4.40E+19 4.36E+19Future 5.01E+07 6.31E+08 20.00 4.38E+19 4.80E+19 4.75E+19 Future 9.47E+07 7.26E+08 23.00 5.05E+19 5.55E+19 5.50E+19 Future 9.47E+07 8.20E+08 26.00 5.73E+19 6.30E+19 6.24E+19 Future 1.26E+08 9.47E+08 30.00 6.64E+19 7.30E+19 7.23E+19 Future 1.26E+08 1.07E+09 34.00 7.54E+19 8.30E+19 8.22E+19 Future 1.26E+08 1.20E+09 38.00 8.44E+19 9.30E+19 9.21E+19 Future 1.26E+08 1.33E+09 42.00 9.35E+19 1.03E+20 1.02E+20 Future 1.89E+08 1.51E+09 48.00 1.07E+20 1.18E+20 1.17E+20 Future 1.89E+08 1.70E+09 54.00 1.21E+20 1.33E+20 1.32E+20 Note: (a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.WCAP-17009-NP April 2009 Revision 1 6-9 Table 6-1 (Continued)
Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)Cumulative Cumulative Iron Atom Displacement Rate Idpa/sl Cycle Irradiation Irradiation Length Time Time Cycle [EFPSI [EFPSI [EFPYI Dual 290 Dual 31.51 Single 31.50 1 3.60E+07 3.60E+07 1.14 1.68E-10 1.80E-10 1.78E-10 2 3.58E+07 7.19E+07 2.28 1.45E-10 1.55E-10 1.53E-10 3 4.19E+07 1.14E+08 3.61 1.49E-10 1.64E-10 1.62E-10 4 3.92E+07 1.53E+08 4.85 1.19E-10 1.29E-10 1.27E-10 5 4.19E+07 1.95E+08 6.17 1.26E-10 1.37E-10 1.35E-10 6 4.24E+07 2.37E+08 7.52 1.28E-10 1.38E-10 1.36E-10 7 3.97E+07 2.77E+08 8.78 1.15E-10 1.24E-10 1.23E-10 8 4.22E+07 3.19E+08 10.11 1.20E-10 1.33E-10 1.31E-10 9 4.61E+07 3.65E+08 11.57 1.43E-10 1.66E-10 1.65E-10 10 4.28E+07 4.08E+08 12.93 1.48E-10 1.62E-10 1.61E-10 11 4.41E+07 4.52E+08 14.33 1.36E-10 1.51E-10 1.50E-10 12 4.28E+07 4.95E+08 15.68 1.28E-10 1.40E-10 1.38E-10 13 4.41E+07 5.39E+08 17.08 1.30E-10 1.41E-10 1.40E-10 14 4.21E+07 5.81E+08 18.41 1.27E-10 1.39E-10 1.37E-10 Future 5.01E+07 6.31E+08 20.00 1.39E-10 1.54E-10 1.53E-10 Future 9.47E+07 7.26E+08 23.00 1.39E-10 1.54E-10 1.53E-10 Future 9.47E+07 8.20E+08 26.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 9.47E+08 30.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 1.07E+09 34.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 1.20E+09 38.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 1.33E+09 42.00 1.39E-10 1.54E-10 1.53E-10 Future 1.89E+08 1.51E+09 48.00 1.39E-10 1.54E-10 1.53E-10 Future 1.89E+08 1.70E+09 54.00 1.39E-10 1.54E-10 1.53E-10 Note: (a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.WCAP- 17009-NP April 2009 WCAP-17009-NP April 2009 Revision I 6-10 Table 6-1 (Continued)
Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)Cumulative Cumulative Iron Atom Displacement Rate Idpa/s]Cycle Irradiation Irradiation Length Time Time Cycle IEFPSI [EFPS] [EFPY] Dual 290 Dual 31.5' Single 31.50 1 3.60E+07 3.60E+07 1.14 6.05E-03 6.48E-03 6.41E-03 2 3.58E+07 7.19E+07 2.28 1.12E-02 1.20E-02 1.19E-02 3 4.19E+07 1.14E+08 3.61 1.75E-02 1.89E-02 1.87E-02 4 3.92E+07 1.53E+08 4.85 2.21E-02 2.39E-02 2.37E-02 5 4.19E+07 1.95E+08 6.17 2.74E-02 2.97E-02 2.94E-02 6 4.24E+07 2.37E+08 7.52 3.28E-02 3.55E-02 3.5 1E-02 7 3.97E+07 2.77E+08 8.78 3.74E-02 4.04E-02 4.OOE-02 8 4.22E+07 3.19E+08 10.11 4.25E-02 4.60E-02 4.55E-02 9 4.61E+07 3.65E+08 11.57 4.91E-02 5.37E-02 5.32E-02 10 4.28E+07 4.08E+08 12.93 5.54E-02 6.06E-02 6.OOE-02 11 4.41E+07 4.52E+08 14.33 6.14E-02 6.73E-02 6.66E-02 12 4.28E+07 4.95E+08 15.68 6.69E-02 7.33E-02 7.26E-02 13 4.41E+07 5.39E+08 17.08 7.26E-02 7.95E-02 7.87E-02 14 4.21E+07 5.81E+08 18.41 7.80E-02 8.53E-02 8.45E-02 Future 5.01E+07 6.31E+08 20.00 8.50E-02 9.30E-02 9.21E-02 Future 9.47E+07 7.26E+08 23.00 9.81E-02 1.08E-01 1.07E-01 Future 9.47E+07 8.20E+08 26.00 1.11E-01 1.22E-01 1.21E-01 Future 1.26E+08 9.47E+08 30.00 1.29E-01 1.42E-01 1.40E-01 Future 1.26E+08 1.07E+09 34.00 1.46E-01 1.61E-01 1.60E-01 Future 1.26E+08 1.20E+09 38.00 1.64E-01 1.81E-01 1.79E-01 Future 1.26E+08 1.33E+09 42.00 1.82E-01 2.OOE-01 1.98E-01 Future 1.89E+08 1.51E+09 48.00 2.08E-01 2.29E-01 2.27E-01 Future 1.89E+08 1.70E+09 54.00 2.34E-01 2.58E-01 2.56E-01 Note: (a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.WCAP- 17009-NP April 2009 Revision 1 6-11 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV) In/cm 2-sI Cycle Irradiation Irradiation Length Time Time Cycle IEFPS] [EFPS] [EFPY] 00 150 300 450 1 3.60E+07 3.60E+07 1.14 1.27E+10 1.89E+10 2.16E+ 10 2.22E+10 2 3.58E+07 7.19E+07 2.28 9.82E+09 1.55E+10 1.87E+10 1.78E+10 3 4.19E+07 1.14E+08 3.61 8.98E+09 1.41E+10 1.92E+10 1.92E+10 4 3.92E+07 1.53E+08 4.85 1.00E+10 1.37E+10 1.58E+10 1.64E+10 5 4.19E+07 1.95E+08 6.17 8.81E+09 1.35E+10 1.64E+10 1.64E+10 6 4.24E+07 2.37E+08 7.52 8.82E+09 1.41E+10 1.68E+10 1.64E+10 7 3.97E+07 2.77E+08 8.78 8.60E+09 1.23E+10 1.51E+10 1.49E+10 8 4.22E+07 3.19E+08 10.11 9.14E+09 1.41E+10 1.59E+10 1.83E+10 9 4.61E+07 3.65E+08 11.57 9.34E+09 1.33E+10 1.89E+10 2.15E+10 10 4.28E+07 4.08E+08 12.93 8.91E+09 1.44E+10 1.91E+10 1.92E+10 11 4.41E+07 4.52E+08 14.33 9.48E+09 1.43E+10 1.78E+10 1.96E+10 12 4.28E+07 4.95E+08 15.68 9.39E+09 1.34E+10 1.68E+10 1.71E+10 13 4.41E+07 5.39E+08 17.08 8.98E+09 1.38E+10 1.70E+10 1.71E+10 14 4.21E+07 5.81E+08 18.41 1.02E+10 1.42E+10 1.66E+10 1.70E+10 Future 5.01E+07 6.31E+08 20.00 9.64E+09 1.46E+10 1.83E+10 2.00E+10 Future 9.47E+07 7.26E+08 23.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10Future 9.47E+07 8.20E+08 26.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 9.47E+08 30.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 1.07E+09 34.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 1.20E+09 38.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 1.33E+09 42.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.89E+08 1.51E+09 48.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.89E+08 1.70E+09 54.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 WCAP- 17009-NP April 2009 Revision 1 6-12 Table 6-2 (Continued)
Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence (E > 1.0 MeV) In/cM 2 I Cycle Irradiation Irradiation Length Time Time Cycle [EFPS] [EFPS] [EFPYJ 00 150 300 450 1 3.60E+07 3.60E+07 1.14 4.58E+17 6.80E+17 7.80E+17 8.00E+17 2 3.58E+07 7.19E+07 2.28 8.10E+17 1.24E+18 1.45E+18 1.44E+18 3 4.19E+07 1.14E+08 3.61 1.19E+18 1.83E+18 2.25E+18 2.24E+18 4 3.92E+07 1.53E+08 4.85 1.57E+18 2.35E+18 2.86E+18 2.86E+18 5 4.19E+07 1.95E+08 6.17 1.94E+18 2.91E+18 3.54E+18 3.55E+18 6 4.24E+07 2.37E+08 7.52 2.31E+18 3.51E+18 4.25E+18 4.24E+18 7 3.97E+07 2.77E+08 8.78 2.65E+18 4.00E+18 4.85E+18 4.83E+18 8 4.22E+07 3.19E+08 10.11 3.03E+18 4.59E+18 5.51E+18 5.60E+18 9 4.61E+07 3.65E+08 11.57 3.46E+18 5.20E+18 6.38E+18 6.59E+18 10 4.28E+07 4.08E+08 12.93 3.84E+18 5.81E+18 7.20E+18 7.41E+18 11 4.41E+07 4.52E+08 14.33 4.26E+18 6.44E+18 7.98E+18 8.27E+18 12 4.28E+07 4.95E+08 15.68 4.66E+18 7.02E+18 8.70E+18 9.OOE+18 13 4.41E+07 5.39E+08 17.08 5.06E+18 7.63E+18 9.45E+18 9.76E+18 14 4.21E+07 5.81E+08 18.41 5.49E+18 8.22E+18 1.02E+19 1.05E+19 Future 5.01E+07 6.31E+08 20.00 5.97E+18 8.95E+18 1.1I1E+19 1.15E+19 Future 9.47E+07 7.26E+08 23.00 6.88E+18 1.03E+19 1.28E+19 1.34E+19 Future 9.47E+07 8.20E+08 26.00 7.79E+18 1.17E+19 1.45E+19 1.53E+19Future 1.26E+08 9.47E+08 30.00 9.01E+18 1.35E+19 1.68E+19 1.78E+19Future 1.26E+08 1.07E+09 34.00 1.02E+19 1.54E+19 1.91E+19 2.03E+19 Future 1.26E+08 1.20E+09 38.00 1.14E+19 1.72E+19 2.14E+19 2.28E+19 Future 1.26E+08 1.33E+09 42.00 1.27E+19 1.91E+19 2.37E+19 2.54E+19 Future 1.89E+08 1.51E+09 48.00 1.45E+19 2.18E+19 2.72E+19 2.92E+19 Future 1.89E+08 1.70E+09 54.00 1.63E+19 2.46E+19 3.07E+19 3.30E+19 WCAP- 17009-NP April 2009 Revision I 6-13 Table 6-2 (Continued)
Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Idpa/sI Cycle Irradiation IrradiationLength Time Time Cycle IEFPSI IEFPS] [EFPY] 00 150 300 450 1 3.60E+07 3.60E+07 1.14 1.97E-11 2.90E-11 3.33E-11 3.51E-11 2 3.58E+07 7.19E+07 2.28 1.53E-11 2.39E-11 2.88E-11 2.82E-11 3 4.19E+07 1.14E+08 3.61 1.40E-11 2.17E-11 2.96E-11 3.03E-11 4 3.92E+07 1.53E+08 4.85 1.56E- 11 2.11E-11 2.44E- 11 2.58E-11 5 4.19E+07 1.95E+08 6.17 1.37E-11 2.08E- 11 2.53E-11 2.59E-11 6 4.24E+07 2.37E+08 7.52 1.37E-11 2.17E-11 2.58E-11 2.59E-11 7 3.97E+07 2.77E+08 8.78 1.34E- 11 1.89E-11 2.33E-11 2.35E-11 8 4.22E+07 3.19E+08 10.11 1.42E-11 2.17E-11 2.45E-11 2.89E-11 9 4.61E+07 3.65E+08 11.57 1.45E-11 2.05E-11 2.91E-11 3.39E-11 10 4.28E+07 4.08E+08 12.93 1.39E-1 1 2.22E-1 1 2.95E-1 1 3.03E-1 1 11 4.41E+07 4.52E+08 14.33 1.48E-11 2.20E-11 2.75E-11 3.09E-11 12 4.28E+07 4.95E+08 15.68 1.46E-I1 2.07E-11 2.59E-11 2.71E-11 13 4.41E+07 5.39E+08 17.08 1.40E-11 2.13E-11 2.62E- 11 2.71E-11 14 4.21E+07 5.81E+08 18.41 1.58E-11 2.19E-11 2.57E-11 2.69E-11 Future 5.01E+07 6.31E+08 20.00 1.50E- 11 2.24E-11 2.82E-11 3.16E-11 Future 9.47E+07 7.26E+08 23.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 Future 9.47E+07 8.20E+08 26.00 1.50E-1 1 2.24E- I1 2.82E- 1I 3.16E-I 1 Future 1.26E+08 9.47E+08 30.00 1.50E-11 2.24E- 11 2.82E- 11 3.16E-11 Future 1.26E+08 1.07E+09 34.00 1.50E- 11 2.24E- 11 2.82E- 11 3.16E-11 Future 1.26E+08 1.20E+09 38.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 Future 1.26E+08 1.33E+09 42.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 Future 1.89E+08 1.51E+09 48.00 1.50E-1I 2.24E- 11 2.82E- 11 3.16E-I 1 Future 1.89E+08 1.70E+09 54.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 WCAP- 17009-NP April 2009 Revision 1 6-14 Table 6-2 (Continued)
Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Idpal Cycle Irradiation Irradiation Length Time Time Cycle [EFPSI IEFPS] [EFPY] 00 150 300 450 1 3.60E+07 3.60E+07 1.14 7.11E-04 1.04E-03 1.20E-03 1.27E-03 2 3.58E+07 7.19E+07 2.28 1.26E-03 1.90E-03 2.23E-03 2.27E-03 3 4.19E+07 1.14E+08 3.61 1.84E-03 2.81E-03 3.47E-03 3.54E-03 4 3.92E+07 1.53E+08 4.85 2.44E-03 3.61E-03 4.40E-03 4.52E-03 5 4.19E+07 1.95E+08 6.17 3.01E-03 4.48E-03 5.46E-03 5.61E-03 6 4.24E+07 2.37E+08 7.52 3.59E-03 5.40E-03 6.56E-03 6.71E-03 7 3.97E+07 2.77E+08 8.78 4.12E-03 6.15E-03 7.48E-03 7.64E-03 8 4.22E+07 3.19E+08 10.11 4.72E-03 7.06E-03 8.50E-03 8.85E-03 9 4.61E+07 3.65E+08 11.57 5.39E-03 8.OOE-03 9.84E-03 1.04E-02 10 4.28E+07 4.08E+08 12.93 5.98E-03 8.95E-03 1.11E-02 1.17E-02 I 1 4.41E+07 4.52E+08 14.33 6.63E-03 9.92E-03 1.23E-02 1.31E-02 12 4.28E+07 4.95E+08 15.68 7.25E-03 1.08E-02 1.34E-02 1.42E-02 13 4.41E+07 5.39E+08 17.08 7.87E-03 1.17E-02 1.46E-02 1.54E-02 14 4.21E+07 5.81E+08 18.41 8.54E-03 1.27E-02 1.57E-02 1.66E-02 Future 5.01E+07 6.31E+08 20.00 9.28E-03 1.38E-02 1.71E-02 1.81E-02 Future 9.47E+07 7.26E+08 23.00 1.07E-02 1.59E-02 1.97E-02 2.11E-02 Future 9.47E+07 8.20E+08 26.00 1.21E-02 1.80E-02 2.24E-02 2.41E-02 Future 1.26E+08 9.47E+08 30.00 1.40E-02 2.09E-02 2.60E-02 2.81E-02 Future 1.26E+08 1.07E+09 34.00 1.59E-02 2.37E-02 2.95E-02 3.21E-02 Future 1.26E+08 1.20E+09 38.00 1.78E-02 2.65E-02 3.31E-02 3.61E-02 Future 1.26E+08 1.33E+09 42.00 1.97E-02 2.93E-02 3.66E-02 4.01E-02 Future 1.89E+08 1.51E+09 48.00 2.25E-02 3.36E-02 4.20E-02 4.61E-02 Future 1.89E+08 1.70E+09 54.00 2.54E-02 3.78E-02 4.73E-02 5.20E-02 WCAP- 17009-NP April 2009 Revision 1 6-15 Table 6-3 Relative Radial Distribution of Neutron Fluence (E > 1.0 MeV) Within the Reactor Vessel Wall(a)Radius Azimuthal Angle (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.571 0.567 0.560 0.558 231.06 0.282 0.277 0.272 0.269 236.54 0.134 0.130 0.127 0.125 242.01 0.064 0.059 0.057 0.056 Base Metal Inner Radius = 220.11 cm Base Metal I/4T = 225.59 cmBase Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Note: (a) Relative radial distribution data are based on the cumulative integrated exposures from Cycles I through 14.WCAP-17009-NP April 2009 Revision 1 6-16 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) Within the Reactor Vessel Wall(a)Radius Azimuthal Angle (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.642 0.637 0.635 0.644 231.06 0.390 0.382 0.381 0.392 236.54 0.237 0.227 0.226 0.234 242.01 0.142 0.128 0.127 0.130 Base Metal Inner Radius = 220.11 cm Base Metal I/4T = 225.59 cm Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Note: (a) Relative radial distribution data are based on the cumulative integrated exposures from Cycles 1 through 14.Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Vogtle Unit 1 Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] In/cm 2] Idpal U 1.14 3.32E+18 6.48E-03 Y 4.85 1.14E+ 19 2.21E-02 V 8.78 1.93E+19 3.74E-02 X 14.33 3.47E+19 6.73E-02 W 18.41 4.36E+19 8.45E-02 WCAP- 17009-NP April 2009 WCAP- 17009-NP April 2009 Revision I 6-17 Table 6-6 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor U (58.5') Withdrawn EOC 1 4.14 Y (24P1) Withdrawn EOC 4 3.97 V (610) Withdrawn EOC 7 3.97X (238.50)
Withdrawn EOC 11 4.19 W (121.5') Withdrawn EOC 14 4.16 Z (301.50) Withdrawn EOC 14; Stored in SFP 4.16(a)Note: (a) Lead factor for Capsule Z, which was withdrawn at the conclusion of Cycle 14 and stored in the spent fuel pool, is based on cycle-specific exposure calculations through the last completed fuel cycle, i.e., Cycle 14.WCAP- 17009-NP April 2009 WCAP-17009-NP April 2009 Revision I 6-18A. W. Vogtle Unrit 1 Reactor R,T Model Meshes- 183R. 916 Ok C ~ ~ h--E~ EU j-Nd-H-I lMaw[Eini Figure 6-1 Vogtle Unit I r,0 Reactor Geometry with a 12.5' Neutron Pad Span at the Core Midplane WCAP- 17009-NP April 2009 Revision I 6-19A. W. Vogfle Unrit 1 Reactor RT Model Moshe 153R. MO ii C -U1-U.--ý moon LU 2~M a ui %R-FlU Figure 6-2 Vogtle Unit 1 rO Reactor Geometry with a 20.00 Neutron Pad Span at the Core Midplane WCAP- 17009-NP April 2009 Revision I 6-20A. W. Vogtle Unit I Reactor RT Model Meshes: 183R. 998---I -m -- W--Rd.U-'I" 1-f.Figure 6-3 Vogtle Unit 1 re Reactor Geometry with a 22.50 Neutron Pad Span at the Core Midplane WCAP- I 7009-NP April 2009 Revision I 6-21A. W. Vogtle UnTI I Reactor RZ Model Meshes: 153U.1U8¥[QN'Figure 6-4 Vogtle Unit I rz Reactor Geometry with Neutron Pad WCAP- 17009-NP April 2009 Revision 1 7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following table summarizes the removal of the six surveillance capsules from the Vogtle Unit 1 reactor vessel, meeting the requirements of ASTM E185-82 [4].Table 7-1 Surveillance Capsule Withdrawal Summary Capsule Capsule Location Lead Factor(a)
Withdrawal EFPY(b) Fluence (n/cm 2)(c)U 58.50 4.14 1.14 3.32E+18 Y 240 3.97 4.85 1.14E+19 V 610 3.97 8.78 1.93E+19 X 238.50 4.19 14.33 3.47E+19 W 121.50 4.16 18.41 4.36E+19 Z 301.50 4.16 18.41 (d)Notes: (a) Updated in Capsule W dosimetry analysis; see Table 6-6.(b) EFPY from plant startup.(c) Updated in Capsule W dosimetry analysis; see Table 6-5.(d) Standby Capsule Z was removed at 18.41 EFPY and placed in the spent fuel pool. No testing or analysis has beenperformed on Capsule Z. This capsule may be re-inserted in the future to fulfill the requirements of the "Additional20-year License Renewal capsule".WCAP-17009-NP April 2009 Revision 1 8-1 8 REFERENCES
: 1. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, 1988.2. 10 CFR 50, Appendix GQ Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.3. WCAP-1 1011, Revision 0, Georgia Power Company Alvin W Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program, L. R. Singer, February 1986.4. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM 1982.5. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G Fracture Toughness Criteria for Protection Against Failure.6. ASTM E208, Standard Test Method for Conducting Drop- Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.7. ASTM E399, Test Method for Plane-Strain Fracture Toughness of Metallic Materials, American Society for Testing Materials.
: 8. Westinghouse Science and Technology Department Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 3, June 6, 2005.9. Westinghouse Science and Technology Department Procedure RMF 8102, Tensile Testing, Revision 3, March 1, 1999.10. Westinghouse Science and Technology Department Procedure RMF 8103, Charpy Impact Testing, Revision 2, August 1, 1998.11. Westinghouse Science and Technology Department Procedure RMF 8804, Opening of Westinghouse Surveillance Capsules, Revision 2, August 1, 2004.12. ASTM E23-06, Standard Test Method for Notched Bar Impact Testing of Metallic Materials, ASTM, 2006.13. General Yielding of Charpy V-Notch and Precracked Charpy Specimens, Journal of Engineering Materials and Technology, Vol. 100, April 1978, pp. 183-188.
: 14. ASTM A370-07, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 2007.WCAP- 17009-NP April 2009 Revision 1 8-2 15. ASTM E8-04, Standard Test Methods for Tension Testing of Metallic Materials, ASTM, 2004.16. ASTM E21-05, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, ASTM, 2005.17. WCAP-12256, Revision 0, Analysis of Capsule Ufrom the Georgia Power Company Vogtle Unit I Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et. al., May 1989.18. WCAP-13931, Revision 1, Analysis of Capsule Y from the Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program, M. J. Malone, et. al., August 1995.19. WCAP-15067, Revision 0, Analysis of Capsule V from Southern Nuclear Vogtle Electric Generating Plant Unit 1 Reactor Vessel Radiation Surveillance Program, T. J. Laubham, et. al., September 1998.20. WCAP-16278-NP, Revision 0, Analysis of Capsule X from the Southern Nuclear Operating Company, Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program, K. G. Knight, et. al.,July 2004.
: 21. ASTM E853-01, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706 (IA), ASTM, 2001 22. ASTM E693-01, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706 (ID), ASTM, 2001.23. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.24. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.25. RSICC Computer Code Collection CCC-650, DOORS 3.2: One, Two- and Three DimensionalDiscrete Ordinates Neutron/Photon Transport Code System, April 1998.26. RSICC Data Library Collection DLC-1 85, BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B- Vfor L WR Shielding and Pressure VesselDosimetry Applications, March 1996.WCAP- 17009-NP April 2009 Revision 1 A-I APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS A.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Vogtle Unit 1 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [A-i]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within +/- 20% as specified by Regulatory Guide 1. 190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report.A.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the five neutron sensor sets analyzed to date as part of the Vogtle Unit 1 Reactor Vessel Materials Surveillance Program are presented.
The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows: Capsule ID Azimuthal Location Withdrawal Time Irradiation Time [EFPY]U 31.5' Dual End of Cycle 1 1.14 Y 29.00 Dual End of Cycle 4 4.85 V 29.00 Dual End of Cycle 7 8.78 X 31.5' Dual End of Cycle 11 14.33 W 31.5' Single End of Cycle 14 18.41 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.The passive neutron sensors included in the evaluations of Surveillance Capsules U, Y, V, X, and W are summarized as follows: WCAP- 17009-NP April 2009 Revision 1 A-2 Since all of the dosimetry monitors located at the radial center of the material test specimen array, radial gradient corrections were not required for these reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table A-1.The use of passive monitors such as those listed above does not yield a direct measure of the energy-dependent neutron flux at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on thevarious monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest: 0 the measured specific activity of each monitor, 0 the physical characteristics of each monitor,* the operating history of the reactor,* the energy response of each monitor, and* the neutron energy spectrum at the monitor location.Results from the radiometric counting of the neutron sensors from Capsules U, Y, V, and X are documented in References A-2 through A-5, respectively.
The radiometric counting of the sensors from Capsule W was carried out by Pace Analytical Services, Inc. In all cases, the radiometric counting followed established ASTM procedures.
Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer.
For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.The irradiation history of the reactor over the irradiation periods experienced by Capsules U, Y, V, X, and W was based on the monthly power generation of Vogtle Unit 1 from initial reactor criticality through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules U, Y, V, X, and W is given in Table A-2.
WCAP- 17009-NP April 2009 Revision I A-3 Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation: A No F Y Z P1 Cj [1- e'ttj] [e-'"d]Pref where: R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus).
A = Measured specific activity (dps/gm).No = Number of target element atoms per gram of sensor.F = Atom fraction of the target isotope in the target element.Y = Number of product atoms produced per reaction.PJ = Average core power level during irradiation period j (MW).Pref = Maximum or reference power level of the reactor (MW).Ci = Calculated ratio of ý(E > 1.0 MeV) during irradiation period j to the time weighted average +(E > 1.0 MeV) over the entire irradiation period.X = Decay constant of the product isotope (1/sec).tj = Length of irradiation period j (sec).td = Decay time following irradiation period j (sec).and the summation is carried out over the total number of monthly intervals comprising the irradiation period.In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref]
accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many WCAP- 17009-NP April 2009 Revision 1 A-4 cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.The fuel cycle specific neutron flux values along with the computed values for Cj are listed in Table A-3.These flux values represent the cycle dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238 U measurements to account for the presence of 235 U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.
Corrections were also made to the 238 U and 237 Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations.
The correction factors applied to the Vogtle Unit 1 fission sensor reaction rates are summarized as follows: Correction Capsule U Capsule Y Capsule V Capsule X Capsule W 235U Impurity/Pu Build-in 0.871 0.840 0.812 0.760 0.73123 8 U(y,f) 0.966 0.967 0.967 0.966 0.969 Net 238 U Correction 0.842 0.813 0.785 0.735 0.708237 Np(y,f) 0.990 0.990 0.990 0.990 0.991 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.
Results of the sensor reaction rate determinations for Capsules U, Y, V, X, and W are given in Tables A-4a through A-4e. In Tables A-4a through A-4e, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for23 8 U impurities, plutonium build-in, and gamma ray induced fission effects.A.1.2 Least Squares Evaluation of Sensor SetsLeast squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties.
Best Estimates for key exposure parameters such as ý(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum.
In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties.
For example, Ri +/- 6Ri = E (g'ig )((Pg +/-5 , )g relates a set of measured reaction rates, Ri, to a single neutron spectrum, 4 g, through the multigroup dosimeter reaction cross-section, trig, each with an uncertainty
: 6. The primary objective of the leastsquares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.
WCAP- 17009-NP April 2009 Revision 1 A-5 For the least squares evaluation of the Vogtle Unit I surveillance capsule dosimetry, the FERRET code[A-6] was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (ý(E
> 1.0 MeV)and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.The application of the least squares methodology requires the following input: 1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.2. The measured reaction rates and associated uncertainty for each sensor contained in the multiplefoil set.3. The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.For the Vogtle Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 6.2 of this report.
The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [A-7]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (1iB)" [A-8].The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances.
The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944,"Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" [A-9].The following provides a summary of the uncertainties associated with the least squares evaluation of the Vogtle Unit 1 surveillance capsule sensor sets.WCAP- 17009-NP April 2009 Revision 1 A-6 Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions.
A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:
Reaction Uncertainty 6 3 Cu(nCC)6 0 Co 5%54 Fe(n,p)54 Mn 5%5 8 Ni(n,p)5 8 Co 5%23 8 U(n,f) 13 7 Cs 10%2 3 7 Np(n,f) 1 3 7 Cs 10%59 Co(n,7)6 0 Co 5%These uncertainties are given at the Icy level.Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, includingcovariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications.
These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations.
Further, the library has been empiricallytested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.For sensors included in the Vogtle Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.WCAP- 17009-NP April 2009 Revision 1 A-7 Reaction Uncertainty 6 3 Cu(n,c()6 0 Co 4.08-4.16%
5 4 Fe(n,p)5 4 Mn 3.05-3.11%58 Ni(n,p)58 Co 4.49-4.56%
2 3 8 U(n,f) 13 7 Cs 0.54-0.64%
2 3 7 Np(n,f) 1 3 7 Cs 10.32-10.97%59 Co(n,7)6°Co 0.79-3.59%
These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.
These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.
Calculated Neutron Snectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:
Calculated Neutron Snectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.
While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:
Mgg = R. + Rg
Mgg = R. + Rg
* Rg *Pgg where R, specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg' specify additional random groupwise uncertainties that are correlated with a correlation matrix
* Rg *Pgg where R, specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg' specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:
Pgg,=[-O]gg,+Oe -H where 2
H - (g -g')
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A-8 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 6 is 1.0 when g = g'
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J --------- L----------- J- ----0A I I-1-I. ,+ --333-Tm.4 4 im)AT36, 100OF WCAP-17009-NP April 2009 Revision 1 B-8 z1xoom I M,0 ra n 10)2 A,- AL -q A'lID0 1.110 2.00 :120 hEm-i (rail AT40O. 110OF&.1fl, SEI 0.1D1 I
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I. 0 1 .C DO ........................... ... ...........A m1m -0ujM 4L.k-.o.00 ,.0O 2.00 11.00 The-I 0"Y)AT 43, 125, WCAP-17009-NP April 2009 Revision 1 B-9---------------------UI I cr ZE .03SI-r~ (MS AT1 .......... ............. .........
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..... ..I I I.!W O.DO~ .... ..................... ... .............. ........O.DD t WN .. ......I .. ............ ......AT31, 180 0 F Reiso I B-10 Thr.-4 i;m)I TqTe-1 (-)j AT34, 200-F Srno.DD.
                                      - I.,+    -                                   -   333-Tm.4 4im)
: x. ..... ...I 4Aoo 1.00 ~ 2.0 I rrr.-1 (fmQ)AT32, 200-F'K SM,WCAP- 17009-NP April 2009 Revision I B-11 20DW.00 ...000 190 ZOO I.O 4.00 5j" OD1.00 2.0030 AAT37, 32.5 0 F 5010 E.00 WCAP-17009-NP April 2009 Revision 1 B- 12!!M 4CO NO:;CD 1CO Ozc ~ ~ ~ --O.Co~ --I 12CC --4 -210-.0 4.0-A 3. -50 0 F 10.00-"3000A, 6LA, A Irap, s,II.TU _ JYa, p -ýW,. -* --4y. A -A ýAý .0 U.UV .6.06 1.00 3.00 3.00 Tfye-1 (-V-)AM0 5100 0.00 1.00 2.0 100 moy-I (mab)AAV37, -20TF WCAP-17009-NP April 2009 Revision 1 B-13 MUD2 I0.~00 N Al a P -R 0.00 1.00 too 0o 0 t o06 Co.Tw-e-1 (M')AW43, F47 000.00 2000.A 10C00.00................ ....... ............ ......
AT36, 100OF WCAP-17009-NP                                                                                                                                                      April 2009 Revision 1
/Li A ý A A A A A A.0.00 1.00 2.00 50 AW38, -F A00 5901 ICCC-.-23 II I I II*---------------4--------- -H- -------------
 
+------ F----------I I I I I I I I I I I I 1 I II I I I I I I I I I I I I I....Ll-- J .I AMtAAV~ -.iL~IJJ ~3.00 AW40, F 433: &X-IC fic" WCAP- 17009-NP April 2009 WCAP-17009-NP April 2009 Revision I B-14Tff.~- (m)
B-8 z1xoom I M,0ra 10)2 n
ANN45, F 3C.O SWD.II...........
A,-                   AL    -q    A
i.00 2200U rqr-t (fm Tm".1 (rnm)AW 36, 0 -F WCAP-17009-NP April 2009 Revision I B-15*c'O.D*D A... ......... .... ...... ....... ..I .% .... .... ... .... .
                      'lID0          1.110            2.00                                      &.1fl,       SEI
II  a nh..0.00 1.00 ? 0 100hrwe-1 (ýM)
:120                                                    0.1D1 hEm-i(rail AT40O. 110OF I............... ........                                                                    ..........
AW42, 10OF&O0 0.11 00I 301 201 W~oo ...00.00 t0o I a0 2.00 Mtr&-l (flu)ANV41, 20-F.4.00 U 3011 T'weI Oni)XAW39, 25 0 F 8. to WCAP-17009-NP April 2009 Revision 1 B-16......... .............. ............4 0 ,00 ........... ............ ........... ..........MOMo 1010c 016 1too AIV31, 125OF 400 S.0 &GOD 0000.08*cGO.09 2000.00 ttOO.00 0.0k 0£000.08~(O0.00.. .. .... .... .... .
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...... .... .......  
Al     A     a.F&           i
....... .... .... .
                      .700 T"o 1015)
..... ... ... .. ..... ... .... .. ... ..... ..........  
AT44, 120-F I. 0 DO 1                  . . . . . . . . . . . . . . . .     . . . ..C. . . . . ... .. . . .   . . . . .   . . .
............ ..... ................ .... .... .. .. ..... ... ..... .... ... ... .... .. .
A  m1m      -0ujM      4L.k-.
...............
o.00          ,.0O              2.00              11.00 The-I 0"Y)
-1.00 2.00.0 TVr- I (mw)~0O 60.w.The- I (em)AW44, 175'F WCAP- 17009-NP April 2009 Revision I B- 17 001 ii 433000-~ ~0~'J.Do tD~O.00... ... .... ..... .... ... .... ...... .... .... ..... .. .... ... .... .... ... .... .... ... .... .... .... .... ..... .... ....... ...... .... .......... .iahA."A.-h JAM im.. ........
AT43, 125, WCAP-17009-NP                                                                                                                          April 2009 Revision 1
0.t I 0.00 1.00 700 ..cc .O0 810 Tme-i OTM)AH41, -150 0 F 5oC II I II I II I I I I I I I I I I I I I---- -------------L----- --,--
 
--- -- -- ----,I I I I---I I I I II I I I I II i I I AH43, -100°F Ice 0.00 0.0 .. e.o ......000 ..................................0.00 t.aO 200 3.00 Tvre4t (M.1 AH40..-60 0 F k00 Sýa0o WCAP- 17009-NP April 2009 Revision 1 B- 18 I ONAc..................... ... .... .... ... .... ... ... ........ .... ... .... .... ..... ... .... ... .... .... ....... ..
B-9 UI                cr      I      ZE                  .03SI
A kU IN .& A,..L.. .. .........A Fn 4- I I.0o 240 3.0V r,-. (m)AH39, -50OF 4.o0 !M&00rI....................
              -r~                                              (MS AT1              .. .. .. . . .. . . .  ..        . . . . . . ..   . . . . . .. . . .         . . ..            . .
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.... .. ..T.-t~ (ffsýAH33,-40 0 F 4.00 ZM0 ADD 100 1P..0. .....................
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    .. O.DO~
......................
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-................... .......... ......... ...... ...........
  .. . .                                           . . . .I    .. .   . . . .     .   . . . . .. . . .   . . .
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AT31, 1800 F Reiso    I
........ ...... .............. ............. .
 
