ML050840318

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Enclosure 1, WCAP-16278-NP, Revision 0, Analysis of Capsule X from the Southern Nuclear Operating Company, Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program
ML050840318
Person / Time
Site: Vogtle Southern Nuclear icon.png
Issue date: 07/02/2004
From: Conermann J, Ghergurovich J, Hagler R, Knight K
Westinghouse
To:
Office of Nuclear Reactor Regulation
References
WCAP-16278-NP, Rev. 0
Download: ML050840318 (198)


Text

Enclosure 1 Vogtle Electric Generating Plant WCAP-16278-NP, Rev. 0

Westinghouse Non-Proprietary Class 3 WCAP-1 6278-NP Revision 0 July 2004 Analysis of Capsule X from the Southern Nuclear Operating Company, Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program

  • Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16278-NP, Revision 0 Analysis of Capsule X from the Southern Nuclear Operating Company, Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program KG Knight R. J. Hagler J. Conermann July 2004 Approved:~

4~4.

J. Gh'gurovich, Manager Reactor Component Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02004 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES.........

iv LIST OF FIGURES..............

vi PREFACE viii EXECUTIVE

SUMMARY

.ix I

SUMMARY

OF RESULTS.1-2 INTRODUCTION.2-1 3

BACKGROUND.3-I 4

DESCRIPTION OF PROGRAM.4-1 5

TESTING OF SPECIMENS FROM CAPSULE X 5-1 5.1 OVERVIEW.5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS.5-3 5.3 TENSILE TEST RESULTS.5-5 5.4 1/2T COMPACT TENSION SPECIMEN TESTS.5-5 6

RADIATION ANALYSIS AND NEUTRON DOSIMETRY

.6-1

6.1 INTRODUCTION

.6-6.2 DISCRETE ORDINATES ANALYSIS.6-2 6.3 NEUTRON DOSIMETRY.6-5 6.4 CALCULATIONAL UNCERTAINTIES.6-6 7

SURVEILLANCE CAPSULE REMOVAL SCHEDULE.7-1 8

REFERENCES.........

8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS.A-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS.B-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR CAPSULE X USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD.C-0 APPENDIX D VOGTLE UNIT I SURVEILLANCE PROGRAM CREDIBILITY EVALUATION.

D-0

iv LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the Vogtle Unit I Reactor Vessel Surveillance Materials (Unirradiated).....................................................

4-3 Table 4-2 Heat Treatment History of the Vogtle Unit 1 Reactor Vessel Surveillance Materials..............

4-4 Table 5-1 Charpy V-Notch Data for the Vogtle Unit I Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 3.53 x 1019 n/cm2 (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 of 3.53 x I o19 n/cm2 (E > 1.0 MeV)

(Transverse Orientation)...............

5-7 Table 5-3 Charpy V-notch Data for the Vogtle Unit I Surveillance Weld Material Irradiated to a Fluence of 3.53 x i0' 9

n/cm2 (E> 1.0 MeV)........................................

5-8 Table 54 Charpy V-notch Data for the Vogtle Unit I Heat-Affected-Zone (HAZ)

Material Irradiated to a Fluence of 3.53 x 1O'9 n/cm2 (E> 1.0 MeV)............................... 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 3.53 x 1019 n/cm2 (E> 1.0 MeV)

(Longitudinal Orientation).................

5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Vogtle Unit I Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 3.53 x 1019 n/cm2 (E> 1.0 MeV)

(Transverse Orientation)...............

5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Vogtle Unit I Surveillance Weld Metal Irradiated to a Fluence of 3.53 x IO' 9 n/cm 2 (E> 1.0 MeV)....................... 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Vogtle Unit I Heat-Affected-Zone (HAZ) Metal Irradiated to a Fluence of 3.53 x 1019 n/cm2 (E> 1.0MeV)............ 5-13 Table 5-9 Effect of Irradiation to 3.53 x 1019 n/cm2 (E> 1.0 MeV) on the Capsule X Notch Toughness Properties of the Vogtle Unit I Reactor Vessel Surveillance 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, Predictions..................

5-15

I ti V

LIST OF TABLES (Cont.)

Table 5-11 Tensile Properties of the Vogtle Unit I Capsule X Reactor Vessel Surveillance Materials Irradiated to 3.53 x 10'9 n/cm2 (E> 1.0MeV).......................................

5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center..................

6-12 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface....................................... 6-16 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall............

6-20 Table 64 Relative Radial Distribution Of Iron Atom Displacements (dpa) Within The Reactor Vessel Wall............

6-20 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Alvin W. Vogtle Unit 1.6-21 Table 6-6 Calculated Surveillance Capsule Lead Factors.6-21 Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule.7-1

vi LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Vogtle Unit I Reactor Vessel.................. 4-5 Figure 4-2 Capsule X Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters.4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit I 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 I Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation).

5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit I 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 Vogtle Unit 1 Reactor Vessel Weld Metal.

5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit I Reactor Vessel Weld Metal.

5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I Reactor Vessel Weld Metal.

5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material.

5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material.

5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I Reactor Vessel Heat-Affected-Zone Material.

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

vii LIST OF FIGURES (Cont.)

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 Vogtle Unit I Reactor Vessel Weld Metal..........

5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit I Reactor Vessel Heat-Affected-Zone Metal.....................

5-32 Figure 5-17 Tensile Properties for Vogtle Unit 1 Reactor Vessel Lower Shell Plate B8805-3 (Longitudinal Orientation)...........................

5-33 Figure 5-18 Tensile Properties for Vogtle Unit 1 Reactor Vessel Lower Shell Plate B8805-3 (Transverse Orientation).5-34 Figure 5-19 Tensile Properties for Vogtle Unit 1 Reactor Vessel Weld Metal.

5-35 Figure 5-20 Fractured Tensile Specimens from Vogtle Unit I 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 Vogtle Unit I Reactor Vessel Weld Metal.

5-38 Figure 5-23 Engineering Stress-Strain Curves for Vogtle Unit 1 Lower Shell Plate B8805-3, Capsule X, Tensile Specimens AL-1 0, AL-I and AL-12 (Longitudinal Orientation)............................................

5-39 Figure 5-24 Engineering Stress-Strain Curves for Vogtle Unit I Lower Shell Plate B8805-3, Capsule X, Tensile Specimens AT-1 0, AT-1I and AT-12 (Transverse Orientation)..............

5-41 Figure 5-25 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens AW-10, ANV-1l andAW -12.........................................................................................................543 Figure 6-1 Alvin NV. Vogtle Unit I rO Reactor Geometry with a 12.50 Neutron Pad at the Core Midplane.6-8 with a 20.0° Neutron Pad at the Core Midplane.6-9 with a 22.50 Neutron Pad at the Core Midplane.6-10 Figure 6-2 Alvin W. Vogdle Unit I rz Reactor Geometry with Neutron Pad.6-1

viii PREFACE This report has been technically reviewed and verified by:

Reviewer:

Sections 1-5, 7, 8, and Appendices B, C, and D Section 6 and Appendix A T. J. Laubham S. L. Anderson 45

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule X from Vogtle Unit 1. Capsule X was removed at 14.33 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI data-base. Capsule X received a fluence of 3.53 x IO"1 n/cm2 (E > 1.0 MeV) after irradiation to 14.33 EFPY. The peak clad/base metal interface vessel fluence after 14.33 EFPY of plant operation was 8.38 x 108 n/cm2 (E >

1.0 MeV).

This evaluation lead to the following conclusions: 1) The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Vogtle Unit I surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions. 2) 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 date of 54 EFPY as required by 1 OCFR50, Appendix G [2]. 3) The Vogtle Unit I surveillance plate data is not credible but the weld data is credible. This evaluation can be found in '

Appendix D.

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.

1-1 1

SUMMARY

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

The Charpy V-notch data presented in herein are based on a re-plot of unirradiated data from WCAP-1101113 ] and STC Report (Capsule X)171. The Charpy plots were developed using CVGRAPH Version 5.0.2, which is a symmetric hyperbolic tangent curve-fitting program. The irradiated capsule data from WCAP-1225614 3 (Capsule U) and WCAP-13931, Rev. 11[5 (Capsule Y) are documented in Appendix C of WCAP-1506716 1 (Capsule V). The results presented in Section 5 are only for the Capsule X test results, which are also based on using CVGRAPH, Version 5.0.2.

Capsule X received an average fast neutron fluence (E> 1.0 MeV) of 3.53 xl019 n/cm2 after 14.33 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 81.61F and an irradiated 50 ft-lb transition temperature of 121.7 0F. This results in a 30 ft-lb transition temperature increase of 96.50F and a 50 ft-lb transition temperature increase of 99.81F 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 77.91F and an irradiated 50 ft-lb transition temperature of 135.27F. This results in a 30 ft-lb transition temperature increase of 60.81F and a 50 ft-lb transition temperature increase of 72.71F for the transverse oriented specimens.

Irradiation of the weld metal (heat number 83653) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-3.8°F and an irradiated 50 ft-lb transition temperature of 12.8°F. This results in a 30 ft-lb transition temperature increase of 53.4°F and a 50 ft-lb transition temperature increase of 43.1 'F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-76.4°F and an irradiated 50 ft-lb transition temperature of -30.8°F. This results in a 30 ft-lb transition temperature increase of 10.6°F and a 50 ft-lb transition temperature increase of 24.6°F.

The average upper shelf energy of the Intermediate Shell Plate B8805-3 (longitudinal orientation) resulted in an average energy decrease of 13 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 109 ft-lb for the longitudinal oriented specimens.

Summary of Results

1-2 The average upper shelf energy of the Intermediate Shell Plate B8805-3 (transverse orientation) resulted in an average energy decrease of 3 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 93 ft-lb for the longitudinal oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 141 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 9 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 127 ft-lb for the weld HAZ metal.

A comparison of the measured 30 ft-lb shift in transition temperature values for the Vogtle Unit I reactor vessel surveillance materials is presented in Table 5-10.

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 IOCFR50, Appendix G [2J*

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

The calculated 36 EFPY (end-of license) and 54 EFPY neutron fluence (E> 1.0 MeV) at the core mid-plane for the Vogtle Unit I 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* = 2.03 x 1019 n/cm 2 (Interpolated From Table 6-2)

Vessel 1/4 thickness= 1.21 x 0I"n/cm2 Vessel 3/4 thickness = 4.30 x 10Os n/cm2 Calculated (54 EFPY):

Vessel inner radius* = 3.03 x 1019 n/cm2 Vessel 1/4 thickness = 1.81 x 1019n/cm 2 Vessel 3/4 thickness = 6.41 x 1018 n/cm2

  • Clad/base metal interface.

Summary of Results

2-1 2

INTRODUCTION This report presents the results of the examination of Capsule X, the fourth 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 I 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-1I01 1, "Georgia Power Company Alvin NV. Vogtle Unit No. I Reactor Vessel Radiation Surveillance Program" 131. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM El 85-82, "Standard Recommended Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Reactor Vessels."l201 Capsule X was removed from the reactor after 14.33 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 and the post-irradiation data obtained from surveillance Capsule X removed from the Southern Nuclear Operating Company Vogtle Unit I reactor vessel and discusses the analysis of the data.

Introduction

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 I (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 1[0]. 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 E-2081 9]) or the temperature 601F 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 (Kk, curve) which appears in Appendix G to the ASME Coderl']. The Ki, 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 Kic 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 I reactor vessel radiation surveillance programs1, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens 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 KI, 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.

Background

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 start-up. 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 Intermediate Shell Plate B8805-3, and weld metal fabricated with 3/16-inch Mil B4 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.

Capsule X was removed after 14.33 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2T-CT fracture mechanics specimens made from Intermediate Shell Plate B8805-3 and submerged arc weld metal representative of the intermediate shell longitudinal weld seams. In addition, this capsule contained Charpy V-notch specimens from the weld Heat-Affected-Zone (HAZ) metal of 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 1/4 and 3/4 thickness locations of the plate after performing a simulated post-weld stress-relieved 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 with the long dimension of the specimen perpendicular to the weld direction.

Compact tension test specimens from Intermediate Shell Plate B8805-3 were machined in the longitudinal and transverse orientations. Compact tension test specimens from the weld metal were machined perpendicular to the weld direction with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399.

Description of Program

4-2 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-110 113,Appendix A.

Capsule X 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 (Np2 3 7) and Uranium (U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.

The capsule contained thermal monitors made from two low-melting-point eutectic alloys and 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: 5790F (3041C) 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590'F (31 0C)

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

Description of Program

4-3 Table 4-1 Chemical Composition (WVt%) of the Vogtle Unit 1 ReactorVessel Surveillance Materials (Unirradiated)(')

Element Intermediate Shell Plate B8805-3 Weld Metal Combustion Westinghouse Westinghouse Engineering Analysis Analysis Analysis C

0.250 0.220 0.130 Mn 1.320 1.320 1.150 P

0.003 0.017 0.017 S

0.010 0.011 0.010 Si 0.260 0.280 0.190 Ni 0.600 0.610 0.100 Mo 0.530 0.570 0.610 Cr 0.040 0.057 0.052 Cu 0.060 0.058 0.037 Al 0.029 0.030 0.002 Co 0.009 0.006 0.005 Pb

<0.00 1

<0.00 1

<0.001 w

<0.010

<0.010

<0.010 Ti

<0.010 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

<0.010

<0.002

<0.002 N2 0.008 0.006 0.003 B

<0.001

<0.001

<0.001 Notes:

(a)

Data obtained from WCAP-11011131 and duplicated herein for completeness.

Description of Program

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

lMaterial Temperature (0f)

Time Coolant Intermediate Shell Plate Austenitized @

4 hrs.

Water-Quench 1600 +/- 25 B8805-3 Tempered @

4 hrs.

Air-cooled 1225

  • 25 Stress Relieved @

17.5 hrs.

Furnace Cooled l_

1150 +/- 50 Weld Metal (heat # 83653)

Post Weld Stress Relieved 12.75 hrs.

Furnace Cooled l

@ 1150*50 Notes:

(a)

This table was taken from WCAP-I 101 1I'.

Description of Program

4-5 0O CAPSULE U (58.5 )

V (a1)

WX.

W (121.5")

REACTOR VESSEL IOos PL ANVEW VESSEL WALL CAPSULE

.CORE NELTRMON PAD OORE BAIM ELEVATKON VIEW Figure 4-1 Arrangement of Surveillance Capsules in the Vogtle Unit 1 Reactor Vessel Description of Program

4-6 LEGEND:

AL-INTEltMEDIATE SIELL PLATE B8805-3,1HEAT NO. C0623-1 (LONGITUDINAL)

AI'- INTERMEDIATE SHELL PLATE 1B8805-3, HEAT NO. C0623-1 (TRANSVERSE)

AW-WELD METAL (HEAT # 83653)

All-HEAT AFFECTED ZONE MATERIAL Cu Fe -

I 3I Al-.15zco I i Al a

n..

-. ISICO (

~I IIII Iem 579.f olt TOR Spacer Tensile Compact Compact Charp Charpy Charpy Compact Block I

Z A

12 7

AW60 Al 160 AW57 AH57 FAs

[

T]

[WI I AW16 AW15 AW14 AW59 A1159 AW56 Arl56 I

[AW3 IA5 AL16 ALIS W 1 l AW58 _Al158 AW 5 2

5 _Al155 AW52 _

AH52 TOP OF VESSEL CENTER Np237 Compact Charpy Charpy Charpy Charpy AW51 I AH51 AW48 l A148 AL14 AL13 AW50 I

AW47 N

A1447 j

AW49 1

A1149 I

AW46 A1146 CENTER -

\\

lATr60 l

AL60 llAT57 I

I AL57 I

lATS9 llAL59 l

AT56 I

AL56 l T5 l[AL58 l

AT55 I

I AL55S 0 BOTTOM OF VESSEL CU 0

t Al-ASSU0

^IIOR I>

I So I I Ia aI I aII h

Cu a s' Al...SSUO Fe I

a Ia mITOR_0F if I_

A a..lSuto (Cd) a oia Charpy Charpy

\\

Charpy Compact Compact Tensile lAT54 AL54 AT53 l l.53

- A~T52 llAl,52l lAT51 llAL51 l1'\\

lAL48 l l

llAT 12ll IAT50 1Al AL47 AT16 AT15 AT14 AT13 I

AT49 AL49 AT46 AL46 Figure 4-2 Capsule X Diagrami Showing The Location of Speciniens, Thermal Monitors and Dosimeters Description of Program

5-l 5

TESTING OF SPECIMENS FROM CAPSULE X 5.1 OVEWRIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Research and Technology Park. Testing was performed in accordance with I OCFR5O, Appendices G and Hid, ASTM Specification El 85-821I", and Westinghouse Procedure RMF 84021't2 Revision 2 as modified by Westinghouse RMF Procedures 81021'31, Revision 1, and 81031'4], Revision 1.

Upon receipt of the capsule at the hot cell laboratory was opened per Procedure RMF 8804(12]. The specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-1 1011 [3. No discrepancies were found.

Examination of the two low-melting point 5800F (3040C) and 590'F (31 0C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 5801F (3040C).

The Charpy impact tests were performed per ASTM Specification E23-02a '51 and Procedure RMF 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Instron Dynatup Impulse instrumentation system, feeding information into an IBM compatible 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 (TGy), the maximum load (PM), and the time to maximum load (TN{)

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 (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load (EN{).

The yield stress (sr) was calculated from the three-point bend formula having the following expression:

B(W-a)2C 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 p = 450 and p = 0.010 in., Equation I is valid with C = 1.21.

Therefore, (for L = 4W),

Testing of Specimens from Capsule X

5-2 L

3.3 05P0 W

'Y = PGY Ba=-a) 2 1 Gy2 (2)

B(W - a) 1.21 B(TY - a)'

For the Charpy specimen, B = 0.394 in., W = 0.394 in., and a = 0.079 in. Equation 2 then reduces to ory = 333PGy (3) where sy is in units of psi and PGy is in units of lbs. 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.

(4)

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM E23-02a 15] and A370-97a '6]. 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 E8-01' 1" and E21-92 (1998)(ISJ and Procedure RMF 8102['3].

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 and final gage length were 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.

Testing of Specimens from Capsule X

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 X, which received a fluence of 3.53 x 10'9 n/cm2(E> 1.0 MeV) in 14.33 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with unirradiated results 41 as shown in Figures 5-1 through 5-12.

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

The Charpy V-notch data presented in WCAP-I 101 1 3], WCAP-1225614 ], WCAP-13931, Rev. I1, and WCAP-1 5067[6] were based on Charpy curves using a hyperbolic tangent curve-fitting routine. The results presented herein are only for the Capsule X test results using CVGRAPH, Version 5.0.2, which is a symmetric hyperbolic tangent curve-fitting program.

Capsule X received an average fast neutron fluence (E> 1.0 MeV) of 3.53 x 1019 n/cm2 after 14.33 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 81.61F and an irradiated 50 ft-lb transition temperature of 121.7 0F. This results in a 30 ft-lb transition temperature increase of 96.50F and a 50 ft-lb transition temperature increase of 99.80F 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 77.91F and an irradiated 50 ft-lb transition temperature of 135.20F. This results in a 30 ft-lb transition temperature increase of 60.81F and a 50 ft-lb transition temperature increase of 72.71F for the transverse oriented specimens.

Irradiation of the weld metal (heat number 83653) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-3.80F and an irradiated 50 ft-lb transition temperature of 12.8°F. This results in a 30 ft-lb transition temperature increase of 53.4°F and a 50 ft-lb transition temperature increase of 43.1 'F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-76.4°F and an irradiated 50 ft-lb transition temperature of -30.8°F. This results in a 30 ft-lb transition temperature increase of 10.6°F and a 50 ft-lb transition temperature increase of 24.6°F.

The average upper shelf energy of the Intermediate Shell Plate B8805-3 (longitudinal orientation) resulted in an average energy decrease of 13 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 109 ft-lb for the longitudinal oriented specimens.

Testing of Specimens from Capsule X

I1_

5-4 The average upper shelf energy of the Intermediate Shell Plate B8805-3 (transverse orientation) resulted in an average energy decrease of 3 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 93 ft-lb for the longitudinal oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 141 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 9 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 127 ft-lb for the weld HAZ metal.

A comparison of the measured 30 ft-lb shift in transition temperature values for the Vogtle Unit I reactor vessel surveillance materials is presented in Table 5-10.

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 (34 EFPY) and a potential license renewal (54 EFPY) as required by IOCFR50, Appendix Gt21.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 13 through 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 A.

Testing of Specimens from Capsule X

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule X irradiated to 3.53 x 10'9 n/cm2 (E> 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated resultsI 31 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 3.53 x 1019 n/cm2 (E> 1.0 MeV) caused approximately a I to 10 ksi increase in the 0.2 percent offset yield strength and approximately a 3 to 10 ksi increase in the ultimate tensile strength when compared to unirradiated data 31. 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 3.53 x 1019 n/cm2 (E> 1.0 MeV) caused approximately a 4 to 7 ksi increase in the 0.2 percent offset yield strength and approximately a 2 to 5 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 3.53 x 1 lO' 9 n/cm2 (E> 1.0 MeV) caused approximately a 2 to 4 ksi increase in the 0.2 percent offset yield strength and approximately a 2 to 6 ksi increase in the ultimate tensile strength when compared to unirradiated datat41. See Figure 5-19 and Table 5-1 1.

The fractured tensile specimens for the intermediate shell plate B8805-3 material are shown in Figures 5-20 and 5-21, 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 Park Hot Cell facility.

Testing of Specimens from Capsule X

11-5-6 Table 5-1 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a Fluence of 3.53 x 1019 n/cm2 (E> 1.0 MeV)

(Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF C T ft-lbs Joules mils J mm AL51

-25

-32 3

4 0

0.00 2

LAL58 25

-4 8

1 1 3

0.08 5

AL55 50 1 0 30 41 1

4 0.36 15 AL54 50 1 0 1 3 1

8 7

0.18 25 AL53 75 24 35 47 2 1 0.53 25 AL52 100 38 42 57 27 0.69 30 AL48 125 52 59 80 40 1.02 40 AL49 150 66 48 65 36 0.91 45 AL60 160 71 66 89 45 1.14 50 AL46 175 79 75 102 50 1.27 65 AL57 200 93 75 102 51 1.30 70 AL47 225 107 117 159 70 1.78 100 AL59 225 107 105 142 68 1.73 100 ALSO 250 121 103 140 67 1.70 100 AL56 275 135 113 153 70 1.78 100 Testing of Specimens from Capsule X

5-7 Table 5-2 Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to a

-4r,2A^

r,

-t 1 n9 -

II-

- n tx--%n_ +^ ^

rAueUMA V1 0.0 AA

1. LU WIL1 kJs 1.U

.V IV) V a1I4fVers VIUiiaLIUU)

Sample Temperature Impact Energy Lateral Expansion Shear Number O

°F cC ft-lbs I

Joules mils J mm AT48

-50

-46 5

7 0

0.00 2

AT49

-25

-32 7

9 0

0.00 2

AT52 25

-4 14 19 7

0.18 5

ATS0 50 10 25 34 15 0.38 10 AT55 75 24 33 45 25 0.64 20 AT60 100 38 42

57.

31 0.79 25 AT54 125 52 47 64 32 0.81 35 AT51 150 66 52 71 38 0.97 50 AT58 175 79 64 87 45 1.14 65 AT56 200 93 64 87 43 1.09 75 AT53 200 93 61 83 44 1.12 75 AT46 225 107 85 115 60 1.52 100 AT47 225 107 66 l

89 50 1.27 95 AT57 250 121 95 129 66 1.68 100 ATS9 275 135 98 133 l

66 1.68 100 Testing of Specimens from Capsule X

5-8 Table 5-3 Charpy V-notch Data for the Vogtle Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 3.53 x 10'9 n/cm2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF

°C ft-lbs Joules mils mm AW57

-100

-73 3

4 0

0.00 2

AW49

-50 46 6

8 0

0.00 10 AW51 0

-18 22 30 15 0.38 40 AW47 10

-12 42 57 31 0.79 45 AW56 25

-4 85 115 25 0.64 65 AW52 50 10 101 137 54 1.37 80 AW50 75 24 127 172 73 1.85 90 AW59 100 38 136 184 85 2.16 95 AW53 125 52 112 152 76 1.93 90 AW48 125 52 128 174 84 2.13 90 AW55 150 66 127 172 82 2.08 95 AW46 150 66 137 186 80 2.03 98 AW58 175 79 142 193 83 2.11 100 AW60 200 93 142 193 81 2.06 100 AW54 225 107 145 197 78 1.98 100 Testing of Specimens from Capsule X

5-9 Table 5-4 Charpy V-notch Data for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of 3.53 x 10l n/cm2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF CC Ft-lbs Joules mils mm AH49

-150

-101 3

4 0

0.00 2

AH56

-100

-73 14 19 1

0.03 5

AH48

-75

-59 23 31 7

0.18 15 AH58

-50

-46 58 79 28 0.71 30 AHS9

-50

-46 41 56 18 0.46 25 AHS4

-25

-32 59 80 30 0.76 40 AH60 25

-4 68 92 43 1.09 70 AH47 75 24 109-148 58 1.47 90 AHS1 100 38 112 152 69 1.75 90 AH55 100 38 106 144 60 1.52 90 AH46 150 66 102 138

70.

1.78 95 AHS3 200 93 93 126 62 1.57 100 AHS0 200 93 122 165 72 1.83 100 AH57 225 107 172 233 71 1.80 100 AH52 225 107 123 167 67 1.70 100 Testing of Specimens from Capsule X

5-10 Table 5-5 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediatc Shell Plate B8805-3 Irradiated to a Fluence of 3.53 x 019I n/CmII 2 (E>1.0 MeV)

(Longitudinal Oricutation)

Charpy Nornalized Energies Time Test Energy (ft-lb/in2 )

Yield to Time Fast YiCId TeSt Load Yield Max.

to Max.

Fract.

Arrest Stress Flow Sample Temp.

ED Charpy Max.

Prop.

PGY tGY Load t*1 Load Load PA cly Stress NO.

(OF)

(ft-lb)

ED/A Ej I/A EP/A (lb)

(IIISCC)

Pt,, (lb)

(mseC)

PF (lb)

(lb)

(kS;)

(kS;)

AL51

-25 3

24 10 14 1374 0.11 1374 0.11 1371 0

46 46 AL58 25 8

64 30 34 3285 0.14 3323 0.15 3318 0

109 110 AL55 50 30 242 200 41 3542 0.14 4652 0.44 4582 0

118 136 ALS4 50 13 105 45 60 3760 0.15 3980 0.18 3961 97 125 129 AL53 75 35 282 180 102 3555 0.14 4808 0.40 4763 368 118 139 AL52 100 42 338 246 92 3390 0.14 4662 0.53 4544 159 113 134 AL48 125 59 475 328 147 3471 0.15 4680 0.67 4543 773 116 136 AL49 150 48 387 232 155 3256 0.14 4484 0.52 4382 1102 108 129 AL60 160 66 532 319 213 3347 0.15 4595 0.67 4295 852 III 132 AL46 175 75 604 230 374 3361 0.14 4507 0.52 4010 930 112 131 AL57 200 75 604 309 296 3241 0.15 4450 0.67 3965 1405 108 128 AL47 225 117 943 321 622 3310 0.15 4665 0.67 n/a n/a 110 133 AL59 225 105 846 316 530 3291 0.14 4632 0.66 n/a n/a 110 132 ALSO 250 103 830 315 515 3284 0.14 4531 0.67 1/a n/a 109 130 AL56 275 113 910 323 587 3172 0.16 4566 0.70 n/a n/a 106 129 Testing of Specimens from Capsule X

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 3.53 x 1019 n/cm2 (E>1.0 MeV)

(Transverse Orientation)

Charpy Normalized Energies Time Charpy

)Yild toTime Fast Yield Energy (ft-lb/in2)

Yield to Test Load Yield Max.

to Max.

Fract.

Arrest Stress Flow Sample Temp.

ED Charpy Max.

Prop.

PGY tcy Load tst Load Load Cy Stress No.

(OF)

(ft-lb)

ED/A EhI/A EW/A (11b)

(msec)

Pr, (lb)

(msec)

PF (lb)

PA (lb)

(ksi)

(ksi)

AT48

-50 S

40 20 20 2495 0.12 2538 0.13 2527 0

83 84 AT49

-25 7

56 28 29 3159 0.13 3251 0.14 3251 0

105 107 AT52 25 14 113 63 50 3768 0.15 4476 0.20 4476 0

125 137 AT5O 50 25 201 159 42 3460 0.14 4562 0.37 4562 0

115 134 AT55 75 33 266 170 96 3439 0.14 4672 0.38 4672 466 115 135 AT60 100 42 338 241 98 3524 0.14 4682 0.52 4596 292 117 137 AT54 125 47 379 240 139 3456 0.14 4654 0.52 4562 396 115 135 ATSI 150 52 419 239 180 3406 0.14 4550 0.52 4410 867 113 132 ATS8 175 64 516 227 289 3365 0.14 4457 0.51 3989 1951 112 130 AT56 200 64 516 228 288 3312 0.14 4459 0.51 4351 1484 110 129 AT53 200 61 491 227 265 3268 0.14 4475 0.51 4060 1971 109 129 AT46 225 85 685 238 447 3389 0.15 4673 0.52 n/a n/a 113 134 AT47 225 66 532 215 316 3321 0.14 4386 0.49 3691 2330 III 128 AT57 250 95 765 236 529 3152 0.15 4479 0.54 n/a n/a 105 127 ATS9 275 98 790 335 455 3593 0.22 4524 0.74 n/a n/a 120 135 Tcsting of Specimens from Capsule X

5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 3.53 x 10i9 n/cm2 (E>1.0 MeV)

Normalized Energies Time Test Energy (ft-lb/in2)

Yield to Time Fast Test Load Yield Max.

to Max.

Fract.

Arrest Yield Flow Sample Temp.

ED Charpy Max.

Prop.

Pcv tGY Load tr1 Load Load Stress Stress No.

(OF)

(ft-lb)

ED/A EN1/A Ep/A (lb)

(msec)

Pr1 (lb)

(msec)

P1 (lb)

PA (lb) ay (ksi)

(ksi)

AW57

-100 3

24 12 12 1576 0.12 1576 0.12 1576 0

52 52 AW49

-50 6

48 20 29 2562 0.13 2562 0.13 2562 0

85 85 AW51 0

22 177 62 115 3877 0.15 4500 0.20 4500 754 129 139 AW47 10 42 338 201 138 3633 0.14 4656 0.44 4656 1570 121 138 AW56 25 85 685 347 338 3746 0.15 4761 0.68 4531 1958 125 142 AW52 50 101 814 338 476 3630 0.14 4630 0.69 4345 2740 121 138 AW50 75 127 1023 345 679 3660 0.14 4741 0.69 3822 2596 122 140 AW59 100 136 1096 326 770 3537 0.15 4586 0.68 2766 2123 118 135 AW53 125 112 902 319 584 3417 0.14 4418 0.68 3608 2795 114 130 AW48 125 128 1031 322 709 3370 0.14 4490 0.68 2570 1900 112 131 AW55 150 127 1023 316 707 3335 0.14 4424 0.68 2509 1910 II1 129 AW46 150 137 1104 333 770 3545 0.21 4416 0.75 3156 2516 118 133 AW58 175 142 1144 309 835 3186 0.14 4392 0.68 n/a n/a 106 126 AW60 200 142 1144 326 818 3142 0.19 4313 0.75 n/a n/a 105 124 AW54 225 145 1168 328 841 3489 0.22 4406 0.74 n/a n/a 116 131 Testing of Specimens from Capsule X

5-13 Table 5-8 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Heat-Affected-Zone (HAZ) Metal Irradiated to a Fluence of 3.53 x 1019 n/cm2 (E>1.0 McV)

Normalized Energies Time CEnery (ft-lb/in2)

Yield to Time Fast Arrest Test Egy Load Yield Max.

to Max.

Fract.

Load Yield Flow Sample Temp.

ED Charpy Max.

Prop.

PGy tcy Load t?,

Load PA Stress Stress No.

(CF)

(ft-lb)

ED/A E,,,/A Ep/A (lb)

(msec)

PN (lb)

(msec)

PF (lb)

(lb) cry (ksi)

(ksi)

AH49

-150 3

24 9

15 1295 0.09 1330 0.10 1330 0

43 44 AH56

-100 14 113 68 44 4732 0.16 5297 0.20 5278 0

158 167 AH48

-75 23 185 82 104 4117 0.15 5084 0.23 4796 0

137 153 AH58

-50 58 467 274 193 4155 0.15 5212 0.52 5112 0

138 156 AH59

-50 41 330 255 75 4164 0.15 5122 0.49 5063 240 139 155 AH54

-25 59 475 370 105 4137 0.15 5076 0.68 4913 123 138 153 AH60 25 68 548 242 306 3807 0.14 4923 0.49 4867 3069 127 145 AH47 75 109 878 300 578 4087 0.21 5050 0.62 4073 2409 136 152 AH51 100 112 902 341 562 3697 0.15 4731 0.68 3713 2152 123 140 AH55 100 106 854 342 512 3693 0.16 4833 0.69 3405 2018 123 142 AH46 150 102 822 327 495 3490 0.14 4613 0.67 3200 2687 116 135 AH153 200 93 749 314 435 3337 0.15 4463 0.68 n/a n/a 111 130 AH50 200 122 983 326 657 3429 0.15 4591 0.69 n/a n/a 114 134 AH57 225 172 1386 330 1056 3388 0.14 4723 0.68 n/a n/a 113 135 Al-152 225 123 991 326 666 3432 0.15 4652 0.68 n/a n/a 114 135 Testing of Specimens from Capsule X

5-14 Table 5-9 Effcct of Irradiation to 3.53 x 1019 n/cm2 (E>1.0 McV) on the Capsule X Notch Toughness Properties of the Vogtle Unit 1 Reactor Vessel Surveillance Materials Average 30 (ft-lb)(2)

Average 35 mil Latcral(b)

Average 50 ft-lb(t )

Average Energy Absorption(')

Material Transition Temperature (°I)

Expansion Temperature (°F)

Transition Temperature (IF) at Full Shear (ft-lb)

Unifrradiatcd Irradiated AT Unirradiatcd Irradiated AT Unirradiated Irradiated AT Unirradiatcd Irradiated AE Intermediate

-14.9 81.6 96.5 18.8 130.5 111.7 21.9 121.7 99.8 122.0 109.0

-13.0 Shell Plate B8805-3 (Long.)

Intermediate 17.1 77.9 60.8 55.0 146.6 91.6 62.5 135.2 72.7 96.0 93.0

-3.0 Shell Plate 1B8805-3 (Trans.)

Weld Metal

-57.2

-3.8 53.4

-32.6 26.9 59.5

-30.3 12.8 43.1 145.0 141.0

-4.0 (Heat # 83653)

IIAZ Metal

-87.0

-76.4 10.6

-49.7

-6.4 43.3

-55.4

-30.8 24.6 136.0 127.0

-9.0 Notes:

a.

"Average" is defined as the value read from the curve fit through 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 read from the curve fit through the data points of the Charpy tests (see Figures 5-2, 5-5, 5-8 and 5-11).

Testing of Specimiens from Capsule X

5-15 c _

j.

t 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 Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence(d)

Predicted Measured Predicted Measured (x 10l" u/cm 2, (OF) (a)

( 0F) (b)

(%) (a)

(%)(c)

E>1.OMeV)

Intermediate Shell U

0.334 26.80 13.56 14.5 0

Plate B8805-3 Y

1.16 39.97 31.94 19.5 0

(Longitudinal)

V 1.97 45.50 42.66 22 3

X 3.53 51.03 96.50 26 11 Intermediate Shell U

0.334 26.80 0.00(e) 14.5 0

Plate B8805-3 Y

1.16 39.97 15.19 19.5 0

(Transverse)

V 1.97 45.50 33.79 22 2

X 3.53 51.03 60.80 26 3

Surveillance U

0.334 23.52 24.98 14.5 0

Program Y

1.16 35.08 7.70 19.5 0

Weld Metal V

1.97 39.93 0.00(0 22 2

X 3.53 44.79 53.40 26 3

Heat Affected Zone U

0.334 O.0(g) 5 Material Y

1.16 20.78 9

V 1.97 42.08 I-I X

3.53 10.60 7

Notes:

a)

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

b)

Calculated using measured Charpy data plotted using CVGRAPH, Version 5.0.2 (See Appendix C) c)

Values are based on the definition of upper shelf energy given in ASTM E185-82.

d)

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

e)

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

f)

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

g)

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

Testing of Specimens from Capsule X

5-16 Table 5-11 Tensile Properties of the Vogtle Unit 1 Capsule X Reactor Vessel Surveillance Materials Irradiated to 3.53 x 1019 n/cm2 (E > 1.0 MeV)

Material Sample Test 0.2%

Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp.

Yield Strength Load Stress (ksi)

Strength Elongation Elongation in Area (OF)

Strength (ksi)

(kip)

(ksi)

(%)

(%)

(%)

(ksi)

Intermediate AL-10 75 78.9 101.9 3.31 183.9 67.4 10.5 23.3 63 Shell Plate AL-1 300 71.5 90.2 3.05 179.6 62.1 10.5 22.5 65 B 8805-3 (Long.)