0.00 1.OD to. ,o0b 4.W c0 6.1c Tre~1 (rm)AH42. 10°F Dl  30 .I ..........
B-10 Thr.-4 i;m)
.. ........... ... ............. .. .. .. ..
I TqTe-1 (-)j AT34, 200-F
........ .0.0 t -..Cc .:.AH38, 25°F.........0,.00 1.05, E.00 S.M.rrr.-1 pfm AH35, 50°F WCAP- 17009-NP April 2009 Revision 1 B-21 Turn-t Orm)AH31, 150-F-WO000 A n 800O 1.00 2.00 18O0 ,*0.0'0 0,.AH34. 200 0 F WCAP- 17009-NP April 2009 Revision 1 C-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Contained in Table C-I are the upper shelf energy (USE) values used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 5.3. The definition for USE is given in ASTM E185-82[C-1], Section 4.18, and reads as follows: "upper shelf energy level -the average energy value for all Charpy specimens (normally three)whose test temperature is above the upper end of the transition region. For specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the upper shelf energy." If there are specimens tested in sets of three at each temperature, Westinghouse reports the set having the highest average energy as the USE (usually unirradiated material).
                        .x.   . . ...     .   . .
If the specimens were not tested in sets of three at each temperature, Westinghouse reports the average of all Charpy data (> 95% shear) as the USE, excluding any values that are deemed outliers using engineering judgement.
Srno.DD I
Hence, the USE values reported in Table C-i, which were used to generate the Charpy V-notch curves, were determined utilizing this methodology.
4Aoo    1.00    ~ 2.0              I      'K SM, rrr.-1 (fmQ)
AT32, 200-F WCAP- 17009-NP                                                        April 2009 Revision I
 
B-11 20DW.00  .   . .
000          190 ZOO            I.O    4.00 5j" OD1.00        2.0030                    5010 E.00 AAT37, 32.50 F WCAP-17009-NP                                                            April 2009 Revision 1
 
B- 12
                          ~ ~ ~
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            !!M O.Co~                                          -            -I 4CO NO 12CC          -    -  4      -210-.0                        4.0-
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0.00          1.00                          100 2.0 moy-I (mab)
AAV37, -20TF WCAP-17009-NP                                                                                                    April 2009 Revision 1
 
B-13 MUD2 I0.~00                                                                  N          Al    a P -       R 0.00                  1.00               too                  0o0                t          o06            Co.
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AW43, F 47 000.00 2000.A 10C00.00
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0.00                  1.00               2.00              50                A00        5901 AW38, -F II                                            I I                                  II I      I I                                      I
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I         IF----------
I                  I 1         I II                                                        I I                  I        I I                                      I                  I        I I                          I I                  I                  I
                                  . .     .         .         Ll--              J                  .
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B-14 Tff.~- (m)
ANN45, F 3C.O i.00 2200U rqr-t (fm SWD.II Tm".1 (rnm)
AW 36, 0-F WCAP-17009-NP                                April 2009 Revision I
 
B-15
              *c'O.D*D                                          ...... . ........... ...... .......
I . %......... .......
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                    .0.00          1.00 ?0          100                              &O0        0.11 hrwe-1
(ýM)
AW42, 10OF 00I I1*1 301W~oo    . . .
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ANV41, 20-F
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XAW39, 250F WCAP-17009-NP                                                                                           April 2009 Revision 1
 
B-16 40 ,00 ......... ........                           ...... .................                       ....
MOMo 1010c 016                1too                                                  400            S.0                &GOD AIV31, 125OF 0000.08.. ................                                       .......... . . .. .....         ...... .........
              *cGO.09      ........... ....................................
2000.00      ........ ..... ............
ttOO.00 0.0k 0                1.00              2.00.0                                            ~0O TVr-I (mw)
              £000.08
              ~(O0.00 60.w.
The- I (em)
AW44, 175'F WCAP- 17009-NP                                                                                                                        April 2009 Revision I
 
B- 17 001                                                ...
ii 433000-        ....
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t  I              iahA."A.-h                  JAM          im 0.
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AH41, -150 0F II                                    I                  I I        I                                II 5oC                            I                  I                                I                                I I                            I                                I                      I I                                          I I                                          I                                          I
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AH43, -100°F 0.00 0.0e.o      ..                                                                         . . . . . .
000                . ... . . ..                     . . . . . . . . .. . . . . . . . .             . .. ... . .
Ice 0.00                t.aO                200                  3.00                k00                  Sýa0o Tvre4t (M.1 AH40..-60 0F WCAP- 17009-NP                                                                                                                                    April 2009 Revision 1
 
B- 18 I ONAc                                ................
                                                                                                .. L.. .. .........
A Fn 4-    I              A kU  IN.& A, I.0o        240                    3.0V        4.o0              !M              &00rI r,-.       (m)
AH39, -50OF
                                                                                    .A-a .. ...
O1002DD                                          ,0 4.00                ZM0 T.-t~ (ffsý AH33,-40 0 F ADD1P..0.                               .....................
100 noo DOD          1.00      ZOO33a                              4.00              S.00          Ga0 Tkfl-1 MU)3 AH37,-30 0 F WCAP-17009-NP                                                                                                            April 2009 Revision 1
 
B- 19 003.0 MUM  33 ZionOn~
20ai' 2
ftm  IIIMI I I      &  J.
                                            -AA~f    mkA-K A  . A A 1.00  2.0OD0                    4.0D fTkP- 1 yaJ
                                                        '300 Timl-I; f*P AH44.. O0F WCAP-17009-NP                                                            April 2009 Revision 1
 
B-20 5 . r, x0 .[*  .............        ......................  -     ...................          .      ......... . . . .........
                  ,3 0.00                           1.OD                          to.                     ,o0b                      4.Wc0                                              6.1c Tre~1 (rm)
AH42. 10°F Dl                                      ..         .
                                                                                    .*I 30  .I      ..........
0.0                                                            t                                                -..Cc                                            .   :.
AH38, 25°F 0,.00                          1.05,                          2.*06                                            AB,10*                      E.00                    S.M.
rrr.-1 pfm AH35, 50°F WCAP- 17009-NP                                                                                                                                                                               April 2009 Revision 1
 
B-21 Turn-t Orm)
AH31, 150-F
              - WO000 An 800O 1.00 2.00          18O0    , 0.0'0 0,.
AH34. 200 0 F WCAP- 17009-NP                                                  April 2009 Revision 1
 
C-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Contained in Table C-I are the upper shelf energy (USE) values used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 5.3. The definition for USE is given in ASTM E185-82
[C-1], Section 4.18, and reads as follows:
          "upper shelf energy level - the average energy value for all Charpy specimens (normally three) whose test temperature is above the upper end of the transition region. For specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the upper shelf energy."
If there are specimens tested in sets of three at each temperature, Westinghouse reports the set having the highest average energy as the USE (usually unirradiated material). If the specimens were not tested in sets of three at each temperature, Westinghouse reports the average of all Charpy data (> 95% shear) as the USE, excluding any values that are deemed outliers using engineering judgement. Hence, the USE values reported in Table C-i, which were used to generate the Charpy V-notch curves, were determined utilizing this methodology.
The lower shelf energy values were fixed at 2.2 ft-lb for all cases.
The lower shelf energy values were fixed at 2.2 ft-lb for all cases.
Table C-1 Upper Shelf Energy Values (ft-lb) Fixed in CVGRAPH Capsule Material Unirradiated U Y V X W Intermediate Shell Plate B8805-3 (Longitudinal Orientation)
Table C-1         Upper Shelf Energy Values (ft-lb) Fixed in CVGRAPH Capsule Material Unirradiated         U           Y         V         X         W Intermediate Shell Plate B8805-3 (Longitudinal Orientation)
Intermediate Shell Plate B8805-3 (Transverse Orientation)
Intermediate Shell Plate B8805-3 (Transverse Orientation)
Weld Metal (Heat
Weld Metal (Heat # 83653)                145              153        139        138        138      125 HAZ Material                      134              129        124        121        122      114 CVGRAPH Version 5.3 plots of all surveillance data are provided in this appendix, on the pages following the reference list.
# 83653) 145 153 139 138 138 125 HAZ Material 134 129 124 121 122 114 CVGRAPH Version 5.3 plots of all surveillance data are provided in this appendix, on the pages following the reference list.C.1 REFERENCES C-i ASTM El 85-82, Standard Practice for Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.WCAP-17009-NP April 2009 Revision 1 C-2 300 250 200 150 100 50 Unirradiated Intermediate Shell Plate B8805-3 (LT)CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:16 AM Page 1 Coefficients of Curve 1 A = 62.1 B = 59.9 C = 93.5 TO = 41.02 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]
C.1      REFERENCES C-i      ASTM El 85-82, StandardPracticefor ConductingSurveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.
Upper Shelf Energy=122.0(Fixed)
WCAP-17009-NP                                                                                      April 2009 Revision 1
Lower Shelf Energy=2.2(Fixed)Temp@30 ft-lbs=-14.9 Deg F Temp@50 ft-lbs=21.9 Deg F Plant: Vogtle I Material: SA533B1 Heat:
 
C0623-1 Orientation: LT Capsule:
C-2 Unirradiated Intermediate Shell Plate B8805-3 (LT)
Unirra Fluence: n/cmA2 o0 0 0o 0 08 0-300.0-200.0 -100.0 0.0 100.0 200.0 300.0 Temperature in Deg F 400.0 500.0 600.0 Charpy V-Notch Data Temperature
CVGRAPH 5.3        Hyperbolic Tangent Curve Printed on 09/19/2008 11:16 AM Page 1 Coefficients of Curve 1 A = 62.1 B = 59.9 C = 93.5 TO = 41.02 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]
-40. 00-40. 00-40, 00-20. 00-20. 00-20. 00.00.00.00Input CVN 11. 00 19.00 9. 00 54. 00 28. 00 12.00 52. 00 47. 00 42. 00 Computed CVN 20. 19 20. 19 20. 19 27. 75 27. 75 27. 75 37. 39 37. 39 37. 39 Differential
Upper Shelf Energy=122.0(Fixed)          Lower Shelf Energy=2.2(Fixed)
-9. 19-1. 19-1 .19 26. 25.25-5 75 14. 61 9.61 4.61 WCAP-17009-NP April 2009 Revision 1 C-3Unirradiated Intermediate Shell Plate B8805-3 (LT)Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule:
Temp@30 ft-lbs=-14.9 Deg F          Temp@50 ft-lbs=21.9 Deg F Plant: Vogtle I Material: SA533B1        Heat: C0623-1 Orientation: LT    Capsule: Unirra      Fluence:      n/cmA2 300 250 200 150 o0    0      0o 100 0
Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature
08 50 0
: 40. 00 40. 00 40. 00 80. 00 80. 00 80. 00 100. 00 100. 00 100. 00 120. 00 120. 00 120. 00 180. 00 180. 00 180. 00 260. 00 260. 00 320. 00 320. 00 Input CVN 48. 00 62. 00 60. 00 93. 00 64. 00 70. 00 84. 00 107. 00 110.00 100. 00 116.00 109.00 126. 00 1.15. 00 116. 00 129.00 121.00 131. 00 119.00 Computed CVN 61.61.61.85.85.85.95.95.95.103.103.103.
                      -300.0 -200.0    -100.0      0.0    100.0 200.0 300.0            400.0    500.0        600.0 Temperature in Deg F Charpy V-Notch Data Temperature                    Input CVN                    Computed CVN                Differential
116.116.116.120.120.121.121.45 45 45 72 72 72 56 56 56 33 33 33 17 17 17 90 90 69 69 Differential
                - 40. 00                  11. 00                            20. 19                    -9. 19
-13.45.55-1.45 7.28-21.72-15. 72-11.56 11. 44 14. 44-3. 33 12.675. 679. 83-1.17-.17 8.10.10 9.31-2. 69Correlation Coefficient
                -40. 00                  19.00                            20. 19                  -1. 19
                -40,    00                    9. 00                            20. 19                -1    . 19
                - 20. 00                  54. 00                            27. 75                    26. 25
                - 20. 00                  28. 00                            27. 75                        . 25
                -20. 00                  12.00                            27. 75                    - 5 75
                      .00                    52. 00                            37. 39                    14. 61
                      .00                    47. 00                            37. 39                      9.61
                      .00                    42. 00                            37. 39                      4.61 WCAP-17009-NP                                                                                                        April 2009 Revision 1
 
C-3 Unirradiated Intermediate Shell Plate B8805-3 (LT)
Page 2 Plant: Vogtle 1 Material: SA533B1        Heat: C0623-1 Orientation: LT      Capsule: Unirra Fluence:      n/cm^2 Charpy V-Notch Data Temperature              Input CVN                  Computed CVN          Differential
: 40. 00              48. 00                          61. 45          -13.45
: 40. 00                62. 00                          61. 45                .55
: 40. 00                60. 00                          61. 45            -1.45
: 80. 00                93. 00                          85. 72              7.28
: 80. 00                64. 00                          85. 72          -21.72
: 80. 00                70. 00                          85. 72          -15. 72 100. 00                84. 00                          95. 56          -11.56 100. 00              107. 00                          95. 56            11. 44 100. 00              110.00                            95. 56            14. 44 120. 00              100. 00                          103. 33            -3. 33 120. 00              116.00                          103. 33            12.67 120. 00              109.00                          103. 33              5. 67 180. 00              126. 00                          116. 17              9. 83 180. 00              1.15. 00                        116. 17            -1.17 180. 00              116. 00                          116. 17              -. 17 260. 00              129.00                          120. 90              8.10 260. 00              121.00                          120. 90                .10 320. 00              131. 00                          121. 69              9.31 320. 00              119.00                          121. 69            -2. 69 Correlation Coefficient = .962 April 2009 WCAP- 17009-NP I7009-NP                                                                                  April 2009 Revision I
 
C-4 Capsule U Intermediate Shell Plate B8805-3 (LT)
CVGRAPH 5.3        Hyperbolic Tangent Curve Printed on 09/19/2008 11:18 AM Page 1 Coefficients of Curve 1 A = 68.1 B = 65.9 C = 101.88 TO = 65.53 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]
Upper Shelf Energy= 134.0(Fixed)          Lower Shelf Energy=2.2(Fixedl)
Temp@30 ft-lbs=-1.6 Deg F        Ternp@50 ft-lbs=36.9 Deg F Plant: Vogtle I Material: SA533BI        I-feat: C0623-1 Orientation: LT    Capsule: U      Fluence:        n/cmA2 300 250 A200 0
IL Lm U-  150 0
100 50 0 f
                -300.0  -200.0  -100.0      0.0      100.0    200.0    300.0      400.0  500.0      600.0 Temperature in Deg F Charpy V-Notch Data Temperature                    Input CVN                      Computed CVN                  Differential
          -75. 00                  11.00                              10.06                              94
          -50. 00                  18.00                              14. 57                        3. 43
          -25. 00                  14. 00                            21.27                      -7. 27
          -10. 00                  25. 00                            26. 59                      -1    59
                .00                    23. 00                            30. 73                      -7. 73
                .00                    38. 00                            30. 73                        7. 27
: 10. 00                    29. 00                            35. 36                      -6. 36
: 25. 00                    44, 00                            43. 19                            81
: 50. 00                    58. 00                            58. 13                          -13 WCAP- 17009-NP                                                                                                  April 2009 Revision 1
 
C-5 Capsule U Intermediate Shell Plate B8805-3 (LT)
Page 2 Plant: Vogtle I      Material: SA533B1    Heat: C0623-1 Orientation: LT        Capsule: U Fluence:        n/cm^2 Charpy V-Notch Data Temperature            Input CVN                    Computed CVN        Differential
: 72. 00              88. 00                          72. 28            15. 72 150. 00              97. 00                          112.91            -15. 91 200. 00            125.00                            125. 22              - . 22 250. 00            136.00                            130. 57              5. 43 350. 00            143.00                            133.51                9. 49 400

Latest revision as of 11:52, 12 March 2020

WCAP-17009-NP, Revision 1, Analysis of Capsule W from the Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program.
ML091550357
Person / Time
Site: Vogtle Southern Nuclear icon.png
Issue date: 04/30/2009
From: Burgos B, Conermann J, Hunter M, Rosier B
Westinghouse
To:
Office of Nuclear Reactor Regulation
References
WCAP-17009-NP, Rev 1
Download: ML091550357 (284)
v • d • e


Text

Westinghouse Non-Proprietary Class 3 WCAP-17009-NP April 2009 Revision 1 Analysis of Capsule W from /

the Vogtle Unit No. I Reactor Vessel Radiation urveillance Program

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17009-NP Revision 1 Analysis of Capsule W from the Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program B. A. Rosier*

B. N. Burgos* for J. M. Conermann M. A. Hunter*

April 2009 Reviewer: N. R. Jurcevich*

Primary Component Asset Management Approved: P. C. Paesano*, Manager Primary Component Asset Management

  • Electronically Approved Records Are Authenticated in the Electronic Document Management System.

Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

© 2009 Westinghouse Electric Company LLC All Rights Reserved

iii Preface Revision 0: Original Issue Revision 1: This report was revised to reflect the fact that surveillance Capsule Z was removed at 18.41 EFPY, the same time that Capsule W was removed. Appropriate changes were made to the relevant portions of this document.

WCAP-17009-NP April 2009 Revision I

iv TABLE OF CONTENTS L IST O F TAB L E S ........................................................................................................................................ v L IST O F F IG U RE S .................................................................................................................................... v ii E X EC U T IVE SU MM A RY ......................................................................................................................... ix 1 SUM M A RY O F RESU LTS ..................................................................................................... 1-1 2 IN T RO D U CT ION ........................................................................................................................ 2-1 3 B A C KG R O U N D .......................................................................................................................... 3-1 4 DESCRIPTION OF PROGRAM ............................................................................................ 4-1 5 TESTING OF SPECIMENS FROM CAPSULE W ................................................................. 5-1 5 .1 O V E RV IE W .................................................................................................................... 5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS ........................................................... 5-3 5.3 T EN SILE TEST RESU LT S ............................................................................................. 5-5 5.4 1/2T COMPACT TENSION SPECIMEN TESTS ........................................................... 5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ................................................. 6-1 6.1 IN T RO DU C TIO N ........................................................................................................... 6-1 6.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 6-2 6.3 N EU TRON D O SIMETRY .............................................................................................. 6-5 6.4 CALCULATIONAL UNCERTAINTIES ........................................................................ 6-5 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE ........................................................ 7-1 8 RE F E R EN C E S ............................................................................................................................. 8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS .......................................................................... A-i APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS ............................... B-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD ...................................... C-1 APPENDIX D VOGTLE UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION .... D-i APPENDIX E VOGTLE UNIT I UPPER SHELF ENERGY EVALUATION ................................. E-I WCAP- 17009-NP April 2009 Revision I

V LIST OF TABLES Table 4-1 Chemical Composition (wt%) of the Vogtle Unit 1 Reactor Vessel Surveillance Materials (U nirradiated)(a) .................. ....................................................................................... 4-3 Table 4-2 Heat Treatment History of the Vogtle Unit 1 Reactor Vessel Surveillance Materials(a)...4-4 Table 5-1 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation) ....................... 5-6 Table 5-2 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of4.36E+ 19 n/cm 2 (E > 1.0 MeV) (Transverse Orientation) .......................... 5-7 Table 5-3 Charpy V-notch Data for the Vogtle Unit 1 Surveillance Weld Metal Irradiated to a Fluence of4.36E+19 n/cm 2 (E > 1.0 M eV) ..................................................................... 5-8 Table 5-4 Charpy V-notch Data for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of4.36E+19 n/cm 2 (E > 1.0 MeV) .............................................. 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Vogtle Unit I Intermediate Shell Plate B8805-3 Irradiated to a Fluence of4.36E+19 n/cm 2 (E > 1.0 MeV)

(L ongitu dinal O rientation) ............................................................................................. 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of4.36E+19 n/cm 2 (E > 1.0 MeV)

(Transverse O rientation) ........................................................................................... 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Vogtle Unit I Surveillance Weld Metal Irradiated to a Fluence of4.36E+19 n/cm 2 (E > 1.0 MeV) ............................................ 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV) ................. 5-13 Table 5-9 Effect of Irradiation to 4.36E+19 n/cm 2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Vogtle Unit 1 Reactor Vessel Surveillance Capsule W Materials ....... 5-14 Table 5-10 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, P red ictio ns ..................................................................................................................... 5 -15 Table 5-11 Tensile Properties of the Vogtle Unit 1 Capsule W Reactor Vessel Surveillance Materials Irradiated to 4.36E+19 n/cm 2 (E > 1.0 M eV) ................................................................ 5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule C enter(a) ..................................................................................................... 6-7 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base M etal Interface .......................................................... 6-11 Table 6-3 Relative Radial Distribution of Neutron Fluence (E > 1.0 MeV) Within the R eactor Vessel Wall(a) ..................................................................................................... 6-15 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) Within the R eactor Vessel Wall(a) ..................................................................................................... 6-16 WCAP- 17009-NP April 2009 Revision 1

vi Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Vo gtle U nit I .................................................................................................................. 6 -16 Table 6-6 Calculated Surveillance Capsule Lead Factors .............................................................. 6-17 Table 7-1 Surveillance Capsule W ithdrawal Summary .................................................................... 7-1 Table A-I Nuclear Parameters used in the Evaluation of Neutron Sensors .............................. A-10 Table A-2 Monthly Thermal Generation During the First Fourteen Fuel Cycles of the Vogtle Unit 1 Reactor (Reactor Power of 3411 MWt from Startup Through the End of Cycle 4, and 3565 M W t for Cycles 5 through 14) ........................................................................ A -11 Table A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation...A-14 Table A-4a Measured Sensor Activities and Reaction Rates Surveillance Capsule U ............... A-15 Table A-4b Measured Sensor Activities and Reaction Rates Surveillance Capsule Y ................ A- 16 Table A-4c Measured Sensor Activities and Reaction Rates Surveillance Capsule V ................ A- 17 Table A-4d Measured Sensor Activities and Reaction Rates Surveillance Capsule X ................ A- 18 Table A-4e Measured Sensor Activities and Reaction Rates Surveillance Capsule W ............... A-19 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance C apsule C enter ......................................................................................... A -20 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates at the Surveillance C apsule C enter ........................................................................................ A -23 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast N eutron Threshold Reactions ................................................................................ A -24 Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... A-24 Table C-I Upper Shelf Energy Values (ft-lb) Fixed in CVGRAPH ........................................... C-I Table D- 1 Calculation of Chemistry Factors using Vogtle Unit 1 Surveillance Capsule Data ........ D-3 Table D-2 Vogtle Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line ...................... D-4 Table E- 1 Predicted Positions 1.2 and 2.2 Upper Shelf Energy Values at 36 EFPY ....................... E-3 Table E-2 Predicted Positions 1.2 and 2.2 Upper Shelf Energy Values at 54 EFPY ....................... E-4 WCAP- 17009-NP April 2009 Revision 1

Vii LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Vogtle Unit 1 Reactor Vessel ................... 4-5 Figure 4-2 Capsule W Diagram Showing the Location of Specimens, Thermal Monitors, an d Do sim eters ................................................................................................................ 4 -6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) ....................................... 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) ....................................... 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) ....................................... 5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) .......................................... 5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) .......................................... 5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) .......................................... 5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld M etal ................................................................................. 5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld M etal ................................................................................. 5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld M etal ................................................................................. 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Vogtle Unit I Reactor Vessel H eat-A ffected-Zone M aterial ......................................................................................... 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Vogtle Unit 1 Reactor Vessel H eat-A ffected-Zone M aterial ......................................................................................... 5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Vogtle Unit 1 Reactor Vessel H eat-A ffected-Zone M aterial ......................................................................................... 5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit I Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) ............................................................ 5-29 Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) ............................................................... 5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for the Vogtle Unit 1 Reactor Vessel Surveillance Program W eld M etal ................................................................................. 5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for the Vogtle Unit 1 Reactor Vessel H eat-A ffected-Zone M aterial ......................................................................................... 5-32 WCAP- 17009-NP April 2009 Revision 1

viii Figure 5-17 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (L ongitudinal O rientation) ............................................................................................. 5-33 Figure 5-18 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse O rientation) ................................................................................................ 5-34 Figure 5-19 Tensile Properties for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal

....................................................................................................................................... 5 -3 5 Figure 5-20 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) .............................................................................. 5-36 Figure 5-21 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) ................................................................................. 5-37 Figure 5-22 Fractured Tensile Specimens from the Vogtle Unit 1 Reactor Vessel Surveillance P rogram Weld M etal ...................................................................................................... 5-38 Figure 5-23 Engineering Stress-Strain Curves for Vogtle Unit 1 Intermediate Shell Plate B8805-3 Tensile Specimens AL7, AL8 and AL9 (Longitudinal Orientation) .............................. 5-39 Figure 5-24 Engineering Stress-Strain Curves for Vogtle Unit I Intermediate Shell Plate B8805-3 Tensile Specimens AT7, AT8 and AT9 (Transverse Orientation) ................................... 5-40 Figure 5-25 Engineering Stress-Strain Curves for Vogtle Unit 1 Surveillance Program Weld Metal Tensile Specimens AW7, AW8 and AW9 ....................................................................... 5-41 Figure 6-1 Vogtle Unit I r,8 Reactor Geometry with a 12.50 Neutron Pad Span at the Core Midplane

....................................................................................................................................... 6 - 18 Figure 6-2 Vogtle Unit I rO Reactor Geometry with a 20.00 Neutron Pad Span at the Core Midplane

....................................................................................................................................... 6 - 19 Figure 6-3 Vogtle Unit 1 r,0 Reactor Geometry with a 22.5' Neutron Pad Span at the Core Midplane

....................................................................................................................................... 6 -2 0 Figure 6-4 Vogtle Unit 1 rz Reactor Geometry with Neutron Pad .................................................. 6-21 Figure E-l Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper Shelf Energy as a Function of Copper and Fluence ..................................................................................... E-2 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the testing results of the surveillance Capsule W from Vogtle Unit 1. Capsule W was removed at 18.41 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database. Capsule W received a fluence of 4.36 x 10' 9 n/cm 2 (E > 1.0 MeV) after irradiation to 18.41 EFPY. The peak clad/base metal interface vessel fluence after 18.41 EFPY of plant operation was 1.05 x 1019 n/cm 2 (E > 1.0 MeV).

This evaluation led to the following conclusions: 1) The measured percent decrease in upper shelf energy for all the surveillance materials contained in the Vogtle Unit 1 Capsule W are less than the Regulatory Guide 1.99, Revision 2 [1] predictions. 2) The Vogtle Unit 1 surveillance plate data is judged to be not credible however the weld data is judged to be credible. This credibility evaluation can be found in Appendix D. 3) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2]. The upper shelf energy evaluation is presented in Appendix E.

Lastly, a brief summary of the Charpy V-notch testing can be found in Section 1. All Charpy V-notch data was plotted using a symmetric hyperbolic tangent curve fitting program.

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1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule W, the fifth capsule removed and tested from the Vogtle Unit 1 reactor pressure vessel, led to the following conclusions:

  • Charpy V-notch test data were plotted using a symmetric hyperbolic tangent curve-fitting program.

Appendix C presents the CVGRAPH, Version 5.3, Charpy V-notch plots for Capsule W and previous capsules, along with the program input data.

" Capsule W received an average fast neutron fluence (E > 1.0 MeV) of 4.36 xl019 n/cm 2 after 18.41 effective full power years (EFPY) of plant operation.

" Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 83.2°F and an irradiated 50 ft-lb transition temperature of 134.9°F. This results in a 30 ft-lb transition temperature increase of 98.1'F and a 50 ft-lb transition temperature increase of I13'F for the longitudinally oriented specimens.

  • Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 111.2°F and an irradiated 50 ft-lb transition temperature of 171.7°F. This results in a 30 ft-lb transition temperature increase of 94.1°F and a 50 ft-lb transition temperature increase of 109.2°F for the transversely oriented specimens.

" Irradiation of the Surveillance Program Weld Metal (Heat #83653) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -22.4°F and an irradiated 50 ft-lb transition temperature of 1.1°F. This results in a 30 ft-lb transition temperature increase of 34.8°F and a 50 ft-lb transition temperature increase of 31.4°F.

  • Irradiation of the Heat-Affected-Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -41.0°F and an irradiated 50 ft-lb transition temperature of 3.5°F.

This results in a 30 ft-lb transition temperature increase of 45. ITF and a 50 ft-lb transition temperature increase of 58.9°F.

" The average upper shelf energy of the Intermediate Shell Plate B8805-3 (longitudinal orientation) resulted in an average energy decrease of 22 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 100 ft-lb for the longitudinally oriented specimens.

  • The average upper shelf energy of the Intermediate Shell Plate B8805-3 (transverse orientation) resulted in an average energy decrease of 14 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 82 ft-lb for the transversely oriented specimens.

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1-2

" The average upper shelf energy of the Surveillance Program Weld Metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 125 ft-lb for the weld metal specimens.

  • The average upper shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 114 ft-lb for the HAZ Material.
  • A comparison of the measured 30 ft-lb shift in transition temperature values for the Vogtle Unit 1 reactor vessel surveillance materials is presented in Table 5-10.

" Based on the credibility evaluation presented in Appendix D, the Vogtle Unit 1 surveillance plate data is not credible but the surveillance weld data is credible.

  • Based on the upper shelf energy evaluation in Appendix E, all beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end of the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2].

" The calculated 36 EFPY (end-of-license) and 54 EFPY (end-of-license renewal) neutron fluence (E > 1.0 MeV) at the core mid-plane for the Vogtle Unit 1 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows:

Calculated (36 EFPY): Vessel inner radius* = 2.155 x 1019 n/cm 2 (Interpolated from Table 6-2) 2 Vessel 1/4 thickness = 1.284 x 1019 n/cm 8 2 Vessel 3/4 thickness = 4.56 x 101 n/cm Calculated (54 EFPY): Vessel inner radius* = 3.30 x 1019 n/cm 2 (Taken from Table 6-2) 2 Vessel 1/4 thickness = 1.967 x 1019 n/cm Vessel 3/4 thickness = 6.99 x 1018 n/cm2

  • Clad/base metal interface.

WCAP- 17009-NP April 2009 Revision 1

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule W, the fifth capsule removed from the reactor in the continuing surveillance program, which monitors the effects of neutron irradiation on the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-1 1011, "Georgia Power Company Alvin W. Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program" [3]. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82 [4], "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Capsule W was removed from the reactor after 18.41 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing of the post-irradiation data obtained from surveillance Capsule W removed from the Southern Nuclear Operating Company Vogtle Unit 1 reactor vessel and discusses the analysis of the data.

April 2009 WCAP- 17009-NP I17009-NP April 2009 Revision I

3-1 3 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Vogtle Unit I reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.

A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code [5]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E208 [6]) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KI, curve) which appears in Appendix G to Section XI of the ASME Code [5]. The K1 , curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel.

When a given material is indexed to the K1 , curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Vogtle Unit 1 reactor vessel radiation surveillance program, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RTNDT Initial + M + ARTNDT) is used to index the material to the KIc curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

WCAP- 17009-NP April 2009 Revision 1

4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Vogtle Unit I reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

The capsules contain specimens made from the following:

  • Intermediate Shell Plate B8805-3 (longitudinal orientation)
  • Intermediate Shell Plate B8805-3 (transverse orientation)

" Weld metal fabricated with 3/16-inch Mil B-4 weld filler wire, Heat Number 83653 Linde Type 0091 flux, Lot Number 3536, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld and all longitudinal weld seams of both the intermediate and lower shell plates of the pressure vessel.

" Weld heat-affected-zone (HAZ) material of Intermediate Shell Plate B8805-1 Test material obtained from the intermediate shell course plate (after thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the '/4 and 3/4 thickness locations of the plate after performing a simulated post-weld stress-relieving treatment on the test material. Test specimens were also removed from weld and heat-affected-zone metal of a stress-relieved weldment joining Intermediate Shell Plate B8805-1 and adjacent Lower Shell Plate B8606-3. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of the Intermediate Shell Plate B8805-1.

Charpy V-notch impact specimens from Intermediate Shell Plate B8805-3 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major rolling direction). The core-region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular to the weld direction. The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.

Tensile specimens from Intermediate Shell Plate B8805-3 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented perpendicular to the weld direction.

Compact Test (CT) specimens from Intermediate Shell Plate B8805-3 were machined in the longitudinal and transverse orientations. Compact test specimens from the weld metal were machined with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399 [7].

All six capsules contained dosimeter wires of pure iron, copper, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Neptunium ( 23 7Np) and Uranium ( 2 38U) were placed in the capsules to measure the integrated flux at specific neutron energy levels.

WCAP- 17009-NP April 2009 Revision 1

4-2 The capsules contained thermal monitors made from two low-melting-point eutectic alloys, which were sealed in Pyrex tubes. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two eutectic alloys and their melting points are as follows:

2.5% Ag, 97.5% Pb Melting Point: 579°F (304'C) 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590'F (310°C)

The chemical composition and heat treatment of the unirradiated surveillance materials are presented in Tables 4-1 and 4-2, respectively. The data in Table 4-1 and 4-2 was obtained from the unirradiated surveillance program report, WCAP-1 1011 [3], Appendix A.

Capsule W was removed after 18.41 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, 1/2T-CT fracture mechanics specimens, dosimeters, and thermal monitors.

The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule W is shown in Figure 4-2.

WCAP- 17009-NP April 2009 Revision 1

4-3 Table 4-1 Chemical Composition (wt%) of the Vogtle Unit 1 Reactor Vessel Surveillance Materials (Unirradiated)(a)

Element Intermediate Shell Plate B8805-3 Weld Metal Combustion Westinghouse Westinghouse Engineering Analysis Analysis Analysis C 0.25 0.22 0.13 Mn 1.320 1.32 1.15 P 0.003 0.017 0.017 S 0.010 0.011 0.010 Si 0.26 0.28 0.19 Ni 0.60 0.61 0.10 Mo 0.53 0.57 0.61 Cr 0.04 0.057 0.052 Cu 0.06 0.058 0.037 Al 0.029 0.030 0.002 Co 0.009 0.006 0.005 Pb < 0.001 < 0.001 < 0.001 W < 0.01 < 0.01 < 0.01 Ti < 0.01 0.004 0.006 Zr < 0.001 < 0.002 < 0.002 V 0.003 < 0.002 0.003 Sn 0.017 0.019 < 0.002 As 0.001 0.003 0.004 Cb < 0.01 < 0.002 < 0.002 N2 0.008 0.006 0.003 B < 0.001 < 0.001 < 0.001 Note:

(a) Data obtained from WCAP- 11011 [3].

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4-4 Table 4-2 Heat Treatment History of the Vogtle Unit 1 Reactor Vessel Surveillance Materials(a)

Material Temperature ( 0F) Time (hr) Cooling Austenitized @ 4 hrs.

Water-quenched 1600 125 Intermediate Shell Plate Tempered @

4 hrs. Air-cooled B8805-3 1225 +/- 25 Stress Relieved @ 17.5 hrs.

Furnace-cooled 1150 +/- 50 Weld Metal (Heat #83653) Post Weld Stress Relieved @ 12.75 hrs. Furnace-cooled 1150 +/- 50 Note:

(a) Data obtained from WCAP-1I01 1 [3].

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4-5 0

f5.6#

w (t2l.S) 11500 PUN VIEW EILEVATION VIEW Figure 4-1 Arrangement of Surveillance Capsules in the Vogtle Unit 1 Reactor Vessel April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

4-6 LEGEND: AL - INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDINAL)

AT - INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE)

AW - WELD METAL (HEAT # 83653)

AH - HEAT AFFECTED ZONE MATERIAL Large Spacer Tensiles Compacts Compacts Charpys Charpys Charpys AW9 AW45 AH45 AW42 AH42 AW39 AH39 AW8 AW44 AH44 AW41 AH41 AW38 AH38 AW7 AW43 AH43 AW40 AH40 AW37 AH37 TOP OF VESSEL CENTER Compacts AL12 ALll CENTER W CENTER CENTER

  • BOTTOM OF VESSEL Figure 4-2 Capsule W Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters WCAP- 17009-NP April 2009 Revision 1

5-1 5 TESTING OF SPECIMENS FROM CAPSULE W 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed at the Westinghouse Science and Technology Department Hot Cell Facility. Testing was performed in accordance with 10 CFR 50, Appendices G and H [2], ASTM Specification E185-82 [4],

and Westinghouse Procedure RMF 8402, Revision 3 [8] as modified by Westinghouse RMF Procedures 8102, Revision 3 [9], and 8103, Revision 2 [10].

The capsule was opened upon receipt at the hot cell laboratory per Procedure RMF 8804, Revision 2 [11].

The specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-11011 [3]. All items were in their proper locations.

Examination of the thermal monitors indicated that none of the melting point monitors had melted. Based on this examination, the maximum temperature to which the specimens were exposed was less than 579°F (304°C).

The Charpy impact tests were performed per ASTM Specification E23-06 [12] and Procedure RMF 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Instron Impulse instrumentation system, feeding information into a computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve, the load of general yielding (PGY), the time to general yielding (ToY), the maximum load (PM), and the time to maximum load (TM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF). If the fast load drop terminates well above zero load, the termination load is identified as the arrest load (PA).

The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (Er) is the difference between the total energy to fracture (ED) and the energy at maximum load (EM).

The yield stress (ay) was calculated from the three-point bend formula having the following expression [13]:

L CTY = PGY 2 (Eqn. 5-1)

B(W - a) C where L = distance between the specimen supports in the impact testing machine; B = the width of the specimen measured parallel to the notch; W = height of the specimen, measured perpendicularly to the notch; a = notch depth. The constant C is dependent on the notch flank angle (p), notch root radius (p) and the type of loading (i.e., pure bending or three-point bending). In three-point bending, for a Charpy specimen in which qp = 450 and p = 0.010 in., Equation 1 is valid with C = 1.21.

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5-2 Therefore, (for L = 4W),

L 3.305 PGYW y --PGY B(W-a)22 1.21 B(W-a) 2 (Eqn. 5-2)

For the Charpy specimen, B = 0.394 in., W = 0.394 in., and a = 0.079 in. Equation 5-2 then reduces to:

aTy= 3 3 . 3 PGY (Eqn. 5-3) where ov is in units of psi and PGY is in units of lb. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

Symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:

A = B(W -a) = 0.1241 sq. in. (Eqn. 5-4)

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM E23-06 [12] and A370-07 [14]. The lateral expansion was measured using a dial gage rig similar to that shown in the same specifications.