AL-12 550 68.8 93.3 3.39 173.1 69.0 10.0 20.6 60 Intermediate AT-10 75 78.2 99.0 3.42 169.8 69.6 11.3 23.6 59 Shell Plate AT-I I 300 71.2 90.3 3.17 171.0 64.5 9.8 20.6 62 B 38805-3 (Trans.)

AT-12 550 68.2 93.3 3.71 150.3 75.6 10.5 18.9 50 Weld Metal AW-l0 75 76.4 89.9 2.58 201.4 52.5 10.5 26.0 74 AW-I I 300 71.1 82.6 2.51 183.7 51.1 9.8 23.3 72 AW-12 550 66.2 85.8 2.65 181.9 54.0 9.8 23.4 70 Testing of Specimens from Capsule X

5-17 INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 01:41 PM Data Set(s) Plotted Curve 1

2 300 -

250 -

th I 200 -

00 U-E150 -

a, z

>100-50 -

-300 Plant VOGTLE I VOGTLE I Capsule UNIRR x

Material SA533B I SA533B I Ori.

Heat #

LT C0623-1 LT C0623-1

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F 0 Set l a Set 2 Results Curve 2

Fluence LSE USE d-USE 2.2 122.0

.0 2.2 109.0

-13.0 T @30 d-T @30 T @50 d-T @50

-14.9

.0 21.9

.0 81.6 96.5 121.7

99. 8 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

_1 5-18 INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 0311912004 02:10 PM Data Set(s) Plotted Curve 2

Plant VOGTLE I VOGTLE 1 Capsule UNIRR X

Material SA533B I SA533BI Ori.

LT LT Heat #

C0623-1 C0623-1 200 150 A

E

.o 2 100 5.

50 0

-300 0

300 Temperature in Deg F o Set 2 600 o Set I Results Curve 2

Fluence LSE USE d-USE

.0 87.4

.0

.0 76.5

-10.9 T @35 d-T @35

18. 8 130. 5

.0 111.7 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit I Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-19 INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/1912004 01:53 PM Data Set(s) Plotted Curve 1

2 Plant VQGTLE I VOGTLE I Capsule UNIRRx Material SA533BI SA533B1 Ori.

LT LT Heat #

C0623-1 C0623-1 D

a)

'E U,

aIL 125-100 -

75-50-25

-300

-200

-1 C 0 Set I 0O 0

100 200 300 Temperature in Deg F 400 500 600 c Set2 Results Curve 2

Fnuence LSE USE

.0 100.0

.0 100.0 d-USE

.0

.0 T @50 64.4 141.7 d-T @50

.0

77. 3 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

'I 5-20 INTERIEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:48 PM Data Set(s) Plotted Curve 2

Plant VOGTLE I VOGTLE I Capsule UNIRR x

Nllaterial SA533B I SA533B I Ori.

TL TL Hleat #

C0623-1 C0623-1 300 250 o

8, 200 00 LL Im 150 C,

z 100 C.

50 -_

00

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F o Set 2 o Set1 Results Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 Curve 2

2.2 96.0 2.2 93.0

.0

- 3. 0

17. 1
77. 9

.0

60. 8
62. 5 135. 2

.0 72.7 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-21 INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 03:02 PM Data Set(s) Plotted Curve 1

2 Plant VOGTLE I VOGTLE I Capsule UNIRR X

Material SA533B I SA533B I Ori.

TL TL Heat #

C0623-1 C0623-1 200 150 u,

=

E 100 C

0 50 0

-300 0

300 Temperature in Deg F a Set 2 600 o Setl Results Curve Fluence LSE USE d-USE T @35 d.T @35 2

.0 96.0

.0 55.0

.0

.0 93.0

-3.0 146.6 91.6 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-22 INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:56 PMI Data Set(s) Plotted Curve I

2 Plant VOGTLE I VOGTLE 1 Capsule UNIRRx Material SA533B 1 SA533BI Ori.

TL TL Ifeat #

C0623-1 C0623-1 M

a) ci Un I-0)I-125 100 -

75 50 25 -

-300

-200

-1(

0 Set I

)0 0

100 200 300 Temperature in Deg F 400 500 600 o Set 2 Results Curve 2

Fluence LSE

.0

.0 USE 100.0 100.0 d-USE

.0

.0 T @50 80.9 145. 8 d-T @50

.0

64. 9 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-23 SURVEIALANCE PROGRAM WNIELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:56 AM Data Set(s) Plotted Curve 12 Plant VOGTLE I VOGTLE I Capsule

.UNIRRx Material SAW SAW Ori.

NA NA Heat #

  • IRE:83653 NNIRE:83653 300 250
, 200

-00U-E 150 c)C1 z

Z; 1 00 50-_

0x

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F c

Set I a Set 2 Results Curve 2

2 Fluence LSE USE 2.2 145.0 2.2 141.0 d-USE T @30 d-T @30 T @50 d-T @50

.0

-57.2

.0

-30.3

.0

-4.0

-3.8 53.4

12. 8
43. 1 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel

'Weld Metal Testing of Specimens from Capsule X

L____i1.

5-24 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 11:04 AM Data Set(s) Plotted Curve I

2 Plant VOGTLE I VOGTLE I Capsule UNIRRx Material SAW SAW Ori.

NA NA Heat #

WIRE:83653 WIRE:83653 200 150 on 0

U, a 100 I-I 2

50 o -

-300 0

300 Temperature in Deg F 0 Seti 0 Set2 600

-Results Curve 2

Fluence LSE

.0

.0 USE

88. 3 82.0 d-USE

.0

-6.2 T @35

- 32. 6

26. 9 d-T @35

.0

59. 5 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-25 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03123/2004 11:01 AM Data Set(s) Plotted Plant Capsule Material Ori.

Heat #

VOGTLE I UNIRR SAW NA WIRE:83653 VOGTLE I X

SAW NA WIRE:83653 Curve 12 125 100 C,a)

  • 0 CL 75 50 25 o

-300

-200

-100 0

100 200 300 400 500 Temperature in Deg F 600 0 Set I a Set 2 Curve 2

Fluence LSE USE

.0 100.0

.0 100.0 d-USE

.0

.0 Results T @50

.- 6. 1 11.9 d.T @50

.0 18.0 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

'I 5-26 SURVEILLANCE PROGRAM HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:25 AM Data Set(s) Plotted Plant Capsule Material Ori.

Heat #

VOGTLE 1 UNIRR SAW NA B8805-1 VOGTLE 1 X

SAW NA B8805-1 Curve 1

2 300 1 250 --

-0 200 -

00o I

m 150 CD uj z

>100 50 -_

-300

-200

-100 0 Set 1 0

100 200 300 400 500 600 Temperature in Deg F o

Set 2 Curve 2

Fluence LSE USE 2.2 136.0 2.2 127.0 d.USE

.0

-9.0 Results T @30

-87.0

-76.4 d-T @30

.0 10.6 T @50

-55.4

-30. 8 d-T @50

.0

24. 6 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule X

5-27 SURVEILLANCE HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03123/2004 10:39 AM Data Set(s) Plotted Plant Capsule Material Ori.

Heat #

VOGTLE 1 UNIRR SAW NA B8805-l VOGTLE 1 X

SAW NA B8805-1 Curve 12 200 150 r-tn 0

U, a 100 a0 50 o 4-

-300 0

300 Temperature in Deq F 600 0 Seti 0 Set2 Curve 2

Fluence LSE USE d-USE

.0 79.1

.0

.0 68.8

-10.3 Results T @35

-49.7

-6.4 d-T @35

.0 43.3 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule X

____ ___________11 5-28 SURVEILLANCE PROGRAM HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:31 ANA Data Set(s) Plotted Curve l2 Plant VOGTLE I VOGTLE I Capsule UNIRR x

Material SAW SAW Ori.

NA NA Heat #

B8805-1 B8805-1 125 100 w

C)a 0l 75 50 25 0

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F

° Set 1 a Set 2 Results Fluence LSE USE d-USE T @50 d-T @50 Curve I

.0 100.0

.0 100.0

.0

- 24.2

.0

-7.5

.0 16.7 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule X

5-29 AL51, -25 0F AL58, 25 0F AL55, 50 0F AL54, 50 0F AL53, 75 0F AL52, 100IF AL48, 125 0F AL49, 150 0F AL60, 160 0F AL46, 175 0F AL57, 200 0F AL47, 225 0F AL59, 225 0F AL50, 250 0F AL56, 275 0F Figure 5-13 Charpy Impact Specimen l;Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate 1B8805-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-30 AT49. -250F AT55 750 F AT56.2000F AF53,200OF AT46,225TF AT47,2250 F AT57,2500F AT59,275TF Figure 5-14 Clharpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805.3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-31 AWN1

°OT AWA7 I lnOl te

-on AWS7",'

sntQt AW)U. 75TF AW59. 1UUF AW53.125'°F AW48.125TF AW55,150TF AW46,150TF AW58,175TF AW60,200TF AW54,22f Figure 5-15 Cliarpy Impact Specimcn Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Veld Metal 5°F Testing of Specimens from Capsule X

5-32 AH58. -5U0F AH59. -500F AH51. 1000F AH55. 100W F

AH46, 1500F AH53, 2000F AH50, 2000F AH57, 225uF AH52, 225WF Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor Vessel Heat-Affected-Zone Metal Testing of Specimens from Capsule X

5-33 120-ULTIMATEYIELD STRENGTH 100 - en 60 -

0.2% YIELD STRENGTH H

20-0 100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated A and

  • are Irradiated to 3.53 x l 0'9 n/cm2 (E > 1.0 MeV)

I-O I--

0 80 70 -

60 -

50 -

40-30 -

20 -

10-0 REDUCTION IN AREA A

A A-A TOTAL ELONGATION UNIFORM ELONGATION A

0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-17 Tensile Properties forVogtle Unit 1 ReactorVessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

Il 5-34 120-100 -ULTIMATE MELD STRENGTH

.;;80

'o 60 Mt 0.2% YIELD STRENGTH co 40 -

20 -

0 II 0

100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated Aand

  • are Irradiated to 3.53 x IlO' n/cm 2 (E > 1.0 MeV)

~~~~~~--

70 REDUCTION IN AREA 60 -

4 50-

> 40 30 TOTAL ELONGATION Q0 a 20.-o 10 UNIFORM UNIFORM 0 -1 I

I I

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)

Testing of Specimens from Capsule X

5-35 100 ULTIMATE YIELD STRENGTH 80 -

^. 60 -

0.2% YIELD STRENGTH C,,

ww 40-20 -

0-

.I 0

100 200 300 400 500 600 TEMPERATURE( F)

Legend: A and o are Unirradiated A and

  • are Irradiated to 3.53 x 1019 n/cm2 (E > 1.0 MeV) 80~

70 -

60 -

REDUCTION IN AREA

,50-5 40-3 30 -

TOTAL ELONGATION 20 -

UNIFORM ELONGATION 10 A

10 A

0 100 200 300 400 500 600 TEMPERATURE ( F)

Figure 5-19 Tensile Properties for Vogtle Unit 1 Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-36 Specimen AL10 Tested at 750F Specimen ALl1 Tested at 300OF Specimen AL12 Tested at 5500F Figure 5-20 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Slicl Plate B8805-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-37 Specimen ATIO Tested at 750F

-.ej

. !0 Specimen ATL 1 Tested at 300OF Specimen AT12 Tested at 5500F Figure 5-21 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-38 Specimen AWIO Tested at 750F Specimen AWl 1 Tested at 300OF Specimen AW12 Tested at 5500F Figure 5-22 Fractured Tensile Specimens from Vogtle Unit I Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-39 VOGTLE UNIT #1

'X CAPSULE 100 I Uso a, 60 -

U)

U) 40 -

20 AL-10 75F 0

0 0.05 0.1 0.15 STRAIN. INIIN 0.2 0.25 0.3 VOGTLE UNIT #1 X CAPSULE 100 So so 70 X

60 U)so UO w

U) 40 30 20 10 0

AL-11 300 F 0.05 0.1 0.15 0.2 0.2S 0.3 STRAIN, INAN Figure 5-23 Engineering Stress-Strain Curves forVogtle Unit 1 Intermediate Shell Plate B8805-3, Capsule X, Tensile Specimens AL-10, AL-1l (Longitudinal Orientation)

Testing of Specimens from Capsule X

540 VOGTLE UNIT # 1

'X' CAPSULE 100 90 80 70 X

60 n

50 to 40 30 20 10 AL-12 550 F 0

0.05 0.1 0.15 0.2 025 STRAIN.

INN 0.3 Figure 5-23 (cont.) Engineering Stress-Strain Curves for Vogtle Unit 1 Intermediate Shell Plate B8805-3, Capsule X, Tensile Specimen AL-12 (Longitudinal Orientation)

Testing of Specimens from Capsule X

541 VOGTLE UNIT #1 X CAPSULE 100 90 so1 70 s

60 U,

50*

M'U U, 40 30 20 10 0-AT-10 75 F 0

0.05 0.1 0.15 STRAIN. INAN 0.2 0.25 0.3 100t 90 SO 70 s

60 U;

U) 50 V) 40 30 20 10 0

VOGTLE UNIT # I AX CAPSULE AT-11 300 F C

0.05 0.1 0.15 STRAIN. INAN 0.2 0.25 0.3 Figure 5-24 Engineering Stress-Strain Curves forVogtle Unit 1 Intermediate Shell Plate B8805-3, Capsule X, Tensile Specimens AT-10, and AT-li (Transverse Orientation)

Testing of Specimens from Capsule X

5-42 VOGTLE UNIT # 1 X CAPSULE 100 90 80 70 W 60 to 0

s U) 50 LU (D40 30 20 10 AT-12 550 F 0

0 05 0.1 0.15 0.2 0.25 STRAIN. INIIN 0.3 Figure 5-24 (cont.) Engineering Stress-Strain Curves for Vogtle Unit 1 Intermediate Shell Plate B8805-3, Capsule X, Tensile Specimen AT-12 (Transverse Orientation)

Testing of Specimens from Capsule X

5-43 VOGTLE UNIT #1 X' CAPSULE 100*

goo 90-80-70 X

60 50 co 40 30 20 10 AW-10 75 F 0

0 0.05 0.1 0.15 STRAIN. ININ 02 0.25 0.3 100 90 eo 70 -

s 60 Cl,

,,, 40 30 20 10 0

VOGTLE UNIT #1 X' CAPSULE AW-il 300 F 0

0.05 0.1 0.15 STRAIN. INAN 0.2 0.25 0.3 Figure 5-25 Engineering Stress-Strain Curves forNVeld Metal, Capsule X, Tensile Specimens AW-10, and AWV-l1 Testing of Specimens from Capsule X

_________1 5-44 VOGTLE UNIT #1 "X'CAPSULE 100-90 80 70 1

60' U) 50 co 40-30-20-10-0-

AW-12 550 F 0

0 05 0.1 0.15 STRAIN, INAN 0.2 0.25 0.3 Figure 5-25 (cont.) Engineering Stress-Strain Curves for Weld Metal, Capsule X, Tensile Specimen ANV-12 Testing of Specimens from Capsule X

6-1 6

RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates Sn transport analysis performed for the Alvin W. Vogtle Unit I 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 X, withdrawn at the end of the eleventh plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the Alvin W. Vogtle Unit I reactor, the sensor sets from the previously withdrawn capsules (U, Y, and V) 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, "Analysis and Interpretation of Light-WVater Reactor Surveillance Results," 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, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." 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."

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 and meet the requirements of Regulatory Guide.1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."122 Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.[23]

Radiation Analysis and Neutron Dosimetry

l1 6-2 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the Alvin \\V. Vogtle Unit I reactor geometry at the core midplane 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 58.50, 610, 121.50, 238.5°, 241°, and 301.5° 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) and 31.5° from the core cardinal axes (for the 121.50 and 301.5° single surveillance capsule holder locations, and for the 58.5° and the 238.50 dual surveillance capsule holder locations). 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 Alvin W. Vogtle Unit I 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:

0(r,9,z) = 0(r,O)

  • 0(r, z) 0(r) where (rO,z) is the synthesized three-dimensional neutron flux distribution, 4(r,0) is the transport solution in r,0 geometry, 4(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and +(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 Alvin NV. Vogtle Unit 1.

For the Alvin WV. Vogtle Unit I transport calculations, the r,0 models depicted in Figure 6-1 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric. These r,0 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 31.5°, 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 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 Radiation Analysis and Neutron Dosimetry

6-3 chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

The rz model used for the Alvin W. 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 r,0 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 rz 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 point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

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

The core power distributions used in the plant specific transport analysis were provided by Southern Nuclear Co and the Nuclear Fuels Division of Westinghouse. for each of the first twelve fuel cycles at Alvin \\V. 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 3.1[241 and the BUGLE-96 cross-section library.E251 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 29° dual capsule, 31.50 dual capsule, and 31.50 single capsule. These results, representative of the axial midplane of the active core, establish the calculated exposure of the Radiation Analysis and Neutron Dosimetry

I L 64 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.

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 6-1 and Table 6-2. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the eleventh fuel cycle as well as future projections to 20, 24, 32, 40, 48, and 54 EFPY. The calculations for Cycle 5 account for an uprate from 3411 MWt to 3565 MWt. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 12 were representative of future plant operation. The future projections are also based on the current reactor power level of 3565 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-3 and 64, respectively. The data, based on the cumulative integrated exposures from Cycles I through 11, 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 four surveillance capsules withdrawn from the Alvin W.

Vogtle Unit I 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 Alvin W. Vogtle Unit I reactor.

Updated lead factors for the Alvin NV. Vogtle Unit I 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 and X) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (W and Z), the lead factor corresponds to the calculated fluence values at the end of Cycle 11, the last completed fuel cycle for Alvin NV. Vogtle Unit 1.

Radiation Analysis and Neutron Dosimetry

6-5 63 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, is documented in Appendix A.

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

Reaction Rates (rpslatom)

M/C Reaction Measured Calculated Ratio 63Cu(n,a)60Co 4.62E-1 7 4.28E-1 7 1.08 5 4 Fe(n,p) 54Mn 4.69E-1 5 4.70E-15 1.00 SSNi(n,p)SSCo 6.47E-15 6.58E-15 0.98 23&U(n p)137Cs (Cd) 2.86E-14 2.50E-14 1.14 "37Np(nj)1 37Cs (Cd) 2.51E-13 2.43E-13 1.03 Average:

1.05

__% Standard Deviation:

6.2 The measured-to-calculated (M/C) reaction rate ratios for the Capsule X threshold reactions range from 0.98 to 1.14, and the average M/C ratio is 1.05 +/- 6.2% (la). 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 Alvin W. Vogtle Unit I reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Alvin W. Vogtle Unit 1.

Radiation Analysis and Neutron Dosimetry

~-

_Ij-6-6 6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Alvin NV. 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:

I -

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.

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 Alvin W. Vogtle Unit I 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 Alvin W. Vogtle Unit I analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Alvin W. Vogtle Unit I 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 Alvin W. Vogtle Unit I 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 2.

Radiation Analysis and Neutron Dosimetry

6-7 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%

.I 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 Alvin W. Vogtle Unit 1.

Radiation Analysis and Neutron Dosimetry

Ij-6-8 Figure 6-1 Alvin W. Vogtle Unit I rO Reactor Geometry with a 12.50 Neutron Pad at the Core Midplane 240 180

._2 x

C):

1 20 60 0

0 75 150 225 300 R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-9 Figure 6-1 (continued)

Alvin W. Vogtle Unit I rO Reactor Geometry with a 20.00 Neutron Pad at the Core Midplane 240-1 8 0 C.L Cn

'X:

1 20 60 -

0 -

I I

A 0

75 150 225 300 R Axis (cm)

Radiation Analysis and Neutron Dosimetry

~-

_________l~

6-10 Figure 6-1 (continued)

Alvin W. Vogtle Unit I r,0 Reactor Geometry with a 22.50 Neutron Pad at the Core Midplane 240 1 80 1 20

._xn M:

60 0

0 75 150 225 300 R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-11 Figure 6-2 Alvin W. Vogtle Unit 1 rz Reactor Geometry with Neutron Pad 300 -

200-100-E 0-

-100-

-200-

-300-

-400-t1

_0-0 I

I I

I I

I I

I I

I l

75 150 225 300 R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-12 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation In/cm2-sl Length Time Time Dual Dual Single Cycle

[EFPS]

[EFPSJ IEFPYI 290 31.50 31.50 I

3.61E+07 3.61E+07 1.14 8.63E+10 9.25E+10 9.16E+10 2

3.58E+07 7.19E+07 2.28 7.55E+10 8.09E+10 8.01E+10 3

4.19E+07 1.14E+08 3.61 7.78E+10 8.58E+10 8.51E+10 4

3.92E+07 1.53E+08 4.85 6.32E+10 6.84E+10 6.77E+10 5

4.19E+07 1.95E+08 6.17 6.69E+10 7.24E+10 7.18E+10 6

4.24E+07 2.37E+08 7.52 6.82E+10 7.33E+10 7.25E+10 7

3.98E+07 2.77E+08 8.78 6.11E+10 6.61E+10 6.55E+10 8

4.22E+07 3.19E+08 10.11 6.14E+10 6.75E+10 6.69E+10 9

4.61 E+07 3.65E+08 11.57 7.53E+10 8.74E+10 8.68E+10 10 4.28E+07 4.08E+08 12.93 7.74E+10 8.49E+10 8.41 E+ 10 11 4.41 E+07 4.52E+08 14.33 7.19E+10 7.99E+10 7.92E+10 Future 1.79E+08 6.31E+08 20.00 6.77E+10 7.34E+10 7.27E+10 Future 1.26E+08 7.57E+08 24.00 6.77E+10 7.34E+10 7.27E+10 Future 2.53E+08 I.01E+09 I 32.00 6.77E+10 7.34E+10 7.27E+10 Future 2.53E+08 1.26E+09 40.00 6.77E+10 7.34E+10 7.27E+10 Future 2.53E+08 1.51E+09 48.00 6.77E+10 7.34E+10 7.27E+10 Future 1.89E+08 1.70E+09 54.00 6.77E+10 7.34E+10 7.27E+10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-13 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation ln/cm2l Length Time Time Dual Dual Single Cycle TEFPSI IEFPSI IEFPYl 290 31.50 31.50 1

3.61E+07 3.61 E+07 1.14 3.1IE+18 3.34E+18 3.30E+18 2

3.58E+07 7.19E+07 2.28 5.82E+18 6.23E+18 6.17E+18 3

4.19E+07 1.14E+08 3.61 9.08E+18 9.83E+18 9.74E+18 4

3.92E+07 1.53E+08 4.85 1.16E+19 1.25E+19 1.24E+19 5

4.19E+07 1.95E+08 6.17 1.44E+19 1.56E+19 1.54E+19 6

4.24E+07 2.37E+08 7.52 1.73E+19 1.87E+19 1.85E+19 7

3.98E+07 2.77E+08 8.78 1.97E+19 2.13E+19 2.11E+19 8

4.22E+07 3.19E+08 10.11 2.23E+19 2.41E+19 2.39E+19 9

4.61E+07 3.65E+08 11.57 2.57E+19 2.82E+19 2.79E+19 10 4.28E+07 4.08E+08 12.93 2.91E+19 3.18E+19 3.15E+19 11 4.41E+07 4.52E+08 14.33 3.22E+19 3.53E+19 3.50E+19 Future 1.79E+08 6.31E+08 20.00 4.43E+19 4.85E+19 4.80E+19 Future 1.26E+08 7.57E+08 24.00 5.29E+19 5.77E+19 5.72E+19 Future 2.53E+08 1.0 IE+09 32.00 6.99E+19 7.63E+19 7.55E+19 Future 2.53E+08 1.26E+09 40.00 8.70E+19 9.48E+19 9.39E+19 Future 2.53E+08 1.51E+09 48.00 1.04E+20 1.13E+20 1.12E+20 Future 1.89E+08 1.70E+09 54.00 1.17E+20 1.27E+20 1.26E+20 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

I'-.

6-14 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation fda/sl Length Time Time Dual Dual Single Cycle

[EFPS]

[EFPSJ IEFPYl 290 31.50 31.50 1

3.61E+07 3.61 E+07 1.14 1.69E-10 1.81E-10 1.79E-10 2

3.58E+07 7.19E+07 2.28 1.47E-10 1.57E-10 1.56E-10 3

4.19E+07 1.14E+08 3.61 1.51E-10 1.67E-10 1.65E-10 4

3.92E+07 1.53E+08 4.85 1.23E-10 1.33E-10 1.31E-10 5

4.19E+07 1.95E+08 6.17 1.30E-10 1.40E-10 1.39E-10 6

4.24E+07 2.37E+08 7.52 1.32E-10 1.42E-10 1.40E-10 7

3.98E+07 2.77E+08 8.78 1.18E-10 1.28E-10 1.27E-10 8

4.22E+07 3.19E+08 10.11 1.20E-10 1.32E-10 1.30E-10 9

4.61 E+07 3.65E+08 11.57 1.46E-10 1.70E-10 1.68E-10 10 4.28E+07 4.08E+08 12.93 1.50E-1O 1.65E-10 1.63E-10 11 4.41 E+07 4.52E+08 14.33 1.40E-10 1.55E-10 1.54E-10 Future 1.79E+08 6.31E+08 20.00 1.31E-10 1.42E-10 1.41E-10 Future 1.26E+08 7.57E+08 24.00 1.31E-10 1.42E-10 1.41E-10 Future 2.53E+08 1.I E+09 32.00 1.31E-10 1.42E-10 1.41 E-I0 Future 2.53E+08 1.26E+09 40.00 1.31E-10 1.42E-10 1.41E-10 Future 2.53E+08 1.51E+09 48.00 1.31E-10 1.42E-10 1.41 E-10 Future 1.89E+08 1.70E+09 54.00 1.311E-10 1.42E-10 1.41E-10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-15 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation Idpal Length Time Time Dual Dual Single Cycle IEFPSI IEFPSI JEFPYJ 290 31.50 31.50 1

3.61E+07 3.61 E+07 1.14 6.08E-03 6.52E-03 6.45E-03 2

3.58E+07 7.19E+07 2.28 1.14E-02 1.22E-02 1.20E-02 3

4.19E+07 1.14E+08 3.61 1.77E-02 1.91E-02 1.89E-02 4

3.92E+07 1.53E+08 4.85 2.25E-02 2.43E-02 2.41E-02 5

4.19E+07 1.95E+08 6.17 2.79E-02 3.02E-02 2.99E-02 6

4.24E+07 2.37E+08 7.52 3.35E-02 3.62E-02 3.58E-02 7

3.98E+07 2.77E+08 8.78 3.82E-02 4.13E-02 4.09E-02 8

4.22E+07 3.19E+08

.10.11 4.33E-02 4.69E-02 4.64E-02 9

4.61E+07 3.65E+08 11.57 5.OOE-02 5.47E-02 5.41E-02 10 4.28E+07 4.08E+08 12.93 5.64E-02 6.17E-02 6.11 E-02 11 4.41E+07 4.52E+08 14.33 6.26E-02 6.85E-02 6.79E-02 Future 1.79E+08 6.31 E+08 20.00 8.61 E-02 9.40E-02 9.31 E-02 Future 1.26E+08 7.57E+08 24.00 1.03E-01 1.12E-01 1.11 E-01 Future 2.53E+08 1.01E+09 32.00 1.36E-01 1.48E-01 1.46E-01 Future 2.53E+08 1.26E+09 40.00 1.69E-01 1.84E-01 1.82E-01 Future 2.53E+08 1.51 E+09 48.00 2.02E-01 2.20E-01 2.18E-01 Future 1.89E+08 1.70E+09 54.00 2.27E-01 2.47E-01 2.44E-01 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

_U-6-16 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 MIeV)

Cycle Irradiation Irradiation

_______sl Length Time Time I

Cycle IEFPS]

IEFPSI IEFPYI 00 150 300 450 3.61E+07 3.61E+07 1.14 1.28E+10 1.90E+10 2.18E+10 2.23E+10 2

3.58E+07 7.19E+07 2.28 9.98E+09 1.57E+10 1.89E+10 1.80E+10 3

4.19E+07 1.14E+08 3.61 9.15SE+09 I.43E+10 1.95E+10 1.94E+10 4

3.92E+07 1.53E+08 4.85 1.04E+10 1.41E+10 1.62E+10 1.67E+10 5

4.19E+07 1.95E+08 6.17 9.05E+09 1.38E+10 1.68E+10 1.67E+10 6

4.24E+07 2.37E+08 7.52 9.13E+09 1.45E+10 1.72E+10 1.68E+10 7

3.98E+07 2.77E+08 8.78 8.88E+09 L.26E+10 1.55E+10 1.52E+10 8

4.22E+07 3.19E+08 10.11 9.08E+09 1.39E+10 1.57E+10 1.80E+10 9