Tensile tests were performed on a 20,000 pound Instron, split console test machine (Model 1115) per ASTM Specification E8-04 [15] and E21-05 [16] and Procedure RMF 8102 [9]. Extension measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer gage length was 1.00 inch. Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air.

The yield load, ultimate load, fracture load, total elongation and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength and fracture strength were calculated using the original cross-sectional area. The final diameter was determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area were computed using the final diameter measurement.

WCAP- 17009-NP April 2009 Revision 1

5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule W, which received a fluence of 4.36 x 10'9 n/cm 2 (E > 1.0 MeV) in 18.41 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with the unirradiated and previously withdrawn capsule results as shown in Figures 5-1 through 5-12. The unirradiated and previously withdrawn capsule results were taken from WCAP-11011 [3], WCAP-12256 [17], WCAP-13931, Revision 1 [18],

WCAP-15067 [19], and WCAP-16278-NP [20].

The transition temperature increases and upper shelf energy decreases for the Capsule W materials are summarized in Table 5-9 and led to the following results:

" Irradiation of the reactor vessel intermediate shell plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation),

resulted in an irradiated 30 ft-lb transition temperature of 83.2°F and an irradiated 50 ft-lb transition temperature of 134.9°F. This results in a 30 ft-lb transition temperature increase of 98.1'F and a 50 ft-lb transition temperature increase of 113F for the longitudinal oriented specimens.

" Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 111 .2°F and an irradiated 50 ft-lb transition temperature of 171.7°F. This results in a 30 ft-lb transition temperature increase of

94. 10F and a 50 ft-lb transition temperature increase of 109.2°F for the transverse oriented specimens.

" Irradiation of the weld metal (Heat # 83653) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -22.4°F and an irradiated 50 ft-lb transition temperature of I.1°F. This results in a 30 ft-lb transition temperature increase of 34.8'F and a 50 ft-lb transition temperature increase of 31.4°F.

  • Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -41°F and an irradiated 50 ft-lb transition temperature of 3.5°F.

This results in a 30 ft-lb transition temperature increase of 45.1 'F and a 50 ft-lb transition temperature increase of 58.9°F.

" The average upper shelf energy of the Intermediate Shell Plate B8805-3 (longitudinal orientation) resulted in an average energy decrease of 22 ft-lb after irradiation to 4.36 x 1019 n/cm 2 (E > 1.0 MeV).

This results in an irradiated average upper shelf energy of 100 ft-lb for the longitudinal oriented specimens.

  • The average upper shelf energy of the Intermediate Shell Plate B8805-3 (transverse orientation) resulted in an average energy decrease of 14 ft-lb after irradiation to 4.36 x 10'9 n/cm 2 (E > 1.0 MeV).

This results in an irradiated average upper shelf energy of 82 ft-lb for the longitudinal oriented specimens.

WCAP- 17009-NP April 2009 Revision 1

5-4

  • The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation to 4.36 x 1019 n/cm 2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 125 ft-lb for the weld metal specimens.
  • The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation to 4.36 x 1019 n/cm 2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 114 ft-lb for the weld HAZ metal.

" Comparisons of the measured 30 ft-lb shift in transition temperature values and upper shelf energy decreases for the Vogtle Unit 1 reactor vessel surveillance materials are presented in Table 5-10.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-13 through 5-16. The fractures show an increasingly ductile or tougher appearance with increasing test temperature. Load-time records for the individual instrumented Charpy specimens are contained in Appendix B.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end of the current license (36 EFPY) and a potential license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [2]. This evaluation can be found in Appendix E.

April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule W irradiated to 4.36E+19 n/cm 2 (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-17 through 5-19.

The results of the tensile tests performed on the intermediate shell plate B8805-3 (longitudinal orientation) indicated that irradiation to 4.36E+19 n/cm2 (E > 1.0 MeV) caused approximately a 8.8 ksi increase in the 0.2 percent offset yield strength and approximately a 11.2 ksi increase in the ultimate tensile strength when compared to unirradiated data [3]. See Figure 5-17 and Table 5-11.

The results of the tensile tests performed on the intermediate shell plate B8805-3 (transverse orientation) indicated that irradiation to 4.36E+19 n/cm 2 (E > 1.0 MeV) caused approximately a 9.3 ksi increase in the 0.2 percent offset yield strength and approximately a 8.1 ksi increase in the ultimate tensile strength when compared to unirradiated data [3]. See Figure 5-18 and Table 5-11.

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 4.36E+19 n/cm 2 (E > 1.0 MeV) caused approximately a 5.7 ksi increase in the 0.2 percent offset yield strength and approximately a 8 ksi increase in the ultimate tensile strength when compared to unirradiated data [3]. See Figure 5-19 and Table 5-11.

The fractured tensile specimens for the intermediate shell plate B8805-3 material are shown in Figures 5-20 and 5-2 1, while the fractured tensile specimens for the surveillance weld metal are shown in Figure 5-22. The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-25.

5.4 1/2T COMPACT TENSION SPECIMEN TESTS Per the surveillance capsule testing contract, the 1/2T Compact Tension Specimens were not tested and are being stored at the Westinghouse Research and Technology Department Hot Cell Facility.

April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

5-6 Table 5-1 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number fF fl-bs

-C Joules mils mm  %

AL43 -30 -34 6 8 5 0.13 5 AL39 0 -18 25 34 22 0.56 5 AL42 50 10 24 33 22 0.56 10 AL36 60 16 20 27 17 0.43 15 AL45 75 24 36 49 31 0.79 15 AL33 100 38 45 61 33 0.84 25 AL31 125 52 35 47 30 0.76 30 AL37 150 66 48 65 44 1.12 50 AL44 170 77 45 61 38 0.97 50 AL32 180 82 55 75 50 1.27 60 AL34 200 93 96 130 71 1.80 98 AL41 200 93 86 117 62 1.57 95 AL35 300 149 98 133 69 1.75 100 AL38 325 163 108 146 83 2.11 100 AL40 350 177 114 155 82 2.08 100 WCAP- 17009-NP April 2009 Revision 1

5-7 Table 5-2 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV) (Transverse Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF °C ft-lbs Joules mils mm  %

AT41 0 -18 15 20 12 0.30 5 AT45 80 27 23 31 20 0.51 20 AT36 100 38 28 38 25 0.64 20 AT40 110 43 28 38 26 0.66 25 AT44 120 49 40 54 31 0.79 25 AT43 125 52 38 52 34 0.86 30 AT31 150 66 36 49 31 0.79 35 AT42 165 74 40 54 38 0.97 40 AT38 180 82 40 54 43 1.09 50 AT39 190 88 50 68 47 1.19 60 AT34 200 93 66 89 54 1.37 75 AT32 200 93 68 92 62 1.57 85 AT35 275 135 80 108 64 1.63 100 AT33 300 149 79 107 63 1.60 100 AT37 325 163 86 117 70 1.78 100 WCAP-17009-NP April 2009 Revision 1

5-8 Table 5-3 Charpy V-notch Data for the Vogtle Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 4.36E+19 n/cm2 (E > 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF °C ft-lbs Joules mils mm  %

AW35 -50 -46 4 5 6 0.15 5 AW34 -20 -29 10 14 14 0.36 15 AW37 -20 -29 14 19 16 0.41 10 AW43 -15 -26 13 18 16 0.41 25 AW38 -15 -26 48 65 40 1.02 25 AW40 -10 -23 53 72 40 1.02 25 AW45 -10 -23 61 83 49 1.24 35 AW33 -5 -21 61 83 49 1.24 30 AW36 0 -18 81 110 60 1.52 45 AW42 10 -12 48 65 41 1.04 40 AW41 20 -7 55 75 44 1.12 50 AW39 25 -4 65 88 51 1.30 60 AW31 125 52 125 170 88 2.24 95 AW32 150 66 126 171 88 2.24 100 AW44 175 79 125 170 89 2.26 100 WCAP- 17009-NP April 2009 Revision 1

5-9 Table 5-4 Charpy V-notch Data for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF C Ft-lbs Joules mils mm  %

AH41 -150 -101 6 8 5 0.13 5 AH43 -100 -73 22 30 13 0.33 10 AH40 -60 -51 34 46 22 0.56 20 AH39 -50 -46 32 43 20 0.51 25 AH33 -40 -40 37 50 26 0.66 25 AH37 -30 -34 18 24 13 0.33 20 AH45 -30 -34 32 43 22 0.56 30 AH32 -10 -23 42 57 30 0.76 25 AH44 0 -18 52 71 40 1.02 50 AH42 10 -12 54 73 37 0.94 40 AH38 25 -4 42 57 31 0.79 40 AH35 50 10 85 115 57 1.45 60 AH31 150 66 106 144 67 1.70 100 AH36 175 79 104 141 73 1.85 100 AH34 200 93 131 178 75 1.91 100 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

5-10 Table 5-5 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation)

Normalized Energies General Test Charpy (ft-lb/in 2) Yield Time to Max. Time to Fract. Arrest Yield Flow ED Load PGy Load, PM Load, Load, Stress Stress No. (OF) (E-lb) Total At PM Prop. PGY (msec) PM (lb) (msec) PF (ib) PA (lb) (ksi) (ksi)

ED/A Em/A Ep/A (lb)

AL43 -30 6 48 33 15 4013 0.09 4013 0.09 4013 0 134 134 AL39 0 25 201 165 36 3239 0.06 4010 0.35 4007 0 108 121 AL42 50 24 193 177 17 3044 0.05 3898 0.36 3891 0 101 116 AL36 60 20 161 118 44 3046 0.05 3883 0.29 3858 0 101 115 AL45 75 36 290 244 46 3132 0.06 4050 0.50 3988 0 104 120 AL33 100 45 363 318 44 3005 0.06 4070 0.62 3819 0 100 118 AL31 125 35 282 268 14 2930 0.05 3849 0.47 3849 606 98 113 AL37 150 48 387 303 84 2831 0.06 4010 0.62 3786 392 94 114 AL44 170 45 363 284 79 2973 0.07 3823 0.63 3572 504 99 113 AL32 180 55 443 294 149 2835 0.06 3845 0.62 3548 957 94 111 AL34 200 96 774 285 489 2855 0.06 4031 0.62 2705 2210 95 115 AL41 200 86 693 278 415 2882 0.05 3924 0.61 2994 2336 96 113 AL35 300 98 790 250 539 2671 0.06 3776 0.62 N/A N/A 89 107 AL38 325 108 870 274 596 2630 0.06 3775 0.62 N/A N/A 88 107 AL40 350 114 919 277 641 2688 0.05 3782 0.62 N/A N/A 90 108 WCAP- 17009-NP April 2009 Revision 1

5-11 Table 5-6 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV) (Transverse Orientation)

Charpy Normalized Energies General Test Yield Flow Energy (ft-lb/in2 ) Yield Time to Max. Time to Fract. Arrest Sample Tern Load, Load, Stress Stress Temp. ED. Load PcY Load, PM (OF) (ft-lb) Total At PM Prop. PGY (msec) PM (lb) (msec) PF (lb) PA (lb) (ksi) (ksi)

ED/A Em/A Ep/A (ib)

AT41 0 15 121 114 7 3067 0.06 3938 0.26 3931 0 102 117 AT45 80 23 185 149 36 3028 0.06 3960 0.31 3939 0 101 116 AT36 100 28 226 218 7 2980 0.06 3943 0.43 3943 0 99 115 AT40 110 28 226 175 51 2981 0.05 3868 0.35 3822 290 99 114 AT44 120 40 322 234 89 2891 0.06 3989 0.47 3762 605 96 115 AT43 125 38 306 267 39 2817 0.06 3972 0.50 3833 494 94 113 AT31 150 36 290 212 78 2970 0.05 3867 0.46 3763 610 99 114 AT42 165 40 322 224 98 2830 0.05 3830 0.47 3614 717 94 111 AT38 180 40 322 224 98 2941 0.06 3768 0.48 3619 358 98 112 AT39 190 50 403 302 101 2778 0.06 3798 0.62 3578 963 93 109 AT34 200 66 532 221 310 2891 0.05 3885 0.48 3495 2669 96 113 AT32 200 68 548 216 332 3027 0.05 3835 0.48 3314 2744 101 114 AT35 275 80 645 287 357 2906 0.06 3932 0.61 N/A N/A 97 114 AT33 300 79 637 264 373 2930 0.06 3807 0.62 N/A N/A 98 112 AT37 325 86 693 265 427 2728 0.06 3840 0.61 N/A N/A 91 109 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV)

Test Charpy Normalized Energies General Sample Test Energy (ft-lb/in 2) Yield Time to Max. Time to Fract. Arrest Yield Flow Temp. Load PGv Load, PM Load, Load, Stress Stress (NF) (E-lb) Total At PM Prop. PcY (msec) PM (lb) (msec) PF (lb) PA (lb) (ksi) (ksi) t ED/A EM/A Ep/A (lb) I AW35 -50 4 32 22 11 3836 0.08 3836 0.08 3836 0 128 128 AW34 -20 10 81 40 40 3010 0.07 4207 0.11 3419 0 100 120 AW37 -20 14 113 35 78 3330 0.07 3946 0.09 3646 0 111 121 AW43 -15 13 105 32 73 3253 0.06 4100 0.09 3574 0 108 122 AW38 -15 48 387 241 146 3040 0.06 4043 0.61 3669 1555 101 118 AW40 -10 53 427 234 193 3226 0.07 4253 0.52 4050 0 107 125 AW45 -10 61 491 324 167 3291 0.06 4079 0.62 3658 736 110 123 AW33 -5 61 491 314 177 3118 0.06 4124 0.62 3680 0 104 121 AW36 0 81 653 306 346 3335 0.05 4080 0.61 3346 813 111 123 AW42 10 48 387 322 64 3125 0.06 4100 0.61 3860 554 104 120 AW41 20 55 443 330 113 3204 0.06 4096 0.61 3884 770 107 122 AW39 25 65 524 321 202 3040 0.05 4044 0.61 3669 1558 101 118 AW31 125 125 1007 278 729 2788 0.06 3826 0.63 2262 1724 93 110 AW32 150 126 1015 272 744 2719 0.06 3749 0.60 N/A N/A 91 108 AW44 175 125 1007 269 738 2817 0.06 3768 0.61 N/A N/A 94 110 WCAP- 17009-NP April 2009 Revision 1

5-13 Table 5-8 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of 4.36E+19 n/cm 2 (E > 1.0 MeV)

Normalized Energies General Charpy Yield Flow Energy (ft-lb/in2 ) Yield Time to Max. Time to Fract. Arrest Sample Test Load PGv Load, PM Load, Load, Stress Stress No. Temp. ED (OF) (ft-lb) Total At PM Prop. PGv (msec) PM (lb) (msec) PF (Ib) PA (lb) (ksi) (ksi)

ED/A EM/A Ep/A (lb)

AH41 -150 6 48 24 25 3604 0.05 4790 0.15 4452 0 120 140 AH43 -100 22 177 161 16 3736 0.07 4299 0.12 4299 0 124 134 AH40 -60 34 274 264 10 3709 0.06 4411 0.50 4366 0 124 135 AH39 -50 32 258 242 16 3225 0.06 4200 0.45 4200 0 107 124 AH33 -40 37 298 234 64 3311 0.07 4259 0.50 4127 0 110 126 AH37 -30 18 145 32 113 3231 0.06 4249 0.09 4027 0 108 125 AH45 -30 32 258 209 48 3158 0.06 4231 0.43 4158 0 105 123 AH32 -10 42 338 290 48 3384 0.07 4284 0.50 4130 0 113 128 AH44 0 52 419 274 145 3131 0.06 4134 0.50 3956 1499 104 121 AH42 10 54 435 330 105 3102 0.06 4296 0.62 4129 1120 103 123 AH38 25 42 338 258 81 3094 0.06 4129 0.48 4094 1066 103 120 AH35 50 85 685 342 342 3184 0.06 4170 0.62 3202 920 106 122 AH31 150 106 854 288 566 2986 0.05 3959 0.60 N/A N/A 99 116 AH36 175 104 838 280 558 3065 0.06 3926 0.62 N/A N/A 102 116 AH34 200 131 1056 458 597 2991 0.05 4038 0.94 N/A N/A 100 117 April 2009 WCAP- 17009-NP 17009-NP April 2009 Revision I

5-14 Table 5-9 Effect of Irradiation to 4.36E+19 n/cm 2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Vogtle Unit 1 Reactor Vessel Surveillance Capsule W Materials Average 30 (ft-lb)(a) Average 35 mil Lateral(b) Average 50 ft-lb(') Average Energy Absorption(a)

Material Transition Temperature (0 F) Expansion Temperature (0F) Transition Temperature (0F) (ft-lb)

Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Intermediate ShellPlate -14.9 83.2 98.1 18.8 114.8 96.0 21.9 134.9 113 122 100 -22 B8805-3 (Longitudinal)

Intermediate ShellPlate 17.1 111.2 94.1 53.7 139.0 85.3 62.5 171.7 109.2 96 82 -14 B8805-3 (Transverse)

Weld WdetalMetal

-57.2 -22.4 34.8 -32.6 -8.0 24.6 -30.3 1.1 31.4 145 (Heat # 83653)11 125 -20 HAZ Material -86.1 -41.0 45.1 -49.7 6.1 55.8 -55.4 3.5 58.9 134 114 -20 Notes:

(a) "Average" is defined as the value mathematically determined by CVGRAPH from the data points of the Charpy tests (see Figures 5-1, 5-4, 5-7 and 5-10).

(b) "Average" is defined as the value mathematically determined by CVGRAPH from the data points of the Charpy tests (see Figures 5-2, 5-5, 5-8 and 5-11).

WCAP-17009-NP April 2009 Revision 1

5-15 Table 5-10 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 Fluence Temperature Shift USE Decrease (x 10' 9n/cm 2, Predicted Measured Predicted Measured Material Capsule E > 1.0 MeV) (OF) (a) (OF) (b) (%) (a) (%)

U 0.332 26.7 13.3 15 ---

Intermediate Shell Y 1.14 39.8 32.4 20 -- -

Plate B8805-3 (Longitudinal) V 1.93 45.3 42.4 22 3 X 3.47 50.9 96.6 26 10 W 4.36 52.8 98.1 27 18 U 0.332 26.7 0.0(c) 15 ---

Intermediate Shell Y 1.14 39.8 14.7 20 ---

Plate B8805-3 V 1.93 45.3 33.5 22 2 (Transverse) X 3.47 50.9 60.8 26 3 W 4.36 52.8 94.1 27 15 U 0.332 23.5 25.1 15 - - -

Y 1.14 34.9 11.4 20 4 Weld Metal (Heat # 83653) V 1.93 39.8 0.0(c) 22 5 X 3.47 44.6 55.4 26 5 W 4.36 46.3 34.8 27 14 U 0.332 --- 0.0(c) --- 4 Y 1.14 --- 20.1 --- 7 HAZ Material V 1.93 -- - 41.0 -- - 10 X 3.47 --- 11.7 -- - 9 W 4.36 - - - 45.1 --- 15 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 by CVGraph Version 5.3 using measured Charpy data (See Appendix C).

(c) Measured ARTNDT value was determined to be negative, but physically a reduction should not occur, therefore a conservative value of zero is used.

WCAP- 17009-NP April 2009 Revision 1

5-16 Table 5-11 Tensile Properties of the Vogtle Unit 1 Capsule W Reactor Vessel Surveillance Materials Irradiated to 4.36E+19 n/cm 2 (E > 1.0 MeV) 0.2%

Test Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Temp. Strength Strength Load Stress Strength Elongation Elongation in Area Material Number (OF) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%)

AL7 100 80.0 102.9 3.45 187.6 70.3 11.4 23.8 63 Intermediate Shell Plate B8805-3 AL8 190 77.7 99.3 3.23 203.6 65.7 10.5 24.1 68 (Longitudinal) AL9 550 70.6 97.8 3.60 151.4 73.3 10.4 20.9 52 AT7 115 79.2 101.3 3.50 147.2 71.3 10.8 22.7 52 Intermediate Shell Plate B8805-3 AT8 190 76.4 97.8 3.35 170.9 68.2 11.4 25.1 60 (Transverse) AT9 550 71.8 97.3 3.80 151.0 77.4 10.5 19.8 49 AW7 30 77.9 93.7 2.84 207.5 57.9 11.3 26.7 72 Weld Metal AW8 75 76.9 90.7 2.68 195.5 54.5 10.8 27.0 72 (Heat # 83653)

AW9 550 66.2 86.6 2.83 186.2 57.6 10.5 24.0 69 WCAP- 17009-NP April 2009 Revision 1

5-17 Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/09/2008 09:46 AM Data Set(s) Plotted Curve Plant Capsule Material Oli. Heat #

1 Vogtle 1 Unirra SA533B1 LT C0623-1 2 Vogtle 1 U SA533B1 LT C0623-1 3 Vogtle I Y SA533BI LT C0623-1 4 Vogtle 1 V SA533B1 LT C0623- 1 5 Vogtle 1 x SA533B1 LT C0623-1 6 Vogtle I w SA533BI LT C0623-1 300 250 4200 150 z

~100 50 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature In Dog F 01 a 2 3 4 v 56 Results Curve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 1 2. 2 122.0 .0 -14.9 .0 21.9 .0 2 2. 2 134.0 12.0 -1.6 13. 3 36.9 15.0 3 2.2 132.0 10.0 17.5 32.4 59. 8 37.9 4 2.2 118.0 -4.0 27. 5 42.4 80. 7 58. 8 5 2. 2 110.0 -12.0 81.7 96. 6 121.9 100.0 6 2. 2 100. 0 -22.0 83. 2 98. 1 134.9 113.0 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-18 Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:15 AM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Vogt e 1 Unirra SA533B1 LT C0623-1 2 Vogtle 1 U SA533B1 LT C0623-1 3 Vogle 1 Y SA533B1 LT C0623-1 4 Vogtle 1 V SA533BI LT C0623-1 5 Vogtle 1 X SA533B1 LT C0623-1 6 Vogtle 1 w SA533B1 LT C0623-1 200 150 Em 100 s

50 0 4-

-300.0 0.0 300.0 600.0 Temperature in Dog F 0 1 a 2 03 A 4 v5 6 Results Curve Fluence LSE USE d-USE T @35 d-T @35 1 .0 87.4 .0 18. 8 .0 2 .0 81.9 -5.4 32.6 13. 8 3 .0 82.3 -5.1 52.0 33.2 4 .0 76.9 -10.5 87. 8 69.0 5 .0 76.5 -10.9 130. 5 111.7 6 .0 90.7 3. 3 114. 8 96.0 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

WCAP-17009-NP April 2009 Revision I

5-19 Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:16 AM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

1 Vogtle 1 Unirra SA533B1 LT C0623-1 2 Vogtle I U SA533B1 LT C0623-1 3 Vogt le 1 Y SA533B1 LT C0623-1 4 VogUe 1 V SA533B1 LT C0623-1 5 Vogtle I x SA533B1 LT C0623-1 6 Vogtle 1 w SA533B1 LT C0623-1 125 100 75 I

?-

50 25 0-

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Dog F 01 02 03 ' 4 V 5 0 6 Results Curve Fluence LSE USE d-USE T @50 d-T @50 1 .0 100.0 .0 64.4 .0 2 .0 100.0 .0 86.5 22.1 3 .0 100.0 .0 70.2 5.8 4 .0 100.0 0 109.9 45.5 5 .0 100.0 .0 141.7 77.3 6 .0 100.0 .0 148.4 84.0 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

WCAP- 17009-NP April 2009 Revision I

5-20 Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:21 AM Data Set(s) Plotted Curve PI3 nt Capsule Material Ori. Heat #

1 Vogtlle 1 Unirra SA533B1 TL C0623-1 2 Vogtlle 1 U SA533B1 TL C0623-1 3 Vogtlle 1 Y SA533B1 TL C0623-1 4 Vogt]le 1 V SA533B1 TL C0623-1 5 Vogt?le 1 x SA533B1 TL C0623-1 6 Vogtlle 1 w SA533B1 TL C0623-1 300 250 200 9150 z

100 50 -

0

-300.0 -200,.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Dog F 0 1 0 2 03 -4 4 v 5 0 6 Results CI urve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 1 2.2 96.0 .0 17. 1 .0 62. 5 .0 2 2.2 98.0 2.0 7. 3 -9.8 62. 1 -. 4 3 2.2 106.0 10.0 31. 8 14. 7 88.8 26. 3 4 2.2 94.0 -2.0 50.6 33.5 108.4 45.9 5 2.2 93.0 -3.0 77. 9 60. 8 135. 2 72. 7 6 2.2 82.0 -14.0 111.2 94. 1 171. 7 109.2 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-21 Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:21 AM Data Set(s) Plotted Curve Plant Capsule Material On. Heat #

1 Vogtle I Unirra SA533B1 TL C0623-1 2 Vogtle 1 U SA533BI TL C0623-1 3 Vogtie 1 Y SA533B1 TL C0623-1 4 Vogtle 1 V SA533B1 TL C0623-1 5 Vogtle 1 X SA533B1 TL C0623-1 6 Vogtle 1 w SA533B1 TL C0623-1 200 150 100 50 0 1

-300.0 0.0 300.0 600.0 Temperature in Dog F 0 1 0 2 03 4 v 5 06 Results Curve Fluence LSE USE d-USE T @35 d-T @35 1 0 79.9 .0 53.7 .0 2 .0 72.8 -7.1 49.6 -4. 1 3 .0 78.0 -1.9 69.9 16.2 4 .0 69.6 -10.3 105. 2 51.5 5 .0 78. 1 -1.8 142. 0 88.3 6 .0 74. 1 -5.8 139. 0 85. 3 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-22 Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:22 AM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Vogtle 1 Unirra SA533B1 TL C0623-1 2 Vogtle 1 U SA533B1 TL C0623-1 3 Vogtle 1 Y SA533B1 TL C0623-1 4 Vogtle 1 V SA533B1 TL C0623-1 5 Vogtle 1 x SA533B1 TL C0623-1 6 Vogtle 1 w SA533B1 TL C0623-1 125 100 I- 75 (0

50 25 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Dog F 0 1 02 03 - 4 v 5 0 6 Results Curve Fluence LSE USE d-USE T @50 d-T @50 1 .0 100.0 .0 80.9 .0 2 .0 100.0 .0 73.5 -7.4 3 .0 100.0 .0 91.3 10.4 4 .0 100.0 .0 107.6 26. 7 5 .0 100.0 .0 145.8 64.9 6 .0 100.0 .0 165.0 84. 1 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-23 Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/09/2008 10:08 AM Data Set(s) Plotted Curve Plant Capsule Material Oril. Heat #

1 Vogtle 1 Unirra SAW NA Wire 83653 2 Vogtle 1 U SAW NA Wire 83653 3 Vogtle 1 Y SAW NA Wire 83653 4 Vogtle 1 V SAW NA Wire 83653 5 Vogtle 1 X SAW NA Wire 83653 6 Vogtle 1 W SAW NA Wire 83653 300 250 7 200 150 z

100 50 0 =

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Dog F 0 1 02 03 . 4 v5 0 6 Results Curve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 1 2. 2 145.0 .0 -57.2 .0 -30.3 .0 2 2. 2 153.0 8.0 -32. 1 25. 1 -14.4 15.9 3 2.2 139.0 -6.0 -45.8 11.4 -24.8 5. 5 4 2. 2 138.0 -7.0 -57.5 -. 3 -37.6 -7.3 5 2. 2 138.0 -7.0 -1.8 55. 4 13. 1 43.4 6 2.2 125.0 -20.0 -22.4 34.8 1.1 31.4 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal WCAP- 17009-NP April 2009 Revision I

5-24 Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:49 AM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Vogte 1 Unirra SAW NA Wire 83653 2 Vogtle I U SAW NA Wire 83653 3 Vogtle 1 Y SAW NA Wire 83653 4 Vogtle V SAW NA Wire 83653 5 Vogtue 1 X SAW NA Wire 83653 6 Vogtle 1 W SAW NA Wire 83653 200 150 0

3100 0 0 5

50 0.

A 0

-300.0 0.0 300.0 600.0 Temperature in Dog F 0 1 a 2 0>3 ' 4 v 5 0 6 Results Curve Fluence LSE USE d-USE T @35 d-T @35 1 .0 88.3 .0 -32. 6 .0 2 .0 83. 1 -5.2 -19.9 12.7 3 .0 76.2 -12.0 -24. 2 8.4 4 .0 84.9 -3.4 -32.9 -. 3 5 .0 82.0 -6.2 26.9 59. 5 6 .0 88.5 .3 -8.0 24.6 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal WCAP- 17009-NP April 2009 Revision 1

5-25 Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:49 AM Data Set(s) Plotted Curve Plant Capsule Mateiial O. Heat #

1 Votle 1 Unirra SAW NA Wire 83653 2 Vogtle 1 U SAW NA Wire 83653 3 Vogtle 1 y SAW NA Wire 83653 4 Vogtle 1 V SAW NA Wire 83653 5 Vogtle 1 x SAW NA Wire 83653 6 Vogtle 1 W SAW NA Wire 83653 125 100 I 75 50 25 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F 0 1 0 2 4 v 5 0 6 Results Curve Fluence LSE USE d-USE T @50 d-T @50 1 .0 100.0 .0 -6.1 .0 2 .0 100.0 .0 -19.3 -13.2 3 .0 100.0 .0 -11.3 -5.2 4 .0 100.0 .0 -24.4 -18.3 5 .0 100.0 .0 11.9 18.0 6 .0 100.0 .0 16.0 22.1 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal WCAP-17009-NP April 2009 Revision 1

5-26 Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/22/2008 04:19 PM Data Set(s) Plotted Curve Pla Jt Capsule Material Oril. Heat #

1 Vogt te 1 Unirra SAW NA B8805-1 2 Vogt te 1 U SAW NA B8805-1 3 Vogt le 1 Y SAW NA B8805-1 4 Vogt le 1 V SAW NA B8805-1 5 Vogt le 1 x SAW NA B8805-1 6 Vogtle 1 W SAW NA B8805-1 300 250 A200 S150 z

100 50 0 .

-300.0 -200 .0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Dog F 0 1 0 2 03

  • 4 v 5 06 Results Cuurve Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 1 2. 2 134,0 .0 -86. 1 .0 -55.4 .0 2 2.2 129.0 -5.0 - 105.9 -19.8 -62. 1 -6.7 3 2.2 124.0 -10.0 -66.0 20. 1 -39.1 16.3 4 2.2 121.0 -13.0 -45. 1 41.0 -13.0 42.4 5 2.2 122.0 -12.0 -74.4 11.7 -32.3 23. 1 6 2.2 114.0 -20.0 -41.0 45. 1 3. 5 58. 9 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material WCAP- 17009-NP April 2009 Revision I

5-27 Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/22/2008 04:16 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Vogtle 1 Unirra SAW NA B8805-1 2 Vogtle I U SAW NA B8805-1 3 Vogtle I Y SAW NA B8805-1 4 Vogtle 1 V SAW NA B8805-1 5 Vogtle 1 X SAW NA B8805-1 6 Vogtle 1 W SAW NA B8805-1 200

.3 150 C

100 3.10so 50 0 4

-300.0 0.0 300.0 600.0 Temperature in Dog F 0 1 a 2 0 3 4 5 0 66 Results Curve Fluence LSE USE d-USE T @35 d-T @35 1 .0 79. 1 .0 -49.7 .0 2 .0 74.3 -4.8 -47.6 2. 1 3 .0 73.0 -6.1 -40.3 9.4 4 .0 71.4 -7.7 -3.9 45.8 5 .0 68. 8 -10.3 -6.4 43.3 6 .0 79.0 -. 1 6. 1 55.8 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Vogtle Unit I Reactor Vessel Heat-Affected-Zone Material WCAP- 17009-NP April 2009 Revision 1

5-28 Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/22/2008 04:17 PM Data Set(s) Plotted Curve Plant Capsule Material Ori. Heat #

1 Vogtle 1 Unirra SAW NA B8805-1 2 Vogtle 1 U SAW NA B8805-1 3 Vogtle 1 Y SAW NA B8805-1 4 Vogtle 1 V SAW NA B8805-1 5 Vogtle 1 x SAW NA B8805-1 6 Vogtle 1 W SAW NA B8805-1 125 100 I- 75 50 25 --

0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Dog F 0 1 a 2 03 N4 v 5 0 6 Results Curve Fluence LSE USE d-USE T @50 d-T @50 1 .0 100.0 .0 -24. 2 .0 2 .0 100.0 .0 -44.6 -20.4 3 .0 100.0 .0 -18.2 6.0 4 .0 100.0 .0 -1.2 23.0 5 .0 100.0 .0 -7.5 16.7 6 .0 100.0 .0 25. 3 49.5 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material WCAP- 17009-NP April 2009 Revision 1

5-29 AL43, -30°F AL39, 0-F AL42, 50°F AL36, 60°F AL45, 75°F AL33, 100-F AL31, 125°F AL37, 150°F AL44,170°F AL32, 180°F AL34, 200°F AL41, 200°F AL35, 300°F AL38, 325°F AL40, 350°F Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

WCAP-17009-NP April 2009 Revision 1

5-30 AT11 nOT7J AT45. 80°F AT36 10OOF AT40- 110OF A Tad 1)Ilc AT43, 125OF AT31, 150OF AT42, 165WF AT38, 180 0F AT39, 190OF AT34, 200OF AT32, 2000 F AT35, 275OF AT33. 300OF AT37, 325OF Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-31 AW35, -50T AW14 -_ 0OF AW'43 -1 *0 F AWIR -150F AW40. -10OF AW45, -lOF AV33. -5F AW36. OTF AW42, 100F AW41, 20OF AW39,25OF AWI,3 1, 1225c AW32.150OF AW44. 17S'F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal WCAP-17009-NP April 2009 Revision I

5-32 AH37, -30 0 F AH45, -30OF A1312. -10F AH44. 00 F AH42, 10cF AH38. 250F AH35. 50OF AH31. 1500 F AI{36. 175OF AH34, 200TF Figure 5-16 Charpy Impact Specimen Fracture Surfaces for the Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material WCAP- 17009-NP April 2009 Revision 1

5-33 120.0 -

100.0 Ultimate Tensile Strength 80.0 C

0.2% Yield Strength C 60.0 40.0 20.0 0.0 0 100 200 300 400 500 600 Temperature (*F)

Legend: Aand

  • and u are unirradiated A and o and o are irradiated to 4.36E+19 n/cm 2 (E > 1.0 MeV) 80.0 Area Reduction 70.0 60.0 50.0

= 40.0 30.0

-noTotal Elongation 20.0 Uniform Elongation 10.0 0.0 1 0 100 200 300 400 500 600 Temperature (*F)

Figure 5-17 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

April 2009 WCAP- 17009-NP WCAP- I17009-NP April 2009 Revision I

5-34 120.0 T Ultimate Tensile Strength 100.0 80.0 0.2% Yield Strength .1 C¢ 60.0 +-

cn 40.0 20.0 0.0 0 100 200 300 400 500 600 Temperature (°F)

Legend: A and

  • and m are unirradiated A and o and oI are irradiated to 4.36E+19 n/cm 2 (E > 1.0 MeV) 70.01 Area Reduction 60.0 50.0 40.0 30.0 20.0 Uniform Elongation 10.0 0.0 0 100 200 300 400 500 600 Temperature (*F)

Figure 5-18 Tensile Properties for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision I

5-35 100.0 Ultimate Tensile Strength 90.0 80.0 70.0 0.2%/ Yield Strength 60.0 0

50.0 40.0 30.0 20.0 10.0 0.0 0 100 200 300 400 500 600 Temperature (*F)

Legend: A and

  • and i are unirradiated A and o and o are irradiated to 4.36E+19 n/cm 2 (E > 1.0 MeV) 80.0 Area Reduction 70.0 60.0 50.0

= 40.0 30.0 Total Elongation 20.0 Uniform Elongation 10.0 0.0 100 200 300 400 500 600 Temperature (*F)

Figure 5-19 Tensile Properties for the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal WCAP- 17009-NP April 2009 Revision 1

5-36 Specimen AL7- Tested at 1000 F

-- - Mwww_______- ~-, ;

Specimen AL9- Tested at 5501F Figure 5-20 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-37 Specimen AT7- Tested at 1151F Specimen AT9- Tested at 5501F Figure 5-21 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-38 Specamen AW7- Tested at 30°F Specimen AW9- Tested at 550IF Figure 5-22 Fractured Tensile Specimens from the Vogtle Unit 1 Reactor Vessel Surveillance Program Weld Metal WCAP- 17009-NP April 2009 Revision 1

5-39 VOGTLE I CAPSULE W 120 100 S80 so 40 AL7 I(40*F 20 0

0.05 M0.1 0.15 0.2 0.25 0.3 STRAIN. INN VOGTLE 1 CAPSULE W 120 ALB 130TF 0

.311 ý.15 0..2 C.3 STIRWN, NI VOGTLE 1 CAPSULE W s0 AUO SSW=:

G.3 BTRX6N. LMAN Figure 5-23 Engineering Stress-Strain Curves for Vogtle Unit 1 Intermediate Shell Plate B8805-3 Tensile Specimens AL7, AL8 and AL9 (Longitudinal Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-40 VOGTLE 1 CAPSULE W 120 40 AT 115TF 20 0

0 OLE -2.1 O1AS 0.2 .25 0.3 STRMN. MY VOGTLE 1 CAPSULE W 120 120 40* ATS 20 0

0* -1 2.11- 0.2 0.25 C-3 VOGTLE 1 CAPSULE W 120 1-20 40 r AT9 550T 20 0

-0.Cs -3.1 -3.1E 02 0.25 -.3 Figure 5-24 Engineering Stress-Strain Curves for Vogtle Unit 1 Intermediate Shell Plate B8805-3 Tensile Specimens AT7, AT8 and AT9 (Transverse Orientation)

WCAP- 17009-NP April 2009 Revision 1

5-41 VOGTLE 1 CAPSULE W 1-70 70 AW7 30'F 10 0

0 STAI.1L VOGTIE*1 CAPSULE W 130 -

Sc0 SO 70.

Mc SO 40 AWM 75F 20.

0 a O.CS 0-1 C.2 STRAL'N, IN04 VOGTLE 1 CAPSULE W 90 so 70 Sc 40 Sc 30 AWS 559'F 10 0

0 2.1 .3j 0.21 STRWox 1mw7 Figure 5-25 Engineering Stress-Strain Curves for Vogtle Unit 1 Surveillance Program Weld Metal Tensile Specimens AW7, AW8 and AW9 WCAP- 17009-NP April 2009 Revision I

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates Sn transport analysis performed for the Vogtle Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules.

In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule W, withdrawn at the end of the fourteenth plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the Vogtle Unit 1 reactor, the sensor sets from the previously withdrawn capsules (U, Y, V, and X) were re-analyzed using the current dosimetry evaluation methodology. These dosimetry updates are presented in Appendix A of this report. Comparisons of the results from these dosimetry evaluations with the analytical predictions served to validate the plant-specific neutron transport calculations. These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY).

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy-dependent damage function for data correlation, ASTM Standard Practice E853-01, "Analysis and Interpretation of Light-Water Reactor Surveillance Results," [21] recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy-dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693-01, "Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom" [22]. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials" [I].

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [23]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 [24].

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6-2 6.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Vogtle Unit 1 reactor vessel is shown in Figure 4-1.

Six irradiation capsules attached to the neutron pad are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.5', 610, 121.5',

238.50, 2410, and 301.50 as shown in Figure 4-1. These full-core positions correspond to the following octant symmetric locations represented in Figure 6-1: 29' from the core cardinal axes (for the 610 and 2410 dual surveillance capsule holder locations found in octants with a 22.5' neutron pad segment) and 31.5' from the core cardinal axes (for the 121.50 and 301.5' single surveillance capsule holder locations found in octants with a 20.00 neutron pad segment, and for the 58.50 and the 238.5' dual surveillance capsule holder locations found in octants with a 22.50 neutron pad segment). The stainless steel specimen containers are 1.182-inch by 1-inch and are approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant.

The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pads and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the Vogtle Unit 1 reactor vessel and surveillance capsules, a series of fuel cycle specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

p(r,0, z) = 9(r,0 (r,z) (Eqn. 6-1) where d(r,0,z) is the synthesized three-dimensional neutron flux distribution, o(rO) is the transport solution in r,0 geometry, o(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and ý(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Vogtle Unit 1.

For the Vogtle Unit I transport calculations, the rO models depicted in Figure 6-1 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric. These rO models include the core, the reactor internals, the neutron pads - including explicit representations of octants not containing surveillance capsules and octants with surveillance capsules at 290 and 31.50, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall.

These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing these analytical models, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and WCAP- 17009-NP April 2009 Revision I

6-3 miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,0 reactor models consisted of 183 radial by 99 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the rO calculations was set at a value of 0.001.

The rz model used for the Vogtle Unit 1 calculations is shown in Figure 6-2 and extends radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation below the lower core plate to above the upper core plate. As in the case of the rO models, nominal design dimensions and full power coolant densities were employed in the calculations.

In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of these reactor models consisted of 153 radial by 188 axial intervals. As in the case of the r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis.

The pointwise inner iteration flux convergence criterion utilized in the rz calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 153 radial mesh intervals included in the rz model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The core power distributions used in the plant-specific transport analysis were provided by the Nuclear Fuels Division of Westinghouse for each of the first fifteen fuel cycles at Vogtle Unit 1. Specifically, the data utilized included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.2 [25] and the BUGLE-96 cross-section library [26]. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and angular discretization was modeled with an S16 order of angular quadrature. Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-6. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 290 dual capsule, 31.50 dual capsule, and 31.50 single capsule. These WCAP- 17009-NP 'April 2009 Revision 1

6-4 results, representative of the axial midplane of the active core, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future. Similar information is provided in Table 6-2 for the reactor vessel inner radius at four azimuthal locations. The vessel data given in Table 6-2 were taken at the clad/base metal interface, and thus, represent maximum calculated exposure levels on the vessel.

From the data provided in Table 6-2 it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the fourteenth fuel cycle (i.e., after 18.41 EFPY of plant operation) was 1.05 x 1019 n/cm 2 .

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Tables 6-1 and 6-2. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the fourteenth fuel cycle as well as future projections to 20, 23, 26, 30, 34, 38, 42, 48, and 54 EFPY. The calculations account for uprates from 3411 MWt to 3565 MWt that occurred at the onset of Cycle 5 and from 3565 MWt to 3626 MWt that occurred at the onset of Cycle 15. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 15 were representative of future plant operation. The future projections are also based on the current reactor power level of 3626 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-3 and 6-4, respectively. The data, based on the cumulative integrated exposures from Cycles 1 through 14, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-3 and 6-4.

The calculated fast neutron exposures for the five surveillance capsules withdrawn from the Vogtle Unit 1 reactor are provided in Table 6-5. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the Vogtle Unit 1 reactor.

From the data provided in Table 6-5, Capsule W received a fluence (E > 1.0 MeV) of 4.36 x 101" n/cm2 after exposure through the end of the fourteenth fuel cycle (i.e., after 18.41 EFPY of plant operation).

Updated lead factors for the Vogtle Unit 1 surveillance capsules are provided in Table 6-6. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-6, the lead factors for capsules that have been withdrawn from the reactor (U, Y,. V, X, and W) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. The lead factor for Capsule Z, which was withdrawn at the end of Cycle 14 and stored in the spent fuel pool, corresponds to the calculated fluence values at the end of Cycle 14, the last completed fuel cycle for Vogtle Unit 1.

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6-5 6.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures previously reported in Section 6.2 is demonstrated by a direct comparison against the measured sensor reaction rates and via a least squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serves to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least squares evaluation comparisons are documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule W, that was withdrawn from Vogtle Unit 1 at the end of the fourteenth fuel cycle, is summarized below.

Reaction Rates (rps/atom)

Reaction Measured Calculated M/C Ratio 63 Cu(n,C) 60Co 4.51 E- 17 4.13E-17 1.09 54Fe(n,p) 54 Mn 4.36E-15 4.51E-15 0.97 58Ni(n,p)58 Co 6.23E- 15 6.31 E- 15 0.99 238U(n,f)137 Cs (Cd) 2.59E-14 2.40E-14 1.08 237Np(n,f)137 Cs (Cd) 2.17E-13 2.33E-13 0.93 Average: 1.01

% Standard Deviation: 7.0 The measured-to-calculated (M/C) reaction rate ratios for the Capsule W threshold reactions range from 0.93 to 1.09, and the average M/C ratio is 1.01 +/- 7.0% (1c). This direct comparison falls well within the

+/- 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the Vogtle Unit I reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Vogtle Unit 1.

6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Vogtle Unit 1 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190.

In particular, the qualification of the methodology was carried out in the following four stages:

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).
2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

WCAP- 17009-NP April 2009 Revision 1

6-6

3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant specific transport calculations used in the neutron exposure assessments.
4. Comparisons of the plant-specific calculations with all available dosimetry results from the Vogtle Unit 1 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the Vogtle Unit 1 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Vogtle Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures previously described in Section 6.2. As such, the validation of the Vogtle Unit 1 analytical model based on the measured plant dosimetry is completely described in Appendix A.

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 24.

Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results.

The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for Vogtle Unit 1.

WCAP- 17009-NP April 2009 Revision 1

6-7 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)

Cumulative Cumulative Neutron Flux (E > 1.0 MeV) [n/cm 2-sl Cycle Irradiation Irradiation Length Time Time Cycle IEFPS] IEFPSI [EFPY] Dual 290 Dual 31.50 Single 31.50 1 3.60E+07 3.60E+07 1.14 8.57E+10 9.20E+10 9.10E+10 2 3.58E+07 7.19E+07 2.28 7.44E+ 10 7.97E+10 7.90E+ 10 3 4.19E+07 1.14E+08 3.61 7.65E+10 8.44E+10 8.36E+10 4 3.92E+07 1.53E+08 4.85 6.13E+10 6.64E+10 6.57E+10 5 4.19E+07 1.95E+08 6.17 6.53E+10 7.07E+10 7.00E+10 6 4.24E+07 2.37E+08 7.52 6.62E+10 7.11E+10 7.04E+10 7 3.97E+07 2.77E+08 8.78 5.94E+10 6.43E+10 6.36E+10 8 4.22E+07 3.19E+08 10.11 6.18E+10 6.80E+10 6.74E+10 9 4.61E+07 3.65E+08 11.57 7.38E+10 8.58E+10 8.52E+10 10 4.28E+07 4.08E+08 12.93 7.63E+10 8.37E+10 8.29E+10 11 4.41E+07 4.52E+08 14.33 7.01E+10 7.79E+10 7.72E+10 12 4.28E+07 4.95E+08 15.68 6.61E+10 7.22E+10 7.15E+10 13 4.41E+07 5.39E+08 17.08 6.71E+10 7.30E+10 7.23E+10 14 4.21E+07 5.81E+08 18.41 6.56E+10 7.15E+10 7.09E+10 Future 5.01E+07 6.31E+08 20.00 7.16E+10 7.93E+10 7.86E+10 Future 9.47E+07 7.26E+08 23.00 7.16E+ 10 7.93E+10 7.86E+10 Future 9.47E+07 8.20E+08 26.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 9.47E+08 30.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 1.07E+09 34.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 1.20E+09 38.00 7.16E+10 7.93E+10 7.86E+10 Future 1.26E+08 1.33E+09 42.00 7.16E+10 7.93E+10 7.86E+10 Future 1.89E+08 1.51E+09 48.00 7.16E+10 7.93E+10 7.86E+10 Future 1.89E+08 1.70E+09 54.00 7.16E+10 7.93E+10 7.86E+10 Note:

(a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP- 17009-NP April 2009 Revision 1

6-8 Table 6-1 (Continued) Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV) In/cm 2]

Cycle Irradiation Irradiation Length Time Time Cycle [EFPSI [EFPSI [EFPY] Dual 290 Dual 31.5' Single 31.50 1 3.60E+07 3.60E+07 1.14 3.09E+18 3.32E+18 3.28E+18 2 3.58E+07 7.19E+07 2.28 5.76E+18 6.17E+18 6.11E+18 3 4.19E+07 1.14E+08 3.61 8.96E+18 9.71E+18 9.62E+18 4 3.92E+07 1.53E+08 4.85 1.14E+19 1.23E+19 1.22E+19 5 4.19E+07 1.95E+08 6.17 1.41E+19 1.53E+19 1.51E+19 6 4.24E+07 2.37E+08 7.52 1.69E+19 1.83E+19 1.81E+19 7 3.97E+07 2.77E+08 8.78 1.93E+19 2.08E+19 2.06E+19 8 4.22E+07 3.19E+08 10.11 2.19E+19 2.37E+19 2.35E+19 9 4.61E+07 3.65E+08 11.57 2.53E+19 2.77E+19 2.74E+19 10 4.28E+07 4.08E+08 12.93 2.85E+19 3.12E+19 3.1OE+19 11 4.41E+07 4.52E+08 14.33 3.16E+19 3.47E+19 3.44E+19 12 4.28E+07 4.95E+08 15.68 3.45E+19 3.78E+19 3.74E+19 13 4.41E+07 5.39E+08 17.08 3.74E+19 4.10E+19 4.06E+19 14 4.21E+07 5.81E+08 18.41 4.02E+19 4.40E+19 4.36E+19 Future 5.01E+07 6.31E+08 20.00 4.38E+19 4.80E+19 4.75E+19 Future 9.47E+07 7.26E+08 23.00 5.05E+19 5.55E+19 5.50E+19 Future 9.47E+07 8.20E+08 26.00 5.73E+19 6.30E+19 6.24E+19 Future 1.26E+08 9.47E+08 30.00 6.64E+19 7.30E+19 7.23E+19 Future 1.26E+08 1.07E+09 34.00 7.54E+19 8.30E+19 8.22E+19 Future 1.26E+08 1.20E+09 38.00 8.44E+19 9.30E+19 9.21E+19 Future 1.26E+08 1.33E+09 42.00 9.35E+19 1.03E+20 1.02E+20 Future 1.89E+08 1.51E+09 48.00 1.07E+20 1.18E+20 1.17E+20 Future 1.89E+08 1.70E+09 54.00 1.21E+20 1.33E+20 1.32E+20 Note:

(a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP-17009-NP April 2009 Revision 1

6-9 Table 6-1 (Continued) Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)

Cumulative Cumulative Iron Atom Displacement Rate Idpa/sl Cycle Irradiation Irradiation Length Time Time Cycle [EFPSI [EFPSI [EFPYI Dual 290 Dual 31.51 Single 31.50 1 3.60E+07 3.60E+07 1.14 1.68E-10 1.80E-10 1.78E-10 2 3.58E+07 7.19E+07 2.28 1.45E-10 1.55E-10 1.53E-10 3 4.19E+07 1.14E+08 3.61 1.49E-10 1.64E-10 1.62E-10 4 3.92E+07 1.53E+08 4.85 1.19E-10 1.29E-10 1.27E-10 5 4.19E+07 1.95E+08 6.17 1.26E-10 1.37E-10 1.35E-10 6 4.24E+07 2.37E+08 7.52 1.28E-10 1.38E-10 1.36E-10 7 3.97E+07 2.77E+08 8.78 1.15E-10 1.24E-10 1.23E-10 8 4.22E+07 3.19E+08 10.11 1.20E-10 1.33E-10 1.31E-10 9 4.61E+07 3.65E+08 11.57 1.43E-10 1.66E-10 1.65E-10 10 4.28E+07 4.08E+08 12.93 1.48E-10 1.62E-10 1.61E-10 11 4.41E+07 4.52E+08 14.33 1.36E-10 1.51E-10 1.50E-10 12 4.28E+07 4.95E+08 15.68 1.28E-10 1.40E-10 1.38E-10 13 4.41E+07 5.39E+08 17.08 1.30E-10 1.41E-10 1.40E-10 14 4.21E+07 5.81E+08 18.41 1.27E-10 1.39E-10 1.37E-10 Future 5.01E+07 6.31E+08 20.00 1.39E-10 1.54E-10 1.53E-10 Future 9.47E+07 7.26E+08 23.00 1.39E-10 1.54E-10 1.53E-10 Future 9.47E+07 8.20E+08 26.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 9.47E+08 30.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 1.07E+09 34.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 1.20E+09 38.00 1.39E-10 1.54E-10 1.53E-10 Future 1.26E+08 1.33E+09 42.00 1.39E-10 1.54E-10 1.53E-10 Future 1.89E+08 1.51E+09 48.00 1.39E-10 1.54E-10 1.53E-10 Future 1.89E+08 1.70E+09 54.00 1.39E-10 1.54E-10 1.53E-10 Note:

(a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

6-10 Table 6-1 (Continued) Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center(a)

Cumulative Cumulative Iron Atom Displacement Rate Idpa/s]

Cycle Irradiation Irradiation Length Time Time Cycle IEFPSI [EFPS] [EFPY] Dual 290 Dual 31.5' Single 31.50 1 3.60E+07 3.60E+07 1.14 6.05E-03 6.48E-03 6.41E-03 2 3.58E+07 7.19E+07 2.28 1.12E-02 1.20E-02 1.19E-02 3 4.19E+07 1.14E+08 3.61 1.75E-02 1.89E-02 1.87E-02 4 3.92E+07 1.53E+08 4.85 2.21E-02 2.39E-02 2.37E-02 5 4.19E+07 1.95E+08 6.17 2.74E-02 2.97E-02 2.94E-02 6 4.24E+07 2.37E+08 7.52 3.28E-02 3.55E-02 3.5 1E-02 7 3.97E+07 2.77E+08 8.78 3.74E-02 4.04E-02 4.OOE-02 8 4.22E+07 3.19E+08 10.11 4.25E-02 4.60E-02 4.55E-02 9 4.61E+07 3.65E+08 11.57 4.91E-02 5.37E-02 5.32E-02 10 4.28E+07 4.08E+08 12.93 5.54E-02 6.06E-02 6.OOE-02 11 4.41E+07 4.52E+08 14.33 6.14E-02 6.73E-02 6.66E-02 12 4.28E+07 4.95E+08 15.68 6.69E-02 7.33E-02 7.26E-02 13 4.41E+07 5.39E+08 17.08 7.26E-02 7.95E-02 7.87E-02 14 4.21E+07 5.81E+08 18.41 7.80E-02 8.53E-02 8.45E-02 Future 5.01E+07 6.31E+08 20.00 8.50E-02 9.30E-02 9.21E-02 Future 9.47E+07 7.26E+08 23.00 9.81E-02 1.08E-01 1.07E-01 Future 9.47E+07 8.20E+08 26.00 1.11E-01 1.22E-01 1.21E-01 Future 1.26E+08 9.47E+08 30.00 1.29E-01 1.42E-01 1.40E-01 Future 1.26E+08 1.07E+09 34.00 1.46E-01 1.61E-01 1.60E-01 Future 1.26E+08 1.20E+09 38.00 1.64E-01 1.81E-01 1.79E-01 Future 1.26E+08 1.33E+09 42.00 1.82E-01 2.OOE-01 1.98E-01 Future 1.89E+08 1.51E+09 48.00 2.08E-01 2.29E-01 2.27E-01 Future 1.89E+08 1.70E+09 54.00 2.34E-01 2.58E-01 2.56E-01 Note:

(a) Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP- 17009-NP April 2009 Revision 1

6-11 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV) In/cm 2-sI Cycle Irradiation Irradiation Length Time Time Cycle IEFPS] [EFPS] [EFPY] 00 150 300 450 1 3.60E+07 3.60E+07 1.14 1.27E+10 1.89E+10 2.16E+ 10 2.22E+10 2 3.58E+07 7.19E+07 2.28 9.82E+09 1.55E+10 1.87E+10 1.78E+10 3 4.19E+07 1.14E+08 3.61 8.98E+09 1.41E+10 1.92E+10 1.92E+10 4 3.92E+07 1.53E+08 4.85 1.00E+10 1.37E+10 1.58E+10 1.64E+10 5 4.19E+07 1.95E+08 6.17 8.81E+09 1.35E+10 1.64E+10 1.64E+10 6 4.24E+07 2.37E+08 7.52 8.82E+09 1.41E+10 1.68E+10 1.64E+10 7 3.97E+07 2.77E+08 8.78 8.60E+09 1.23E+10 1.51E+10 1.49E+10 8 4.22E+07 3.19E+08 10.11 9.14E+09 1.41E+10 1.59E+10 1.83E+10 9 4.61E+07 3.65E+08 11.57 9.34E+09 1.33E+10 1.89E+10 2.15E+10 10 4.28E+07 4.08E+08 12.93 8.91E+09 1.44E+10 1.91E+10 1.92E+10 11 4.41E+07 4.52E+08 14.33 9.48E+09 1.43E+10 1.78E+10 1.96E+10 12 4.28E+07 4.95E+08 15.68 9.39E+09 1.34E+10 1.68E+10 1.71E+10 13 4.41E+07 5.39E+08 17.08 8.98E+09 1.38E+10 1.70E+10 1.71E+10 14 4.21E+07 5.81E+08 18.41 1.02E+10 1.42E+10 1.66E+10 1.70E+10 Future 5.01E+07 6.31E+08 20.00 9.64E+09 1.46E+10 1.83E+10 2.00E+10 Future 9.47E+07 7.26E+08 23.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 9.47E+07 8.20E+08 26.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 9.47E+08 30.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 1.07E+09 34.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 1.20E+09 38.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.26E+08 1.33E+09 42.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.89E+08 1.51E+09 48.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 Future 1.89E+08 1.70E+09 54.00 9.64E+09 1.46E+10 1.83E+10 2.OOE+10 WCAP- 17009-NP April 2009 Revision 1

6-12 Table 6-2 (Continued) Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence (E > 1.0 MeV) In/cM2I Cycle Irradiation Irradiation Length Time Time Cycle [EFPS] [EFPS] [EFPYJ 00 150 300 450 1 3.60E+07 3.60E+07 1.14 4.58E+17 6.80E+17 7.80E+17 8.00E+17 2 3.58E+07 7.19E+07 2.28 8.10E+17 1.24E+18 1.45E+18 1.44E+18 3 4.19E+07 1.14E+08 3.61 1.19E+18 1.83E+18 2.25E+18 2.24E+18 4 3.92E+07 1.53E+08 4.85 1.57E+18 2.35E+18 2.86E+18 2.86E+18 5 4.19E+07 1.95E+08 6.17 1.94E+18 2.91E+18 3.54E+18 3.55E+18 6 4.24E+07 2.37E+08 7.52 2.31E+18 3.51E+18 4.25E+18 4.24E+18 7 3.97E+07 2.77E+08 8.78 2.65E+18 4.00E+18 4.85E+18 4.83E+18 8 4.22E+07 3.19E+08 10.11 3.03E+18 4.59E+18 5.51E+18 5.60E+18 9 4.61E+07 3.65E+08 11.57 3.46E+18 5.20E+18 6.38E+18 6.59E+18 10 4.28E+07 4.08E+08 12.93 3.84E+18 5.81E+18 7.20E+18 7.41E+18 11 4.41E+07 4.52E+08 14.33 4.26E+18 6.44E+18 7.98E+18 8.27E+18 12 4.28E+07 4.95E+08 15.68 4.66E+18 7.02E+18 8.70E+18 9.OOE+18 13 4.41E+07 5.39E+08 17.08 5.06E+18 7.63E+18 9.45E+18 9.76E+18 14 4.21E+07 5.81E+08 18.41 5.49E+18 8.22E+18 1.02E+19 1.05E+19 Future 5.01E+07 6.31E+08 20.00 5.97E+18 8.95E+18 1.1I1E+19 1.15E+19 Future 9.47E+07 7.26E+08 23.00 6.88E+18 1.03E+19 1.28E+19 1.34E+19 Future 9.47E+07 8.20E+08 26.00 7.79E+18 1.17E+19 1.45E+19 1.53E+19 Future 1.26E+08 9.47E+08 30.00 9.01E+18 1.35E+19 1.68E+19 1.78E+19 Future 1.26E+08 1.07E+09 34.00 1.02E+19 1.54E+19 1.91E+19 2.03E+19 Future 1.26E+08 1.20E+09 38.00 1.14E+19 1.72E+19 2.14E+19 2.28E+19 Future 1.26E+08 1.33E+09 42.00 1.27E+19 1.91E+19 2.37E+19 2.54E+19 Future 1.89E+08 1.51E+09 48.00 1.45E+19 2.18E+19 2.72E+19 2.92E+19 Future 1.89E+08 1.70E+09 54.00 1.63E+19 2.46E+19 3.07E+19 3.30E+19 WCAP- 17009-NP April 2009 Revision I

6-13 Table 6-2 (Continued) Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Idpa/sI Cycle Irradiation Irradiation Length Time Time Cycle IEFPSI IEFPS] [EFPY] 00 150 300 450 1 3.60E+07 3.60E+07 1.14 1.97E-11 2.90E-11 3.33E-11 3.51E-11 2 3.58E+07 7.19E+07 2.28 1.53E-11 2.39E-11 2.88E-11 2.82E-11 3 4.19E+07 1.14E+08 3.61 1.40E-11 2.17E-11 2.96E-11 3.03E-11 4 3.92E+07 1.53E+08 4.85 1.56E- 11 2.11E-11 2.44E- 11 2.58E-11 5 4.19E+07 1.95E+08 6.17 1.37E-11 2.08E- 11 2.53E-11 2.59E-11 6 4.24E+07 2.37E+08 7.52 1.37E-11 2.17E-11 2.58E-11 2.59E-11 7 3.97E+07 2.77E+08 8.78 1.34E- 11 1.89E-11 2.33E-11 2.35E-11 8 4.22E+07 3.19E+08 10.11 1.42E-11 2.17E-11 2.45E-11 2.89E-11 9 4.61E+07 3.65E+08 11.57 1.45E-11 2.05E-11 2.91E-11 3.39E-11 10 4.28E+07 4.08E+08 12.93 1.39E-1 1 2.22E-1 1 2.95E-1 1 3.03E-1 1 11 4.41E+07 4.52E+08 14.33 1.48E-11 2.20E-11 2.75E-11 3.09E-11 12 4.28E+07 4.95E+08 15.68 1.46E-I1 2.07E-11 2.59E-11 2.71E-11 13 4.41E+07 5.39E+08 17.08 1.40E-11 2.13E-11 2.62E-11 2.71E-11 14 4.21E+07 5.81E+08 18.41 1.58E-11 2.19E-11 2.57E-11 2.69E-11 Future 5.01E+07 6.31E+08 20.00 1.50E- 11 2.24E-11 2.82E-11 3.16E-11 Future 9.47E+07 7.26E+08 23.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 Future 9.47E+07 8.20E+08 26.00 1.50E-1 1 2.24E- I1 2.82E- 1I 3.16E-I 1 Future 1.26E+08 9.47E+08 30.00 1.50E-11 2.24E- 11 2.82E- 11 3.16E-11 Future 1.26E+08 1.07E+09 34.00 1.50E- 11 2.24E- 11 2.82E- 11 3.16E-11 Future 1.26E+08 1.20E+09 38.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 Future 1.26E+08 1.33E+09 42.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 Future 1.89E+08 1.51E+09 48.00 1.50E-1I 2.24E-11 2.82E- 11 3.16E-I 1 Future 1.89E+08 1.70E+09 54.00 1.50E-11 2.24E-11 2.82E-11 3.16E-11 WCAP- 17009-NP April 2009 Revision 1

6-14 Table 6-2 (Continued) Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Idpal Cycle Irradiation Irradiation Length Time Time Cycle [EFPSI IEFPS] [EFPY] 00 150 300 450 1 3.60E+07 3.60E+07 1.14 7.11E-04 1.04E-03 1.20E-03 1.27E-03 2 3.58E+07 7.19E+07 2.28 1.26E-03 1.90E-03 2.23E-03 2.27E-03 3 4.19E+07 1.14E+08 3.61 1.84E-03 2.81E-03 3.47E-03 3.54E-03 4 3.92E+07 1.53E+08 4.85 2.44E-03 3.61E-03 4.40E-03 4.52E-03 5 4.19E+07 1.95E+08 6.17 3.01E-03 4.48E-03 5.46E-03 5.61E-03 6 4.24E+07 2.37E+08 7.52 3.59E-03 5.40E-03 6.56E-03 6.71E-03 7 3.97E+07 2.77E+08 8.78 4.12E-03 6.15E-03 7.48E-03 7.64E-03 8 4.22E+07 3.19E+08 10.11 4.72E-03 7.06E-03 8.50E-03 8.85E-03 9 4.61E+07 3.65E+08 11.57 5.39E-03 8.OOE-03 9.84E-03 1.04E-02 10 4.28E+07 4.08E+08 12.93 5.98E-03 8.95E-03 1.11E-02 1.17E-02 I1 4.41E+07 4.52E+08 14.33 6.63E-03 9.92E-03 1.23E-02 1.31E-02 12 4.28E+07 4.95E+08 15.68 7.25E-03 1.08E-02 1.34E-02 1.42E-02 13 4.41E+07 5.39E+08 17.08 7.87E-03 1.17E-02 1.46E-02 1.54E-02 14 4.21E+07 5.81E+08 18.41 8.54E-03 1.27E-02 1.57E-02 1.66E-02 Future 5.01E+07 6.31E+08 20.00 9.28E-03 1.38E-02 1.71E-02 1.81E-02 Future 9.47E+07 7.26E+08 23.00 1.07E-02 1.59E-02 1.97E-02 2.11E-02 Future 9.47E+07 8.20E+08 26.00 1.21E-02 1.80E-02 2.24E-02 2.41E-02 Future 1.26E+08 9.47E+08 30.00 1.40E-02 2.09E-02 2.60E-02 2.81E-02 Future 1.26E+08 1.07E+09 34.00 1.59E-02 2.37E-02 2.95E-02 3.21E-02 Future 1.26E+08 1.20E+09 38.00 1.78E-02 2.65E-02 3.31E-02 3.61E-02 Future 1.26E+08 1.33E+09 42.00 1.97E-02 2.93E-02 3.66E-02 4.01E-02 Future 1.89E+08 1.51E+09 48.00 2.25E-02 3.36E-02 4.20E-02 4.61E-02 Future 1.89E+08 1.70E+09 54.00 2.54E-02 3.78E-02 4.73E-02 5.20E-02 WCAP- 17009-NP April 2009 Revision 1

6-15 Table 6-3 Relative Radial Distribution of Neutron Fluence (E > 1.0 MeV) Within the Reactor Vessel Wall(a)

Radius Azimuthal Angle (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.571 0.567 0.560 0.558 231.06 0.282 0.277 0.272 0.269 236.54 0.134 0.130 0.127 0.125 242.01 0.064 0.059 0.057 0.056 Base Metal Inner Radius = 220.11 cm Base Metal I/4T = 225.59 cm Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Note:

(a) Relative radial distribution data are based on the cumulative integrated exposures from Cycles I through 14.

WCAP-17009-NP April 2009 Revision 1

6-16 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) Within the Reactor Vessel Wall(a)

Radius Azimuthal Angle (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.642 0.637 0.635 0.644 231.06 0.390 0.382 0.381 0.392 236.54 0.237 0.227 0.226 0.234 242.01 0.142 0.128 0.127 0.130 Base Metal Inner Radius = 220.11 cm Base Metal I/4T = 225.59 cm Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Note:

(a) Relative radial distribution data are based on the cumulative integrated exposures from Cycles 1 through 14.

Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Vogtle Unit 1 Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] In/cm 2] Idpal U 1.14 3.32E+18 6.48E-03 Y 4.85 1.14E+ 19 2.21E-02 V 8.78 1.93E+19 3.74E-02 X 14.33 3.47E+19 6.73E-02 W 18.41 4.36E+19 8.45E-02 April 2009 17009-NP WCAP- 17009-NP April 2009 Revision I

6-17 Table 6-6 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor U (58.5') Withdrawn EOC 1 4.14 Y (24P1) Withdrawn EOC 4 3.97 V (610) Withdrawn EOC 7 3.97 X (238.50) Withdrawn EOC 11 4.19 W (121.5') Withdrawn EOC 14 4.16 Z (301.50) Withdrawn EOC 14; Stored in SFP 4.16(a)

Note:

(a) Lead factor for Capsule Z, which was withdrawn at the conclusion of Cycle 14 and stored in the spent fuel pool, is based on cycle-specific exposure calculations through the last completed fuel cycle, i.e., Cycle 14.

April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

6-18 A. W. Vogtle Unrit 1 Reactor R,T Model Meshes- 183R. 916 Ok C ~ ~ h

- - E~ EU j-Nd-H-

I lMaw

[Eini Figure 6-1 Vogtle Unit I r,0 Reactor Geometry with a 12.5' Neutron Pad Span at the Core Midplane WCAP- 17009-NP April 2009 Revision I

6-19 A. W. Vogfle Unrit 1 Reactor RT Model Moshe 153R. MO

-ý moon ii C-U1-U.-

a ui LU 2~M %R

-FlU Figure 6-2 Vogtle Unit 1 rO Reactor Geometry with a 20.00 Neutron Pad Span at the Core Midplane WCAP- 17009-NP April 2009 Revision I

6-20 A. W. Vogtle Unit I Reactor RT Model Meshes: 183R. 998 m --

--- I - W--

  • ,.M Rd.

U-

'I" 1-f.

Figure 6-3 Vogtle Unit 1 re Reactor Geometry with a 22.50 Neutron Pad Span at the Core Midplane WCAP- I 7009-NP April 2009 Revision I

6-21 A. W. Vogtle UnTI I Reactor RZ Model Meshes: 153U.1U8¥

[QN' Figure 6-4 Vogtle Unit I rz Reactor Geometry with Neutron Pad WCAP- 17009-NP April 2009 Revision 1

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following table summarizes the removal of the six surveillance capsules from the Vogtle Unit 1 reactor vessel, meeting the requirements of ASTM E185-82 [4].

Table 7-1 Surveillance Capsule Withdrawal Summary 2

Capsule Capsule Location Lead Factor(a) Withdrawal EFPY(b) Fluence (n/cm )(c)

U 58.50 4.14 1.14 3.32E+18 Y 240 3.97 4.85 1.14E+19 V 610 3.97 8.78 1.93E+19 X 238.50 4.19 14.33 3.47E+19 W 121.50 4.16 18.41 4.36E+19 Z 301.50 4.16 18.41 (d)

Notes:

(a) Updated in Capsule W dosimetry analysis; see Table 6-6.

(b) EFPY from plant startup.

(c) Updated in Capsule W dosimetry analysis; see Table 6-5.

(d) Standby Capsule Z was removed at 18.41 EFPY and placed in the spent fuel pool. No testing or analysis has been performed on Capsule Z. This capsule may be re-inserted in the future to fulfill the requirements of the "Additional 20-year License Renewal capsule".

WCAP-17009-NP April 2009 Revision 1

8-1 8 REFERENCES

1. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, 1988.
2. 10 CFR 50, Appendix GQ Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.
3. WCAP-1 1011, Revision 0, Georgia Power Company Alvin W Vogtle Unit No. 1 Reactor Vessel RadiationSurveillanceProgram,L. R. Singer, February 1986.
4. ASTM E185-82, Standard Practicefor Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM 1982.
5.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G Fracture Toughness Criteriafor ProtectionAgainst Failure.
6. ASTM E208, Standard Test Methodfor ConductingDrop- Weight Test to Determine Nil-Ductility Transition Temperature of FerriticSteels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.
7. ASTM E399, Test Method for Plane-StrainFracture Toughness of Metallic Materials,American Society for Testing Materials.
8. Westinghouse Science and Technology Department Procedure RMF 8402, Surveillance Capsule Testing Program,Revision 3, June 6, 2005.
9. Westinghouse Science and Technology Department Procedure RMF 8102, Tensile Testing, Revision 3, March 1, 1999.
10. Westinghouse Science and Technology Department Procedure RMF 8103, Charpy Impact Testing, Revision 2, August 1, 1998.
11. Westinghouse Science and Technology Department Procedure RMF 8804, Opening of Westinghouse Surveillance Capsules, Revision 2, August 1, 2004.
12. ASTM E23-06, Standard Test Method for Notched Bar Impact Testing of Metallic Materials, ASTM, 2006.
13. General Yielding of Charpy V-Notch and Precracked Charpy Specimens, Journal of Engineering Materials and Technology, Vol. 100, April 1978, pp. 183-188.
14. ASTM A370-07, Standard Test Methods and Definitions for Mechanical Testing of Steel Products,ASTM, 2007.

WCAP- 17009-NP April 2009 Revision 1

8-2

15. ASTM E8-04, StandardTest Methods for Tension Testing of Metallic Materials,ASTM, 2004.
16. ASTM E21-05, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials,ASTM, 2005.
17. WCAP-12256, Revision 0, Analysis of Capsule Ufrom the Georgia Power Company Vogtle Unit I Reactor Vessel RadiationSurveillanceProgram, S. E. Yanichko, et. al., May 1989.
18. WCAP-13931, Revision 1, Analysis of Capsule Y from the Georgia Power Company Vogtle Unit 1 Reactor Vessel RadiationSurveillance Program,M. J. Malone, et. al., August 1995.
19. WCAP-15067, Revision 0, Analysis of Capsule V from Southern Nuclear Vogtle Electric GeneratingPlant Unit 1 Reactor Vessel Radiation Surveillance Program, T. J. Laubham, et. al.,

September 1998.

20. WCAP-16278-NP, Revision 0, Analysis of Capsule X from the Southern Nuclear Operating Company, Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program, K. G. Knight, et. al.,

July 2004.

21. ASTM E853-01, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706 (IA), ASTM, 2001
22. ASTM E693-01, StandardPracticefor CharacterizingNeutron Exposures in Iron and Low Alloy Steels in Terms of DisplacementsPerAtom (DPA), E706 (ID), ASTM, 2001.
23. Regulatory Guide 1.190, Calculationaland Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.
24. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.
25. RSICC Computer Code Collection CCC-650, DOORS 3.2: One, Two- and Three Dimensional Discrete OrdinatesNeutron/Photon Transport Code System, April 1998.
26. RSICC Data Library Collection DLC-1 85, BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section LibraryDerivedfrom ENDF/B- Vfor L WR Shielding and Pressure Vessel Dosimetry Applications, March 1996.

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A-I APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS A.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Vogtle Unit 1 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [A-i]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report.

A.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the five neutron sensor sets analyzed to date as part of the Vogtle Unit 1 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Capsule ID Azimuthal Location Withdrawal Time Irradiation Time [EFPY]

U 31.5' Dual End of Cycle 1 1.14 Y 29.00 Dual End of Cycle 4 4.85 V 29.00 Dual End of Cycle 7 8.78 X 31.5' Dual End of Cycle 11 14.33 W 31.5' Single End of Cycle 14 18.41 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules U, Y, V, X, and W are summarized as follows:

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A-2 Since all of the dosimetry monitors located at the radial center of the material test specimen array, radial gradient corrections were not required for these reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table A-1.