4.61 E+07 3.65E+08 11.57 9.54E+09 1.36E+10 1.92E+10 2.18E+10 10 4.28E+07 4.08E+08 12.93 9.04E+09 1.46E+10 1.93E+10 1.94E+10 11 4.41 E+07 4.52E+08 14.33 9.74E+09 1.46E+10 1.82E+10 2.OOE+10 Future 1.79E+08 6.31 E+08 20.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10 Future 1.26E+08 7.57E+08 24.00 9.92E+09 1.45E+10 1.71 E+10 1.75E+10 Future 2.53E+08 1.01E+09 32.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10 Future 2.53E+08 1.26E+09 40.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10 Future 2.53E+08 1.51 E+09 48.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10 Future 1.89E+08 1.70E+09 54.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10 Radiation Analysis and Neutron Dosimetry

6-17 Table 6-2 cont'd 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)

Cycle Irradiation Irradiation InJ ml-Length Time Time Cycle IEFPS]

IEFPSI IEFPYI 00 150 300 450 I

3.61E+07 3.61E+07 1.14 4.61E+17 6.84E+17 7.84E+17 8.04E+17 2

3.58E+07

-7.19E+07 2.28 8.18E+17 1.25E+18 1.46E+18 1.45E+18 3

4.19E+07 1.14E+08 3.61 1.20E+18 1.85E+18 2.28E+18 2.26E+18 4

3.92E+07 1.53E+08 4.85 1.60E+18 2.38E+18 2.90E+18 2.90E+18 5

4.19E+07 1.95E+08 6.17 1.98E+18 2.96E+18 3.60E+18 3.60E+18 6

4.24E+07 2.37E+08 7.52 2.36E+18 3.58E+18 4.33E+18 4.31E+18 7

3.98E+07 2.77E+08 8.78 2.71E+18 4.08E+18 4.94E+18 4.911E+18 8

4.22E+07 3.19E+08 10.11 3.1 OE+1 8 4.66E+18 5.60E+1 8 5.67E+18 9

4.61E+07 3.65E+08 11.57 3.54E+18 5.29E+18 6.49E+18 6.67E+18 10 4.28E+07 4.08E+08 12.93 3.92E+18 5.911E+18 7.311E+18 7.50E+18 11 4.41E+07 4.52E+08 14.33 4.35E+18 6.55E+18 8.11E+18 8.38E+18 Future 1.79E+08 6.31E+08 20.00 6.13E+18 9.16E+18 1.12E+19 1.15E+19 Future 1.26E+08 7.57E+08 24.00 7.38E+18 1.1OE+19 1.33E+19 1.37E+19 Future 2.53E+08 1.01E+09 32.00 9.89E+18 1.47E+19 1.77E+19 1.81E+19 Future 2.53E+08 1.26E+09 40.00 1.24E+19 1.83E+19 2.20E+19 2.25E+19 Future 2.53E+08 1.51E+09 48.00 1.49E+19 2.20E+19 2.63E+19 2.70E+19 Future 1.89E+08 1.70E+09 54.00 1.68E+19 2.48E+19 2.95E+19 3.03E+19 Radiation Analysis and Neutron Dosimetry

ll U-6-18 Table 6-2 cont'd Calculated Azimuthal Variation of Fast Neutron Exposure Rates and Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation Id alsl Length Time Time Cycle IEFPSJ IEFPSl IEFPY1 00 150 300 450 I

3.61 E+07 3.61E+07 1.14 1.99E-11 2.91E-1 I 3.35E-11 3.53E-1 1 2

3.58E+07 7.19E+07 2.28 1.55E-11 2.42E-11 2.91E-11 2.85E-I I 3

4.19E+07 1.14E+08 3.61 1.42E-11 2.20E-11 3.01E-1 I 3.07E-I 4

3.92E+07 1.53E+08 4.85 1.61E-11 2.17E-11 2.51E-11 2.64E-1 1 5

4.19E+07 1.95E+08 6.17 1.41E-11 2.13E-11 2.59E-11 2.64E-I I 6

4.24E+07 2.37E+08 7.52 1.42E-11 2.23E-11 2.66E-11 2.66E-II 7

3.98E+07 2.77E+08 8.78 1.38E-11 1.95E-1I 2.39E-1I 2.40E-I I 8

4.22E+07 3.19E+08 10.11 1.41E-11 2.14E-11 2.42E-1I 2.84E-I1 9

4.61E+07 3.65E+08 11.57 1.48E-11 2.09E-11 2.96E-11 3.44E-I I 10 4.28E+07 4.08E+08 12.93 1.41E-11 2.25E-11 2.98E-11 3.06E-I I I1 4.41E+07 4.52E+08 14.33 1.52E-11 2.25E-11 2.81E-11 3.15E-I I Future 1.79E+08 6.31 E+08 20.00 1.54E-11 2.24E-11 2.64E-11 2.76E-I I Future 1.26E+08 7.57E+08 24.00 1.54E-1 I 2.24E-I 2.64E-I 2.76E-1 I Future 2.53E+08 1.O1E+09 32.00 1.54E-11 2.24E-1I 2.64E-11 2.76E-1I I Future 2.53E+08 1.26E+09 40.00 1.54E-11 2.24E-11 2.64E-1I 2.76E-I I Future 2.53E+08 1.51E+09 48.00 1.54E-11 2.24E-1I 2.64E-11 2.76E-I1 I Future 1.89E+08 1.70E+09 54.00 1.54E-11 2.24E-11 2.64E-11 2.76E-I I Radiation Analysis and Neutron Dosimetry

6-19 Table 6-2 cont'd Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation Id pal Length Time Time Cycle IEFPSI IEFPSI

[EFPYJ 00 150 300 450 1

3.61E+07 3.61E+07 1.14 7.16E-04 1.05E-03 1.21E-03 1.27E-03 2

3.58E+07 7.19E+07 2.28 1.27E-03 1.92E-03 2.25E-03 2.29E-03 3

4.19E+07 1.14E+08 3.61 1.87E-03 2.84E-03 3.51 E-03 3.58E-03 4

3.92E+07 1.53E+08 4.85 2.48E-03 3.66E-03 4.46E-03 4.58E-03 5

4.19E+07 1.95E+08 6.17 3.07E-03 4.55E-03 5.55E-03 5.68E-03 6

4.24E+07 2.37E+08 7.52 3.67E-03 5.50E-03 6.67E-03 6.811E-03 7

3.98E+07 2.77E+08 8.78 4.22E-03 6.27E-03 7.62E-03 7.76E-03 8

4.22E+07 3.19E+08 10.11 4.81 E-03 7.17E-03 8.64E-03 8.96E-03 9

4.61 E+07 3.65E+08 11.57 5.50E-03 8.14E-03 I.OOE-02 1.05E-02 10 4.28E+07 4.08E+08 12.93 6.1OE-03 9.10E-03 1.13E-02 1.19E-02 I

4.41 E+07 4.52E+08 14.33 6.77E-03 1.O1E-02 1.25E-02 1.32E-02 Future 1.79E+08 6.31E+08 20.00 9.53E-03 1.411E-02 1.72E-02 1.82E-02 Future 1.26E+08 7.57E+08 24.00 1.15E-02 1.69E-02 2.06E-02 2.17E-02 Future 2.53E+08 1.01E+09 32.00 1.54E-02 2.26E-02 2.72E-02 2.86E-02 Future 2.53E+08 1.26E+09 40.00 1.93E-02 2.82E-02 3.39E-02 3.56E-02 Future 2.53E+08 1.5 1 E+09 48.00 2.32E-02 3.39E-02 4.06E-02 4.26E-02 Future 1.89E+08 1.70E+09 54.00 2.61E-02 3.81 E-02 4.56E-02 4.78E-02 Radiation Analysis and Neutron Dosimetry

lI-6-20 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.11 1.000 1.000 1.000 1.000 225.59 0.571 0.567 0.561 0.557 231.06 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 Note:

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 Table 6-4 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall 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 0.645 231.06 0.390 0.382 0.381 0.392 236.54 0.237 0.227 0.227 0.235 242.01 0.142 0.128 0.127 0.130 Note:

Base Metal Inner Radius = 220.11 cm Base Metal 1/4T

= 225.59 cm Base Metal 1/2T

= 231.06 cm Base Metal 3/4T

= 236.54 cm Base Metal Outer Radius = 242.01 cm Radiation Analysis and Neutron Dosimetry

6-21 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Alvin W. Vogtle Unit 1 Irradiation Time Fluence (E > 1.0 MeV)

Iron Displacements Capsule fEFPY]

I In/cm21 I

fdpal U

1.14 3.34E+18 6.52E-03 Y

4.85 1.16E+19 2.25E-02 V

8.78 1.97E+19 3.82E-02 X

14.33 3.53E+19 6.85E-02 Table 6-6 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor U(31.50 Dual)

Withdrawn EOC 1 4.15 Y (29.00 Dual)

Withdrawn EOC 4 3.99 V (29.00 Dual)

Withdrawn EOC 7 3.98 X (31.5° Dual)

Withdrawn EOC 11 4.21 W (31.5° Single)

In Reactor 4.17 Z (31.5° Single)

In Reactor 4.17 Note:

Lead factors for capsules remaining in the reactor are based on cycle specific exposure calculations through the last completed fuel cycle, i.e., Cycle 11.

Radiation Analysis and Neutron Dosimetry

7-1 7

SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM El 85-82 and is recommended for future capsules to be removed from the Vogtle Unit I reactor vessel. This recommended removal schedule is applicable to 36 EFPY of operation.

Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule Capsule Capsule Location Lead Factor (a)

Withdrawal EFPY (b)

Fluence (n/cm 2) (a)

U 58.50 4.15 1.14 3.34 x 10" (c)

Y 241° 3.99 4.85 1.16 x 10'9 (c)

V 610 3.98 8.78 1.97x I09 (c)(e)

X 238.50 4.21 14.33 3.53 x IO'9 (c)(f)

W 121.50 4.17 18.54 (Standby)(g) 4.01 x IO"1 (d)

Z 301.50 4.17 Standby(g)

Notes:

(a) Updated in Capsule X dosimetry analysis.

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) This projected fluence is not less than once or greater than twice the peak EOL fluence for an additional 20-year license renewal term to 80 years.

(e) This capsule was withdrawn at approximately the current end-of-license (36 EFPY) peak fluence.

(f) This capsule was withdrawn at approximately (54 EFPY) peak fluence.

(g) Since the lead factor for both capsule W and Z are same, either one may be withdrawn for 80 years license renewal Surveillance Capsule Removal Schedule

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. Code of Federal Regulations, IOCFR50, Appendix G. Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
3. WCAP-1 1011, "Georgia Power Company Alvin W. Vogtle Unit No. I Reactor Vessel Radiation Surveillance Program", L.R. Singer, February 1986.
4. WCAP-12256, "Analysis of Capsule U from the Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko, et. al., May 1989.
5. WCAP-13931, Revision 1, "Analysis of Capsule Yfrom the Georgia Power Company Vogtle Unit I Reactor Vessel Radiation Surveillance Program", M. J. Malone, et. al., August 1995.
6. WCAP-1 5067, "Analysis ofCapsule Vfrom Southern Nuclear Vogtle Electric Generating Plant Unit I Reactor Vessel Radiation Surveillance Program." T. J. Laubham, et. al., September 1998
7. STD-MCE-04-9, "Vogtle Unit I Capsule X Test Report," J. Conermann, 3/16/04.
8. Procedure RMF 8804, Opening of Westinghouse Surveillance Capsules, Revision 0.
9. ASTM E208, Standard Test Methodfor 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.
10.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G. Fracture Toughness Criteria for Protection Against Failure
11. ASTM El 85-82, Standard Practice/or Conducting Surveillance Testsfor Light-Water Cooled Nuclear Power Reactor Vessels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.
12. Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 2.
13. Procedure RMF 8102, Tensile Testing, Revision 1.
14. Procedure RMF 8103, Charpy Impact Testing, Revision 1.
15. ASTM E23-02a, Standard Test Method/or Notched Bar Impact Testing o0Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 2002.

References

-~

l-8-2

16. ASTM A370-97a, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1997.
17. ASTM E8-01, Standard Test Methods for Tension Testing of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 2001.
18. ASTM E21-92 (1998), Standard Test Methods for Elevated Temperature Tension Tests of Metallic Mfaterials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1998.
19. ASTM E83-93, Standard Practice for Verification and Classification of Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
20. ASTM El 85-82, Standard Recommended Practicefor Conducting Surveillance TestsforLight-Water Cooled Nuclear Reactor Vessels.
21. WCAP-14370, Use of the Hyperbolic Tangent Function for Fitting Transition Temperature Toughness Data, T. R. Mager, et al, May 1995.
22. 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.
23. WCAP-14040-NP-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.
24. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two-and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System, August 1996.
25. RSIC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

References

A-O APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS Appendix A

A-1 A.1 Neutron Dosimetrv Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Alvin W. 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.I'[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. This information may also be useful in the future, in particular, as least squares adjustment techniques become accepted in the regulatory environment.

A.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the four neutron sensor sets withdrawn to date as part of the Alvin X'. Vogtle Unit I 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:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPYJ U

31.50 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

31.50 Dual End of Cycle 11 14.33 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, and X are summarized as follows:

Appendix A

A-2 Reaction Sensor Material Of Interest Capsule U Capsule Y Capsule V Capsule X Copper 63Cu(na)60Co X

X X

X Iron S4Fe(n p)54Mn X

X X

X Nickel s&Ni(n,p)5sCo X

X X

X Uranium-238 238U(n, f) 37cs X

X X

X Neptunium-237 23 7Np(n,f) 137CS X

X X

X Cobalt-Aluminum*

59Co(ny)0Co X

X X

X The cobalt-aluminum measurements for this plant include both bare wire and cadmium-covered sensors.

Since all of the dosimetry monitors were accommodated within the dosimeter block centered at the radial, azimuthal, and axial center of the material test specimen array, 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:

  • the measured specific activity of each monitor,
  • 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 Capsule U, Y, and V are documented in References A-2, A-5, and A-6. The radiometric counting of the sensors from Capsule X was carried out by Pace Analytical Services, Inc., located at the Westinghouse Waltz Mill Site. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and Appendix A

A-3 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, and X was based on the monthly power generation of Alvin W. Vogtle Unit I 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, and X is given in Table A-2.

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:

1?

A No F Y X C [1-e4"] [e Af]

P'~f where:

R

=

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

A

=

Measured specific activity (dps/gm).

No

=

Number of target element atoms per gram of sensor.

F

=

Weight fraction of the target isotope in the sensor material.

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).

Cj

=

Calculated ratio of c(E > 1.0 MeV) during irradiation period j to the time weighted average j(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.

Appendix A

A-4 In the equation describing the reaction rate calculation, the ratio [Pj]/[P,,f] 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 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 23sU 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 Alvin W. Vogtle Unit I fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V Capsule X 23SU Impurity/Pu Build-in 0.871 0.836 0.806 0.758 238U(y,f) 0.967 0.969 0.969 0.967 Net 238U Correction 0.842 0.810 0.781 0.733 Np(7,f) 0.990 0.991 0.991 0.990 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, and X are given in Table A-4. In Table A-4, 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 28U impurities, plutonium build-in, and gamma ray induced fission effects.

Appendix A

A-5 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 ¢(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 + 8R, =Z(o, 8 + Ar )(Og -+60 g

relates a set of measured reaction rates, R., to a single neutron spectrum, fg, through the multigroup dosimeter reaction cross-section, ag, each with an uncertainty 5. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the Alvin NV. Vogtle Unit I surveillance capsule dosimetry, the FERRET code!A-3] 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

(¢(E > 1.0 MeV) and dpa) along with associated uncertainties for the four in-vessel capsules withdrawn to date.

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

I - 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 Alvin W. 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.I.I. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library[A4]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations byASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".

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 E 944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance."

Appendix A

A-6 The following provides a summary of the uncertainties associated with the least squares evaluation of the Alvin NV. Vogtle Unit I surveillance capsule sensor sets.

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 63Cu(n,a)6Co 5%

54Fe(n,p) 54Mn 5%

58Ni(n,p)58Co 5%

238U(n,f) 37Cs 10%

237Np(n,f)137Cs 10%

59Co(n,y)6'Co 5%

These uncertainties are given at the I a 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 Alvin W. Vogtle Unit I surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Appendix A

A-7 Reaction Uncertainty 63Cu(n,a)6OCo 4.08-4.16%

54 Fe(np)54 Mn 3.05-3.11%

5"Ni(n,p) 58Co 4.494.56%

238U(nf)137Cs 0.54-0.64%

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

59Co(n,y)60Co 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 Spectrum 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 + R* 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:

Pg,

[1-aks, + 9 e-where H=(g-g')2 H=272 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 5 is 1.0 when g = g', and is 0.0 otherwise.

Appendix A

A-8 The set of parameters defining the input covariance matrix for the Alvin W. Vogtle Unit 1 calculated spectra was as follows:

Flux Normalization Uncertainty (R.)

15%

Flux Group Uncertainties (Rg, Rg')

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

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 Appendix A

A-9 A.13 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Alvin W. Vogtle Unit 1 surveillance capsules withdrawn to date are provided in Tables A-S 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 M/BE 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 Ila 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 8% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the Ica level.