The use of passive monitors such as those listed above does not yield a direct measure of the energy-dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

0 the measured specific activity of each monitor, 0 the physical characteristics of each monitor,

  • the operating history of the reactor,
  • the energy response of each monitor, and
  • the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules U, Y, V, and X are documented in References A-2 through A-5, respectively. The radiometric counting of the sensors from Capsule W was carried out by Pace Analytical Services, Inc. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules U, Y, V, X, and W was based on the monthly power generation of Vogtle Unit 1 from initial reactor criticality through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules U, Y, V, X, and W is given in Table A-2.

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A-3 Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:

A No F Y Z P1 Cj [1- e'ttj] [e-'"d]

Pref where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus).

A = Measured specific activity (dps/gm).

No = Number of target element atoms per gram of sensor.

F = Atom fraction of the target isotope in the target element.

Y = Number of product atoms produced per reaction.

PJ = Average core power level during irradiation period j (MW).

Pref = Maximum or reference power level of the reactor (MW).

Ci = Calculated ratio of ý(E > 1.0 MeV) during irradiation period j to the time weighted average +(E > 1.0 MeV) over the entire irradiation period.

X = Decay constant of the product isotope (1/sec).

tj = Length of irradiation period j (sec).

td = Decay time following irradiation period j (sec).

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many WCAP- 17009-NP April 2009 Revision 1

A-4 cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.

The fuel cycle specific neutron flux values along with the computed values for Cj are listed in Table A-3.

These flux values represent the cycle dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Vogtle Unit 1 fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V Capsule X Capsule W 235U Impurity/Pu Build-in 0.871 0.840 0.812 0.760 0.731 238U(y,f) 0.966 0.967 0.967 0.966 0.969 Net 238U Correction 0.842 0.813 0.785 0.735 0.708 237 Np(y,f) 0.990 0.990 0.990 0.990 0.991 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules U, Y, V, X, and W are given in Tables A-4a through A-4e. In Tables A-4a through A-4e, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma ray induced fission effects.

A.1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as ý(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, Ri +/- 6Ri = E (g'ig )((Pg, +/- 5 )

g relates a set of measured reaction rates, Ri, to a single neutron spectrum, 4 g, through the multigroup dosimeter reaction cross-section, trig, each with an uncertainty 6. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

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A-5 For the least squares evaluation of the Vogtle Unit I surveillance capsule dosimetry, the FERRET code

[A-6] was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (ý(E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.

The application of the least squares methodology requires the following input:

1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.
2. The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.
3. The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Vogtle Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [A-7]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (1iB)" [A-8].

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" [A-9].

The following provides a summary of the uncertainties associated with the least squares evaluation of the Vogtle Unit 1 surveillance capsule sensor sets.

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A-6 Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

Reaction Uncertainty 63 60 Cu(nCC) Co 5%

54 Fe(n,p)54Mn 5%

58 Ni(n,p) 58Co 5%

23 8 U(n,f) 13 7Cs 10%

2 37 Np(n,f) 137 Cs 10%

59 60 Co(n,7) Co 5%

These uncertainties are given at the Icy level.

Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the Vogtle Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

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A-7 Reaction Uncertainty 63 60 Cu(n,c() Co 4.08-4.16%

54 Fe(n,p) 54Mn 3.05-3.11%

58 Ni(n,p) 58Co 4.49-4.56%

238 U(n,f) 137 Cs 0.54-0.64%

237 Np(n,f) 137 Cs 10.32-10.97%

59 Co(n,7) 6°Co 0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Snectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

Mgg = R. + Rg

  • Rg *Pgg where R, specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg' specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Pgg,=[-O]gg,+Oe -H where 2

H - (g -g')

27,2 WCAP-17009-NP April 2009 Revision I

A-8 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 6 is 1.0 when g = g', and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Vogtle Unit 1 calculated spectra was as follows:

Flux Normalization Uncertainty (Re) 15%

Flux Group Uncertainties (Rg, Rg.)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 A.1.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Vogtle Unit 1 surveillance capsules withdrawn to date are provided in Tables A-5 and A-6. In Table A-5, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates.

These ratios of M/C and MIBE illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table A-6, comparison of the calculated and best estimate values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the BE/C ratios observed for each of the capsules.

The data comparisons provided in Tables A-5 and A-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the lcy level. From Table A-6, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6% for neutron flux (E > 1.0 MeV) and 7-8% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the IcY level.

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A-9 Further comparisons of the measurement results (from Tables A-5 and A-6) with calculations are given in Tables A-7 and A-8. These comparisons are given on two levels. In Table A-7, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table A-8, calculations of fast neutron exposure rates in terms of 4(E > 1.0 MeV) and dpa/s are compared with the best estimate results obtained from the least squares evaluation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.90 to 1.16 for the 25 samples included in the data set.

The overall average M/C ratio for the entire set of Vogtle Unit 1 data is 1.03 with an associated standard deviation of 7.1%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.98 to 1.06 for neutron flux (E > 1.0 MeV) and from 0.97 to 1.05 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 1.01 with a standard deviation of 3.5% and 1.01 with a standard deviation of 3.2%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.2 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region of the Vogtle Unit 1 reactor pressure vessel.

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A-10 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors Reaction of Target Atom 90% Response Product Fission Yield Monitor Material Interest Fraction Range (MeV)(a) Half-life (%)

63 Copper Cu (n,a) 0.6917 4.9- 11.9 5.271 y 54 Iron Fe (n,p) 0.0585 2.1 -8.5 312.1 d 5

Nickel 8Ni (n,p) 0.6808 1.5 - 8.3 70.82 d 238 Uranium-238 U (n,f) 1.0000 1.3 -6.9 30.07 y 6.02 23 7 Neptunium-237 Np (n,f) 1.0000 0.3 -3.8 30.07 y 6.17 59 Cobalt-Aluminum Co (n,y) 0.0015 non-threshold 5.271 y Note:

(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Vogtle Unit I surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

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A-lI Table A-2 Monthly Thermal Generation During the First Fourteen Fuel Cycles of the Vogtle Unit 1 Reactor (Reactor Power of 3411 MWt from Startup Through the End of Cycle 4, and 3565 MWt for Cycles 5 through 14)

Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Mar-87 68766 Jul-89 2443387 Sep-91 969976 Nov-93 2558233 Apr-87 797491 Aug-89 2286024 Oct-91 0 Dec-93 2646046 May-87 1044332 Sep-89 2450229 Nov-91 215953 Jan-94 2639758 Jun-87 759746 Oct-89 2142954 Dec-91 2466013 Feb-94 2156617 Jul-87 1835718 Nov-89 2391716 Jan-92 2534684 Mar-94 2581209 Aug-87 2509822 Dec-89 2535607 Feb-92 2371364 Apr-94 2557327 Sep-87 2452829 Jan-90 2374089 Mar-92 2528590 May-94 2554173 Oct-87 707673 Feb-90 1811171 Apr-92 2239948 Jun-94 2561379 Nov-87 1927388 Mar-90 0 May-92 1866712 Jul-94 2646904 Dec-87 2467702 Apr-90 591136 Jun-92 2452840 Aug-94 2448946 Jan-88 1365280 May-90 2311713 Jul-92 2534681 Sep-94 629927 Feb-88 1387377 Jun-90 2299026 Aug-92 2535008 Oct-94 1099701 Mar-88 2456340 Jul-90 2196834 Sep-92 2188889 Nov-94 2564465 Apr-88 1907244 Aug-90 2512580 Oct-92 2538900 Dec-94 2631652 May-88 2531355 Sep-90 2452206 Nov-92 2454211 Jan-95 2650377 Jun-88 2444967 Oct-90 2534258 Dec-92 2536190 Feb-95 2130621 Jul-88 2220349 Nov-90 2428733 Jan-93 2536730 Mar-95 2650182 Aug-88 2415264 Dec-90 1692955 Feb-93 2273143 Apr-95 2561802 Nov-88 52233 Jan-91 2534837 Mar-93 849752 May-95 2630821 Dec-88 2135007 Feb-91 2260779 Apr-93 166750 Jun-95 2564944 Jan-89 1771903 Mar-91 2495386 May-93 2401502 Jul-95 2381719 Feb-89 1905573 Apr-91 2449552 Jun-93 2564437 Aug-95 2650844 Mar-89 2533004 May-91 2533685 Jul-93 2499130 Sep-95 2519961 Apr-89 2380073 Jun-91 2449889 Aug-93 2645970 Oct-95 2651520 May-89 2264902 Jul-91 2534501 Sep-93 2560140 Nov-95 2564913 Jun-89 2452382 Aug-91 2483204 Oct-93 2649962 Dec-95 2650729 WCAP- 17009-NP April 2009 Revision 1

A-12 Table A-2 (Continued) Monthly Thermal Generation During the First Fourteen Fuel Cycles of the Vogtle Unit 1 Reactor (Reactor Power of 3411 MWt from Startup Through the End of Cycle 4, and 3565 MWt for Cycles 5 through 14)

Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Jan-96 2650608 Mar-98 2650194 May-00 2651160 Jul-02 2651322 Feb-96 2255312 Apr-98 2563238 Jun-00 2375260 Aug-02 2627559 Mar-96 130446 May-98 2547428 Jul-00 2649880 Sep-02 2565444 Apr-96 648324 Jun-98 2563238 Aug-00 2601536 Oct-02 2654808 May-96 2258085 Jul-98 2648218 Sep-00 1185996 Nov-02 2038792 Jun-96 1467397 Aug-98 2632408 Oct-00 1062698 Dec-02 2380764 Jul-96 2651000 Sep-98 2563238 Nov-00 2565553 Jan-03 2618718 Aug-96 2651013 Oct-98 2650194 Dec-00 2531649 Feb-03 2394679 Sep-96 2565318 Nov-98 2561262 Jan-01 2651403 Mar-03 2651255 Oct-96 2654401 Dec-98 2640313 Feb-01 2394810 Apr-03 2561256 Nov-96 2442399 Jan-99 2648218 Mar-01 2618784 May-03 2409539 Dec-96 2648271 Feb-99 2211460 Apr-O 1 2562381 Jun-03 2560407 Jan-97 2648498 Mar-99 299858 May-01 2650897 Jul-03 2646705 Feb-97 2393961 Apr-99 2560910 Jun-01 2565700 Aug-03 2647404 Mar-97 2392019 May-99 2651230 Jul-01 2650920 Sep-03 2279739 Apr-97 1086834 Jun-99 2565431 Aug-01 2297481 Oct-03 635553 May-97 2489873 Jul-99 2651031 Sep-01 2553201 Nov-03 2486092 Jun-97 2565296 Aug-99 2650940 Oct-01 2644259 Dec-03 2643852 Jul-97 2645858 Sep-99 2565539 Nov-01 2557060 Jan-04 2643067 Aug-97 2650538 Oct-99 2654884 Dec-01 2644057 Feb-04 2472553 Sep-97 503695 Nov-99 2565850 Jan-02 2646581 Mar-04 2297330 Oct-97 677865 Dec-99 2650654 Feb-02 2385611 Apr-04 2546135 Nov-97 2565214 Jan-00 2651323 Mar-02 400069 May-04 2639731 Dec-97 2652171 Feb-00 2479813 Apr-02 728341 Jun-04 2553984 Jan-98 2650194 Mar-00 2650657 May-02 2651604 Jul-04 2639535 Feb-98 2395254 Apr-00 2562272 Jun-02 2564080 Aug-04 2640124 WCAP- 17009-NP April 2009 Revision I

A-13 Table A-2 (Continued) Monthly Thermal Generation During the First Fourteen Fuel Cycles of the Vogtle Unit 1 Reactor (Reactor Power of 3411 MWt from Startup Through the End of Cycle 4, and 3565 MWt for Cycles 5 through 14)

Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Sep-04 2556142 Aug-08 2643398 Jul-08 2644378 Jun-08 2497557 Oct-04 2645029 Sep-08 2558930 Aug-08 2646338 Jul-08 2578206 Nov-04 2556338 Oct-08 2475834 Sep-08 1342669 Aug-08 2579187 Dec-04 2641693 Nov-08 2556971 Oct-08 45917 Sep-08 2421029 Jan-05 2445867 Dec-08 2644574 Nov-08 2502070 Oct-08 2578010 Feb-05 2382881 Jan-08 2644574 Dec-08 2600183 Nov-08 2496576 Mar-05 977170 Feb-08 2390974 Jan-08 2597436 Dec-08 2575459 Apr-08 1624099 Mar-08 2647318 Feb-08 2343717 Jan-08 2574281 May-08 2574609 Apr-08 1866136 Mar-08 2589195 Feb-08 2406901 Jun-08 2557755 May-08 2642811 Apr-08 2510116 Mar-08 1192854 Jul-08 2642811 Jun-08 2532669 May-08 2587036 WCAP- 17009-NP April 2009 Revision 1

A-14 Table A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Cycle q(E > 1.0 MeV) In/cm 2-sl Length Fuel Cycle IEFPS] Capsule U Capsule Y Capsule V Capsule X Capsule W 1 3.60E+07 9.20E+10 8.57E+10 8.57E+10 9.20E+10 9.10E+10 2 3.58E+07 7.44E+10 7.44E+10 7.97E+10 7.90E+10 3 4.19E+07 7.65E+10 7.65E+10 8.44E+10 8.36E+10 4 3.92E+07 6.13E+10 6.13E+10 6.64E+10 6.57E+10 5 4.19E+07 6.53E+10 7.07E+10 7.OOE+10 6 4.24E+07 6.62E+10 7.11E+10 7.04E+10 7 3.97E+07 5.94E+10 6.43E+10 6.36E+10 8 4.22E+07 6.80E+10 6.74E+10 9 4.61E+07 8.58E+10 8.52E+10 10 4.28E+07 8.37E+10 8.29E+10 11 4.41E+07 7.79E+10 7.72E+10 12 4.28E+07 7.15E+10 13 4.41E+07 7.23E+10 14 4.21E+07 7.09E+10 Average 9.20E+10 7.43E+10 6.95E+10 7.67E+10 7.50E+10 Cycle Cj Length Fuel Cycle [EFPSI Capsule U Capsule Y Capsule V Capsule X Capsule W 1 2.89E+07 1.000 1.154 1.233 1.199 1.214 2 2.43E+07 1.001 1.070 1.040 1.053 3 2.30E+07 1.029 1.100 1.101 1.115 4 3.70E+07 0.825 0.882 0.866 0.877 5 4.24E+07 0.938 0.922 0.934 6 4.14E+07 0.951 0.927 0.938 7 4.41E+07 0.854 0.838 0.849 8 4.37E+07 0.887 0.899 9 4.39E+07 1.119 1.136 10 4.01E+07 1.091 1.105 11 4.47E+07 1.016 1.029 12 4.28E+07 0.953 13 4.41E+07 0.964 14 4.21E+07 0.945 Average 1.000 1.000 1.000 1.000 1.000 WCAP-17009-NP April 2009 Revision 1

A-15 Table A-4a Measured Sensor Activities and Reaction Rates Surveillance Capsule U Measured Activity(a) Saturated Activity Adjusted Reaction Reaction Location (dps/g) (dps/g) Rate(d) (rps/atom) 63 60 CU (n,CC) Co Top 4.82E+04 3.68E+05 5.61E-17 Middle 4.38E+04 3.34E+05 5.10E-17 Bottom 4.44E+04 3.39E+05 5.17E- 17 Average 5.29E-17 54 5 Fe (np) 4Mn Top 1.49E+06 3.56E+06 5.64E- 15 Middle 1.34E+06 3.20E+06 5.07E-15 Bottom 1.36E+06 3.25E+06 5.14E- 15 Average 5.28E-15 5

8Ni (n,p) "Co Top 1.26E+07 5.40E+07 7.73E-15 Middle 1.16E+07 4.97E+07 7.12E-15 Bottom 1.17E+07 5.O0E+07 7.18E- 15 Average 7.34E-15 238 37 U (n,f) 1 Cs (Cd) Middle 1.29E+05 5.02E+06 3.30E-14 Including 131U, 239pu, and y fission corrections: 2.77E-14(b) 237 37 Np (n,f) 1 Cs (Cd) Middle 1.24E+06 4.82E+07 3.08E-13 Including y fission corrections: 3.05E-13(c) 59 60 Co (nY) Co Top 1.03E+07 7.86E+07 5.13E-12 Middle 1.01E+07 7.7 1E+07 5.03E- 12 Bottom 1.05E+07 8.O1E+07 5.23E-12 Average 5.13E-12 59 60 Co (nY) Co (Cd) Top 5.21E+06 3.98E+07 2.59E-12 Middle 5.46E+06 4.17E+07 2.72E- 12 Bottom 5.58E+06 4.26E+07 2.78E-12 Average 2.70E-12 Notes:

(a) Measured specific activities are indexed to a counting date of February 22, 1989.

(b) The average 238U (n,f) reaction rate of 2.77E-14 includes a correction factor of 0.871 to account for plutonium build-in and an additional factor of 0.966 to account for photo-fission effects in the sensor.

23 7 (c) The average Np (n,f) reaction rate of 3.05E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

(d) Reaction rates referenced to the Cycle I Rated Reactor Power of 3411 MWt.

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A-16 Table A-4b Measured Sensor Activities and Reaction Rates Surveillance Capsule Y Measured Activity(a) Saturated Activity Adjusted Reaction Reaction Location (dps/g) (dps/g) Rate (rps/atom)(d) 63 CU (n,a) 60Co Top 1.38E+05 3.3 1E+05 5.04E- 17 Middle 1.21E+05 2.90E+05 4.42E- 17 Bottom 1.23E+05 2.95E+05 4.50E-17 Average 4.65E-17 54 Fe (np) 54Mn Top 1.63E+06 2.90E+06 4.59E-15 Middle 1.47E+06 2.61E+06 4.14E- 15 Bottom 1.48E+06 2.63E+06 4.17E- 15 Average 4.30E-15 58 5 Ni (n,p) "Co Top 8.43E+06 4.54E+07 6.50E-15 Middle 7.75E+06 4.17E+07 5.97E-15 Bottom 7.63E+06 4.11E+07 5.88E-15 Average 6.12E-15 238 13 7 U (n,f) Cs (Cd) Middle 5.07E+05 4.91E+06 3.22E-14 Including 235 U, 2 39 Pu, and 7 fission corrections: 2 .6 2 E-14(b) 37 237Np (n,f) 1 Cs (Cd) Middle 3.38E+06 3.27E+07 2.09E-13 Including y fission corrections: 2.07E-13(c) 59 60 Co (n,7) Co Top 2.34E+07 5.61E+07 3.66E-12 Middle 2.35E+07 5.63E+07 3.67E-12 Bottom 2.34E+07 5.6 1E+07 3.66E-12 Average 3.66E-12 59 Co (n,7) 60Co (Cd) Top 1.20E+07 2.88E+07 1.96E-12 Middle 1.29E+07 3.09E+07 1.90E-12 Bottom 1.29E+07 3.09E+07 1.99E-12 Average 1.97E-12 Notes:

(a) Measured specific activities are indexed to a counting date of August 8, 1993.

(b) The average 238U (n,f) reaction rate of 2.62E-14 includes a correction factor of 0.840 to account for plutonium build-in and an additional factor of 0.967 to account for photo-fission effects in the sensor.

(c) The average 237Np (n,f) reaction rate of 2.07E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

(d) Reaction rates referenced to the Cycles 1-4 Average Rated Reactor Power of 3411 MWt.

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A-17 Table A-4c Measured Sensor Activities and Reaction Rates Surveillance Capsule V Measured Activity(a) Saturated Activity Adjusted Reaction Reaction Location (dps/g) (dps/g) Rate(d) (rps/atom) 63 60 Cu (n, a) CO Top 1.75E+05 3.08E+05 4.70E-17 Middle 1.55E+05 2.73E+05 4.16E-17 Bottom 1.55E+05 2.73E+05 4.16E- 17 Average 4.34E-17 54 Fe(n,p) 54Mn Top 1.35E+06 2.75E+06 4.36E-15 Middle 1.24E+06 2.53E+06 4.OOE-15 Bottom 1.23E+06 2.51E+06 3.97E-15 Average 4.11E-15 58 58 Ni (n,p) Co Top 4.20E+06 4.40E+07 6.29E-15 Middle 3.89E+06 4.07E+07 5.83E-15 Bottom 3.88E+06 4.06E+07 5.81E-15 Average 5.98E-15 37 238U (n,f) 1 Cs (Cd) Middle 8.45E+05 4.83E+06 3.17E-14 235 239 2.49E-14(b)

Including U, Pu, and y fission corrections:

237 37 Np (n,f) 1 Cs (Cd) Middle 6.27E+06 3.58E+07 2.28E-13 Including y fission corrections: 2.26E-13(c) 59 Co (n,7) 60 Co Top 2.88E+07 5.07E+07 3.31E-12 Middle 2.88E+07 5.07E+07 3.31E-12 Bottom 2.87E+07 5.05E+07 3.29E-12 Average 3.30E-12 59 60 Co (n,7) Co (Cd) Top 1.45E+07 2.55E+07 1.66E-12 Middle 1.50E+07 2.64E+07 1.72E-12 Bottom 1.53E+07 2.69E+07 1.76E-12 Average 1.71E-12 Notes:

(a) Measured specific activities are indexed to a counting date of April 13, 1998.

238 (b) The average U (n,f) reaction rate of 2.49E-14 includes a correction factor of 0.812 to account for plutonium build-in and an additional factor of 0.967 to account for photo-fission effects in the sensor.

(c) The average 237Np (n,f) reaction rate of 2.26E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

(d) Reaction rates referenced to the Cycles 1-7 Average Rated Reactor Power of 3480 MWt.

April 2009 WCAP- 17009-NP I 7009-NP April 2009 Revision I

A-18 Table A-4d Measured Sensor Activities and Reaction Rates Surveillance Capsule X Measured Activity(a) Saturated Activity Adjusted Reaction Reaction Location (dps/g) (dps/g) Rate(d) (rps/atom) 63 60 CU (n,'a) Co Top 2.52E+05 3.27E+05 4.98E-17 Middle 2.23E+05 2.89E+05 4.41E-17 Bottom 2.26E+05 2.93E+05 4.47E-17 Average 4.62E-17 54 Fe (np) 54Mn Top 2.41E+06 3.16E+06 5.01E-15 Middle 2.20E+06 2.89E+06 4.57E-15 Bottom 2.19E+06 2.87E+06 4.55E-15 Average 4.71E-15 5 5 8Ni (n,p) 8Co Top 1.55E+07 4.72E+07 6.75E-15 Middle 1.44E+07 4.38E+07 6.27E-15 Average 6.51E-15 238 37 U (n,f) 1 Cs (Cd) Middle 1.63E+06 5.94E+06 3.90E-14 Including 235 U, 239 pu, and y fission corrections: 2.86E-14(b) 237Np (n,f) 37 1 Cs (Cd) Middle 1.09E+07 [ 3.97E+07 2.53E-13 Including y fission corrections: 2.51E-13(c) 59 60 Co (n,y) Co Top 4.59E+07 5.95E+07 3.88E-12 Middle 4.68E+07 6.07E+07 3.96E-12 Bottom 4.76E+07 6.17E+07 4.03E-12 Average 3.95E-12 59 60 Co (n,y) Co (Cd) Top 2.39E+07 3.IOE+07 2.02E-12 Middle 2.56E+07 3.32E+07 2.17E-12 Bottom 2.60E+07 3.37E+07 2.20E-12 Average 2.13E-12 Notes:

(a) Measured specific activities are indexed to a counting date of January 20, 2004.

238 (b) The average U (n,f) reaction rate of 2.86E-14 includes a correction factor of 0.760 to account for plutonium build-in and an additional factor of 0.966 to account for photo-fission effects in the sensor.

(c) The average 237 Np (n,f) reaction rate of 2.511E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

(d) Reaction rates referenced to the Cycles 1-11 Average Rated Reactor Power of 3513 MWt.

WCAP- 17009-NP April 2009 Revision I

A-19 Table A-4e Measured Sensor Activities and Reaction Rates Surveillance Capsule W Measured Activity(') Saturated Activity Adjusted Reaction Reaction Location (dps/g) (dps/g) Rate(d) (rps/atom) 63 60 Cu (n,aX) Co Top 2.50E+05 3.18E+05 4.85E-17 Middle 2.23E+05 2.84E+05 4.33E-17 Bottom 2.24E+05 2.85E+05 4.35E-17 Average 4.51E-17 54 Fe (n,p) 54Mn Top 1.69E+06 2.93E+06 4.65E-15 Middle 1.53E+06 2.66E+06 4.21E-1 5 Bottom 1.53E+06 2.66E+06 4.21E- 15 Average 4.36E-15 5

SNi (n,p) 5SCo Top 5.96E+06 4.59E+06 6.57E-15 Middle 5.48E+06 4.22E+06 6.04E- 15 Bottom 5.52E+06 4.25E+06 6.08E-15 Average 6.23E-15 37 238U (n,f) 1 Cs (Cd) Middle 1.86E+06 5.57E+06 3.65E-14 Including 231U, 2 39 pu, and y fission corrections: 2.59E-14(b) 23 7 37 Np (n,f) 1 Cs (Cd) Middle 1 1.15E+07 I 3.44E+07 2.20E-13 Including y fission corrections: 2.18E-13(c)

'9Co (n,7,) 60 Co Top 4.27E+07 5.43E+07 3.54E-12 Middle 4.23E+07 5.38E+07 3.5 1E-12 Bottom 4.3 1E+07 5.48E+07 3.58E-12 Average 3.54E-12 9 60

' Co (n,7) Co (Cd) Top 2.10E+07 2.67E+07 1.74E-12 Middle 2.18E+07 2.77E+07 1.81E-12 Bottom 2.24E+07 2.85E+07 1.86E-12 Average 1.80E-12 Notes:

I. Measured specific activities are indexed to a counting date of October 01, 2008.

2. The average 238U (n,f) reaction rate of 2.59E-14 includes a correction factor of 0.731 to account for plutonium build-in and an additional factor of 0.969 to account for photo-fission effects in the sensor.
3. The average 237Np (n,f) reaction rate of 2.18E-1 3 includes a correction factor of 0.991 to account for photo-fission effects in the sensor.
4. Reaction rates referenced to the Cycles 1-14 Average Rated Reactor Power of 3524 MWt.

WCAP-17009-NP April 2009 Revision 1

A-20 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance Capsule Center Capsule U Reaction Rate Irps/atoml Reaction Measured Calculated Best Estimate M/C M/BE 63 Cu(n,Q) 60Co 5.29E- 17 4.78E-17 5.1IE-17 1.11 1.04 54 4 Fe(n,p)1 Mn 5.28E-15 5.40E-15 5.37E-15 0.98 0.98 58 Ni(n,p)5 SCo 7.34E- 15 7.58E-15 7.49E-15 0.97 0.98 238 137 U(n,f) Cs (Cd) 2.77E-14 2.92E-14 2.86E-14 0.95 0.97 237 37 Np(nf)1 Cs (Cd) 3.05E-13 2.88E-13 2.93E-13 1.06 1.04 59 Co(n,7) 60Co 5.13E-12 4.12E- 12 5.03E-12 1.25 1.02 59 Co(n,7) 60Co (Cd) 2.70E-12 2.87E-12 2.74E-12 0.94 0.99 Note:

See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

Capsule Y Reaction Rate Irps/atom]

Reaction Measured Calculated Best Estimate M/C M/BE 63 0 Cu(n,c)6 Co 4.65E-17 4.05E-17 4.45E-17 1.15 1.04 54 54 Fe(n,p) Mn 4.30E-15 4.45E-15 4.48E-15 0.97 0.96 58 Ni(n,p)58Co 6.11 E-15 6.24E-15 6.25E-15 0.98 0.98 238 U(n,f)137Cs (Cd) 2.62E-14 2.38E-14 2.35E-14 1.10 1.11 237 37 Np(n,f)1 Cs (Cd) 2.07E-13 2.31E-13 2.16E-13 0.90 0.96 59 Co(n,7) 60Co 3.66E-12 3.23E-12 3.60E-12 1.13 1.02 59 Co(n,7) 60Co (Cd) 1.97E-12 2.26E-12 2.OOE-12 0.87 0.99 Note:

See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

WCAP- 17009-NP April 2009 Revision I

A-21 Table A-5 (Continued) Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance Capsule Center Capsule V Reaction Rate Irps/atoml Reaction Measured Calculated Best Estimate M/C M/BE 63 60 Cu(n, t) Co 4.34E-17 3.86E-17 4.19E-17 1.12 1.04 54 54 Fe(n,p) Mn 4.11 E-15 4.20E-15 4.30E-15 0.98 0.96 5 1.02 0.99 8Ni(n,p)s8Co 5.98E-15 5.88E-15 6.05E-15 238 37 U(n,f)1 Cs (Cd) 2.49E-14 2.23E-14 2.29E-14 1.12 1.09 237 37 Np(n,f)1 Cs (Cd) 2.26E-13 2.16E-13 2.24E-13 1.05 1.01 59 Co(n,y) 60Co 3.30E-12 3.OOE-12 3.25E-12 1.10 1.02 59 6 Co(n,7) °Co (Cd) 1.71E-12 2.10E-12 1.75E-12 0.81 0.98 Note:

See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

Capsule X Reaction Rate [rps/atoml Reaction Measured Calculated Best Estimate M/C M/BE 63 Cu(n, a) 6°Co 4.62E-17 4.22E- 17 4.52E-17 1.09 1.02 54 54 Fe(n,p) Mn 4.71E-15 4.62E-15 4.81E-15 1.02 0.98 58 Ni(n,p)5 8Co 6.51E-15 6.47E-15 6.69E-15 1.01 0.97 23 37 8U(n,f)1 Cs (Cd) 2.86E-14 2.46E-14 2.58E-14 1.16 1.11 237 137 Np(n,f) Cs (Cd) 2.51E-13 2.38E-13 2.51E-13 1.05 1.00 59 60 Co(n,7) Co 3.95E-12 3.36E-12 3.89E- 12 1.18 1.02 59 60 Co(n,y) Co (Cd) 2.13E-12 2.33E-12 2.16E-12 0.91 0.99 Note:

See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

WCAP- 17009-NP April 2009 Revision 1

A-22 Table A-5 (Continued) Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance Capsule Center Capsule W Reaction Rate Irps/atoml Reaction Measured Calculated Best Estimate M/C M/BE 63Cu(n,ct) 6 °Co 4.51E-17 4.13E-17 4.38E-17 1.09 1.03 54 Fe(n,p)54 Mn 4.36E-15 4.51E-15 4.47E-15 0.97 0.98 58 58 Ni(np) Co 6.23E-15 6.31E-15 6.27E-15 0.99 0.99 238 U(n,f)137Cs (Cd) 2.59E-14 2.40E-14 2.38E-14 1.08 1.09 237 137 Np(nf) Cs (Cd) 2.17E-13 2.33E-13 2.24E-13 0.93 0.97 59 Co(n,,) 6°Co 3.54E-12 2.98E-12 3.47E- 12 1.19 1.02 59 Co(n,7) 60Co (Cd) 1.80E-12 2.11E-12 1.84E-12 0.85 0.98 Note:

See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

WCAP-17009-NP April 2009 Revision 1

A-23 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates at the Surveillance Capsule Center (p(E > 1.0 MeV) In/cm 2-sI Capsule ID Calculated Best Estimate Uncertainty (Ia) BE/C U 9.20E+10 9.04E+10 6% 0.98 Y 7.43E+10 7.29E+10 6% 0.98 V 6.95E+10 7.19E+ 10 6% 1.03 X 7.67E+10 8.13E+10 6% 1.06 W 7.50E+10 7.45E+10 6% 0.99 Note:

Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule.

See Section A. 1.2 for details describing the Best Estimate (BE) exposure rates.

Iron Atom Displacement Rate Idpa/s]

Capsule ID Calculated Best Estimate Uncertainty (Ic) BE/C U 1.80E-10 1.78E-10 8% 0.99 Y 1.45E-10 1.41E-10 7% 0.97 V 1.35E-10 1.39E-10 8% 1.03 X 1.49E-10 1.56E-10 7% 1.05 W 1.45E-10 1.43E-10 7% 0.98 Note:

Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule.

See Section A. 1.2 for details describing the Best Estimate (BE) exposure rates.

WCAP-17009-NP April 2009 Revision 1

A-24 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule U Capsule Y Capsule V Capsule X Capsule W 63 Cu(n,a) 60Co 1.09 1.11 1.15 1.12 1.09 54 Fe(n,p) 54Mn 0.97 0.98 0.97 0.98 1.02 58 Ni(n,p) 58Co 0.99 0.97 0.98 1.02 1.01 238 U(n,f)137Cs (Cd) 1.08 0.95 1.10 1.12 1.16 237 Np(n,f) 137Cs (Cd) 0.93 1.06 0.90 1.05 1.05 Average 1.01 1.02 1.06 1.07 1.01

% Standard Deviation 6.7 10.2 6.0 5.9 7.0 Note:

The overall average M/C ratio for the set of 25 sensor measurements is 1.03 with an associated standard deviation of 7.1%.

Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID O(E > 1.0 MeV) dpa/s U 0.98 0.99 Y 0.98 0.97 V 1.03 1.03 X 1.06 1.05 W 0.99 0.98 Average 1.01 1.01

% Standard Deviation 3.5 3.2 WCAP-17009-NP April 2009 Revision I

A-25 A.2 REFERENCES A-1 Regulatory Guide RG- 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

A-2 WCAP-12256, Analysis of Capsule U from the Georgia Power Company Vogtle Unit 1 Reactor Vessel RadiationSurveillance Program,May 1989.

A-3 WCAP-13931, Revision 1, Analysis of Capsule Yfrom the Georgia Power Company Vogtle Unit I Reactor Vessel Radiation Surveillance Program,August 1995.

A-4 WCAP-15067, Analysis of Capsule V from Southern Nuclear Vogtle Electric GeneratingPlant Unit 1 Reactor Vessel Radiation Surveillance Program, September 1998.

A-5 WCAP-16278, Analysis of Capsule Xfrom the Southern Nuclear Operating Company, Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program, July 2004.

A-6 A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-7 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.

A-8 ASTM Standard El018, Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (JIB).

A-9 ASTM Standard E944, Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance.

WCAP- 17009-NP April 2009 Revision 1

B-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS

" Specimen prefix "AL" denotes Intermediate Shell Plate B8805-3, Longitudinal Orientation

  • Specimen prefix "AT" denotes Intermediate Shell Plate B8805-3, Transverse Orientation
  • Specimen prefix "AW" denotes Surveillance Program Weld Metal
  • Specimen prefix "AH" denotes Heat-Affected Zone Material April 2009 WCAP- 17009-NP I17009-NP April 2009 Revision I

B-2

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13C IM.0  ?.11 300 too -.01. ".0D AL42, 500F WCAP- 17009-NP April 2009 Revision 1

B-3 1C00600t .. ...

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zoo 3.90 too0

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B4 I

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B-5

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,AL41, 200-F WCAP-17009-NP April 2009 Revision 1

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AL4O, 350 ,OF WCAP- 17009-NP April 2009 Revision 1

B-7 2000.00

.... ....... ...... .......A ICOGGO . .........

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AT36, 100OF WCAP-17009-NP April 2009 Revision 1

B-8 z1xoom I M,0ra 10)2 n

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AT43, 125, WCAP-17009-NP April 2009 Revision 1

B-9 UI cr I ZE .03SI

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AT32, 200-F WCAP- 17009-NP April 2009 Revision I

B-11 20DW.00 . . .

000 190 ZOO I.O 4.00 5j" OD1.00 2.0030 5010 E.00 AAT37, 32.50 F WCAP-17009-NP April 2009 Revision 1

B- 12

~ ~ ~

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AAV37, -20TF WCAP-17009-NP April 2009 Revision 1

B-13 MUD2 I0.~00 N Al a P - R 0.00 1.00 too 0o0 t o06 Co.

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B-14 Tff.~- (m)

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AW 36, 0-F WCAP-17009-NP April 2009 Revision I

B-15

  • c'O.D*D ...... . ........... ...... .......

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a nh.II A

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(ýM)

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B-16 40 ,00 ......... ........ ...... ................. ....

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AW44, 175'F WCAP- 17009-NP April 2009 Revision I

B- 17 001 ...

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B- 18 I ONAc ................

.. L.. .. .........

A Fn 4- I A kU IN.& A, I.0o 240 3.0V 4.o0 !M &00rI r,-. (m)

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B- 19 003.0 MUM 33 ZionOn~

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B-20 5 . r, x0 .[* ............. ...................... - ................... . ......... . . . .........

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AH38, 25°F 0,.00 1.05, 2.*06 AB,10* E.00 S.M.

rrr.-1 pfm AH35, 50°F WCAP- 17009-NP April 2009 Revision 1

B-21 Turn-t Orm)

AH31, 150-F

- WO000 An 800O 1.00 2.00 18O0 , 0.0'0 0,.

AH34. 200 0 F WCAP- 17009-NP April 2009 Revision 1

C-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Contained in Table C-I are the upper shelf energy (USE) values used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 5.3. The definition for USE is given in ASTM E185-82

[C-1], Section 4.18, and reads as follows:

"upper shelf energy level - the average energy value for all Charpy specimens (normally three) whose test temperature is above the upper end of the transition region. For specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the upper shelf energy."

If there are specimens tested in sets of three at each temperature, Westinghouse reports the set having the highest average energy as the USE (usually unirradiated material). If the specimens were not tested in sets of three at each temperature, Westinghouse reports the average of all Charpy data (> 95% shear) as the USE, excluding any values that are deemed outliers using engineering judgement. Hence, the USE values reported in Table C-i, which were used to generate the Charpy V-notch curves, were determined utilizing this methodology.

The lower shelf energy values were fixed at 2.2 ft-lb for all cases.

Table C-1 Upper Shelf Energy Values (ft-lb) Fixed in CVGRAPH Capsule Material Unirradiated U Y V X W Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

Intermediate Shell Plate B8805-3 (Transverse Orientation)

Weld Metal (Heat # 83653) 145 153 139 138 138 125 HAZ Material 134 129 124 121 122 114 CVGRAPH Version 5.3 plots of all surveillance data are provided in this appendix, on the pages following the reference list.

C.1 REFERENCES C-i ASTM El 85-82, StandardPracticefor ConductingSurveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.

WCAP-17009-NP April 2009 Revision 1

C-2 Unirradiated Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:16 AM Page 1 Coefficients of Curve 1 A = 62.1 B = 59.9 C = 93.5 TO = 41.02 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=122.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-14.9 Deg F Temp@50 ft-lbs=21.9 Deg F Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: Unirra Fluence: n/cmA2 300 250 200 150 o0 0 0o 100 0

08 50 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 40. 00 11. 00 20. 19 -9. 19

-40. 00 19.00 20. 19 -1. 19

-40, 00 9. 00 20. 19 -1 . 19

- 20. 00 54. 00 27. 75 26. 25

- 20. 00 28. 00 27. 75 . 25

-20. 00 12.00 27. 75 - 5 75

.00 52. 00 37. 39 14. 61

.00 47. 00 37. 39 9.61

.00 42. 00 37. 39 4.61 WCAP-17009-NP April 2009 Revision 1

C-3 Unirradiated Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

40. 00 48. 00 61. 45 -13.45
40. 00 62. 00 61. 45 .55
40. 00 60. 00 61. 45 -1.45
80. 00 93. 00 85. 72 7.28
80. 00 64. 00 85. 72 -21.72
80. 00 70. 00 85. 72 -15. 72 100. 00 84. 00 95. 56 -11.56 100. 00 107. 00 95. 56 11. 44 100. 00 110.00 95. 56 14. 44 120. 00 100. 00 103. 33 -3. 33 120. 00 116.00 103. 33 12.67 120. 00 109.00 103. 33 5. 67 180. 00 126. 00 116. 17 9. 83 180. 00 1.15. 00 116. 17 -1.17 180. 00 116. 00 116. 17 -. 17 260. 00 129.00 120. 90 8.10 260. 00 121.00 120. 90 .10 320. 00 131. 00 121. 69 9.31 320. 00 119.00 121. 69 -2. 69 Correlation Coefficient = .962 April 2009 WCAP- 17009-NP I7009-NP April 2009 Revision I

C-4 Capsule U Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:18 AM Page 1 Coefficients of Curve 1 A = 68.1 B = 65.9 C = 101.88 TO = 65.53 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 134.0(Fixed) Lower Shelf Energy=2.2(Fixedl)

Temp@30 ft-lbs=-1.6 Deg F Ternp@50 ft-lbs=36.9 Deg F Plant: Vogtle I Material: SA533BI I-feat: C0623-1 Orientation: LT Capsule: U Fluence: n/cmA2 300 250 A200 0

IL Lm U- 150 0

100 50 0 f

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-75. 00 11.00 10.06 94

-50. 00 18.00 14. 57 3. 43

-25. 00 14. 00 21.27 -7. 27

-10. 00 25. 00 26. 59 -1 59

.00 23. 00 30. 73 -7. 73

.00 38. 00 30. 73 7. 27

10. 00 29. 00 35. 36 -6. 36
25. 00 44, 00 43. 19 81
50. 00 58. 00 58. 13 -13 WCAP- 17009-NP April 2009 Revision 1

C-5 Capsule U Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: U Fluence: n/cm^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

72. 00 88. 00 72. 28 15. 72 150. 00 97. 00 112.91 -15. 91 200. 00 125.00 125. 22 - . 22 250. 00 136.00 130. 57 5. 43 350. 00 143.00 133.51 9. 49 400. 00 130. 00 133.81 -3. 81 Correlation Coefficient = .987 WCAP-17009-NP April 2009 Revision 1

C-6 Capsule Y Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:19 AM Page 1 Coefficients of Curve 1 A = 67.1 B = 64.9 C = 111.24 TO = 89.72 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1 32.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 ft-lbs=17.5 Deg F Temp@50 ft-lbs=59.8 Deg F Plant: Vogtle I Material: SA533B 1 Heat: C0623-1 Orientation: LT Capsule: Y Fluence: n/cm^2 300 250 S200 0

U-LM150 S

z8100 0 0

00/

50 0

0 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 25. 00 6. 00 16. 84 -10. 84 00 23. 00 23. 77 - .77

25. 00 17. 00 33. 09 -16. 09
25. 00 23. 00 33.09 -10. 09
35. 00 53. 00 37. 52 15.48
50. 00 57. 00 44. 86 12. 14
60. 00 41. 00 50. 16 -9.16
75. 00 82. 00 58.56 23.44 100. 00 74. 00 73.08 .92 WCAP- 17009-NP April 2009 Revision 1

C-7 Capsule Y Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: Y Fluence: n/cmA^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 125. 00 79. 00 87.02 -8. 02 150. 00 96. 00 99. 18 -3. 18 t75. 00 94. 00 108. 96 -14. 96 225. 00 132. 00 121.52 10. 48 275. 00 130. 00 127. 52 2. 48 300. 00 133. 00 129. 10 3. 90 Correlation Coefficient = .962 WCAP- 17009-NP April 2009 Revision 1

C-8 Capsule V Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:20 AM Page 1 Coefficients of Curve I A = 60.1 B = 57.9 C = 133.03 TO = 104.09 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1 18.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=27.5 Deg F Temp@50 ft-lbs=80.7 Deg F Plant: Vogtle I Material: SA533B 1 Heat: C0623-1 Orientation: LT Capsule: V Fluence: n/cmA2 300 250 j200 0

Lm 150 100 50 0 n

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-80. 00 2. 00 9. 04 -7. 04

-50. 00 17. 00 12.59 4. 41

-25. 00 28. 00 16. 74 11 .26

-15.00 10. 00 18. 76 -8. 76

. 00 9. 00 22. 23 -13. 23

5. 00 24. 00 23. 50 .50
20. 00 34. 00 27. 71 6. 29
50. 00 49. 00 37 .78 11.22 100. 00 59. 00 58. 32 .68 WCAP-17009-NP April 2009 Revision I

C-9 Capsule V Ilitermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 150.00 65. 00 79. 33 -14. 33 175. 00 90. 00 88. 34 1.66 200. 00 95. 00 95. 86 - . 86 250. 00 112.00 106. 38 5. 62 300. 00 123. 00 112.22 10.78 375. 00 118.00 116. 06 1.94 Correlation Coefficient = .981 WCAP-17009-NP April 2009 Revision 1

C-10 Capsule X Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:20 AM Page 1 Coefficients of Curve 1 A = 56.1 B = 53.9 C = 97.09 TO = 132.92 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 1I0.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=81.7 Deg F Temnp@50 ft-lbs=121.9 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: X Fluence: n/cmA2 300 250 j200

,LL Lm 150 0 o 100 0 0/

50 1 0 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 25. 00 3.00 6.21 -3.21

25. 00 8. 00 12. 73 -4.73
50. 00 30. 00 18. 74 11.26
50. 00 13. 00 18. 74 -5. 74
75. 00 35. 00 27. 28 7.72 100. 00 42. 00 38. 49 3. 5t 125. 00 59. 00 51. 71 7. 29 150. 00 48. 00 65. 48 -17.48 160. 00 66. 00 70. 75 -4.75 WCAP-17009-NP April 2009 Revision 1

C-11 Capsule X Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: X Fluence: n/cmA^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 175.00 75. 00 78. 10 -3.10 200. 00 75. 00 88.36 -13. 36 225. 00 117. 00 95.93 21. 07 225. 00 105. 00 95.93 9. 07 250. 00 103.00 101. 13 1.87 275.00 1 13. 00 104. 52 8.48 Correlation Coefficient = .966 WCAP-17009-NP April 2009 Revision 1

C-12 Capsule W Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/09/2008 09:56 AM Page 1 Coefficients of Curve 1 A = 51.1 B = 48.9 C = 117.67 TO = 137.5 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 100.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=83.2 Deg F Temp@50 ft-lbs=134.9 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: W Fluence: n/cm^2 300 250

- 200 0

0%

150 0

100 0

50 0 0 0 0

0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-30. 00 6. 00 7. 56 -1.56 00 25. 00 10. 82 14. 18

50. 00 24. 00 20. 23 3. 77
60. 00 20. 00 22. 86 -2. 86
75. 00 36. 00 27. 32 8.68 100. 00 45. 00 36. 02 8.98 125. 00 35. 00 45. 92 -10.92 t50. 00 48. 00 56. 27 -8. 27 170. 00 45. 00 64. 27 -19.27 WCAP- 17009-NP April 2009 Revision I

C- 13 Capsule W Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: W Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 180.00 55. 00 68. 03 -13. 03 200. 00 96. 00 74. 88 21. 12 200. 00 86. 00 74. 88 11. 12 300. 00 98. 00 94. 19 3. 81 325. 00 108.00 96. 12 1 1.88 350. 00 114.00 97 .43 16. 57 Correlation Coefficient = .942 WCAP-17009-NP April 2009 Revision 1

C-14 Unirradiated Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:31 PM Page 1 Coefficients of Curve 1 A = 43.68 B = 43.68 C = 101.09 TO = 39.13 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E,=87.4 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=18.8 Deg F Plant: Vogtle 1 Material: SA533B I Heat: C0623-1 Orientation: LT Capsule: Unirra Fluence: n/cm^2 200 150 Em

._o B. o00 0

15 50

ý/

0 0 00 nI U

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-40. 00 7.00 15. 10 -8.10

-40. 00 12.00 15. 10 -3.10

-40. 00 6. 00 15. 10 -9. 10

- 20. 00 38. 00 20. 69 17. 31

-20. 00 22. 00 20. 69 1.31

-20. 00 8. 00 20. 69 -12.69 00 39. 00 27. 57 11.43 00 34. 00 27. 57 6.43 00 33. 00 27. 57 5. 43 WCAP-17009-NP April 2009 Revision I

C-15 Unirradiated Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature hiput L.E. Computed L.E. Differential

40. 00 36. 00 44. 05 -8.05
40. 00 43. 00 44. 05 -1.05
40. 00 44. 00 44, 05 -. 05
80. 00 66. 00 60. 43 5.57
80. 00 45. 00 60. 43 -15.43
80. 00 55. 00 60. 43 -5.43 100. 00 62. 00 67. 20 -5.20 100. 00 69. 00 67. 20 1.80 100.00 77. 00 67. 20 9. 80 120. 00 72. 00 72 68 - .68 120.00 8 1.00 72. 68 8.32 120. 00 74. 00 72. 68 1.32 180. 00 84. 00 82. 29 1.71 180. 00 82. 00 82. 29 - . 29 180.00 80. 00 82. 29 -2. 29 260. 00 92. 00 86. 27 5.73 260. 00 83. 00 86. 27 -3.27 320. 00 85. 00 87. 02 -2.02 320. 00 85. 00 87. 02 -2. 02 Correlation Coefficient = .964 WCAP-17009-NP April 2009 Revision I

C- 16 Capsule U Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:38 PM Page 1 Coefficients of Curve I A = 40.96 B = 40.96 C = 91.86 TO = 46.06 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=81.9 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=32.6 Deg F Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: U Fluence: n/cmA2 200 150 C

.100 50 0 4--

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-75. 00 7. 00 5. 48 I. 52

- 50. 00 12.00 9. 01 2. 99

- 25. 00 9. 00 14. 38 -5. 38

- 10. 00 17. 00 18. 67 -1. 67

.00 23. 00 21. 99 1. 01 00 28. 00 21. 99 6. 01

10. 00 22. 00 25. 66 -3. 66
25. 00 32. 00 31. 73 27
50. 00 41. 00 42. 72 -1. 72 WCAP-17009-NP April 2009 Revision 1

C-17 Capsule U Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Jnput L.E. Computed L.E. Differential

72. 00 55. 00 52. 23 2.77 150. 00 70. 00 74. 20 4. 20 200. 00 85. 00 79. 15 5. 85 250. 00 80. 00 80.97 -. 97 350. 00 81.00 81.81 -. 81 400. 00 81.00 81.89 -. 89 Correlation Coefficient = .993 April 2009 WCAP- 17009-NP WCAP- I7009-NP April 2009 Revision I

C-18 Capsule Y Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:35 PM Page 1 Coefficients of Curve 1 A = 41.14 B = 41.14 C = 104.22 TO = 67.58 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=82.3 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=52.0 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: Y Fluence: n/cm^2 200 150 0C 0

Ialo100 R

50 0 - i i

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 25. 00 5. 00 11.91 -6. 91

. 00 18. 00 17. 66 34

25. 00 17. 00 25. 21 -8. 21
25. 00 20. 00 25. 21 -5. 21
35. 00 39. 00 28. 68 10. 32
50. 00 44. 00 34. 26 9. 74
60. 00 31. 00 38. 15 -7. 15
75. 00 51. 00 44. 06 6. 94 100. 00 56. 00 53. 54 2. 46 WCAP-17009-NP April 2009 Revision 1

C- 19 Capsule Y Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: Y Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.. Differential 125. 00 60. 00 61.76 -1.76 150. 00 62. 00 68.24 -6. 24 175.00 69. 00 72.98 -3. 98 225. 00 78.00 78.45 -. 45 275. 00 80. 00 80.76 -. 76 300. 00 88. 00 81.33 6. 67 Correlation Coefficient = .970 WCAP- 17009-NP April 2009 Revision 1

C-20 Capsule V Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:42 PM Page 1 Coefficients of Curve 1 A = 38.44 B = 38.44 C = 121.17 TO = 98.66 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=76.9 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=87.8 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: V Fluence: n/cm^2 200 150

.2 I, 100 s

50 0 i - "

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 80. 00 1.00 3. 83 -2.83

-50. 00 6. 00 6. 09 - . 09

-25. 00 16.00 8. 84 7.16

- 15. 00 4. 00 10.21 -6.21 00 4. 00 12.61 -8. 61

5. 00 13.00 13.51 -. 51
20. 00 22. 00 16. 49 5.51
50. 00 31. 00 23.79 7. 21 100. 00 38.00 38. 87 - .87 WCAP-17009-NP April 2009 Revision 1

C-21 Capsule V Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 150.00 45. 00 53. 82 -8. 82 175.00 61.00 59. 90 1.10 200. 00 69. 00 64. 73 4.27 250. 00 74. 00 71.04 2.96 300. 00 76. 00 74. 21 1.79 375. 00 72. 00 76. 09 -4.09 Correlation Coefficient = .983 WCAP- 17009-NP April 2009 Revision 1

C-22 Capsule X Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:46 PM Page 1 Coefficients of Curve I A = 38.24 B = 38.24 C = 104.72 TO = 139.31 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=76.5 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=130.5 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: X Fluence: n/cm^2 200 150

.2 0C 2L 100 5l 50 0 0___________

n

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-25. 00 00 3.18 3.18

25. 00 3 00 7.75 -4. 75
50. 00 14 00 11.76 2. 24
50. 00 7 00 11.76 -4. 76
75. 00 21 00 17. 32 3. 68 100. 00 27. 00 24. 52 2.48 125.00 40. 00 33.05 6. 95
50. 00 36. 00 42. 13 -6. 13 160. 00 45. 00 45. 70 - .70 WCAP- 17009-NP April 2009 Revision I

C-23 Capsule X Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: X Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 175. 00 50. 00 50. 79 - . 79 200. 00 51.00 58. 21 -7.21 225. 00 70. 00 64. 02 5.98 225. 00 68. 00 64. 02 3.98 250. 00 67. 00 68. 24 -1.24 275. 00 70. 00 71. 15 -1.15 Correlation Coefficient = .985 WCAP-17009-NP April 2009 Revision 1

C-24 Capsule W Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:52 PM Page 1 Coefficients of Curve I A = 45.33 B = 45.33 C = 172.61 TO = 154.82 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=90.7 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 rnils=114.8 Deg F Plant: Vogtle I Material: SA533B 1 Heat: C0623-1 Orientation: LT Capsule: W Fluence: n/cmA2 200 150 E,

C C

& 100 so 50 0 1-

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 30. 00 5. 00 9.53 -4. 53

. 00 22. 00 12. 93 9. 07

50. 00 22. 00 20. 75 1 .25
60. 00 17. 00 22. 66 -5. 66
75. 00 31.00 25. 74 5. 26 100. 00 33. 00 31.40 1 .60 125. 00 30. 00 37. 57 -7. 57 150. 00 44. 00 44. 06 - . 06 170. 00 38.00 49. 30 -11 30 WCAP- 17009-NP April 2009 Revision 1

C-25 Capsule W Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: W Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 180. 00 50. 00 51.89 -1. 89 200. 00 71.00 56. 93 14. 07 200. 00 62. 00 56.93 5.07 300. 00 69. 00 76. 44 -7.44 325. 00 83. 00 79.58 3. 42 350. 00 82. 00 82. 10 - .10 Correlation Coefficient = .961 WCAP- 17009-NP April 2009 Revision 1

C-26 Unirradiated Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:32 PM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 91.96 TO = 64.37 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+/-DT))]

Termperature at 50% Shear = 64.4 Plant: Vogtle 1 Material: SA533B 1 Heat: C0623-1 Orientation: LT Capsule: Unirra Fluence: n/cn^2 125 100 i

U, 75 0

2

a. 50 25 0 4-i - --- !1-1 --

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-40.00 5. 00 9. 36 -4. 36

-40.00 14. 00 9. 36 4. 64

-40.00 9.00 9. 36 - .36

-20. 00 27. 00 13. 76 13. 24

-20. 00 14. 00 13. 76 . 24

-20. 00 5. 00 t3. 76 -8. 76 00 30. 00 19. 78 10. 22 0o 30. 00 19. 78 10. 22 00 25. 00 19. 78 5. 22 WCAP- 17009-NP April 2009 Revision 1

C-27 Unirradiated Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: Unirra Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

40. 00 30. 00 37. 05 -7. 05
40. 00 36. 00 37 .05 -1 05
40. 00 36. 00 37 05 -1 05
80. 00 60. 00 58. 42 1 58
80. 00 45. 00 58. 42 -13. 42
80. 00 40. 00 58. 42 -18. 42 100. 00 55. 00 68. 46 -13. 46 100. 00 75. 00 68. 46 6. 54 100.00 75. 00 68. 46 6. 54 120.00 80. 00 77 .03 2. 97 120.00 85. 00 77. 03 7. 97 120. 00 85. 00 77 .03 7. 97 180. 00 [00. 00 92. 52 7. 48 180.00 100. 00 92. 52 7. 48 180.00 100. 00 92. 52 7. 48 260. 00 t00. 00 98. 60 1 40 260. 00 100.00 98. 60 1 40 320. 00 t00. 00 99 62 38 320. 00 100. 00 99. 62 38 Correlation Coefficient = .975 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

C-28 Capsule U Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:38 PM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 105.71 TO = 86.48 D =O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 86.5 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: U Fluence: n/cm^2 125 100 16 75 U) 50 25 0 iI

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-75. 00 5. 00 4. 50 50

-50. 00 5. 00 7. 03 -2. 03

- 25. 00 5. 00 10. 82 -5. 82

-10. 00 15.00 13. 88 1 12 00 15.00 16. 30 -1 30 00 20. 00 16. 30 3. 70

10. 00 20. 00 19. 05 .95
25. 00 25. 00 23. 81 1.19
50. 00 35. 00 33. 40 1.60 WCAP-17009-NP April 2009 Revision 1

C-29 Capsule U Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: U Fluence: n/cmn^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

72. 00 45. 00 43. 20 1. 80 150. 00 65. 00 76. 88 11.88 200. 00 t00. 00 89,55 10. 45 250. 00 100. 00 95. 66 4. 34 350. 00 100. 00 99. 32 .68 400. 00 100. 00 99. 74 26 Correlation Coefficient = .992 WCAP- 17009-NP April 2009 Revision I

C-30 Capsule Y Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:36 PM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 74.61 TO = 70.11 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 70.2 Plant: Vogfle t Material: SA533B 1 Heat: C0623-1 Orientation: LT Capsule: Y Fluence: n/cma2 125 100 L..

75 U) 50 25 0 - -' I-"-

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-25. 00 10. 00 7. 25 2. 75 00 15. 00 13.25 1. 75

25. 00 20. 00 22. 98 -2.98
25. 00 20. 00 22. 98 -2.98
35. 00 35. 00 28. 07 6.93
50. 00 40. 00 36. 84 3.16
60. 00 30. 00 43. 27 - 13. 27
75. 00 60. 00 53. 27 6.73 100. 00 70. 00 69. 02 .98 April 2009 WCAP- 17009-NP WCAP- I17009-NP April 2009 Revision I

C-31 Capsule Y Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: Y Fluence: n/cm^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125. 00 80. 00 81 33 -1 33 150. 00 90. 00 89 49 551 175.00 95. 00 94. 33 67 225. 00 100. 00 98. 45 1 55 275. 00 100. 00 99. 59 41 300. 00 100. 00 99. 79 21 Correlation Coefficient = .991 WCAP- 17009-NP April 2009 Revision I

C-32 Capsule V Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:43 PM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 95.23 TO = 109.89 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 109.9 Plant: Vogtle I Material: SA533B 1 Heat: C0623-1 Orientation: LT Capsule: V Fluence: n/cm^2 125 100 I-ce, 75 0

U)

I 0.

50 25 0

0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 80. 00 2. 00 1. 82 18

-50. 00 2. 00 3. 36 -1 36

-25. 00 5. 00 5. 56 56

-15.00 5. 00 6.77 -I 77

.00 10. 00 9. 05 95

5. 00 10. 00 9.95 05
20. 00 20. 00 13. 15 6. 85
50. 00 20. 00 22. 14 -2. 14 100. 00 45. 00 44. 83 .17 WCAP-17009-NP April 2009 Revision 1

C-33 Capsule V Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 150.00 65. 00 69. 90 -4.90 175. 00 80. 00 79. 70 .30 200. 00 90. 00 86. 90 3. 10 250. 00 100. 00 94. 99 5. 01 300. 00 100. 00 98. 19 1.81 375. 00 100. 00 99 62 .38 Correlation Coefficient = .998 WCAP- 17009-NP April 2009 Revision 1

C-34 Capsule X Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:47 PM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 90.87 TO = 141.65 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 141.7 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: X Fluence: n/cmA2 125 100 U) 75 I-50 25 0 - -- 1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-25. 00 2. 00 2. 49 - . 49

25. 00 5. 00 7 .13 -2. 13
50. 00 15. 00 11. 74 3. 26
50. 00 25. 00 11. 74 13. 26
75. 00 25. 00 18. 74 6. 26 100. 00 30. 00 28. 56 1 44 125. 00 40. 00 40. 94 94 150. 00 45. 00 54. 58 -9. 58 160. 00 50. 00 59. 96 -9. 96 WCAP- 17009-NP April 2009 Revision 1

C-35 Capsule X Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: X Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 175. 00 65. 0(0 67. 57 -2. 57 200. 00 70. 00 78.32 -8 .32 225. 00 100. 00 86. 23 13. 77 225. 00 too. 00 86. 23 13. 77 250. 00 100. 00 91.57 8. 43 275. 00 100. 00 94. 96 5. 04 Correlation Coefficient = .975 WCAP- 17009-NP April 2009 Revision 1

C-36 Capsule W Intermediate Shell Plate B8805-3 (LT)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/18/2008 03:53 PM Page I Coefficients of Curve 1 A = 50. B = 50. C = 72.26 TO = 148.32 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 148.4 Plant: Vogte I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: W Fluence: n/cn^A2 125 100 I- 75 2) 50 25 0 ! - i - 4w! !1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 30. 00 5. 00 .71 4.29 00 5. 00 1 .62 3.38

50. 00 10. 00 6. 17 3.83
60. 00 15. 00 7 98 7.02
75. 00 15.00 11 62 3.38 100. 00 25. 00 20. 79 4.21 125. 00 30. 00 34. 40 -4.40 150. 00 50. 00 51 .16 -1.16 170 00 50. 00 64. 56 -14. 56 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

C-37 Capsule W Intermediate Shell Plate B8805-3 (LT)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: W Fluence: n/cm^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 180.00 60. 00 70. 61 -0l . 61 200. 00 98. 00 80. 69 17. 31 200. 00 95. 00 80. 69 14. 31 300. 00 100. 00 98. 52 1.48 325. 00 100. 00 99. 25 75 350. 00 100. 00 99. 62 .38 Correlation Coefficient = .978 WCAP-17009-NP April 2009 Revision 1

C-38 Unirradiated Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:23 AM Page I Coefficients of Curve I A = 49.1 B = 46.9 C = 100.61 TO = 60.52 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=96.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=17.1 Deg F Temp@50 ft-lbs=62.5 Deg F Plant: Vogtle I Material: SA533B 1 Heat: C0623-1 Orientation: TL Capsule: Unirra Fluence: n/cmA2 300 250 S200 "a

U.

LM150 100 5o 0 i -i i---1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-40. 00 9. 00 13. 40 -4.40

-40. 00 16. 00 13. 40 2. 60 00 26. 00 23. 86 2. 14 00 24. 00 23. 86 14 00 24. 00 23. 86 14

40. 00 35. 00 39. 67 -4.67
40. 00 46. 00 39. 67 6.33
40. 00 53. 00 39. 67 13. 33
80. 00 55. 00 58. 07 -3. 07 WCAP-17009-NP April 2009 Revision 1

C-39 Unirradiated Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature hiput CVN Computed CVN Differential

80. 00 45. 00 58. 07 -13.07
80. 00 56. 00 58. 07 -2. 07 120.00 69. 00 73.99 -4.99 120.00 73. 00 73.99 - . 99 120.00 71.00 73.99 -2.99 140. 00 84. 00 79.98 4. 02 140. 00 79. 00 79.98 - .98 180.00 94. 00 88. 02 5.98 180.00 98. 00 88.02 9.98 180.00 90. 00 88.02 1.98 240. 00 97. 00 93.43 3.57 240. 00 102. 00 93. 43 8. 57 320. 00 97. 00 95. 46 1. 54 320. 00 96. 00 95. 46 54 Correlation Coefficient = ,982 WCAP- 17009-NP April 2009 Revision 1

C-40 Capsule U Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:23 AM Page 1 Coefficients of Curve 1 A = 50.1 B = 47.9 C = 123.09 TO = 62.29 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=98.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Teinp@30 ft-lbs=7.3 Deg F Ternp@50 ft-lbs=62.1 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: U Fluence: n/cmA2 300 250 A200 IL LM150 100 50 0

n

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-75. 00 14. 00 11. 49 2. 51

-50. 00 18.00 15. 51 2. 49 725.00 15.00 20. 87 -5. 87

10. 00 32. 00 30. 89 1 l1
10. 00 33.00 30. 89 2. 11
25. 00 37.00 36. 02 .98
25. 00 33.00 36. 02 -3. 02
50. 00 49.00 45. 33 3. 67
50. 00 47.00 45. 33 1 .67 WCAP-17009-NP April 2009 Revision I

C-41 Capsule U Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

72. 00 50. 00 53. 87 -3.87 150. 00 75. 00 79.43 -4.43 200. 00 93 .00 88. 76 4. 24 250. 00 96. 00 93.67 2. 33 350. 00 101. 00 97. 11 3. 89 400. 00 103. 00 97. 61 5. 39 Correlation Coefficient = .995 WCAP- 17009-NP April 2009 Revision 1

C-42 Capsule Y Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:24 AM Page 1 Coefficients of Curve 1 A = 54.1 B = 51.9 C = 134.49 TO = 99.4 D = O.OOE+O0 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1060(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=31.8 Deg F Temp@50 ft-lbs=88.8 Deg F Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: Y Fluence: nrcm^2 300 250 A200 IL 0 150 0

U-z

> 100 f0 50 _ _

0~

0 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-50. 00 17.00 12. 35 4. 65

-25. 00 28. 00 16. 30 11. 70

-20. 00 21.00 17. 23 3. 77

.00 24. 00 21. 48 2. 52

25. 00 20. 00 28. 00 -8. 00
35. 00 33. 00 30. 99 2. 01
75. 00 45. 00 44. 79 . 21 100. 00 47. 00 54. 33 -7. 33 110. 00 55. 00 58. 18 -3. 18 WCAP- 17009-NP April 2009 Revision 1

C-43 Capsule Y Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533Bl Heat: C0623-I Orientation: TL Capsule: Y Fluence: n/cmA^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 125. 00 65. 00 63. 86 1. 14 150. 00 67. 00 72. 75 -5. 75 175. 00 81.00 80. 55 .45 225. 00 100. 00 92. 11 7. 89 275. 00 109. 00 98. 90 10. 10 300. 00 109. 00 101. 00 8. 00 Correlation Coefficient = .983 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

C-44 Capsule V Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:25 AM Page 1 Coefficients of Curve 1 A = 48.1 B = 45.9 C = 126.1 TO = 103.15 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=94.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=50.6 Deg F Temp@50 ft-lbs=108.4 Deg F Plant: Vogte 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: V Fluence: n/cmA2 300 250 A200 1L Lm 150 100 50 0 ,i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-25. 00 4. 00 12. 83 -8. 83

.00 19. 00 17. 16 1. 84

25. 00 32. 00 22.81 9.19
50. 00 35. 00 29. 82 5. 18
72. 00 47. 00 36.99 10.01
72. 00 24. 00 36. 99 -12.99 100. 00 42. 00 46. 95 -4.95 110.00 48. 00 50. 59 -2. 59 125.00 61.00 55. 98 5. 02 WCAP-17009-NP April 2009 Revision 1

C-45 Capsule V Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: V Fluence: n/cmA^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 160. 00 64. 00 67. 50 -3.50 200. 00 62. 00 77.74 -15. 74 225.00 99. 00 82. 39 16.61 250. 00 90. 00 85. 85 4. 15 300. 00 94. 00 90. 13 3. 87 375. 00 94. 00 92.79 1.21 Correlation Coefficient = .956 April 2009 17009-NP WCAP- 17009-NP April 2009 Revision I

C-46 Capsule X Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/21/2008 11:42 AM Page 1 Coefficients of Curve I A = 47.6 B = 45.4 C = 124.11 TO = 128.59 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=93.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=77.9 Deg F Temnp@50 ft-lbs=135.2 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: X Fluence: n/cmA2 300 250 I200 0

IL 150 100 1° 0o 50 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-50.00 5. 00 7.04 -2. 04

-25. 00 7. 00 9.25 -2. 25

25. 00 14. 00 16. 59 -2. 59
50. 00 25. 00 22. 16 2. 84
75. 00 33. 00 29. 13 3. 87 100. 00 42. 00 37.32 4. 68 125. 00 47. 00 46. 29 71 150. 00 52. 00 55. 36 -3. 36 175.00 64. 00 63. 83 17 WCAP-17009-NP April 2009 Revision I

C-47 Capsule X Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: X Fluence: n/cm^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 200. 00 64. 00 71.18 -7.18 200. 00 61.00 71.18 -10. 18 225. 00 85. 00 77. 15 7. 85 225. 00 66. 00 77. 15 -11.15 250. 00 95. 00 81.75 13. 25 275. 00 98. 00 85. 16 12. 84 Correlation Coefficient = .970 WCAP-17009-NP April 2009 Revision 1

C-48 Capsule W Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 11:27 AM Page 1 Coefficients of Curve 1 A = 42.1 B = 39.9 C = 117,75 TO = 148.02 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=82.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=1 11.2 Deg F Temp@50 ft-lbs=171.7 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: W Fluence: n/cm^2 300 250 j200

0%

LA 150 w

z 100 50 0 - 1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

. 00 15.00 8. 18 6. 82

80. 00 23. 00 21. 31 1 69 100. 00 28. 00 26. 68 1 32 110.00 28.00 29. 65 -1 65 120. 00 40. 00 32. 78 7. 22 125. 00 38.00 34. 40 3. 60 150. 00 36. 00 42. 77 -6. 77 t65. 00 40. 00 47. 81 -7. 81 180.00 40. 00 52. 68 -12. 68 WCAP- 17009-NP April 2009 Revision 1

C-49 Capsule W Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: W Fluence: n/cm^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 190. 00 50. 00 55. 75 -5.75 200. 00 66. 00 58. 65 7.35 200. 00 68. 00 58. 65 9. 35 275. 00 80. 00 73. 72 6.28 300. 00 79. 00 76. 39 2.61 325. 00 86. 00 78.24 7.76 Correlation Coefficient = .954 WCAP- 17009-NP April 2009 Revision 1

C-50 Unirradiated Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 09:34 AM Page 1 Coefficients of Curve 1 A = 39.95 B = 39.95 C = 112.27 TO = 67.67 D = O.OOE+0O Equation is A + B * [Tanh((T-To)/(C+DT))1 Upper Shelf L.E.=79.9 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils= 5 3.7 Deg F Plant: Vogtle I Material: SA533B I Heat: C0623-1 Orientation: TL Capsule: Unirra Fluence: n/cmA2 200 150 E

00 10 50 0 -

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-40. 00 5. 00 10. 23 -5.23

-40. 00 11. 00 10. 23 77 00 19. 00 18. 42 .58 00 17. 00 18. 42 1.42 00 19. 00 18. 42 .58

40. 00 27. 00 30. 30 -3.30
40. 00 37.00 30. 30 6.70
40. 00 41. 00 30. 30 10. 70
80. 00 40. 00 44. 32 -4.32 WCAP-17009-NP April 2009 Revision 1

C-51 Unirradiated Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: Unirra Fluence: n/cmA2 Charpy V-Notch Data Temperature InputL.E. Computed L.E. Differential

80. 00 37. 00 44. 32 -7. 32
80. 00 44. 00 44. 32 - . 32 120.00 53. 00 57. 33 -4. 33 120. 00 56. 00 57. 33 -1 33 120.00 58. 00 57. 33 67 140. 00 63. 00 62. 63 37 140. 00 62. 00 62. 63 63 180.00 73. 00 70. 38 2. 62 180.00 71.00 70. 38 .62 180.00 80. 00 70. 38 9. 62 240. 00 77. 00 76. 35 65 240. 00 74. 00 76. 35 35 320. 00 80. 00 79.02 98 320. 00 74. 00 79.02 -5. 02 Correlation Coefficient =.983 WCAP- 17009-NP April 2009 Revision 1

C-52 Capsule U Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:08 AM Page I Coefficients of Curve 1 A = 36.4 B = 36.4 C = 121.65 TO = 54.24 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=72.8 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mnils=49.6 Deg F Plant: Vogtle 1 Material: SA533B 1 Heat: C0623-1 Orientation: TL Capsule: U Fluence: n/cmA2 200 150 0_9 Em100 5 0l 50 00 0

0

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-75. 00 10.00 7. 77 2.23

-50. 00 12. 00 11. 12 .88

-25. 00 10. 00 15. 56 5.56 10.00 28. 00 23. 72 4. 28

10. 00 24. 00 23. 72 .28
25. 00 30. 00 27. 82 2.18
25. 00 26. 00 27. 82 -1.82
50. 00 36. 00 35. 13 .87
50. 00 36. 00 35. 13 87 WCAP- 17009-NP April 2009 Revision 1

C-53 Capsule U Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: U Fluence: n/cmA^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 72.00 36. 00 41.68 -5.68 150. 00 64. 00 60.31 3 69 200. 00 67. 00 66. 72 28 250. 00 68. 00 70. 00 -2. 00 350. 00 72. 00 72. 24 -. 24 400. 00 73. 00 72. 55 .45 Correlation Coefficient = .992 WCAP-17009-NP April 2009 Revision 1

C-54 Capsule Y Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:11 AM Page 1 Coefficients of Curve 1 A = 38.99 B = 38.99 C = 145.17 TO = 84.77 D = O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=78.0 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=69.9 Deg F Plant: Voglle 1 Material: SA533B 1 Heat: C0623-1 Orientation: TL Capsule: Y Fluence: n/cmA2 200 150 0

10 soo 0 ý-

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-50.00 12. 00 10.53 1 47

-25. 00 20. 00 14, 08 5. 92

-20. 00 15. 00 14. 90 .10

.00 20. 00 18. 50 1. 50

25. 00 18. 00 23. 79 -5. 79
35. 00 26. 00 26. 13 -. 13
75. 00 34. 00 36.37 -2. 37 100. 00 40. 00 43. 07 -3. 07 110. 00 48. 00 45. 70 2. 30 WCAP- 17009-NP April 2009 Revision 1

C-55 Capsule Y Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 125. 00 50. 00 49. 53 47 150. 00 53. 00 55. 42 -2. 42 175. 00 66. 00 60, 53 5. 47 225. 00 73. 00 68. 12 4. 88 275. 00 66. 00 72. 70 -6. 70 300. 00 75. 00 74. 17 .83 Correlation Coefficient = .986 April 2009 WCAP- 17009-NP I 7009-NP April 2009 Revision I

C-56 Capsule V Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:13 AM Page 1 Coefficients of Curve 1 A = 34.82 B = 34.82 C = 126.79 TO = 104.47 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=69.6 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=105.2 Deg F Plant: Voglle I Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: V Fluence: n/cm^2 200 150 E

.2 100 IL1o so_

50

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.. Differential

-25. 00 2. 00 8.00 -6. 00

.00 14. 00 11.24 2. 76

25. 00 21.00 15. 47 5. 53
50. 00 22. 00 20. 72 1.28
72. 00 31.00 26. 09 4.91
72. 00 19.00 26. 09 -7. 09 100. 00 31.00 33. 60 -2. 60 110.00 35. 00 36. 34 -1.34 125.00 44. 00 40.41 3. 59 WCAP- 17009-NP April 2009 Revision 1

C-57 Capsule V Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 160. 00 46. 00 49. 17 -3.17 200. 00 53. 00 57. 01 -4.01 225. 00 66. 00 60. 59 5.41 250. 00 67. 00 63. 27 3. 73 300. 00 67. 00 66. 60 .40 375. 00 65. 00 68. 68 3.68 Correlation Coefficient = .980 WCAP- 17009-NP April 2009 Revision I

C-58 Capsule X Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:16 AM Page 1 Coefficients of Curve I A = 39.03 B = 39.03 C = 149.02 TO = 157.41 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=78.l Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=142.0 Deg F Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: X Fluence: n/cm^2 200 150

.o_

a. 100 15 50 0 4-

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-50. 00 .00 4. 54 -4.54

-25. 00 .00 6. 21 -6.21

25. 00 7.00 11. 29 -4.29
50. 00 15.00 14. 93 .07
75. 00 25. 00 19. 41 5. 59 100. 00 31. 00 24. 70 6.30 125. 00 32. 00 30. 67 1.33 150. 00 38. 00 37.09 .91 175.00 45. 00 43.61 1.39 WCAP- 17009-NP April 2009 Revision 1

C-59 Capsule X Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: X Huence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 200. 00 43. 00 49. 89 -6. 89 200. 00 44. 00 49. 89 -5. 89 225. 00 60. 00 55.61 4. 39 225. 00 50. 00 55. 61 -5. 61 250. 00 66. 00 60. 57 5. 43 275. 00 66. 00 64. 70 1. 30 Correlation Coefficient = .977 WCAP-17009-NP April 2009 Revision 1

C-60 Capsule W Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:18 AM Page 1 Coefficients of Curve 1 A = 37.03 B = 37.03 C = 137.21 TO = 146.47 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=74. t Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=139.0 Deg F Plant: Vogole 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: W Fluence: n/cmn^2 200 150 C

0 C

& 100 50 0J +

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

. 00 12.00 7. 83 4. 17

80. 00 20. 00 20. 37 -. 37 100.00 25. 00 24. 95 .05 110.00 26. 00 27 .41 -1.41 120. 00 31.00 29. 97 1.03 125. 00 34. 00 31. 28 2. 72 150. 00 31.00 37 .98 -6.98 165.00 38.00 42. 00 -4.00 180.00 43.00 45. 90 -2.90 WCAP-17009-NP April 2009 Revision 1

C-61 Capsule W Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-I Orientation: TL Capsule: W Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 190. 00 47. 00 48. 40 - 1. 40 200. 00 54. 00 50. 78 3. 22 200. 00 62. 00 50. 78 1 122 275. 00 64. 00 64. 20 20 300. 00 63.00 66. 92 -3. 92 325.00 70. 00 68.95 1 .05 Correlation Coefficient = .972 WCAP- 17009-NP April 2009 Revision I

C-62 Unirradiated Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 09:34 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C= 88.03 TO = 80.81 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 80.9 Plant: Vogtle I Material: SA533B 1 Heat: C0623-1 Orientation: TL Capsule: Unirra Fluence: ncnMA2 125 100 75 U'

0 20 50 a-25 0 k-I 1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 40. 00 9.00 6. 04 2.96

-40. 00 14. 00 6. 04 7.96

.00 20. 00 13. 75 6.25

.00 14. 00 13 75 .25

.00 25. 00 13 75 11.25

40. 00 18. 00 28. 35 -10. 35
40. 00 36. 00 28. 35 7. 65
40. 00 40. 00 28. 35 11.65
80. 00 30. 00 49 54 -19. 54 WCAP-17009-NP April 2009 Revision 1

C-63 Unirradiated Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

80. 00 41. 00 49. 54 -8. 54
80. 00 48. 00 49. 54 -1 .54 120. 00 65. 00 70. 90 -5. 90 120.00 70. 00 70. 90 90 120.00 70. 00 70. 90 90 140. 00 90. 00 79. 33 10. 67 140. 00 80. 00 79. 33 .67 180.00 100. 00 90. 50 9. 50 180.00 100. 00 90. 50 9. 50 180.00 100. 00 90. 50 9. 50 240. 00 t00. 00 97. 38 2. 62 240. 00 too. O0 97. 38 2. 62 320. 00 100. 00 99. 57 43 320.00 100. 00 99. 57 .43 Correlation Coefficient = .974 WCAP-17009-NP April 2009 Revision 1

C-64 Capsule U Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:09 AM Page I Coefficients of Curve 1 A = 50. B = 50. C = 78.41 TO = 73.46 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temrperature at 50% Shear = 73.5 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: U Fluence: n/cm^2 125 100 75 U,

0 2

0 50 a-25 0 - ,  !