Further comparisons of the measurement results 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 (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 0.88 to 1.14 for the 20 samples included in the data set. The overall average M/C ratio for the entire set of Alvin NV. Vogtle Unit I data is 1.02 with an associated standard deviation of 7.3%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BEIC comparisons for the capsule data sets range from 0.95 to 1.04 for neutron flux (E > 1.0 MeV) and from 0.96 to 1.04 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.99 with a standard deviation of 3.5% and 1.00 with a standard deviation of 3.3%, 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 Alvin W. Vogtle Unit 1 reactor pressure vessel.

Appendix A

A-10 Table A-i Nuclear Parameters Used In The Evaluation Of Neutron Sensors Monitor Reaction of Target 90% Response Product Fission Material Interest Atom Range (MeV)

Half-life Yield Fraction

(%)

Copper 63Cu (n,a) 0.6917 4.9 - 11.9 5.271 y Iron "Fe (n,p) 0.0585 2.1 - 8.5 312.1 d Nickel "Ni (n,p) 9.6808 1.5 - 8.3 70.82 d Uranium-238

' 8U (n,f) 1.0000 1.3 -6.9 30.07 y 6.02 Neptunium-237 137Np (n,f) 1.0000 0.3 -3.8 30.07 y 6.17 Cobalt-Aluminum 5 9Co (n,y) 0.0015 non-threshold 5.271 y Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the Alvin W. 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.

Appendix A

A-li Table A-2 Monthly Thermal Generation During The First Eleven Fuel Cycles Of The Alvin W. Vogtle Unit I Reactor (Reactor power of 3411 MWt from startup through Cycle 4 (3/13/93) and 3565 MWt from Cycle 5 (4/27/93) through the End of Cycle 11)

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

Year Month (MWt-hr)

Year Month (MWt-hr) 1987 3

68766 1990 3

.0 1993 3

849752 1987 4

797491 1990 4

591136 1993 4

166750 1987 5

1044332 1990 5

2311713 1993 5

2401502 1987 6

759746 1990 6

2299026 1993 6

2564437 1987 7

1835718 1990 7

2196834 1993 7

2499130 1987 8

2509822 1990 8

2512580 1993 8

2645970 1987 9

2452829 1990 9

2452206 1993 9

2560140 1987 10 707673 1990 10 2534258 1993 10 2649962 1987 11 1927388 1990 11 2428733 1993 11 2558233 1987 12 2467702 1990 12 1692955 1993 12 2646046 1988 1

1365280 1991 1

2534837 1994 1

2639758 1988 2

1387377 1991 2

2260779 1994 2

2156617 1988 3

2456340 1991 3

2495386 1994 3

2581209 1988 4

1907244 1991 4

2449552 1994 4

2557327 1988 5

2531355 1991 5

2533685 1994 5

2554173 1988 6

2444967 1991 6

2449889 1994 6

2561379 1988 7

2220349 1991 7

2534501 1994 7

2646904 1988 8

2415264 1991 8

2483204 1994 8

2448946 1988 9

2370737 1991 9

969976 1994 9

629927 1988 10 483956 1991 10 0

1994 10 1099701 1988 11 52233 1991 11 215953 1994 11 2564465 1988 12 2135007 1991 12 2466013 1994 12 2631652 1989 1

1771903 1992 1

2534684 2001 1

2650377 1989 2

1905573 1992 2

2371364 2001 2

2130621 1989 3

2533004 1992 3

2528590 1995 3

2650182 1989 4

2380073 1992 4

2239948 1995 4

2561802 1989 5

2264902 1992 5

1866712 1995 5

2630821 1989 6

2452382 1992 6

2452840 1995 6

2564944 1989 7

2443387 1992 7

2534681 1995 7

2381719 1989 8

2286024 1992 8

2535008 1995 8

2650844 1989 9

2450229 1992 9

2188889 1995 9

2519961 1989 10 2142954 1992 10 2538900 1995 10 2651520 1989 11 2391716 1992 11 2454211 1995 11 2564913 1989 12 2535607 1992 12 2536190 1995 12 2650729 1990 1

2374089 1993 1

2536730 1996 1

2650608 1990 2

1811171 1993 2

2273143 1996 2

2255312 Appendix A

iI-A-12 Table A-2 cont'd Monthly Thermal Generation During The First Eleven Fuel Cycles Of The Alvin W. Vogtle Unit 1 Reactor (Reactor power of 3411 MWt from startup through Cycle 4 (3/13/93) and 3565 MWt from Cycle 5 (4/27/93) through the End of Cycle 11)

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

Year Month (MWt-hr)

Year Month (MWt-hr) 1996 3

130446 1999 3

299858 2002 3

400069 1996 4

648324 1999 4

2560910 2002 4

728341 1996 5

2258085 1999 5

2651230 2002 5

2651604 1996 6

1467397 1999 6

2565431 2002 6

2564080 1996 7

2651000 1999 7

2651031 2002 7

2651322 1996 8

2651013 1999 8

2650940 2002 8

2627559 1996 9

2565318 1999 9

2565539 2002 9

2565444 1996 10 2654401 1999 10 2654884 2002 10 2654808 1996 11 2442399 1999 11 2565850 2002 11 2038792 1996 12 2648271 1999 12 2650654 2002 12 2380764 1997 1

2648498 2000 1

2651323 2003 1

2618718 1997 2

2393961 2000 2

2479813 2003 2

2394679 1997 3

2392019 2000 3

2650657 2003 3

2651255 1997 4

1086834 2000 4

2562272 2003 4

2561256 1997 5

2489873 2000 5

2651160 2003 5

2409539 1997 6

2565296 2000 6

2375260 2003 6

2560407 1997 7

2645858 2000 7

2649880 2003 7

2646705 1997 8

2650538 2000 8

2601536 2003 8

2647404 1997 9

503695 2000 9

1185996 2003 9

2279739 1997 10 677865 2000 10 1062698 1997 11 2565214 2000 11 2565553 1997 12 2652171 2000 12 2531649 1998 1

2650194 2001 1

2651403 1998 2

2395254 2001 2

2394810 1998 3

2650194 2001 3

2618784 1998 4

2563238 2001 4

2562381 1998 5

2547428 2001 5

2650897 1998 6

2563238 2001 6

2565700 1998 7

2648218 2001 7

2650920 1998 8

2632408 2001 8

2297481 1998 9

2563238 2001 9

2553201 1998 10 2650194 2001 10 2644259 1998 11 2561262 2001 11 2557060 1998 12 2640313 2001 12 2644057 1999 1

2648218 2002 1

2646581 1999 2

2211460 2002 2

2385611 Appendix A

A-13 t III, Table A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel

  • (E > 1.0 MeV) [n/cm2-sJ C ycle Cycle Capsule U Capsule Y Capsule V Capsule X I

9.25E+10 8.63E+10 8.63E+10 9.25E+10 2

7.55E+10 7.55E+10 8.09E+10 3

7.78E+10 7.78E+10 8.58E+10 4

6.32E+10 6.32E+10 6.84E+10 5

6.69E+10 7.24E+10 6

6.82E+10 7.33E+10 7

6.1lE+10 6.61E+10 8

6.75E+10 9

8.74E+1 0 10 8.49E+1 0 11 7.99E+10 Average 9.25E+10 7.58E+10 7.1lE+10 7.81E+10 Fuel C,

Cycle Capsule U Capsule Y l

Capsule V Capsule X I

1.00 1.14 1.21 1.19 2

1.00 1.06 1.04 3

1.03 1.10 1.10 4

0.84 0.89 0.88 5

0.94 0.93 6

0.96 0.94 7

0.86 0.85 8

0.86 9

1.12 10 1.09 1 1 1.02 Average 1.00 1.00 1.00 1.00 Appendix A

A-14 Table A-4 Measured Sensor Activities And Reaction Rates Surveillance Capsule U Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g)

(dpslg)

(dps/g)

(rps/atom) 63Cu (n,a) 6Co Top 4.82E+04 3.68E+05 3.68E+05 5.61E-17 Middle 4.38E+04 3.34E+05 3.34E+05 5.1OE-17 Bottom 4.44E+04 3.39E+05 3.39E+05 5.17E-17 Average 5.29E-17 54Fe (n,p) 4Mn Top 1.49E+06 3.56E+06 3.56E+06 5.64E-15 Middle 1.34E+06 3.20E+06 3.20E+06 5.07E-15 Bottom 1.36E+06 3.25E+06 3.25E+06 5.14E-15 Average 5.28E-15 58Ni (n,p) 8Co Top 1.26E+07 5.40E+07 5.40E+07 7.73E-15

)

Middle l1.16E+07 4.97E+07 4.97E+07 7.12E-15 Bottom 1.17E+07 5.01E+07 5.01 E+07 7.18E-15 Average 7.34E-15

-23U (n,f) 137CS (Cd) l Middle l

1.29E+05 l 5.02E+06 l 5.02E+06 I 3.30E-14 U (n,f)

Cs (Cd)

Including 235U, 2 9Pu, and y,fission corrections:

2.78E-14 237Np (n,f) '37Cs (Cd) I Middle l

1.24E+06 l 4.82E+07 I 4.82E+07 l

3.08E-13 231Np (n,f)

Cs (Cd)

Including yfission correction:

3.05E-13 59Co (n,y) OCo l

Top 1.03E+07 7.86E+07 7.86E+07 5.13E-12 Middle l

l.OIE+07 7.71E+07 l

7.71E+07 l

5.03E-12 Bottom l

1.05E+07 8.01E+07 8.01E+07 j 5.23E-12 Average l

5.13E-12 59Co (n Y) 60Co (Cd)

Top 5.21E+06 3.98E+07 3.98E+07 2.59E-12

.Y Middle 5.46E+06 4.17E+07 4.17E+07 2.72E-12 Bottom 5.58E+06 4.26E+07 4.26E+07 2.78E-12 Average 2.70E-12 Notes:

1) Measured specific activities are indexed to a counting date of February 22, 1989.
2) The average 238U (n,f) reaction rate of 2.78E-14 includes a correction factor of 0.871 to account for plutonium build-in and an additional factor of 0.967 to account for photo-fission effects in the sensor.
3) The average 237Np (n,f) reaction rate of 3.04E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

Appendix A

A-15 Tal At o

Table A~4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule Y Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g)

(dps/g)

(dpslg)

(rps/atom) 63Cu (n,a) 6Co Top 1.38E+05 3.30E+05 3.30E+05 5.03E-17 Middle 1.21E+05 2.89E+05 j 2.89E+05 4.41E-17 Bottom 1.23E+05 2.94E+05 T 2.94E+05 4.48E-17 Average 4.64E-17 14Fe (n,p) 54Mn Top 1.63E+06 2.87E+06 2.87E+06 4.55E-15 Middle 1.47E+06 2.59E+06 2.59E+06 4.1OE-15 Bottom 1.48E+06 2.60E+06 2.60E+06 4.13E-15 Average 4.26E-15 58Ni (n,p) 58Co Top 8.43E+06 4.47E+07 4.47E+07 6.40E-1 5 Middle 7.75E+06 4.11 E+07 4.11 E+07 5.89E-15 Bottom 7.63E+06 4.05E+07 4.05E+07 5.79E-15 Average 6.03E-15 238U (n,f) "37Cs (Cd)

Middle 5.07E+05 I 4.90E+06 I 4.90E+06 3.22E-14 "8U (n,f) '37Cs (Cd)

Including 235U, 2 9Pu, and y,fission corrections:

2.61E-14

'Np (n,f) 3 7 Cs (Cd Middle 3.38E+06 I 3.27E+07 I 3.27E+07 2.08E-13 Np (n,f)

Cs (Cd)

Including y,fission correction:

2.07E-13 5 9 Co (ny) 6 0Co Top 2.34E+07 5.59E+07 5.59E+07 3.65E-12 Middle 2.35E+07 5.62E+07 5.62E+07 3.66E-12 Bottom 2.34E+07 5.59E+07 5.59E+07 3.65E-12 Average 3.65E-12 5 9 Co (n y) 6 0Co (Cd)

Top 1.20E+07 2.87E+07 2.87E+07 1.87E-12 Middle 1.29E+07 3.08E+07 3.08E+07 2.01E-12 Bottom 1.29E+07 3.08E+07 3.08E+07 2.01E-12 Average 1.96E-12 Notes:

1) Measured specific activities are indexed to a counting date ofAugust 8, 1993.
2) The average 238U (n,f) reaction rate of 2.61E-14 includes a correction factor of 0.836 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 2.07E-13 includes a correction factor of 0.991 to account for photo-fission effects in the sensor.

Appendix A

__1 2

A-16 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule V Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g)

(dps/g)

I (dps/g)

(rps/atom) 63Cu (n,a) WCo Top 1.75E+05 3.07E+05 3.07E+05 4.68E-17 Middle 1.55E+05 2.72E+05 2.72E+05 4.15E-17 Bottom 1.55E+05 2.72E+05 2.72E+05 4.15E-17 Average 4.33E-17 54Fe (n,p) 54Mn Top 1.35E+06 2.73E+06 2.73E+06 4.33E-15 Middle 1.24E+06 2.51E+06 2.51 E+06 3.98E-15 Bottom 1.23E+06 2.49E+06 2.49E+06 3.94E-15 Average 4.08E-15 58Ni (n,p) 5 8Co Top 4.20E+06 4.37E+07 4.37E+07 6.25E-15 Middle 3.89E+06 4.04E+07 4.04E+07 5.79E-15 Bottom 3.88E+06 4.03E+07 4.03E+07 5.77E-15 Average 5.94E-15 238U (n,f) '37Cs (Cd) ]

Middle l

8.45E+05 l

4.82E+06 I 4.82E+06 l

3.17E-14 23SU (nf) '37Cs (Cd) l Including..5U, " 9Pu, and y,fission corrections:

2.47E-14 27Np (nf) '"CS (Cd)

Middle l

6.27E+06 I 3.58E+07 I 3.58E+07 2.28E-13 237Np (nf) '-"Cs (Cd)

Including yfission correction:

2.26E-13 59Co (ny) 6wCo Top 2.88E+07 5.05E+07 5.05E+07 3.30E-12 Middle 2.88E+07 5.05E+07 5.05E+07 3.30E-12 Bottom 2.87E+07 5.03E+07 5.03E+07 3.28E-12 Average 3.29E-12 59Co (n,'y) 60Co (Cd)

Top l

1.45E+07 2.54E+07 2.54E+07 1.66E-12 Middle l

1.50E+07 2.63E+07 2.63E+07 1.72E-12 Bottom l

1.53E+07 2.68E+07 2.68E+07 1.75E-12 Average l

1.71E-12 Notes:

1) Measured specific activities are indexed to a counting date of April 13, 1998.
2) The average 238U (n,f) reaction rate of 2.47E-14 includes a correction factor of 0.806 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 2.26E-13 includes a correction factor of 0.991 to account for photo-fission effects in the sensor.

Appendix A

A-17 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule X Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g)

(dps/g)

(dps/g)

(rps/atom) 63Cu (n,a) 60Co Top 2.52E+05 3.27E+05 3.27E+05 4.98E-17 Middle 2.23E+05 2.89E+05 2.89E+05 4.41E-17 Bottom 2.26E+05 2.93E+05 2.93E+05 4.47E-17 Average 4.62E-17 54Fe (n, p) 54Mn Top 2.41E+06 3.15E+06 3.15E+06 4.99E-15 Middle 2.20E+06 2.88E+06 2.88E+06 4.56E-15 Bottom 2.19E+06 2.86E+06 2.86E+06 4.54E-15 Average 4.70E-15 "Ni (n,p) 58Co Top 1.55E+07 4.69E+07 4.69E+07 6.71E-15 r

Middle 1.44E+07 4.35E+07 4.35E+07 6.23E-15 Average l

6.47E-15 U (nf) "3 7Cs (Cd) I Middle l

1.63E+06 l

5.94E+06 l

5.94E+06 3.90E-14 2sU (nf) "Cs (Cd)

Including 235U, 239Pu, and y,fission corrections:

2.86E-14 237Np (n,f) I37Cs (Cd) I Middle 1.0-9E+07 l

3.97E+07 l

3.97E+07 [ 2.53E-13

'3'Np (nf) '37Cs (Cd)

Including,fission correction:

2.51E-13 "Co (ny) 60Co Top 4.59E+07 5.95E+07 5.95E+07 3.88E-12 Middle 4.68E+07 6.07E+07 6.07E+07 3.96E-12 Bottom 4.76E+07 6.17E+07 6.17E+07 4.02E-12 Average 3.95E-12 59CO (n,Y) 60Co (Cd)

Top 2.39E+07 3.10E+07 3.10E+07 2.02E-12 Middle 2.56E+07 3.32E+07 3.32E+07 2.16E-12 Bottom 2.60E+07 3.37E+07 3.37E+07 2.20E-12 Average l

2.13E-12 Notes:

1) Measured specific activities are indexed to a counting date of January 20, 2004.
2) The average 238U (n,f) reaction rate of 2.86E-14 includes a correction factor of 0.758 to account for plutonium build-in and an additional factor of 0.967 to account for photo-fission effects in the sensor.
3) The average 237Np (n,f) reaction rate of 2.51 E-1 3 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

Appendix A

1L-A-18 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule U Reaction Rate Irns/atoml Best Reaction Measured Calculated Estimate IUC MIBE 63Cu(n,a)6Co 5.29E-17 4.80E-17 5.11E-17 1.10 1.04 "Fe(n,p)4 Mn 5.28E-15 5.43E-15 5.37E-15 0.97 0.98 5Ni(np)"Co 7.34E-15 7.63E-15 7.49E-15 0.96 0.98 238U(n, f)137Cs (Cd) 2.77E-14 2.94E-14 2.86E-14 0.94 0.97 237Np(n, f) 37Cs (Cd) 3.05E-13 2.90E-13 2.93E-13 1.05 1.04 59Co(n,y)60Co 5.13E-12 4.15E-12 5.03E-12 1.24 1.02 59Co(ny)60Co (Cd) 2.70E-12 2.89E-12 2.74E-12 0.93 0.99 Capsule V Reaction Rate I s/atoml Best Reaction Measured Calculated Estimate MIC MIBE 63Cu(n,a)6OCo 4.64E-17 4.11E-17 4.43E-17 1.13 1.05 54Fe(n,p)54Mn 4.26E-15 4.52E-15 4.44E-15 0.94 0.96 58Ni(n,p)53Co 6.03E-15 6.33E-15 6.19E-15 0.95 0.97 238U(nf)137Cs (Cd) 2.62E-14 2.41E-14 2.33E-14 1.09 1.12 237Np(n, f) 37Cs (Cd) 2.06E-13 2.35E-13 2.15E-13 0.88 0.96 59Co(n,y)60Co 3.65E-12 3.28E-12 3.59E-12 1.11 1.02 59Co(n,y) 6 Co (Cd) 1.96E-12 2.30E-12 2.OOE-12 0.85 0.98 Appendix A

A-19 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule Y Reaction Rnte Irmq/atnm1 Best Reaction Measured Calculated Estimate M/C M/BE 6 3Cu(n,a)"Co 4.32E-17 3.93E-17 4.17E-17 1.10 1.04 5Fe(n,p)54 Mn 4.08E-15 4.29E-15 4.27E-15 0.95 0.96 "Ni(n,p)"Co 5.93E-15 6.OOE-15 6.01E-15 0.99 0.99 238u(njf)137Cs (Cd) 2.49E-14 2.28E-14 2.28E-14 1.09 1.09 137Np(n f)137Cs (Cd) 2.26E-13 2.21E-13 2.24E-13 1.02 1.01 "Co(n,y)"Co 3.29E-12 3.07E-12 3.24E-12 1.07 1.02 "Co(n,y) 60Co (Cd) 1.71E-12 2.1 5E-12 1.74E-12 0.80 0.98 Capsule X Reaction Rate [rps/atoml Best Reaction Measured Calculated Estimate MWC M/BE 63Cu(n,a)60Co 4.62E-17 4.28E-17 4.51E-17 1.08 1.02 54Fe(n,p)54 Mn 4.69E-15 4.70E-15 4.79E-15 1

0.98 5"Ni(n,p)"Co 6.47E-15 6.58E-15 6.67E-15 0.98 0.97 23 8U(nf)137Cs (Cd) 2.86E-14 2.50E-14 2.57E-14

-1.14 1.11 137Np(nf) 37Cs (Cd) 2.51E-13 2.43E-13 2.51E-13 1.03 1.00

' 9Co(n,y) Co 3.95E-12 3.42E-12 3.89E-12 1.16 1.02 "Co(n,y) 0Co (Cd) 2.13E-12 2.38E-12 2.16E-12 0.89 0.98 Appendix A

IL A-20 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center R(E > 1.0 MeV) fn/cm2-sl l

Best Uncertainty Capsule ID Calculated Estimate (1a)

BE/C U

9.34E+10 9.04E+10 6%

0.97 Y

7.62E+10 7.25E+10 6%

0.95 V

7.17E+10 7.17E+10 6%

1.00 X

7.88E+10 8.11E+10 6%

1.03 Note:

Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Iron Atom Displac ment Rate [dpa/sl J

Best Uncertainty Capsule ID Calculated Estimate (10 BE/C U

1.81E-10 1.78E-10 8%

0.99 Y

1.47E-10 1.40E-10 8%

0.95 V

1.38E-10 1.38E-10 8%

1.01 X

1.52E-10 1.52E-10 8%

1.03 Note:

Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Appendix A

A-21 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 63Cu(n,a)60Co 1.10 1.13 1.10 1.08 5 4 Fe(n,p)54Mn 0.97 0.94 0.95 1.00 58Ni(n,p)58Co 0.96 0.95 0.99 0.98 23'U(np)' 37Cs (Cd) 0.94 1.09 1.09 1.14

'37Np(n, f)137Cs (Cd) 1.05 0.88 1.02 1.03 Average 1.00 1.00 1.03 1.05

% Standard Deviation 5.2 10.6 6.3 6.2 Note:

The overall average MWC ratio for the set of 20 sensor measurements is 1.02 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

  • (E > 1.0 MeV) dpa/s U

0.98 0.99 Y

0.96 0.95 V

1.01 1.01 X

1.04 1.03 Average 1.00 0.99

% Standard Deviation 3.5 3.3 Appendix A

1.

A-22 Appendix A 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 1995.

A-2.

WCAP-12256, "Analysis of Capsule U from the Georgia Power Company Vogtle Unit I Reactor Vessel Radiation Surveillance Program," May 1989.

A-3.

A. Schmittroth, FERRETData Analysis Core, HEDL-TME 7940, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-4.

RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.

A-5 WCAP-1393 1-RI, "Analysis of Capsule Y from the Georgia Power Company Vogtle Unit I Reactor Vessel Radiation Surveillance Program," August 1995.

A-6 WCAP-15067, "Analysis of Capsule V from the Georgia Power Company Vogtle Unit I Reactor Vessel Radiation Surveillance Program," September 1998 Appendix A

B-O' APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS Specimen prefix "AL" denotes Intermediate Shell Plate, Longitudinal Orientation

  • Specimen prefix "AT' denotes Intermediate Shell Plate, Transverse Orientation Specimen prefix "AW" denotes Weld Material
  • Specimen prefix "AH" denotes Heat-Affected Zone material
  • Load (1) is in units of lbs
  • Time (1) is in units of milli-seconds Appendix B

B-I '

5000.00 4000.00 6 3000.00 0

-J 2000.00 1000.00 0.00 1.00 2.00 300 Time-1 (is)

AL51, -250 F 4.00 5.00 6.00 5000.00 4000.00 6-3000.00 0

-J 2000.00 1000.00

/.

%AD 111111 4

l 0.00 1.00 2.00 3.00 Time-I (ms)

AL58, 250 F 4.00 5.00 6.00 Appendix B

B-2 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tirne-(ms)

AL55, 500F 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AL54, 500 F Appendix B

B-3 5000.00 _

4000.00-3 3000.00

-J 2000.00 1000.00I 0.00-0.00 1.00 2.00 3.00 Tine-1 (ms)

AL53, 750F 4.00 5.00 6.00 5000.00 4000.00 a, 3000.00 2000.00 1000.00 n0n.

i F -- -

0.00 1.00 2.00 3.00 Tine-I (ms)

AL52, 100F 4.00 5.00 6.00 Appendix B

M-BA4 4000.00 4 3000.00.

.0

-j 2000.00 1000.00 0.00 0.0 6.00 0

1.00 2.00 3.00 rAne81 (ms2 AL48, 125°F 4.00 5.00 5000.00 4000.00

.7 3000.00 coa 0

-j 2000.00 1000.00 In_

nmF,.

0.00 1.00 2.00 3.00 rtme-I (ms)

AL49, 1500 F 4.00 5.00 6.00 Appendix B

B-5 4000.00

-, 3000.00 0

0

-j 2000.00 0.00 1.00 2.00 3.00 4.00 5.00 6me,0 (ms)

AL60, 160°F 6.00 5000.00 4000.00 3000.00 03 2000.00 0.00 1.00 2.00 3.00 4.00 5.00 Tine-I (ms)

AL46, 1750F 6.00 Appendix B

B-6 5000.00 4000.00

,33000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AL57, 2000F 5000.00.

4000.00 x 3000.00 2000.00 1000.001 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time.1 (m)

AL47, 225TF Appendix B

B-7 5000.00 4000.00 a

-, 3000.00 0

2000.00 1000.00 I 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 ryne-1 (ms)

AL59, 2250 F 5000.00 4000.00 n

3000.00 2000.00 1000.00 X

0.00 0.o 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

AL50, 2500F Appendix B

B-8 5000.00 3000.00 30M0 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AL56, 2750 F 5000.00.

4000.00.

.0.

.3 3000.00 0

-j 2000.00 1000.00 N

0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AT48, -50°F Appendix B

B-9 S000.00 4000.00

.3000.

X 2000.00 1000.00 0.00 l 0.00 1 00 2.00 3.00 4.00 S.00 6.00 Tune-I (ms)

AT49, -250F 5000.00 4000.00-30000 2000.00 1000.00 i 0.00 00 1.00 2.00 3.00 4.00 5.00 6.00

T52, (ms)

AT52, 25°F Appendix B

B-10'

-j 2000.00 1000.00 0.00 0.

5000.00, 4000.00-

.0

-j 2000.00 1 000..00 00 1.00 2.00 3.00 4.00 5.00 lime-I (ms)

AT50, 500F 6.00 0.00 1.00 2.00 3.00 4.00 5.00 rwe-1 (Ms)

AT55, 750F 6.00 Appendix B

B-lI 5000.00 4000.00 6-3000.00 a

-j 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 Tfne-1 (Ms)

AT60, 100IF 5000.00.

4000.00 6.00 6.00 Q

-6 C5 3000.00 2000.00 1 000.00

.00 3.00 Time-1 (ms)

AT54, 125°F Appendix B

B-12 5000.00.

4000.00C

,3 3000.00\\

2000.00W 1000.00.

ofoo 0.00 1.00 2.00 3.00 4.00 5.00 6.00 rime-1 (ms)

AT51, 1500 F 5000.00 4000.00

, 3000.00 0

2000.00 1000.001 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tine-I (ms)

AT58, 1750F Appendix B

B-13 so000.00 4000.00 n

300.00.

0-J 2000.00.

1000.00 0.00 0.0 0

I I

I I

I T a I

I I

I 1.00 2.00 3.00 rme-1 (ms)

T56, 2000F 4.00 5.00 6.00 5000.00 4000.00 (a' 3000.00 0-J 2000.00 1000.00 0.00 0.00 P

i

-i

- F-1.00 2..00 3.00 Te-i (ms)

AT53, 2000F 4.00 5.00 6.00 Appendix B

It-B-14 5000.00 4000.00 2 3000.00 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tome-i (ms)

AT46, 225-F 5000.00 4000.00 3000.00 2000.00.

1000.00 I 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 rme- (iMs)

AT47, 2250F Appendix B

B-15 5000.00.

400.00 C

r 3000.00i 2000.00 1000.00I 0.00.

0.00 1.00 2.00 3.00 4D0 5.00 6.00 Trne-1 (ms)

AT57, 2500F 5000.001.

400.00.

v 3000.00

/

2000.00 1000.00 I 0.00 0.00 1.00 2.00 3.00 4.00 S.00 6.00 rne-I (ms)

AT59, 275TF Appendix B

B-16 5000.00 4000.00 X

3000.00 0

2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tme-1 (ms)

AW57, -1000F 5000.001 4000.00.

'a'0 3000.00 0

2000.00 1000.00 0.00l 1OS

-vP r

o l

0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tine-I (ms)

AW49, -500 F Appendix B

B-17 5000.00 4000.00

-6 3000.00-0

-J 0.00 1.00 2.00 3.00 4.00 5.00 Tune-1 (ms)

AW51, 0F 6.00 5000.00-4000.00-a Z073000.00-0

-i2000.0O01 1 000.00 0.00' 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tune-I (ms)

AW47, 100 F Appendix B

a-B-18 a

0-A Time-1 (Ms)

AW56, 250 F 5000.00*

4000.00 a 3000.00 03 0

-J 2000.00 1000.00.

00 3.00 Time-i (ms)

AW52, 50F 6.00 Appendix B

B-19' 5000.00 4000.00 3000.00 01

-j 2000.00 1 000.00 K

E _

11 ULI I

I I

I I

I I

I I

w 0.00 1DO 2.00 3.00 Time-1 (Ms)

AW50, 75TF 4.00 5.00 6.00 5000.00*

4000.00-30M.00-0-J 2000.00 1 000.00 Inni,

0.00 1.00 2.00 3.00 Time-i (ms)

AW59, 100TF 4.00 5s00 6.00 Appendix B

It l B-20 as 03 J

0

-j 5000.00 4000.00 3000.00 2000.00 1000.00 0.00 1.00 2.00 3.00 4.00 5.00 Time-1 (ms)

AW53, 125TF 6.00

.00 3.00 Trie-I (ms)

AW48, 125TF 6.00 Appendix B

B-21 5000.00 4000.00

.0 3000.00'

~.0 2000.00 1000.00' 0.00 1.00 2.00 3.00 4.00 5.00 Time-1 (ms)

AW55, 1500F 6.00 5000.007 4000.00 1s 3000.00 0

2000.00-1 0 0 0L

.0 I 0.00 1.00 2.00 3.00 4.00 5.00 6.00 lMfe-l (ins)

AW46, 150 0F Appendix B

B-22 5000.00 4000.00

\\

x3000.00 2000.00I 1000.00 t Ls 0.W 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (Ms)

AW58, 1750F 4000.00.

Ca 3000.00 0-j 2000.00 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tme-1 (ms)

AW60, 2000F Appendix B

B-23 a

0-J 5000.00 4000.00 3000.00 2000.00 1000.00 0.00I DO 1 Do 2.00 3.00 Time-1 (ms)

AW54, 225F 4.00 sO0 6.00 5000S.

4000.00-n a, 3000.00

-J 2000.00.

1000.00.

IJLI Few 0.00 1.00 2.00 3.00 Twne-I (ms)

AH49, -1500F 4.00 5.00 6H0 Appendix B

l.

B-24 5000.00 4000.00 X7 3000.00 0

-J 2000.00 1000.00*

0.00.

0.

5000.00.

4000.00-2000.00 1000.00*

00 1.00 2.00 3.00 4.00 5.00 Time-i Cms)

AH56, -1000F 0.00 1.00 2.00 3.00 4.00 5.00 Time-I (ms)

AH48, -750F 6.00 Appendix B

B-25 5000.00 4000.00

, 3000.00 0

-J 2000.00 1000.00.

0.00 0.

5000.00 4000.00 3000.00' 0

-J 2000.00 1000.00 00 1.00 2.00 3.00 4.00 5.00 Tre-I (is)

AHS8, -500F 6.00 0.00 1.00 2.00 3.00 4.00 5.00 Timfe-1 (Ms)

AH59, -500F 6.00 Appendix B

it-B-26

.0 a6 00

-j

.01

-j rmie-1 (ms)

AH54, -250F 0.00 1.00 2.00 3.00 4.00 5.00 Time-1 (ms)

AH60, 250 F 6.00 Appendix B

B-27 3000.00.

2000.00l 1000.00 0.00I 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AH47, 75TF 5000.001 4000.00 m

2000.00 {

2000.00 1000.00.

0.00 0.00 1.00 2.00 3.00 4.00 500 6.00 Time-i (ms)

AH51, 1000 F Appendix B

i-B-28 5000.00 4000.00 3000.00 2000.00 1000.00I 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 rime.1 (ms)

AH55, 1000F 5000.00 4000.00 2 3000.00 2000.00

\\

1000.00 I 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

AH46, 1500 F Appendix B

B-29 5000.00 4000.00 m 3000.00 0-a 2000.00 1000.00' ono.

0.

5000.00 4000.00 10

'm' 3000.00' 0

-J 2000.00 1000.00 00 1.00 2.00 3.00 4.00 5.00 lime-i (ms)

AH53, 200TF 6.00 0.00 1.00 2.00 3.00 4.00 5.00 lime-i (Ms)

AH50, 200TF 6.00 Appendix B

t.

B-30 5000.00 4000.00 a

7 3000.00 J

2000.00 1000.00*

0.00.

0.i 5000.00 4000.00 la 3000.00 0

-j I

.00 3.00 Time-I (ms)

AH57, 2250F 6.00 0.00 1.00 2.00 3.00 4.00 5.00 Trne-1 (ms)

AH52, 2250F 6.00 Appendix B

C-o.

APPENDIX C CHARPY V-NOTCH PLOTS FOR CAPSULE X USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Appendix C

c-i.

Contained in Table C-I are the upper shelf energy values used as input for the generation of the Charpy V-notch plots for Capsule X using CVGRAPH, Version 5.0.2. Applicable Charpy V-notch plots for Capsule U, Y, and V are included in WCAP-15067 61. The definition for Upper Shelf Energy (USE) is given inASTM E185-82, 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 set 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 100% shear Charpy data as the USE. Hence, the USE values reported in Table C-1 and 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 Fixed in CVGRAPH [ft-bb]

Material Unirradiated Capsule U Capsule Y Capsule V Capsule X (ft-lbs)

(ft-lbs)

(ft-lbs)

(ft-lbs)

(ft-lbs)

Intermediate Shell Plate 122 133.5 131.6 118 109 B8805-3 (Long.)

Intermediate Shell Plate 96 98 106 94 93 B8805-3 (Trans.)

Weld Metal 145 156 144 142 141 (Heat 1 83653)

HAZ Material 136 129 124 121 127 Appendix C

UNIRRADIATED (LONGITUDNAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 01:35 PM Page 1 Coefficients of Curve I A = 62.1 B = 59.9 C = 93.5 TO = 41.02 D = O.OOE+00 Equation is A + B

  • fTanh((T-To)/(C+DT))]

Upper Shelf Energy= I 22.0(Fixed)

Lowver 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: UNIRR Fluence:

n/cm^2 300 250

('

200 8

0 0

IL M 150 a) r-z

> 100 50 0

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN

-40. 00

-40. 00

-40. 00

- 20. 00

- 20. 00

-20. 00

.00

.00

.00 1 1. 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 Differenti3l

-9.

19

-1. 19 1I 1. 19

26. 25

. 25

15. 75 14.61 9.61 4.61 I

C1-2

a-UNIRRADIATED (LONGITUDNAL ORIENTATION)

Page 2 Plant: VOGTLE I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: UNLR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input CVN

40. 00
40. 00
40. 00
80. 00
80. 00
80. 00 100. 00 1 00. 00 1 00. 00 120.00 120.00 120.00 180. 00 180. 00 180. 00 260. 00 260. 00 320. 00 320. 00 48.00
62. 00
60. 00 93.00
64. 00
70. 00 84.00 107.00 1 10. 00 100. 00 116. 00 109. 00 126. 00 115.00 11 6.00 129. 00 12 1. 00 131. 00 1 9.00 Computed CVN 61.45 61.45 61.45
85. 72
85. 72
85. 72 95.56
95. 56 95.56 103. 33 103.33 103. 33 116.17 116. 17

] 16.17 120. 90 120.90 121. 69 121.69 Differential

- 13.45

.55

- 1.45 7.28

-21.72

- 15. 72 I 1.56 11.44 14.44

- 3. 33 12.67 5.67

9. 83

-1.17

-.17

8. 10

.10 9.31

- 2.69 Correlation Coefficient =.962

(,-,3

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 01:36 PM Page 1 Coefficients of Curve 2 A = 55.6 B = 53.4 C = 95.98 TO = 131.71 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=] 09.0(Fixed)

Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=81.6 Deg F Temp@50 ft-lbs=1221.7 Deg F Plant: \\'OGTLE I Material: SA533B]

Heat: C0623-1 Orientation: LT Capsule: X Fluence:

n1cmA2 300 250 w

200

, -00 IL i 150 0

z

> 100 0

50 0 k-

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature

-25. 00

25. 00
50. 00
50. 00 75.00 100. 00 125. 00 150. 00 1 60. 00 Input CVN 3.00 8.00
30. 00 13.00 35.00 42.00 59.00
48. 00
66. 00 Computed CVN Differential
6. 13
12. 63 18.66 1S. 66
27. 27
38. 57
51. 87
65. 66
70. 90

- 3. 13

-4. 63 1]. 34

-5. 66

7. 73 3.43
7. 13

- 17. 66

-4. 90 I

P-A

CAPSULE X (LONGITUDINAL ORIENTATION)

Page 2 Plant: VOGTLE I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: X Fluence:

n/cmA2 Charpy V-Notch Data Temperature 175.00 200. 00 225. 00 225. 00 250. 00 275. 00 Input CVN

75. 00
75. 00 117. 00 105. 00 103. 00 113..00 Computed CVN
78. 18
88. 26
95. 63
95. 63 100. 63 103. 87 Differential

- 3.18

-13.26 21.37 9.37 2.37

9. 13 Correlation Coefficient =.965

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03119/2004 01:57 PM Page 1 Coefficients of Curve I A = 43.68 B = 43.68 C = 101.09 TO = 39.13 D = O.OOE+0O 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 DegF Plant: VOGTLE I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: UNIRR Fluence:

n/cmA2 200 150 2

0 2 100

'I-a) 50 0

-300 0

300 Temperature in Deg F 600 Charpy V-Notch Data Temperature Input L.E.

-40. 00

-40. 00

-40. 00

-20. 00

-20. 00

-20. 00

. 00

. 00

.00 7.00

12. 00 6.00
38. 00 22.00 8.00
39. 00
34. 00 33.00 Computed L.E.
15. 10
15. 10
15. 10
20. 69 20.69
20. 69
27. 57
27. 57
27. 57 Differential

- 8. 10

- 3. 10

-9.

10 17.31 1.31

12. 69 11.43 6.43 5.43 C-6

UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: VOGTLE I Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input 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

81. 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

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 01:58 PM Page 1 Coefficients of Curve 2 A = 38.24 B = 38.24 C = 104.72 TO = 139.31 D = O.OOE+0O 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 niils=130.5 Deg F Plant: VOGTLE I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: X Fluence:

n/cmA2 200 150 2

E aa 0

2.100 50 0

-300 0

300 Temperature in Deg F 600 Charpy V-Notch Data Temperature Input L.E.

-25. 00

25. 00
50. 00
50. 00
75. 00 100. 00 125. 00 150.00 160. 00

. 00

3. 00
14. 00
7. 00
21. 00
27. 00
40. 00
36. 00
45. 00 Computed L.E.
3. 18 7.75 11.76 11.76 17.32
24. 52
33. 05
42. 13
45. 70 Differential

-3.

18

-4.75

2. 24

-4. 76

3. 68
2. 48
6. 95

- 6.13

-. 70 C-8

CAPSULE X (LONGITUDINAL ORIENTATION)

Plant: VOGTLE I Orientation: LT Page 2 Material: SA533B1 Capsule: X Fluence:

Heat: C0623-1 n/cmA2 Charpy V-Notch Data Temperature Input L.E.

175. 00 200. 00 225. 00 225. 00 250. 00 275.00

50. 00 51.00
70. 00
68. 00 67.00
70. 00 Computed L.E.
50. 79 58.21 64.02
64. 02
68. 24 71.15 Differential

-. 79

-7.

21 5.98

3. 98

- 1. 24

- 1. 15 Correlation Coefficient =.985

. 1.

C-9

UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 01:49 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))]

Temperature at 50% Shear = 64.4 Plant: VOGTLE 1 Material: SA533B1 Heat: C0623-1 Orientation: LT Capsule: UNIRR Fluence:

n/cmA2 125 100 3.-

co r-0i 0P 75 50 25 o 4-

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear

-40. 00

-40.00

-40.00

- 20. 00

- 20.00

- 20.00

.00

.00

.00 5.00 14.00 9.00 27.00 14.00 5.00

30. 00
30. 00 25.00 Computed Percent Shear
9. 36
9. 36
9. 36 13.76 13.76 13.76
19. 78
19. 78
19. 78 Differential

-4.36

4. 64

-. 36 13.24

.24

- 8.76 10.22 10.22

5. 22

.i C-10

a-UNIRRADIATED (LONGITUDINAL ORIENTATION)

Page 2 Plant: VOGTLE 1 Material: SA533B1 H

Orientation: LT Capsule: UNIRR Fluence:

.eat: C0623-1 n/cmA2 Charpy V-Notch Data Temperature

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 Percent Shear 30.00 36.00
36. 00
60. 00
45. 00
40. 00
55. 00
75. 00
75. 00
80. 00
85. 00
85. 00 100. 00 100. 00 100. 00 100. 00 100. 00 100. 00 100. 00 Computed Percent Shear
37. 05
37. 05 37.05 58.42
58. 42 58.42 68.46 68.46 68.46 77.03
77. 03
77. 03 92.52 92.52
92. 52 98.60 98.60
99. 62
99. 62 Differential

-7. 05

- 1. 05

- I. 05 1.58

- 13.42

- 18.42

- 13. 46 6.54 6.54 2.97 7.97 7.97 7.48 7.48 7.48 1.40 1.40

.38

.38 Correlation Coefficient =.975 C-1l1

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 01:50 PM Page 1 Coefficients of Curve 2 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 I Material: SA533BI Heat: C0623-1 Orientation: LT Capsule: X Fluence:

nlcmA2 125 100 4-a cm 0

coCa) 2 0~

75 50 25 0 4-

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear

- 25. 00 25.00

50. 00
50. 00
75. 00 100. 00 125.00 150. 00 160. 00
2. 00
5. 00
15. 00
25. 00
25. 00
30. 00
40. 00
45. 00
50. 00 Computed Percent Shear 2.49
7. 13 11.74
11. 74 18.74 28.56
40. 94 54.58
59. 96 Differential

-. 49

- 2. 13 3.26 13.26 6.26

1. 44

-. 94

-9.58

-9. 96 C-12 I

I

CAPSULE X (LONGITUDINAL ORIENTATION)

Plant: VOGTLE 1 Orientation: LT Page 2 Material: SA533BI Capsule: X Fluence:

Heat: C0623-1 n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear 175.00 200. 00 225. 00 225. 00 250. 00 275. 00 65.00

70. 00 100. 00 100. 00 100. 00 100. 00 Computed Percent Shear 67.57
78. 32 86.23
86. 23 91.57
94. 96 Differential

- 2. 57

-8.

32 13.77 13.77

8. 43
5. 04 Correlation Coefficient =.975 C-13

VNIRRADIATED (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:41 PM Page 1 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: SA533B1 Heat: C0623-1 Orientation: TL Capsule: UNIRR Fluence:

n/cmA2 300 250

-D 200 9

0 0

M150 a) z i

100 50 0

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN

-40. 00

-40. 00

. 00

. 00

. 00

40. 00
40. 00
40. 00
80. 00 9.00 16.00 26.00
24. 00
24. 00 35.00
46. 00 53.00
55. 00 Computed CVN 13.40
13. 40
23. 86
23. 86
23. 86
39. 67
39. 67
39. 67
58. 07

. Differential

-4.

40

2. 60
2. 14

.14

.14

-4.67

6. 33
13. 33

-3.

07 C-14

1_