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-75. 00 5. 00 2.22 2. 78

-50. 00 10.00 4.11 5.89

-25. 00 10. 00 7.51 2. 49

10. 00 20. 00 16.54 3.46
10. 00 20. 00 16.54 3.46
25. 00 25. 00 22.51 2. 49
25. 00 20. 00 22.51 -2.51
50. 00 35. 00 35.47 - .47
50. 00 30. 00 35. 47 -5.47 WCAP-17009-NP April 2009 Revision 1

C-65 Capsule U Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: U Fluence: n/cmA^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

72. 00 45. 00 49. 07 -4.07 150. 00 95. 00 87.57 7.43 200. 00 100. 00 96. 19 3.81 250. 00 o00. 00 98.90 1. 10 350. 00 100. 00 99.91 09 400. 00 100. 00 99.98 02 Correlation Coefficient = .996 WCAP-17009-NP April 2009 Revision 1

C-66 Capsule Y Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:11 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 84.83 TO = 91.26 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 91.3 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: Y Fluence: n/cmA2 125 100 75 A

0.

50 25 0 -! --- i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 50. 00 5. 00 3. 45 1.. 55

- 25. 00 15. 00 6. 06 8. 94

-20. 00 10. 00 6. 77 3. 23

.00 10. 00 10. 42 42

25. 00 15.00 17. 33 -2. 33
35. 00 15. 00 20. 98 -5. 98
75. 00 45. 00 40. 53 4. 47 100. 00 45. 00 55, 14 -10. 14 110. 00 70. 00 60. 87 9. 13 WCAP-17009-NP April 2009 Revision 1

C-67 Capsule Y Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125. 00 70. 00 68. 90 1 10 150. 00 75. 00 79.98 4. 98 175. 00 90. 00 87.81 2. 19 225. 00 100. 00 95.90 4. 10 275. 00 100. 00 98.70 1 30 300. 00 100. 00 99.28 72 Correlation Coefficient = .990 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

C-68 Capsule V Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:14 A-M Page 1 Coefficients of Curve I A = 50. B = 50. C= 91.4 TO = 107.58 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 107.6 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: V Fluence: nrcmA2 125 100 75 U) 11 50 25 0 1 't "-4-'

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 25. 00 5. 00 5. 21 -. 21 00 10.00 8.67 1. 33

25. 00 15. 00 14. 10 .90
50. 00 15. 00 22. 10 7. 10
72. 00 40. 00 31.46 8. 54
72. 00 30. 00 31. 46 -1.46 100. 00 40. 00 45. 86 -5.86 110. 00 50. 00 51.32 -1.32 125. 00 70. 00 59. 42 10. 58 WCAP- 17009-NP April 2009 Revision I

C-69 Capsule V Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: V Fluence: nIcm^A2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 160. 00 70. 00 75. 90 -5.90 200. 00 80. 00 88. 31 -8.31 225. 00 too. oo 92. 89 7.11 250. 00 100. 00 95. 76 4. 24 300. 00 1oo. 00 98. 54 1.46 375. 00 100. 00 99. 71 29 Correlation Coefficient = .987 WCAP-17009-NP April 2009 Revision 1

C-70 Capsule X Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:16 AIM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 77.11 TO = 145.71 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 145.8 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: X Fluence: n/crnA2 125 100 i

U, 75 2ai 50 a.

25 0 +/- i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-50. 00 2. 00 *62 1. 38

-25. 00 2. 00 1. 18 82

25. 00 5. 00 4. 19 81
50. 00 10. 00 7. 71 2. 29
75. 00 20. 00 13. 78 6. 22 100. 00 25. 00 23. 40 60 125. 00 35. 00 36. 88 2. 88 150. 00 50. 00 52. 78 78 175. 00 65. 00 - 3.
68. 13 13 WCAP-17009-NP April 2009 Revision I

C-71 Capsule X Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle I Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: X Fluence: ncra^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 200. 00 75. 00 80. 34 -5.34 200. 00 75. 00 80. 34 -5. 34 225. 00 100. 00 88. 66 11. 34 225. 00 95. 00 88. 66 6. 34 250. 00 100. 00 93.73 6. 27 275. 00 100. 00 96. 62 3.38 Correlation Coefficient = .993 WCAP- 17009-NP April 2009 Revision 1

C-72 Capsule W Intermediate Shell Plate B8805-3 (TL)

CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 10:18 AM Page 1 Coefficients of Curve I A = 50. B = 50. C = 81.97 TO = 164.97 D =0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 165.0 Plant: Vogtle t Material: SA533B 1 Heat: C0623-1 Orientation: TL Capsule: W Fluence: n/cn^A2 125 100 75 U,

0 2 50 0

0.

25 0 iT-+--, 4

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

.00 5. 00 1 75 3. 25

80. 00 20. 00 11. 17 8 83 100.00 20. 00 17. 00 3 00 110. 00 25. 00 20. 73 4. 27 120. 00 25. 00 25. 02 -. 02 125.00 30. 00 27 .38 2. 62 150.00 35. 00 40. 97 -5. 97 165. 00 40. 00 50.02 - 10, 02 180.00 50. 00 59.07 -9. 07 WCAP- 17009-NP April 2009 Revision 1

C-73 Capsule W Intermediate Shell Plate B8805-3 (TL)

Page 2 Plant: Vogtle 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: W Fluence: n/cm^2 Charpy V-Notch Data Tenmperature Input Percent Shear Computed Percent Shear Differential 190. 00 60. 00 64. 81 -4.81 200. 00 75. 00 70. 16 4. 84 200. 00 85. 00 70. 16 14. 84 275. 00 too. 00 93.61 6. 39 300. 00 100.00 96.42 3.58 325. 00 100. 00 98.03 1.97 Correlation Coefficient =.979 WCAP-17009-NP April 2009 Revision 1

C-74 Unirradiated Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:52 AM Page I Coefficients of Curve 1 A = 73.6 B = 71.4 C = 73.31 TO = -5.18 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 145.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temnp@30 ft-lbs=-57.2 Deg F Temp@50 ft-lbs=-30.3 Deg F Plant: Vogle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Unirra Fluence: n/cm^2 300 250

-200 0

LM150 0

-0 0~

0 100 0o 0A 0

0

.0 50 00

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-120. 00 7.00 8. 17 -1.17

-120. 00 5. 00 8. 17 -3.17

- 80. 00 12. 00 18.61 -6.61

-80. 00 15. 00 18.61 -3.61

-80. 00 16.00 18.61 -2.61

-60.00 10.00 28. 34 -18.34

-60.00 11.00 28. 34 -17. 34

-60.00 24. 00 28.34 -4.34

-40.00 58. 00 42. 02 15.98 WCAP- 17009-NP April 2009 Revision 1

C-75 Unirradiated Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 40. 00 22. 00 42. 02 -20. 02

-40. 00 35. 00 42. 02 -7. 02

-20. 00 94. 00 59. 36 34. 64

-20.00 76. 00 59. 36 16. 64

-20. 00 86. 00 59. 36 26. 64 o0 68.00 78. 64 -10. 64 00 95. 00 78. 64 16. 36 00 60. 00 78. 64 -18. 64 40 00 101.00 112. 77 -t 1.77 40 00 101.00 112. 77 -11 .77 40 00 100.00 112. 77 -12. 77 60 00 130.00 124. 36 5. 64 60 00 118.00 124.36 -6. 36 60 00 123. 00 124.36 - . 36 80 o0 128. 00 132.27 -4. 27 80 00 126. 00 132.27 -6. 27 80 00 153. 00 132. 27 20. 73 120 00 141.00 140. 46 54 120 00 140.00 140. 46 -. 46 120 00 135. 00 140. 46 -5. 46 180 00 144. 00 144. 09 - . 09 180 00 158. 00 144. 09 13.91 180 00 154. 00 144. 09 9.91 240 00 144. 00 144. 82 - . 82 240 00 135,00 144. 82 -9.82 240 00 145. 00 144. 82 .18 320 00 144. 00 144. 98 - .98 320. 00 143.00 144. 98 -1.98 320. 00 143. 00 144. 98 -1. 98 Correlation Coefficient = .972 WCAP- 17009-NP April 2009 Revision 1

C-76 Capsule U Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:55 AM Page 1 Coefficients of Curve 1 A = 77.6 B = 75.4 C = 49.05 TO = 4.36 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 153.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 ft-lbs=-32.1 Deg F Temp@50 ft-lbs=-14.4 Deg F Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: U Fluence: n/cinA2 300 250 A200 0

IL Lx 150 100 50

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 100. 00 3. 00 4. 31 -1. 31

- 50. 00 12. 00 17. 02 -5. 02

30. 00 12. 00 32. 00 -20. 00
30. 00 17. 00 32. 00 -15. 00
30. 00 18. 00 32. 00 -14. 00
25. 00 70. 00 37. 18 32. 82
25. 00 16. 00 37. 18 -21. 18
25. 00 90. 00 37. 18 52. 82 00 52. 00 70. 91 -18. 91 WCAP- 17009-NP April 2009 Revision 1

C-77 Capsule U Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

25. 00 115.00 107.58 7.42
72. 00 142. 00 144. 01 -2.01 200. 00 137.00 152. 95 -15. 95 250. 00 158.00 152. 99 5.01 350. 00 181.00 153.00 28. 00 350. 00 147.00 153. 00 -6.00 Correlation Coefficient = .940 WCAP- 17009-NP April 2009 Revision 1

C-78 Capsule Y Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:57 AM Page 1 Coefficients of Curve 1 A = 70.6 B = 68.4 C = 56.34 TO = -7.34 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 139.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Ternp@30 ft-lbs=-45.8 Deg F Temp@50 ft-lbs=-24.8 Deg F Plant: Vogule 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Y Fluence: n/cm^2 300 250 A200 0

2' 150 o/0° 0 C 0O uJ 00 100 50 Oc 0 Oi i 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-50. 00 13. 00 26. 86 -13. 86

-35. 00 17. 00 39 .47 -22. 47

-25. 00 15. 00 49. 83 -34. 83

-25. 00 67. 00 49. 83 17. 17

-10. 00 98. 00 67. 37 30. 63

-5. 00 80. (0 73. 44 6. 56 00 94. 00 79. 46 14. 54

5. 00 1 14. 00 85. 34 28. 66
25. 00 70. 00 106. 05 -36. 05 WCAP-17009-NP April 2009 Revision 1

C-79 Capsule Y Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

40. 00 109. 00 117.52 -8. 52
60. 00 112. 00 127. 52 -15.52 100. 00 124. 00 136.04 -12. 04 175. 00 134. 00 138.79 -4. 79 275. 00 153. 00 138.99 14.01 300. 00 145. 00 139. 00 6. 00 Correlation Coefficient = .889 WCAP- 17009-NP April 2009 Revision 1

C-80 Capsule V Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:59 AM Page 1 Coefficients of Curve I A = 70.1 B = 67.9 C = 53.34 TO = -21.34 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1 38.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-57.5 Deg F Temp@50 ft-lbs=-37.6 Deg F Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: V Fluence: n/cm^2 300 250 S200 0

LL 150

'L 0

100 50 o °/2

'0ý 0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-125. 00 10. 00 4. 93 5. 07

-100. 00 26. 00 8. 96 17. 04

-75. 00 31.00 18. 22 12. 78

- 60. 00 3.00 28. 01 -25. 01

-50. 00 15.00 36. 76 -21. 76

-45. 00 14. 00 41. 81 -27. 81

-40. 00 125. 00 47. 27 77. 73

-25. 00 85. 00 65. 44 19. 56

-25. 00 7. 00 65. 44 -58. 44 WCAP- 17009-NP April 2009 Revision 1

C-81 Capsule V Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: V Fluence: n/cm^2 Charpy V-Notch Data Temperature input CVN Computed CVN Differential

-10. 00 102. 00 84. 32 17.68

50. 00 124.00 129. 24 -5. 24 100. 00 138.00 136. 58 1.42 150. 00 140. 00 137. 78 2. 22 200. 00 146. 00 137. 97 8. 03 300. 00 143. 00 138. 00 5. 00 Correlation Coefficient = .863 WCAP- 17009-NP April 2009 Revision 1

C-82 Capsule X Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/09/2008 10:12 AM Page 1 Coefficients of Curve 1 A = 70.1 B = 67.9 C = 40.03 TO = 25.27 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 138.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-1.8 Deg F Temp@50 ft-lbs=13.1 Deg F Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: X Fluence: n/cmA2 300 250 A200 0

Lm 150 100 50 0 1 i

-300.0 -200.0 -100.0 0.0 100.0 200.0 30 0.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 100. 00 3. 00 2. 46 54

- 50. 00 6. 00 5. 29 .71 00 22. 00 32. 15 -10. 15

10. 00 42. 00 45. 38 -3.38
25. 00 85. 00 69. 63 15. 37
50. 00 101. 00 107. 41 -6.41
75. 00 127. 00 127. 55 - . 55 100. 00 136. 00 t34. 83 1.17 125. 00 112. 00 137, 07 -25. 07 WCAP- 17009-NP April 2009 Revision 1

C-83 Capsule X Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: X Fluence: n/crn^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 125. 00 128.00 137.07 -9.07 150. 00 127. 00 137.73 -10.73 150. 00 137.00 137.73 - .73 175.00 142. 00 137.92 4.08 200. 00 142. 00 137.98 4.02 225. 00 145. 00 137.99 7.01 Correlation Coefficient = .984 WCAP- 17009-NP April 2009 Revision 1

C-84 Capsule W Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:07 AM Page 1 Coefficients of Curve 1 A = 63.6 B = 61.4 C = 60.5 TO = 14.68 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 125.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-22.4 Deg F Temp@50 ft-lbs=l. 1 Deg F Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: W Fluence: n/cr'n2 300 250

,200

0 LA10 IL 100

______ a 5o n

U

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-50. 00 4. 00 15. 15 - 11. 15

- 20. 00 10. 00 31.81 -21. 81

- 20. 00 14. 00 31.81 -17. 81

- 15. 00 13. 00 35. 68 -22. 68

-15.00 48. 00 35.68 12. 32

-10.00 53. 00 39. 86 13. 14

- 10. 00 61.00 39. 86 21. 14

-5. 00 61.00 44. 30 16. 70

.00 8 1. 00 48. 99 32. 01 WCAP-17009-NP April 2009 Revision 1

C-85 Capsule W Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: W Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 10.00 48. 00 58. 86 -10. 86

20. 00 55. 00 68. 99 -13.99
25. 00 65. 00 73. 97 -8. 97 125. 00 125.00 121.88 3.12 150. 00 126.00 123.62 2.38 175. 00 125. 00 124. 39 .61 Correlation Coefficient = .914 WCAP- 17009-NP April 2009 Revision 1

C-86 Unirradiated Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:53 AM Page 1 Coefficients of Curve I A = 44.13 B = 44.13 C = 54.42 TO = -21.21 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=88.3 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mnils=-32.6 Deg F Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Unirra Fluence: n/cmn^2 200 150 0

.o_

S 2Llo R 100 51 50 0 -

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 120. 00 3. 00 2.28 .72

- 120. 00 2. 00 2. 28 .28

- 80. 00 7. 00 9.12 -2. 12

- 80. 00 10. 00 9.12 .88

- 80. 00 10. 00 9.12 .88

-6.0 0 7. 00 17. 10 10. 10

-60. 00 7. 00 17. 10 10. 10

-60. 00 18. 00 17. 10 190

-40. 00 40. 00 29. 46 10.54 WCAP- 17009-NP April 2009 Revision 1

C-87 Unirradiated Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Unirra Fluence: rIcm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-40.00 16.00 29. 46 -13. 46

-40.00 26. 00 29. 46 -3. 46

-20. 00 67. 00 45. 10 21. 90

-20. 00 56. 00 45. 10 10. 90

-20. 00 52. 00 45. 10 6. 90

.00 50. 00 60. 50 -10. 50

,00 68. 00 60. 50 7. 50

.00 45. 00 60. 50 -15. 50

40. 00 72. 00 79. 83 -7 .83
40. 00 74. 00 79 .83 -5. 83
40. 00 72. 00 79. 83 -7. 83
60. 00 8 7. 00 84. 00 3. 00
60. 00 77. 00 84. 00 -7. 00
60. 00 86. 00 84. 00 2. 00
80. 00 87. 00 86. 16 84
80. 00 84. 00 86. 16 -2. 16
80. 00 107. 00 86. 16 20. 84 120.00 90. 00 87. 76 2. 24 120.00 88. 00 87.76 .24 120.00 89. 00 87.76 1.24 180.00 97. 00 88.20 8. 80 180.00 82. 00 88.20 -6.20 180.00 91. 00 88. 20 2. 80 240. 00 90. 00 88. 25 1.75 240. 00 88. 00 88.25 -. 25 240. 00 88. 00 88.25 -. 25 320. 00 86. 00 88.25 -2. 25 320. 00 92. 00 88. 25 3.75 320. 00 77. 00 88. 25 -11.25 Correlation Coefficient = .969 WCAP-17009-NP April 2009 Revision I

C-88 Capsule U Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:56 AM Page 1 Coefficients of Curve I A = 41.54 B = 41.54 C = 42.6 TO = -13.16 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=83.1 Lower Shelf L.E.=.0(Fixed)

Telnp.@L.E. 35 mils=-19.9 Deg F Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: U Fluence: n/cin^2 200 150 C

.2 2L100 0

00 f0Q 0

0o n

0

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 100. 00 2. 00 1. 38 .62

- 50. 00 12. 00 12. 51 - .51

-30. 00 12. 00 25. 93 13. 93

-30. 00 17. 00 25. 93 -8.93

-30. 00 15.00 25. 93 -10. 93

-25. 00 50. 00 30. 29 19. 71

-25. 00 16. 00 30. 29 - 14. 29

-25. 00 66. 00 30. 29 35.71

.00 40. 00 53. 99 -13.99 WCAP-17009-NP April 2009 Revision 1

C-89 Capsule U Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: U Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

25. 00 80. 00 71.22 8.78
72. 00 75. 00 81.59 -6.59 200. 00 85. 00 83. 09 1.91 250. 00 85. 00 83. 09 1.91 350. 00 80. 00 83.09 -3. 09 350. 00 86. 00 83. 09 2.91 Correlation Coefficient = .910 WCAP-17009-NP April 2009 Revision 1

C-90 Capsule Y Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:57 AM Page 1 Coefficients of Curve 1 A = 38.12 B = 38.12 C = 22.94 TO = -22.36 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=76.2 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 mils=-24.2 Deg F Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Y Fluence: n/cn^2 200 a 150 i

.2 100O 50 0 __--,__ _ 4

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-50. 00 11. 00 6. 29 4.71

-35. 00 14. 00 19. 02 -5.02

-25. 00 14. 00 33.76 -19.76

-25. 00 52. 00 33.76 18.24

-10. 00 67. 00 56. 88 10. 12

-5. 00 57. 00 62. 49 -5. 49 00 68. 00 66. 74 1.26

5. 00 71. 00 69. 82 1. 18
25. 00 52. 00 75. 03 -23.03 WCAP-17009-NP April 2009 Revision 1

C-91 Capsule Y Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Y Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

40. 00 72. 00 75. 91 -3.91
60. 00 74. 00 76. 18 -2. 18 100. 00 86. 00 76. 24 9. 76 175. 00 87. 00 76. 24 10. 76 275. 00 76. 00 76. 24 - . 24 300. 00 81. 00 76. 24 4. 76 Correlation Coefficient = .906 WCAP- 17009-NP April 2009 Revision I

C-92 Capsule V Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:59 AM Page 1 Coefficients of Curve I A = 42.44 B = 42.44 C = 41.35 TO = -25.62 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=84.9 Lower Shelf L.E.=.O(Fixed)

Temp.@L.E. 35 inls=-32.9 Deg F Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: V Fluence: n/cin^2 200 150 i i

'a 0

.o_

Ia 100 4,-

0 so 0

50 0/

n'

-300.0 0

00 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 125. 00 2. 00 69 1.31 100. 00 11. 00 2 26 8. 74

-75. 00 18. 00 7 14 10. 86

60. 00 1 00 13 53 -12.53
50. 00 9. 00 19 97 -10. 97
45. 00 8. 00 23 89 -15. 89

-40. 00 70. 00 28 25 41. 75

- 2.5. 00 56. 00 43 08 12. 92

- 25. 00 7. 00 43. 08 -36. 08 WCAP- 17009-NP April 2009 Revision 1

C-93 Capsule V Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: V Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-10. 00 69. 00 57. 75 11. 25

50. 00 77. 00 82. 74 -5. 74 100. 00 86. 00 84. 68 1. 32 150. 00 88. 00 84. 86 3. 14 200. 00 84. 00 84. 88 88 300. 00 85. 00 84. 88 12 Correlation Coefficient = .884 WCAP- 17009-NP April 2009 Revision 1

C-94 Capsule X Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/21/2008 11:35 AM Page 1 Coefficients of Curve I A = 41.02 B = 41.02 C = 44.72 TO = 33.45 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=82.0 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=26.9 Deg F Plant: Vogte 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: X Fluence: n/cm^2 200 150 E

00 ulo 2L 100 5

50 0 ý-

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 100. 00 00 21 -. 21

-50. 00 00 1. 92 -1.92 00 15. 00 15. 01 -. 01

10. 00 31. 00 21. 28 9.72
25. 00 25. 00 33. 35 -8. 35
50. 00 54. 00 55. 53 -1.53
75. 00 73.00 70. 96 2. 04 100. 00 85. 00 78. 05 6.95 125. 00 76. 00 80. 69 -4.69 WCAP- 17009-NP April 2009 Revision I

C-95 Capsule X Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: X Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 125. 00 84. 00 80. 69 3. 31 150. 00 82. 00 81.59 . 41 150. 00 80. 00 81.59 - 1.59 175. 00 83. 00 81. 89 11 200. 00 81.00 81.99 -. 99 225. 00 78. 00 82. 02 -4.02 Correlation Coefficient = .991 WCAP-17009-NP April 2009 Revision 1

C-96 Capsule W Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:07 AM Page 1 Coefficients of Curve I A = 44.26 B = 44.26 C = 62.68 TO = 5.28 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=88.5 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 rnils=-8.0 Deg F Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: W Fluence: n/cmnA2 200 150 00 E

.2 100 5.

10o 50 0

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.F. Differential

- 50. 00 6. 00 12.95 -6. 95

-20. 00 14. 00 27. 32 -13. 32

-20. 00 16. 00 27. 32 -11. 32

-15. 00 16. 00 30. 42 - 14. 42

- 15. 00 40. 00 30. 42 9. 58

-10.00 40. 00 33. 68 6. 32

-10.00 49. 00 33. 68 15. 32

-5. 00 49. 00 37.07 11. 93

.00 60. 00 40. 55 19. 45 WCAP- 17009-NP April 2009 Revision 1

C-97 Capsule W Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: W Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

10. 00 41.00 47. 59 -6. 59
20. 00 44. 00 54. 47 10.47
25. 00 51. 00 57. 75 -6.75 125. 00 88. 00 86. 63 1.37 150. 00 88. 00 87. 66 34 175. 00 89. 00 88. 13 87 Correlation Coefficient = .915 WCAP- 17009-NP April 2009 Revision 1

C-98 Unirradiated Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:53 AM Page 1 Coefficients of Curve I A = 50. B = 50. C = 74.21 TO = -6.16 D =0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -6.1 Plant: Vogtle. I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Unirra Fluence: n/cmA2 125 100 75 U) 50 25 0 J -- i M -

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 120. 00 .00 4. 44 -4. 44

- 120. 00 .00 4. 44 -4. 44

- 80. 00 5.00 12. 03 -7. 03

- 80. 00 5. 00 12. 03 -7. 03

-80. 00 5.00 12. 03 -7. 03

-60. 00 5. 00 18. 98 -13. 98

- 60. 00 5. 00 18. 98 -13. 98

-60. 00 25. 00 18. 98 6. 02

-40. 00 43. 00 28. 66 14. 34 WCAP-17009-NP April 2009 Revision I

C-99 Unirradiated Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Unirra Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 40. 00 15. 00 28.66 -13.66

-40. 00 33. 00 28. 66 4. 34

- 20. 00 65. 00 40. 78 24. 22

-20. 00 56. 00 40, 78 15.22

-20. 00 50. 00 40. 78 9.22

.00 48. 00 54. 14 -6.14

.00 65. 00 54. 14 10. 86

.00 35. 00 54. 14 -19. 14

40. 00 75. 00 77.62 -2. 62
40. 00 70. 00 77. 62 -7.62
40. 00 75. 00 77. 62 -2. 62
60. 00 85. 00 85.61 -. 61
60. 00 85. 00 85.61 -. 61
60. 00 85 . 00 85.61 -. 61
80. 00 80. 00 91.07 -11.07
80. 00 80. 00 91.07 -11..07
80. 00 100.00 91. 07 8.93 120. 00 100. 00 96. 77 3. 23 120. 00 100. 00 96.77 3.23 120. 00 100. 00 96,77 3.23 180. 00 100. 00 99. 34 66 180. 00 100.00 99. 34 66 180 00 100.00 99. 34 66 240 00 100. 00 99. 87 13 240 00 100. 00 99. 87 13 240 00 100. 00 99. 87 13 320 00 100. 00 99. 98 02 320 00 100. 00 99. 98 02 320. 00 100. 00 99. 98 02 Correlation Coefficient = .972 April 2009 17009-NP WCAP- 17009-NP April 2009 Revision I

C-100 Capsule U Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:56 AM Page 1 Coefficients of Curve I A = 50. B = 50. C = 36.29 TO = -19.32 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -19.3 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: U Fluence: n/cmA2 125 100 I 75 0

20 50 a.

25 0 i i . f i

-300.0 -200.0 -100.0 0.0 100.0 200.0 3C0.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 100. 00 2. 00 1. 16 .84

- 50. 00 10.00 15.57 -5.57

-30. 00 20. 00 35.70 -15. 70

-30. 00 25. 00 35.70 -10.70

-30. 00 25. 00 35. 70 -10. 70

-25. 00 65. 00 42. 24 22. 76

-25. 00 25. 00 42. 24 -17. 24

-2.5. 00 85. 00 42. 24 42. 76

.00 60. 00 74.36 -14.36 WCAP- 17009-NP April 2009 Revision 1

C-101 Capsule U Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

25. 00 95. 00 92. 00 3. 00
72. 00 95. 00 99. 35 -4. 35 200. 00 10o. oo too. O0 00 250. 00 100. 00 100. 00 00 350. 00 tOO. 00 t0o. O0 00 350. 00 100. 00 t00. 00 00 Correlation Coefficient = .914 WCAP- 17009-NP April 2009 Revision 1

C-102 Capsule Y Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 09:58 AM Page 1 Coefficients of Curve I A = 50. B = 50. C = 55.21 TO = -11.38 D = 0.00E+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -11.3 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Y Fluence: n/cmA2 125 100 75 U,

0 20 50 a-25 0 -!1i i i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-50. 00 10. 00 19 .79 -9. 79

-35. 00 15.00 29. 82 -14. 82

-25. 00 20. 00 37. 91 -17. 91

-25. 00 50. 00 37. 91 12. 09

-10. 00 70. 00 51. 25 18. 75

-5. 00 60. 00 55. 75 4. 25

.00 70. 00 60. 16 9. 84

5. 00 80. 00 64. 41 15. 59
25. 00 55. 00 78. 88 -23. 88 WCAP-17009-NP April 2009 Revision I

C- 103 Capsule Y Surveillance Program Weld Page 2 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: Y Fluence: n/cm"2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

40. 00 80. 00 86. 54 -6. 54
60. 00 80. 00 92. 99 -12. 99 too. 00 95. 00 98. 26 -3. 26 175. 00 100. 00 99. 88 12 275. 00 100. 00 100. 00 00 300. 00 tOO. O0 100. 00 00 Correlation Coefficient = .912 WCAP- 17009-NP April 2009 Revision 1

C-104 Capsule V Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:00 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 53.11 TO = -24.47 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -24.4 Plant: Vogde 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: V Fluence: n/cmA2 125 100 75 I1.

50 25 0 1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125. 00 5. 00 2.22 2. 78 100. 00 15. 00 5. 50 9. 50

-75. 00 15. 00 12. 98 2. 02

60. 00 10. 00 20. 78 -10. 78
50. 00 15. 00 27. 66 -12. 66
45. 00 15. 00 31. 58 -16. 58

-40. 00 80. 00 35. 78 44. 22

-25. 00 60. 00 49. 50 10. 50

-25. 00 25. 00 49. 50 -24. 50 WCAP-17009-NP April 2009 Revision 1

C-105 Capsule V Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-10. 00 65. 00 63.29 1 71

50. 00 95. 00 94. 29 71 100. 00 100. 00 99. 09 .91 150.00 too. O0 99.86 14 200. 00 100. 00 99.98 02 300. 00 100. 00 100. 00 00 Correlation Coefficient = .922 WCAP- 17009-NP April 2009 Revision I

C-106 Capsule X Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/22/2008 11:02 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 60.36 TO = 11.87 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 11.9 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: X Fluence: n/cm^2 125 100 75 U,

S 20 50 a.

25 0 -1------.--1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 100. 00 2. 00 2. 40 - . 40

- 50. 00 10. 00 11.40 -1.40 00 40. 00 40. 29 -. 29

10. 00 45. 00 48. 45 -3.45
25. 00 65. 00 60. 71 4. 29
50. 00 80. 00 77 .96 2. 04
75. 00 90. 00 89. 01 .99 100. 00 95. 00 94. 88 .12 125. 00 90. 00 97 70 -7.70 WCAP-17009-NP April 2009 Revision 1

C-107 Capsule X Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: X Fluence: n/cm^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 125. 00 90. 00 97 . 70 -7.70 150.00 95. 00 98.98 -3.98 150.00 98. 00 98.98 .98 175.00 100. 00 99. 55 .45 200. 00 100. 00 99. 80 20 225.00 100. 00 99.91 09 Correlation Coefficient = .995 WCAP- 17009-NP April 2009 Revision I

C-108 Capsule W Surveillance Program Weld CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/23/2008 10:08 AM Page 1 Coefficients of Curve I A = 50. B = 50. C = 51.08 TO = 15.91 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 16.0 Plant: Vogtle I Material: SAW Heat: Wire 83653 Orientation: NA Capsule: W Fluence: n/cm^2 125 100 I

U, 75 0

20 50 0.

25 0 I i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

50. 00 5. 00 7. 04 -2. 04
20. 00 15. 00 19. 69 -4. 69
20. 00 10. 00 19. 69 -9. 69
15. 00 25. 00 22. 97 2. 03
15. 00 25. 00 22. 97 2. 03
10. 00 25. 00 26. 61 -I . 61
10. 00 35. 00 26. 61 8. 39

-5. 00 30. 00 30. 61 - ,61

.00 45. 00 34. 91 10. 09 WCAP-17009-NP April 2009 Revision I

C-109 Capsule W Surveillance Program Weld Page 2 Plant: Vogtle 1 Material: SAW Heat: Wire 83653 Orientation: NA Capsule: W Fluence: n/cm12 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 1 0 00 40. 00 44. 24 -4.24

20. 00 50. 00 54. 00 -4.00
25. 00 60. 00 58.81 1.19 125.00 95. 00 98. 62 -3. 62 150.00 100. 00 99.48 .52 175. 00 100. 00 99. 80 20 Correlation Coefficient = .987 WCAP- 17009-NP April 2009 Revision 1

C-110 Unirradiated Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:26 PM Page 1 Coefficients of Curve 1 A = 68.1 B = 65.9 C = 81.18 TO = -32.59 D " O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 134.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-86.1 Deg F Temp@50 ft-lbs=-55.4 Deg F Plant: Vogfle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: Unirra Fluence: n/cmA2 300 250 r 200 L150 Lu 100 50 0 1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-180.00 4. 00 5.60 -1 60

-180. 00 6. 00 5. 60 40

-140. 00 4. 00 10.93 -6. 93

-140. 00 5. 00 10. 93 -5. 93

- 110. 00 17.00 19. 24 -2. 24

-110.00 10. 00 19. 24 -9. 24

-110.00 16. 00 19.24 -3. 24

-80. 00 25. 00 33.47 -8. 47

-80. 00 29. 00 33.47 -4.47 WCAP- 17009-NP April 2009 Revision 1

C-Ill Unirradiated Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Unirra Fluence: II/cmA2 Charpy V-Notch Data Temperature input CVN Computed CVN Differential

-80. 00 27. 00 33. 47 -6.47

-60.00 45. 00 46. 66 -1 .66

-60.00 60. 00 46. 66 13. 34

-60.00 42. 00 46. 66 -4. 66

-40. 00 29.00 62. t1 -33. 11

-40. 00 58. 00 62. 11 -4. 11

-40. 00 119. 00 62. 11 56. 89

-20. 00 112.00 78. 24 33. 76

-20. 00 60. 00 78. 24 -18. 24

-20. 00 79. 00 78 .24 - 76 00 76. 00 93. 23 17. 23 00 94. 00 93 .23 77 00 108.00 93. 23 14. 77 40 00 125. 00 115. 12 9. 88 40 00 95. 00 115.12 -20. 12 40 00 103.00 1 15. 12 -12. 12 80 00 147. 00 126.26 20. 74 80 00 113. 00 t26. 26 -13 .26 80 00 93. 00 126. 26 -33 .26 120 00 136. 00 131.00 5. 00 120 00 125. 00 131.00 -6. 00 120 00 136.00 13 1. 00 5. 00 220 00 140. 00 133.74 6. 26 220 00 126. 00 133.74 -7. 74 220. 00 t40. 00 133.74 6. 26 Correlation Coefficient = .939 WCAP- 17009-NP April 2009 Revision 1

C-112 Capsule U Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:29 PM Page 1 Coefficients of Curve I A = 65.6 B = 63.4 C = 114.08 TO = -33.51 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 129.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-105.9 Deg F Temp@50 ft-lbs=-62.1 Deg F Plant: Vogtle 1 Material: SAW Heat: 138805-1 Orientation: NA Capsule: U Fluence: n/cm^2 300 250 j200 Lx 150 0

0 100

/0 ~ ~_

0 50 0

n

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 125. 00 12.00 23. 43 -11. 43

- 75. 00 27. 00 43. 51 -16. 51

-75. 00 24. 00 43. 51 -19. 51

-75. 00 38. 00 43. 51 - 5. 51

-70. 00 67. 00 45. 98 21. 02

-70. 00 43. 00 45. 98 - 2. 98

-60. 00 96. 00 51 .13 44. 87

-50. 00 59. 00 56. 50 2. 50 00 62. 00 83. 70 -21. 70 WCAP-17009-NP April 2009 Revision 1

C-113 Capsule U Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: U Fluence: n/cm^2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

25. 00 89. 00 95.53 -6. 53
72. 00 128.00 111.77 16.23 150. 00 116.00 124. 11 -8.11 250. 00 132. 00 128. 13 3. 87 350. 00 138. 00 128. 85 9. 15 Correlation Coefficient = .909 WCAP-17009-NP April 2009 Revision I

C-114 Capsule Y Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:31 PM Page 1 Coefficients of Curve 1 A = 63.1 B = 60.9 C = 68.69 TO = -24.19 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 124.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-66.0 Deg F Temp@50 ft-lbs=-39.1 Deg F Plant: Vogde I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Y Fluence: n/cn^A2 300 250 A200 i.4 0

Lx10 U- 01 0

0 100 0

50 ý/0 0

0

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-100. 00 19. 00 14. 27 4. 73

-50. 00 28. 00 41. 23 -13. 23

-35. 00 15. 00 53 .59 -38. 59

-30. 00 75 .00 57 .96 17. 04

-25. 00 89. 00 62. 38 26. 62

-25. 00 78. 00 62. 38 15. 62

.00 65. 00 83. 70 -18. 70 2.5. 00 121. 00 100. 52 20. 48

45. 00 114. 00 109. 67 4. 33 WCAP-17009-NP April 2009 Revision I

C-115 Capsule Y Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

60. 00 78. 00 114.34 -36. 34 100. 00 134. 00 120. 81 13. 19 150. 00 109.00 123.24 -[4. 24 200. 00 114.00 123. 82 -9. 82 250. 00 141. 00 123.96 17. 04 300. 00 123.00 123.99 - .99 Correlation Coefficient =.869 WCAP- 17009-NP April 2009 Revision 1

C-116 Capsule V Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:34 PM Page 1 Coefficients of Curve I A = 61.6 B = 59.4 C = 81.29 TO = 3.07 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 121.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-45.1 Deg F Temp@50 ft-lbs=-13.0 Deg F Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: V Fluence: n/cm^2 300 250 S200

>6 0

U-Lx 0

150 z -o0 100 0 00 50 0

n, U

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 100. 00 3. 00 10. 92 -7. 92

- 75. 00 9. 00 17. 38 -8.38

- 50. 00 22. 00 27. 53 -5.53

-40. 00 32. 00 32. 77 - . 77

-35. 00 40. 00 35. 65 4.35

-25. 00 46. 00 41. 87 4. 13 00 70. 00 59. 36 10. 64

25. 00 104. 00 77. 25 26. 75 40 00 64. 00 86. 87 -22. 87 WCAP- 17009-NP April 2009 Revision 1

C-117 Capsule V Heat Affected Zone Page 2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

50. 00 76. 00 92. 53 -16.53
60. 00 82. 00 97. 51 -15.51 100. 00 136. 00 110. 98 25. 02 150. 00 120. 00 117.89 2. 11 225. 00 133. 00 t20. 50 12.50 300. 00 94. 00 120. 92 -26.92 Correlation Coefficient = .930 WCAP- 17009-NP April 2009 Revision 1

C-118 Capsule X Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 10/22/2008 10:53 A-M Page 1 Coefficients of Curve I A = 62.1 B = 59.9 C = 106.79 TO = -10.51 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy= 122.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-74.4 Deg F Temp@50 ft-lbs=-32.3 Deg F Plant: Vogde 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence: n/cinA2 300 250 200 a

L 150 w

100 50 0 r V i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 150. 00 3. 00 10.39 -7. 39

-100. 00 14. 00 21. 08 -7. 08

- 75. 00 23. 00 29. 77 -6. 77

-50. 00 58. 00 40.91 17. 09

-50. 00 41. 00 40.91 09

-25. 00 59. 00 54. 02 4. 98

25. 00 68. 00 81.31 -13. 31
75. 00 109 .00 101. 90 7 .10 100. 00 1 12. 00 108. 57 3 43 WCAP- 17009-NP April 2009 Revision 1

C-119 Capsule X Heat Aftected Zone Page 2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 100. 00 106.000 108.57 -2. 57 150. 00 102.00 116.35 -14. 35 200. 00 93. 00 119.72 -26. 72 200. 00 122.00 119.72 2.28 225. 00 172.00 120.56 51.44 225. 00 123.00 120. 56 2. 44 Correlation Coefficient = .931 April 2009 WCAP- 17009-NP WCAP- I17009-NP April 2009 Revision I

C-120 Capsule W Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:39 PM Page 1 Coefficients of Curve I A = 58.1 B = 55.9 C = 109.41 TO = 19.45 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=114.0(Fixed) Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-41.0 Deg F Temp@50 ft-lbs=3.5 Deg F Plant: Vogfe 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: W Fluence: n/cMA2 300 250 4200 0

LL Lm 150 aa 100 50 0  !ý. - - i + 4

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

- 150. 00 6. 00 7. 03 -1. 03

- 100. 00 22. 00 13.52 8.48

- 60. 00 34. 00 23. 40 10. 60

-50. 00 32. 00 26. 72 5.28

-40. 00 37. 00 30. 40 6. 60

-30. 00 18.00 34. 43 - 16. 43

-30. 00 32. 00 34. 43 -2.43

-10.00 42. 00 43. 41 -1.41

.00 52. 00 48. 27 3. 73 WCAP-17009-NP April 2009 Revision 1

C-121 Capsule W Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: W Fluence: n/cmA2 Charpy V-Notch Data Temperature Input CVN Computed CVN Differential 10.00 54. 00 53. 29 .71

25. 00 42. 00 60. 94 -18. 94
50. 00 85. 00 73. 32 1.1. 68 150. 00 106. 00 104. 59 1 .41 175. 00 104. 00 107. 85 -3. 85 200. 00 131.00 110.02 20. 98 Correlation Coefficient = .961 WCAP- 17009-NP April 2009 Revision 1

C-122 Unirradiated Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:26 PM Page 1 Coefficients of Curve I A = 39.56 B = 39.56 C = 70.03 TO = -41.62 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=79.1 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=-49.7 Deg F Plant: Vogdle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: Unirra Fluence: n/cmA2 200 150 0

Ia 100 50 0 4-

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 180. 00 1. 00 1. 49 -. 49 180. 00 1. O0 1. 49 -. 49 140. 00 2. 00 4. 49 -2.49 140. 00 3. 00 4. 49 -1.49 110. 00 9. 00 9. 83 -. 83 110. 00 5. 00 9. 83 -4.83 110. 00 9. 00 9. 83 -. 83

- 80. 00 15. 00 19. 81 -4. 81

-80. 00 18. 00 19. 81 -1.81 WCAP- 17009-NP April 2009 Revision 1

C-123 Unirradiated Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Unirra Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-80. 00 19.00 19. 81 -. 81

- 60. 00 28. 00 29. 40 -1.40

-60. 00 40. 00 29. 40 10. 60

-60. 00 29. 00 29. 40 - .40

-40. 00 20. 00 40. 47 20. 47

-40. 00 36. 00 40. 47 -4. 47

-40. 00 66. 00 40. 47 25. 53

-20. 00 67. 00 51. 39 15. 61

-20. 00 44. 00 51 .39 -7. 39

-20. 00 51.00 51. 39 - . 39 00 49. 00 60. 64 11. 64 00 61. 00 60. 64 36 00 70. 00 60. 64 9. 36

40. 00 74. 00 72. 10 1 .90
40. 00 64. 00 72. 10 -8. 10
40. 00 64. 00 72. 10 -8. 10
80. 00 81.00 76. 73 4. 27
80. 00 74. 00 76. 73 -2. 73
80. 00 70. 00 76. 73 -6. 73 120. 00 79. 00 78. 34 .66 120. 00 77. 00 78. 34 I-. 34 120. 00 78.00 78. 34 - . 34 2 20. 00 84. 00 79. 07 4. 93 220. 00 81.00 79 .07 1 .93 220. 00 87. 00 79. 07 7. 93 Correlation Coefficient = .964 April 2009 WCAP- I17009-NP 17009-NP April 2009 Revision I

C-124 Capsule U Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:29 PM Page 1 Coefficients of Curve 1 A = 37.17 B = 37.17 C = 120.7 TO = -40.57 D = 0.OOE+00 Equation is A + B * [Tanih((T-To)/(C+DT))]

Upper Shelf L.E.=74.3 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mnils=-47.6 Deg F Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: U Fluence: nrcmA2 200 150

.o 0

50 0 0 0

0 n

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.F. Differential

- 125. 00 8. 00 14. 72 -6. 72

-75. 00 14. 00 26. 85 -12. 85

-75. 00 16.00 26. 85 -10. 85

-75. 00 25. 00 26. 85 -1. 85

-70. 00 41.00 28.28 12. 72

-70. 00 30. 00 28. 28 1. 72

-60. 00 54. 00 31.24 22. 76

-50. 00 33. 00 34. 27 -1. 27 00 41.00 49.21 -8. 21 WCAP-17009-NP April 2009 Revision 1

C-125 Capsule U Heat Affected Zone Page 2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: U Fluence: n/crMA2 Charpy V-Notch Data Temperature Input L.E. Computed LE. Differential

25. 00 56. 00 55. 59 . 41
72. 00 69. 00 64. 37 4. 63 150. 00 58.00 71.31 -13.31 250. 00 82. 00 73. 74 8. 26 350. 00 76. 00 74. 22 1.78 Correlation Coefficient = .901 April 2009 WCAP- 17009-NP April 2009 Revision I

C-126 Capsule Y Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:32 PM Page 1 Coefficients of Curve 1 A = 36.5 B = 36.5 C = 74.19 TO = -37.31 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=73.0 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=-40.3 Deg F Plant: Vogfle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: Y Fluence: n/cinm2 200 150 Ew 0

.o C

& 100 50 0 4-

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 100. 00 14. 00 11. 37 2. 63

- 50. 00 21.00 30. 32 -9. 32

-35. 00 22. 00 37. 64 -15. 64

-30. 00 49. 00 40. 09 8. 91

- 25. 00 55. 00 42. 50 12. 50

-25. 00 52. 00 42. 50 9. 50

.00 43.00 53. 45 -10. 45

25. 00 69. 00 61. 53 7. 47
45. 00 69. 00 65. 84 3. 16 WCAP- 17009-NP April 2009 Revision I

C-127 Capsule Y Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Y Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

60. 00 51. 00 68. 06 -17.06 100. 00 77. 00 71.24 5.76 150. 00 71. 00 72.54 -1.54 200. 00 71.00 72. 88 -1. 88 250. 00 77. 00 72. 97 4. 03 300. 00 75. 00 72.99 2.01 Correlation Coefficient = .902 WCAP- 17009-NP April 2009 Revision 1

C-128 Capsule V Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:34 PM Page 1 Coefficients of Curve I A = 35.69 B = 35.69 C = 75.15 TO = -2.49 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=71.4 Lower Shelf L.E.=.0(Fixed)

Temnp.@L.E. 35 mils=-3.9 Deg F Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: V Fluence: n/cm^2 200 150 Em 0

.o a

50 V ,r 0

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

- 100. 00 1. 00 4. 96 -3. 96

-75. 00 4. 00 9. 05 -5. 05

-50. 00 13.00 15. 72 -2. 72

-40. 00 20. 00 19. 22 .78

-35. 00 23. 00 21. 15 1 .85

-25. 00 28. 00 25. 31 2. 69 00 41.00 36. 87 4. 13

25. 00 58. 00 48. 19 9. 81
40. 00 46. 00 53. 96 -7 96 WCAP- 17009-NP April 2009 Revision 1

C-129 Capsule V Heat Affected Zone Page 2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: V Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

50. 00 49.00 57.23 -8.23
60. 00 53. 00 60. 01 -7. 01 100. 00 77. 00 67. 00 10. 00 150. 00 75. 00 70. 17 4. 83 225. 00 75. 00 71.21 3.79 300. 00 63. 00 71.36 -8. 36 Correlation Coefficient = .970 WCAP-17009-NP April 2009 Revision 1

C-130 Capsule X Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:37 PM Page 1 Coefficients of Curve I A = 34.39 B = 34.39 C = 85.38 TO = -7.93 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=68.8 Lower Shelf L.E.=.0(Fixed)

Temp.@L.E. 35 mils=-6.4 Deg F Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence: n/cm^2 200 150 0

a.

E 100 U1 T J0° 50 0

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

-150. 00 00 2. 38 -2. 38

- 100. 00 1. 00 7. 13 -6. 13

-75. 00 7. 00 11. 83 -4. 83

-50. 00 28. 00 18. 69 9. 31

-50. 00 18. 00 18. 69 - . 69

-25. 00 30. 00 27. 60 2. 40

25. 00 43. 00 47. 03 -4. 03
75. 00 58. 00 60, 15 -2 15 100. 00 69. 00 63. 69 5. 31 WCAP- 17009-NP April 2009 Revision 1

C-131 Capsule X Heat Affected Zone Page 2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence: n/cm^2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential 100. 00 60. 00 63. 69 -3. 69 150. 00 70. 00 67. 11 2. 89 200. 00 62. 00 68.25 -6. 25 200. 00 72. 00 68. 25 3. 75 225. 00 71.00 68.48 2.52 225. 00 67. 00 68.48 -1.48 Correlation Coefficient = .986 April 2009 17009-NP WCAP- I17009-NP April 2009 Revision I

C-132 Capsule W Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:39 PM Page 1 Coefficients of Curve 1 A = 39.51 B = 39.51 C = 128.54 TO = 20.74 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=79.0 Lower Shelf L.E.=.O(Fixed)

Ternp.@L.E. 35 mils=6.1 Deg F Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: W Fluence: n/cMA^2 200 150

.o_

2100 l0 50 0 1 i i i1

-300.0 0.0 300.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E. Computed LE. Differential

-150. 00 5. 00 5. 18 -. 18

- 100. 00 13.00 10 47 2.53

-60. 00 22. 00 17. 51 4. 49

50. 00 20. 00 19. 72 .28
40. 00 26. 00 22. 11 3. 89
30. 00 13.00 24. 67 -11.67
30. 00 22. 00 24. 67 -2. 67 10.00 30. 00 30. 24 - . 24 00 40. 00 33. 19 6. 81 WCAP- 17009-NP April 2009 Revision 1

C-133 Capsule W Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: W Fluence: n/cmA2 Charpy V-Notch Data Temperature Input L.E. Computed L.E. Differential

10. 00 37. 00 36.21 .79
25. 00 31.00 40. 82 -9.82
50. 00 57.00 48. 35 8.65 150. 00 67. 00 69. 69 -2. 69 175. 00 73. 00 72. 45 .55 200. 00 75.00 74. 44 56 Correlation Coefficient = .971 WCAP- 17009-NP April 2009 Revision 1

C-134 Unirradiated Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:27 PM Page I Coefficients of Curve I A = 50. B = 50. C = 75.25 TO = -24.25 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -24.2 Plant: Vogfle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Unirra Fluence: n/cm^2 125 100 75 50 a.

25

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-180. 00 00 1 57 -1 57

-180. 00 00 1 57 -1 57

- 140. 00 00 4.41 -4. 41

- 140. 00 00 4.41 -4. 41

-110. 00 5. 00 9. 29 -4. 29

-110. 00 5.00 9. 29 -4. 29

-110.00 5.00 9. 29 -4. 29

- 80. 00 10. 00 18. 52 -8 .52

- 80. 00 25. 00 18. 52 6. 48 April 2009 WCAP- 17009-NP WCAP-17009-NP April 2009 Revision I

C-135 Unirradiated Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: Unirra Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 80. 00 18.00 18. 52 - .52

-60. 00 20. 00 27. 88 -7.88

- 60. 00 34. 00 27. 88 6.12

-60. 00 20. 00 27. 88 -7.88

-40. 00 25. 00 39. 68 14.68

-40. 00 30. 00 39 .68 -9.68

-40. 00 80. 00 39 .68 40. 32

- 20. 00 80. 00 52. 82 27. 18

-20. 00 35. 00 52. 82 17. 82

-20. 00 65. 00 52. 82 12. 18 00 65. 00 65. 58 - .58 00 60. 00 65.58 -5.58 00 56. 00 65.58 -9.58

40. 00 100. 00 84. 65 15.35
40. 00 65. 00 84. 65 19.65
40. 00 75. 00 84. 65 -9. 65
80. 00 100. 00 94.11 5. 89
80. 00 90. 00 94.11 -4. 11
80. 00 90. 00 94.11 -4. 1 1.

120. 00 100. 00 97.88 2. 12 120. 00 100.00 97. 88 2 .12 120. 00 100. 00 97.88 2. 12 220. 00 too. 00 99. 85 15 220. 00 100. 00 99. 85 15 220. 00 too. oo 99. 85 15 Correlation Coefficient = .954 WCAP- 17009-NP April 2009 Revision I

C-136 Capsule U Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:29 PM Page 1 Coefficients of Curve I A = 50. B = 50. C = 88.26 TO = -44.61 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -44.6 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: U Fluence: n/cm^2 125 100 I

U, 75 0

2 50 0.

25 0 T- -- -f - l . - -

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 125. 00 10.00 13. 93 -3.93

- 75. 00 15. 00 33. 44 -18.44

- 75. 00 15. 00 33. 44 -18.44

-75. 00 30. 00 33. 44 -3.44

-70. 00 55. 00 36. 00 19. 00

-70. 00 30. 00 36. 00 -6.00

-60. 00 70. 00 41. 37 28. 63

- 50. 00 55. 00 46. 95 8. 05

.00 60. 00 73. 32 -13. 32 WCAP-17009-NP April 2009 Revision 1

C-137 Capsule U Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: U Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

25. 00 85. 00 82.88 2. 12
72. 00 90. 00 93. 35 -3.35 150. 00 too. 00 98. 80 1.20 250. 00 too. 00 99. 87 .13 350. 00 100. 00 99.99 01 Correlation Coefficient = .923 WCAP- 17009-NP April 2009 Revision 1

C-138 Capsule Y Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:32 PM Page I Coefficients of Curve I A = 50. B = 50. C = 76.42 TO = -18.28 D = 0.00E+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -18.2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: Y Fluence: n/cm^2 125 100 I

U) 75 50 0..

25 0 1 -, I i

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 100. 00 5. 00 10. 54 -5. 54

- 50. 00 15.00 30. 36 -15. 36

-35. 00 20. 00 39. 23 -19. 23

-30. 00 55. 00 42. 39 12. 61.

- 25. 00 65. 00 45. 62 19. 38

-25. 00 60. 00 45. 62 14. 38 00 50. 00 61. 74 -11. 74

25. 00 85. 00 75. 63 9. 37
45. 00 85. 00 83. 97 1 .03 WCAP- 17009-NP April 2009 Revision I

C-139 Capsule Y Heat Affected Zone Page 2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: Y Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

60. 00 65. 00 88. 58 -23.58 100.00 100. 00 95. 67 4. 33 1.50. 00 100. 00 98. 79 1 . 2t 200. 00 100.00 99. 67 .33 250. 00 t00. 00 99. 91 09 300. 00 100. 00 99. 98 02 Correlation Coefficient = .927 WCAP- 17009-NP April 2009 Revision 1

C-140 Capsule V Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:34 PM Page I Coefficients of Curve I A = 50. B = 50. C = 85.42 TO = -1.3 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -1.2 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: V Fluence: nrcmA2 125 100 75 0

0 50 a-25

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 100. 00 10. 00 9. 02 .98

- 75. 00 20. 00 15.11 4. 89

-50. 00 25 00 24. 23 .77

-40. )0 40. 00 28.78 11.22

-35. 00 25. 00 31.24 -6. 24

- 25. 00 30. 00 36. 47 -6. 47 00 40. 00 50. 76 -10. 76

25. 00 85. 00 64. 92 20. 08
40. 00 60. 00 72. 45 -12.45 WCAP-17009-NP April 2009 Revision I

C-141 Capsule V Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: V Fluence: n/cm^2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Stear Differential

50. 00 75. 00 76. 87 -1. 87
60. 00 80. 00 80. 77 -. 77 100. 00 tO0. 00 91.47 8. 53 150.00 100. 00 97. 19 2.81 225. 00 tOO. oo 99.50 50 300. 00 100. 00 99.91 09 Correlation Coefficient = .969 WCAP- 17009-NP April 2009 Revision 1

C-142 Capsule X Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:37 PM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 82.03 TO = -7.57 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -7.5 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence: n/cmA2 125 100 75 U) 50 25

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 150. 00 2. 00 3.01 -1.01

-100. 00 5. 00 9. 50 -4.50

-75. 00 15. 00 16. 19 -1.19

-50. 00 30. 00 26. 22 3.78

-50. 00 25. 00 26. 22 -1.22

-25. 00 40. 00 39. 53 .47

25. 00 70. 00 68. 87 1.13
75. 00 90. 00 88. 22 1.78 100. 00 90. 00 93. 23 -3.23 WCAP-17009-NP April 2009 Revision I

C-143 Capsule X Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 100. 00 90. 00 93.23 -3. 23 150. 00 95. 00 97. 90 -2. 90 200. 00 100.00 99.37 63 200. 00 t00. 00 99.37 63 225. 00 100. 00 99.66 34 225. 00 100. 00 99. 66 34 Correlation Coefficient = .998 WCAP-17009-NP April 2009 Revision 1

C-144 Capsule W Heat Affected Zone CVGRAPH 5.3 Hyperbolic Tangent Curve Printed on 09/19/2008 01:40 PM Page 1 Coefficients of Curve I A = 50. B = 50. C = 104.87 TO = 25.21 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 25.3 Plant: Vogtle 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: W Fluence: n/cm^2 125 100 75 U) 50 25 0 1-1 1

-300.0 -200.0 -100.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

- 150. 00 5. 00 3 42 1 58

-100. 00 10. 00 8.41 1 59

- 60. 00 20. 00 16. 45 3 55

- 50. 00 25. 00 19. 24 5. 76

-40. 00 25. 00 22. 38 2. 62

- 30. 00 20. 00 25. 87 -5. 87

-30. 00 30. 00 25. 87 4. 13

-10.00 25. 00 33. 82 -8. 82

.00 50. 00 38 .21 11 .79 WCAP-17009-NP April 2009 Revision 1

C-145 Capsule W Heat Affected Zone Page 2 Plant: Vogtle I Material: SAW Heat: B8805-1 Orientation: NA Capsule: W Fluence: n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

10. 00 40. 00 42. 80 -2. 80
25. 00 40. 00 49. 90 -9. 90
50. 00 60. 00 61. 61 -I .61 150. 00 100. 00 91.53 8. 47 175. 00 100. 00 94. 57 5. 43 200. 00 100. 00 96. 56 3. 44 Correlation Coefficient = .983 April 2009 WCAP- I17009-NP 7009-NP April 2009 Revision I

D-1 APPENDIX D VOGTLE UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION D.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [D-l] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there have been five surveillance capsules removed from the Vogtle Unit 1 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Vogtle Unit 1 reactor vessel surveillance data and determine if that surveillance data is credible.

D.2 EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements" [D-2], as follows:

"the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predictedto experience sufficient neutron radiationdamage to be consideredin the selection of the most limiting materialwith regard to radiationdamage."

The Vogtle Unit 1 reactor vessel consists of the following beltline region materials:

  • Intermediate Shell Plates B8805-1, 2, and 3 (Heat # C0613-1, C0613-2, and C0623-1)

" Lower Shell Plates B8606-1, 2, and 3 (Heat # C2146-1, C2146-2, and C2085-2)

  • Intermediate Shell Longitudinal Weld Seams 101-124A, B & C (Heat # 83653)

" Lower Shell Longitudinal Weld Seams 101-142A, B & C (Heat # 83653)

  • Circumferential Weld Seam 101-171 (Heat # 83653)

WCAP- 17009-NP April 2009 Revision 1

D-2 At the time when the Vogtle Unit 1 surveillance program material was selected it was believed that copper and phosphorous were the elements most important to embrittlement of the reactor vessel steels.

Intermediate Shell Plate B8805-3 had the highest initial RTNDT and one of the lowest initial USE of all plate materials in the beltline region. In addition, Intermediate Shell Plate B8805-3 had approximately the same copper and phosphorous content as the other beltline plate materials. Therefore, based on the highest initial RTNDT and one of the lowest USE, Intermediate Shell Plate B8805-3 was chosen for the surveillance program.

The weld material in the Vogtle Unit 1 surveillance program was made of weld wire (Heat # 83653) identical to all the beltline weld seams. Thus, it was chosen as the surveillance weld material.

Hence, Criterion 1 is met for the Vogtle Unit 1 reactor vessel.

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and USE unambiguously.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the USE of the Vogtle Unit 1 surveillance materials unambiguously. Hence, the Vogtle Unit 1 surveillance program meets this criterion.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in USE if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [D-3].

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 28°F for welds and less than 17'F for the plate.

The Vogtle Unit I intermediate shell plate and surveillance weld material will be evaluated for credibility.

The weld is made from weld wire heat 83653; Vogtle Unit I does not have a sister plant that shares the same weld wire heat and thus, does not utilize data from other surveillance programs. Therefore, the method of Regulatory Guide 1.99, Revision 2 will be followed for determining credibility of the weld as well as the plate material.

Credibility Assessment:

Since all surveillance data is from one vessel (Vogtle Unit 1), the measured ARTNDT and fluence factor (FF) should be used to calculate the chemistry factor to determine if the Vogtle Unit 1 surveillance material test results are credible.

WCAP- 17009-NP April 2009 Revision 1

D-3 The chemistry factors for the Vogtle Unit 1 surveillance plate and weld material contained in the surveillance program were calculated in accordance with Regulatory Guide 1.99, Revision 2, Position 2.1 and presented in Table D-1. The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2.

Table D-1 Calculation of Chemistry Factors using Vogtle Unit 1 Surveillance Capsule Data Capsule f ARTNDT FF*ARTNDT Material Capsule (x1019 n/cm 2) FF (OF) (OF) FF' U 0.332 0.697 13.3 9.26 0.485 Intermediate Shell Y 1.14 1.037 32.4 33.59 1.075 Plate B8805-3 V 1.93 1.180 42.4 50.02 1.392 (Longitudinal)

X 3.47 1.325 96.6 127.96 1.755 W 4.36 1.375 98.1 134.84 1.889 U 0.332 0.697 -9.8 -6.83 0.485 Intermediate Shell 1.14 1.037 14.7 15.24 1.075 Plate B8805-3 (Transverse) V 1.93 1.180 33.5 39.52 1.392 X 3.47 1.325 60.8 80.54 1.755 W 4.36 1.375 94.1 129.35 1.889 SUM: 613.49 13.191 CFB88 05- 3 = Z(FF

  • ARTNDT) + Z( FF 2) -(613.49) + (13.191) = 46.50 F U 0.332 0.697 25.1 17.48 0.485 Y 1.14 1.037 11.4 11.82 1.075 Surveillance V 1.93 1.180 -0.3 -0.35 1.392 Weld Material _____

X 3.47 1.325 55.4 73.38 1.755 W 4.36 1.375 34.8 47.83 1.889 SUM: 150.17 6.596 CF Su*. Weld = Y'7(FF

  • ARTNDT) + Y( FF 2) = (150.17) + (6.596) = 22.8 0 F WCAP-17009-NP April 2009 Revision 1

D-4 Table D-2 Vogtle Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line CF Measured Predicted Scatter <17 0 F (Slopebest fit) Capsule f ARTNDT ARTNDT ARTNDT (Base Metal) 9 2 Material Capsule (OF) (xl01 n/cm ) FF (OF) (OF) (OF) <28 0 F (Weld)

Intermediate U 46.5 0.332 0.697 13.3 32.4 19.1 No Shell Plate Y 46.5 1.14 1.037 32.4 48.2 15.8 Yes B8805-3 V 46.5 1.93 1.180 42.4 54.9 12.5 Yes (Longitudinal) X 46.5 3.47 1.325 96.6 61.6 35.0 No W 46.5 4.36 1.375 98.1 63.9 34.2 No Intermediate U 46.5 0.332 0.697 -9.8 32.4 42.2 No Shell Plate Y 46.5 1.14 1.037 14.7 48.2 33.5 No B8805-3 V 46.5 1.93 1.180 33.5 54.9 21.4 No (Transverse) X 46.5 3.47 1.325 60.8 61.6 0.8 Yes W 46.5 4.36 1.375 94.1 63.9 30.2 No U 22.8 0.332 0.697 25.1 15.9 9.2 Yes Surveillance Y 22.8 1.14 1.037 11.4 23.6 12.2 Yes Weld Material V 22.8 1.93 1.180 -0.3 26.9 27.2 Yes X 22.8 3.47 1.325 55.4 30.2 25.2 Yes W 22.8 4.36 1.375 34.8 31.3 3.5 Yes Table D-2 indicates that seven of the ten surveillance data points fall outside the +- lcy of 17F scatter band for surveillance base metals, therefore the plate data is deemed "not credible" per the third criterion.

Table D-2 indicates that five of the five surveillance data points are within the +/- 1cF of 28°F scatter band for the surveillance weld material, therefore the surveillance weld data is deemed "credible" per the third criterion.

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The capsule specimens are located in the reactor between the neutron pad and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad.

The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.

April 2009 WCAP- 17009-NP 17009-NP April 2009 Revision 1

D-5 Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The Vogtle Unit 1 surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to the Vogtle surveillance program.

D.3 CONCLUSION Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B, the Vogtle Unit 1 surveillance data is deemed credible for the weld specimens and non-credible for the plate specimens.

D.4 REFERENCES D-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1998.

D-2 10 CFR 50, Appendix G, Fracture Toughness Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.

D-3 ASTM E185-82, StandardPracticefor Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society for Testing and Materials.

WCAP- 17009-NP April 2009 Revision 1

E-1 APPENDIX E VOGTLE UNIT 1 UPPER SHELF ENERGY EVALUATION Per Regulatory Guide 1.99, Revision 2 [E-1], the Charpy upper shelf energy is assumed to decrease as a function of fluence and copper content as indicated in Figure 2 of the Guide (Figure E- 1 of this appendix) when surveillance data is not used. Linear interpolation is permitted. In addition, if surveillance data is to be used, the decrease in upper shelf energy may be obtained by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure E- 1 of this appendix) and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.

The 36 EFPY (end-of-license) and 54 EFPY (end-of-license renewal) upper shelf energy of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the beltline materials and/or the results of the capsules tested to date using Figure 2 in Regulatory Guide 1.99, Revision 2. The maximum vessel clad/base metal interface fluence value was used to determine the corresponding 1/4T fluence value at 36 and 54 EFPY The Vogtle Unit I reactor vessel beltline region minimum thickness is 8.625 inches. Calculation of the 1/4T vessel surface fluence values at 36 and 54 EFPY for the beltline materials is shown as follows:

Maximum Vessel Fluence @ 36 EFPY = 2.155 x 1019 n/cm 2 (E > 1.0 MeV) 2 0 24 1/4T Fluence @ 36 EFPY (2.155 x 1019 n/cm )

  • e(- . * (8.625/4))

= 1.284 x 1019 n/cm 2 (E > 1.0 MeV)

Maximum Vessel Fluence @ 54 EFPY = 3.30 x 1019 n/cm 2 (E > 1.0 MeV) 1/4T Fluence @ 54 EFPY (3.30 x 2

1019 n/cm )

  • e(-°0 2 4 * (8.625 / 4))

= 1.967 x 10'9 n/cm 2 (E > 1.0 MeV)

The following pages present the Vogtle Unit 1 upper shelf energy evaluation. Figure E- 1, as indicated above, is used in making predictions in accordance with Regulatory Guide 1.99, Revision 2. Table E-I provides the predicted upper shelf energy values for 36 EFPY (end-of-license). Table E-2 provides the predicted upper shelf energy values for 54 EFPY (end-of-license renewal).

WCAP- 17009-NP April 2009 Revision I

E-2 a Surveillance Material: Intermediate Shell Plate B8805-3 A Surveillance Material: Circumferential Weld 101-171 100.0 10.0 1.0 1.OOE+17 1.OOE+18 1.OOE+19 1.OOE+20 Neutron Fluence, n/cm 2 (E > 1 MeV)

Figure E-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper Shelf Energy as a Function of Copper and Fluence WCAP- 17009-NP April 2009 Revision 1

E-3 Table E-1 Predicted Positions 1.2 and 2.2 Upper Shelf Energy Values at 36 EFPY I/4T EOL Fluence Unirradiated Projected Projected Weight (x 1019 n/cm 2, USE USE Decrease EOL USE Material  % of Cu E > 1.0 MeV) (ft-lb) (%) (ft-lb)

Position 1.2 Intermediate Shell Plate B8805-1 0.083 1.284 90 20 72.0 Intermediate Shell Plate B8805-2 0.083 1.284 100 20 80.0 Intermediate Shell Plate B8805-3 0.062 1.284 107 20 85.6 Lower Shell Plate B8606-1 0.053 1.284 116 20 92.8 Lower Shell Plate B8606-2 0.057 1.284 113 20 90.4 Lower Shell Plate B8606-3 0.067 1.284 118 20 94.4 Intermediate Shell Longitudinal 0.042 0.41.284(a) 124a)1420 134 107.2 Weld Seam 101-124A, B & C Lower Shell Longitudinal Weld 0.042 1.284(a) 134 20 107.2 Seam 101-142A, B & C Circumferential Weld 101-171 0.042 1.284 134 20 107.2 Position 2 .2(b)

Intermediate Shell Plate B8805-3 0.062 1.284 107 14 92.0 Intermediate Shell Longitudinal 0.042 1284(a) 002128a)134 12 117.9 Weld Seams 101-124A, B & C Lower Shell Longitudinal Weld 0.042 1.284(a) 134 12 117.9 Seams 101-142A, B & C Circumferential Weld 101-171 0.042 1.284 134 12 117.9 Notes:

(a) The fluence values listed for Intermediate Shell Longitudinal Weld Seams 101-124A, B & C and Lower Shell Longitudinal Weld Seams 101-142A, B & C conservatively pertain to the maximum vessel fluence value (450 location), though the welds vary in location.

(b) Calculated using surveillance capsule measured percent decrease in USE from Table 5-10 and Regulatory Guide 1.99, Revivion 2, Position 2.2; see Figure E-1.

WCAP- 17009-NP April 2009 Revision 1

E-4 Table E-2 Predicted Positions 1.2 and 2.2 Upper Shelf Energy Values at 54 EFPY 1/4T EOLR Projected Fluence Unirradiated USE Projected Weight (x 1019 n/cm 2, USE Decrease EOLR Material  % of Cu E > 1.0 MeV) (ft-lb) (%) USE (ft-lb)

Position 1.2 Intermediate Shell Plate B8805-1 0.083 1.967 90 22 70.2 Intermediate Shell Plate B8805-2 0.083 1.967 100 22 78.0 Intermediate Shell Plate B8805-3 0.062 1.967 107 22 83.5 Lower Shell Plate B8606-1 0.053 1.967 116 22 90.5 Lower Shell Plate B8606-2 0.057 1.967 113 22 88.1 Lower Shell Plate B8606-3 0.067 1.967 118 22 92.0 Intermediate Shell Longitudinal 0.042 1.967(a, 134 22 104.5 Weld Seam 101-124A, B & C Lower Shell Longitudinal Weld 0.042 1.967(a) 134 22 104.5 Seam 101-142A, B & C Circumferential Weld 101-171 0.042 1.967 134 22 104.5 Position 2 .2(b)

Intermediate Shell Plate B8805-3 0.062 1.967 107 16 89.9 Intermediate Shell Longitudinal 0.042 1.967(al 134 14 115.2 Weld Seams 101-124A, B & C Lower Shell Longitudinal Weld 0.042 1.967(a) 134 14 115.2 Seams 101-142A, B & C Circumferential Weld 101-171 0.042 1.967 134 14 115.2 Notes:

(a) The fluence values listed for Intermediate Shell Longitudinal Weld Seams 101-124A, B & C and Lower Shell Longitudinal Weld Seams 101-142A, B & C conservatively pertain to the maximum vessel fluence value (450 location),

though the welds vary in location.

(b) Calculated using surveillance capsule measured percent decrease in USE from Table 5-10 and Regulatory Guide 1.99, Revivion 2, Position 2.2; see Figure E-1.

E.1 REFERENCES E-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials,May 1988.

WCAP- 17009-NP ADril 2009 Revision I

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