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2 Plant: VOGTLE 1 Material: SA533BI H

Orientation: TL Capsule: UNIRR Fluence:

Seat: C0623-1 nlcmA2 Charpy V-Notch Data Temperature

80. 00
80. 00 120.00 120.00 120. 00 140. 00 140.00 180. 00 180.00 180. 00 240. 00 240. 00 320. 00 320. 00 Input CVN
45. 00 56.00
69. 00 73.00
71. 00
84. 00 79.00
94. 00
98. 00
90. 00
97. 00 102. 00
97. 00
96. 00 Computed CVN
58. 07
58. 07
73. 99
73. 99 73.99
79. 98 79.98
88. 02 88.02
88. 02 93.43 93.43
95. 46
95. 46 Differential

-13. 07

-2.

07

-4.

99

-.99

-2.

99 4.02

-.98 5.98 9.98 1.98

3. 57 8.57 1.54

.54 Correlation Coefficient =.982 C-15

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:41 PM Page 1 Coefficients of Curve 2 300 250 o

,- 200 0

0LL E0 150 a) r-w z

> 100 50 0

A = 47.6 B = 45.4 C = 124.11 TO = 128.59 D = 0.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 Temp@50 ft-lbs=135.2 Deg F Plant: VOGTLE I Material: SA533BI Heat: C0623-l Orientation: TL Capsule: X Fluence:

n/cm^2 0 ~--

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN Computed CVN Differential

-50. 00

- 25.00

25. 00
50. 00 75.00 100. 00 125. 00 150. 00 175. 00 5.00 7.00
14. 00
25. 00
33. 00
42. 00 47.00
52. 00 64.00 7.04 9.25 16.59
22. 16
29. 13 37.32 46.29 55.36 63.83

-2.04

-2.25

-2.59

2. 84
3. 87
4. 68

.71

-3.

36

.17

  • i C-16

CAPSULE X (TRANSVERSE ORIENTATION)

Plant: VOGTLE 1 Orientation: TL Page 2 Material: SA533B1 Heat: C0623-1 Capsule: X Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input CVN 200. 00 200. 00 225. 00 225. 00 250. 00 275. 00

64. 00 61.00
85. 00
66. 00
95. 00 98.00 Computed CVN 71.18 71.18
77. 15
77. 15
81. 75
85. 16 Differential

-7.

18

- 10. 1 8

7. 85

-I 1. 15

13. 25
12. 84 Correlation Coefficient =.970 C-17

UNIRRADIATED (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 03:01 PM Page 1 Coefficients of Curve I A =48. B = 48. C = 160.8 TO = 99.62 D = O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=96.0(Fixed)

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

Temp. @L.E. 35 niils=55.0 Deg F Plant: VOGTLE I Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: UNIRR Fluence:

n/cmA2 200 150 n0 2 100 1..

0 9

-300 0

300 Temperature in Deg F 600 Charpy V-Notch Data Temperature Input L.E.

-40. 00

-40. 00

.00

.00

. 00

40. 00
40. 00
40. 00
80. 00 5.00 II.

00 19.00

17. 00
19. 00
27. 00
37. 00
41. 00
40. 00 Computed L.E.

14.38 14.38 21.56 21.56 21.56 30.98 30.98 30.98

42. 17

-9. 38

-3. 38

- 2. 5 6

-4. 56

-2. 56

- 3. 9 8

6. 02
10. 02

- 2. 17 Differential C-18

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2 Plant: VOGTLE 1 Material: SA533B1 Heat: C0623-1 Orientation: TL Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input L.E.

80. 00
80. 00 120. 00 120. 00 120. 00 140. 00 140. 00 180.00 180. 00 180. 00 240. 00 240. 00 320. 00 320. 00
37. 00
44. 00 53.00
56. 00
58. 00
63. 00 62.00 73.00
71. 00
80. 00
  • 77. 00
74. 00
80. 00
74. 00 Computed L.E.
42. 17
42. 17
54. 05
54. 05
54. 05 59.81
59. 81
70. 18
70. 18
70. 18
81. 74 81.74
90. 18
90. 18 Differential

-5.

17

1. 83

- 1. 05

1. 95
3. 95
3. 19
2. 19
2. 82

.82

9. 82

- 4. 74

-7.

74

- 10. 18

- 16. 18 Cornelation Coefficient =.966 C-19

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 03:01 PM Page 1 200 0

r-B.n CE 150 100 50 Coefficients of Curve 2 A = 46.5 B = 46.5 C = 174.11 TO = 190.53 D = 0.0OE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf L.E.=93.0(Fixed)

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

Temp.@L.E. 35 mils= 146.6 Deg F Plant: VOGTLE I Material: SA533BI Heat: C0623-1 Orientation: TL Capsule: X Fluence:

nlcmA2 fa 1P 0

-300 0

300 600 Temperature in Deg F Charpy V-Notch Data Temperature

-50. 00

- 25.00

25. 00
50. 00
75. 00 100. 00 125. 00 150. 00 175. 00 Input L.E.

.00

.00

7. 00
15. 00 25.00 31.00 32.00 38.00
45. 00 Computed L.E.

5.52 7.21

12. 09 15.44 19.50
24. 29
29. 78
35. 87 42.36 Differential

-5.

52

-7.21

-5. 09 44 5.50 6.71 2.22

2. 13
2. 64 C-20

itQ CAPSULE X (TRANSVERSE ORIENTATION)

Plant: VOGTLE I Orientation: TL Page 2 Material: SA533BI Heat: C0623-1 Capsule: X Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input L.E.

200. 00 200. 00 225. 00 225. 00 250. 00 275. 00

43. 00 44.00
60. 00
50. 00
66. 00
66. 00 Computed L.E.
49. 03
49. 03 55.59 55.59 61.79
67. 44 Differential

-6.

03

-5. 03 4.41

-5. 59 4.21

- 1. 44 Correlation Coefficient =.976 E iC-21

UNIRRADIATED (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:54 PM Page 1 Coefficients of Curve I 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: SA533B1 Heat: C0623-1 Orientation: TL Capsule: UNIRR Fluence:

n/cMA2 125 100 a)

U, a)0) 0)

a.~

75 50 25 O 4-

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear

-40.00

-40.00

. 00

. 00

. 00

40. 00
40. 00
40. 00
80. 00 9.00 14.00 20.00 14.00 25.00 18.00
36. 00
40. 00
30. 00 Computed Percent Shear
6. 04
6. 04 13.75 13.75 13.75
28. 35
28. 35 28.35 49.54 Differential 2.96 7.96
6. 25 6 2 c

.2 5

11. 25

- 10.35

7. 65
11. 65

- 19.54

.i C-22

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2 Plant: VOGTLE 1 Material: SA533B1 H

Orientation: TL Capsule: UNIRR Fluence:

reat: C0623-1 n/cmA2 Charpy V-Notch Data Temperature

80. 00
80. 00 120.00 120. 00 120. 00 140. 00 140. 00 180. 00 180. 00 180.00 240. 00 240. 00 320. 00 320. 00 Input Percent Shear
41. 00 48.00
65. 00
70. 00
70. 00
90. 00
80. 00 1 0 0. 0 0 1 00. 00 1 0 0. 00 1 00. 00 1 0 0. 00 1 00. 00 100.00 Computed Percent Shear 49.54 49.54 70.90
70. 90
70. 90
79. 33
79. 33
90. 50
90. 50
90. 50
97. 38
97. 38
99. 57
99. 57 Differential

- 8. 54

-1. 54

-5.90 90

-. 90

10. 67

.67 9.50

9. 50
9. 50 2.62 2.62

.43

.43 Correlation Coefficient =.974 C-23

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:54 PM Page 1 Coefficients of Curve 2 A = 50. B

50. C= 77.11 TO = 145.71 D =O.OOE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = 145.8 Plant: VOGTLE I Material: SA533B I Heat: C0623-1 Orientation: TL Capsule: X Fluence:

nlcmA2 125 100 1-s0 CO

  • -a a) 0 0) 75 50 D '.

II I

II IIIjb II0 11 II J-I III II I

AIIa0 a0I I C;1A PI 1-

--- 9 I

I I

25 0

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature

-50. 00

- 25. 00

25. 00 50.00
75. 00
00. 00 125.00 150. 00 175.00 Input Percent Shear 2.00 2.00 5.00 10.00
20. 00
25. 00 35.00 50.00
65. 00 Computed Percent Shear

. 62 1.18

4. 19 7.71 13.78 23.40
36. 88 52.78
68. 13 Differential 1.38

.82

.81 2.29 6.22

1. 60

- 1.88

-2.78

-3. 13

.C-24

a-CAPSULE X (TRANSVERSE ORIENTATION)

Plant: VOGTLE 1 Orientation: TL Page 2 Material: SA533B1 Heat: C0623-1 Capsule: X Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear 200. 00 200. 00 225. 00 225. 00 250. 00 275. 00

75. 00 75.00 1 00. 00
95. 00 1 00. 0 0 1 00. 00 Computed Percent Shear
80. 34
80. 34
88. 66
88. 66
93. 73
96. 62 Differential

-5.

34

-5. 34

11. 34
6. 34 6.27
3. 38 Correlation Coefficient =.993 C-25

UNIRRADIATED WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:54 AM Page 1 Coefficients of Curve I 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)

Temp@30 ft-lbs=-57.2 Deg F Temp@50 ft-lbs=-30.3 Deg F Plant: VOGTLE 1 Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: UNIRR Fluence:

nlcmA2 300 250 200 4-00IL El 150 a)a, ULi z

> 100 50 0

o

-300

-200

-100 0

100 200 300 400 500 Temperature in Deg F Charpy V-Notch Data 600 Temperature Input CVN

- 120. 00

- 120. 00

- 80. 00

- 80. 00

- 80. 00

- 60. 00

- 60.00

- 60.00

-40. 00

7. 00 5.00 12.00
15. 00 16.00
10. 00
11. 00
24. 00 58.00 Computed CVN
8. 17
8. 17 18.61 18.61 18.61 28.34
28. 34
28. 34
42. 02 Differential

-1. 17

- 3.17

- 6. 61

-3.61

-2.61

- 18. 34

- 17. 34

-4.34 15.98

-C-26

UNIRRADIATED WELD METAL Page 2 Plant: VOGTLE I Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input CVN

40. 00
40. 00

-20. 00

-20. 00

-20. 00 00 00

.00

40. 00
40. 00
40. 00
60. 00
60. 00
60. 00
80. 00 80.00
80. 00 120. 00 120.00 120.00 180. 00 180. 00 180. 00 240. 00 240. 00 240. 00 320. 00 320. 00 320. 00
22. 00 35.00
94. 00
76. 00
86. 00 68.00
95. 00
60. 00 101.00 101.00 100. 00 130. 00 118. 00 123. 00 128. 00 126. 00 153. 00 141.00 140. 00 135.00 144. 00 158. 00 154. 00 144. 00 135. 00 145. 00 144.00 143.00 143. 00 Computed CVN 42.02 42.02
59. 36
59. 36
59. 36
78. 64
78. 64 78.64 112.77 112.77 112. 77 124. 36 124. 36 124.36 132. 27 132.27 132.27 140.46 140.46 140. 46 144.09 144. 09 144. 09 144. 82 144. 82 144. 82 144.98 144. 98 144. 98

-20. 02

-7. 02

34. 64 16.64
26. 64 I10. 64
16. 36 18.64

-11. 77

-11. 77

- 12.77 5.64

- 6.36

- 1.36

-4.27

-6.

27

20. 73

.54

-. 46

-5. 46 09 13.91

9. 91

-. 82

- 9. 82

.18

- 1.98

- 1. 98 Differential Correlation Coefficient =.972 C-27

CAPSULE X WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed.on 03/23/2004 10:55 AM Page 1 Coefficients of Curve 2 A = 71.6 B = 69.4 C = 44.79 TO = 27.14 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=1 41.0(Fixed)

Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-3.8 Deg F Temp@50 ft-lbs= 12.8 Deg F Plant: VOGTLE I Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: X Fluence:

n/cmA2 300 250 W

-. 200 100 LL El 150 0

z

> 100 C.

0E 113 l- ---------- l--

1 50 0

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN

- 100. 00

-50. 00

.00 I0. 00

25. 00
50. 00
75. 00 100. 00 125. 00 3.00 6.00 22.00
42. 00 85.00 1 01. 00 127.00 136.00 112.00 Computed CVN
2. 67 6.49 34.03 46.26 68.28 104.23 126. 35 135. 84 139. 27 Differential

.33

-. 49

- 12.03

-4.

26 16.72

- 3.23

.65

.16

- 27. 27 C-28

it-CAPSULE X WELD METAL Plant: VOGTLE I Orientation: NA Page 2 Material: SAW Heat: WIRE:83653 Capsule: X Fluence:

nIcmA2 Charpy V-Notch Data Temperature 125. 00 150. 00 1.50. 00 175. 00 200. 00 225. 00 Input CVN 128. 00 127. 00 137. 00 142. 00 142. 00 145. 00 Computed CVN 1 39. 27 140.43 140.43 140. 81 140. 94 140.98 Differential

- 11.27

- 13.43

- 3. 43 1.19

1. 06 4.02 Correlation Coefficient =.983 C-29

UNIRRADIATED WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 11:03 AM Page 1 Coefficients of Curve 1 A = 44.13 B = 44.13 C = 54.42 TO = -21.21 D = O.OOE+O0 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 mils=-32.6 Deg F Plant: VOGTLE I Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: UNTRR Fluence:

nlcmA2 200 150 u,

0 I-0 la 100 50 o 4-

-300 0

300 Temperature in Deg F Charpy V-Notch Data 600 Temperature Input L.E.

- 120. 00

- 12 0. 00

-80. 00

- 80. 00

-so. 00

- 60. 00

- 60. 00

- 60. 00

-40. 00 3.00 2.00 7.00

10. 00 10.00 7.00 7.00 18.00 40.00 Computed L.E.

2.28

2. 28
9. 12
9. 12
9. 12
17. 10
17. 10
17. 10 29.46 Differential

.72

-. 28

-2.

12

.88

.88

- 10. 10

-10. 10

.90 10.54 C-30

ii-UNIRRADIATED WELD METAL Page 2 Plant: VOGTLE 1 Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input L.E.

-40. 00

-40. 00

-20. 00

-20. 00

-20. 00

.00 00

.00 40.00

40. 00
40. 00
60. 00
60. 00
60. 00
80. 00
80. 00
80. 00 120. 00 120. 00 120. 00 180.00 180. 00 180. 00 240. 00 240. 00 240. 00 320. 00 320. 00 320. 00 16.00
26. 00
67. 00
56. 00
52. 00
50. 00
68. 00
45. 00
72. 00
74. 00
72. 00
87. 00
77. 00
86. 00 87.00
84. 00 107. 00
90. 00
88. 00
89. 00
97. 00
82. 00
91. 00
90. 00
88. 00
88. 00
86. 00
92. 00
77. 00 Computed L.E.

29.46 29.46

45. 10
45. 10
45. 10
60. 50 60.50
60. 50
79. 83
79. 83
79. 83
84. 00
84. 00
84. 00
86. 16
86. 16
86. 16
87. 76 87.76
87. 76 88.20 88.20 88.20 88.25 88.25
88. 25 88.25 88.25
88. 25 Differential

-13.46

-3. 46

21. 90
10. 90 6.90

- 10. 50 7.50

- 15.50

-7. 83

-5. 83

-7. 83 3.00

-7. 00

2. 00

.84

-2.

16

20. 84
2. 24

.24

1. 24
8. 80

-6. 20

2. 80 1.75

-. 25

-. 25

-2.25 3.75 I11.25 Correlation Coefficient =.969 C-31

CAPSULE X WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 11:03 AM Page 1 Coefficients of Curve 2 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: VOGTLE I Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: X Fluence:

n/cm^2 200 C

In0 r-L.

150 100 50 1

0

-300 0

300 600 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E.

- 100. 00

-50. 00

.00

10. 00
25. 00
50. 00
75. 00 1 00. 00 125. 00

. 00

. 00

15. 00
31. 00
25. 00
54. 00
73. 00
85. 00 76.00 Computed L.E.

.21

1. 92 15.01 21.28
33. 35 55.53
70. 96 78.05
80. 69 Differential

-.21

- 1. 92

-. 01 9.72

-8.35

-1.53 2.04 6.95

-4.69 C-32

CAPSULE X WELD METAL Page 2 Plant: VOGTLE 1 Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: X Fluence:

nlcmA2 Charpy V-Notch Data Temperature Input L.E.

1 25. 00 150. 00 150. 00 175.00 200. 00 225. 00 84.00

82. 00
80. 00
83. 00
81. 00 78.00 Computed L.E.
80. 69 81.59 81.59
81. 89 81.99
82. 02 Differential 3.31

.41

- 1.59 1.11 99

- 4. 02 Correlation Coefficient =.991 C-33

UNIRRADIATED WELD METAL CVGRAPH 5.0.2 Hyperbo]ic Tangent Curve Printed on 03/23/2004 10:59 AM Page 1 Coefficients of Curve 1 A = 50. B = 50. C = 74.21 TO = -6.16 D = O.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: UNIRR Fluence:

n/cmA2 125 100 S-C, IL 75 50 25 -_

0 -

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear

- 120. 00

- 120. 00

- 8 0. 00

-80. 00

- 80. 00

- 60. 00

- 60. 00

- 60. 00

-40. 00

.00

.00 5.00 5.00 5.00 5.00 5.00 25.00

43. 00 Computed Percent Shear
4. 44
4. 44 12.03
12. 03
12. 03 18.98
18. 98 18.98
28. 66 Differential

-4. 44

-4. 44

-7.

03

-7.

03

-7.

03.

- 13. 98

- 13. 98

6. 02
14. 34 C-34

UNIRRADIATED WELD METAL Page 2 Plant: VOGTLE 1 Material: SAW Heat: WIRE:83653 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature

-40. 00

-40. 00

-20. 00

- 20. 00

- 20. 00

.00

.00

.00

40. 00
40. 00
40. 00
60. 00
60. 00
60. 00 80.00
80. 00
80. 00 120.00 120. 00 120.00 180. 00 1 80. 00 180. 00 240. 00 240. 00 240. 00 320. 00 320. 00 320.00 Input Percent Shear 15.00
33. 00
65. 00
56. 00
50. 00
48. 00
65. 00
35. 00
75. 00
70. 00
75. 00
85. 00
85. 00
85. 00
80. 00
80. 00 100. 00 100. 00 100. 00 100. 00 100. 00 100. 00 100. 00 1 00. 0 0 100. 00 1 00. 0 0 100. 00 100. 00 100.00 Computed Percent Shear
28. 66
28. 66
40. 78
40. 78
40. 78
54. 14
54. 14
54. 14
77. 62
77. 62
77. 62 85.61
85. 61
85. 61 91.07 91.07 91.07 96.77
96. 77
96. 77
99. 34
99. 34
99. 34
99. 87
99. 87
99. 87
99. 98
99. 98
99. 98 13.66
4. 34 24.22 15.22 9.22

-6.

14

10. 86
19. 14

-2.

62

-7.

62

-2.

62

. 61

. 61

11. 07
11. 07 8.93 3.23 3.23
3. 23 66 66 66

.13

.13

.13 02 02 02 Differential Correlation Coefficient =.972 C-35

CAPSULE X WELD METAL CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed~on 03/23/2004 10:59 AM Page 1 Coefficients of Curve 2 A

50. B = 50. C = 60.36 TO = 11.87 D = 0.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/cmA2 125 100 I) s CO 1..

r-e) 0.

75 50 25 0 !

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input Percent Shear

- 100. 00

-50. 00

.00

10. 00
25. 00
50. 00
75. 00 100. 00 125. 00
2. 00
10. 00
40. 00
45. 00
65. 00
80. 00
90. 00 95.00
90. 00 Computed Percent Shear 2.40 11.40
40. 29 48.45 60.71
77. 96 89.01
94. 88 97.70 Differential

-.40

- 1.40

-. 29

- 3.45

4. 29 2.04

.99

.12

-7.70 C-36

It CAPSULE X WELD METAL Page 2 Plant: VOGTLE I Material: SAW Heat: WERE:83653 Orientation: NA Capsule: X Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear 125. 00 150. 00 150. 00 175.00 200. 00 225. 00

90. 00
95. 00
98. 00 100. 00 100. 00 100. 00 Computed Percent Shear 97.70
98. 98 98.98
99. 55
99. 80 99.91 Differential

-7. 70

-3. 98

-.98

.45

.20

.09 Correlation Coefficient =.995 i

C-37

UNIRRADIATED HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:23 AM Page I Coefficients of Curve I A = 69.1 B = 66.9 C = 84.2 TO = -30.72 D = O.0OE+OO Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=136.0(Fixed)

Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-87.0 Deg F Temp@50 ft-lbs=-55.4 Deg F Plant: VOGTLE I Material: SAW Heat: B8805-1 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 300 250 Ln 200 a00 50 Im150 z

>100 50 0

=

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature Input CVN

-180.00

-180.00

- 140. 00

- 140. 00

-110.

00

-I I 0. 00

-110. 00

- 80.00

-80. 00

4. 00
6. 00
4. 00
5. 00
17. 00
10. 00 16.00
25. 00
29. 00 Computed CVN
5. 95 5.95
11. 49 11.49
19. 87
19. 87
19. 87 33.88
33. 88 Differential

- 1. 95

.05

-7.

49

- 6.49

-2.87

-9. 87

- 3. 87

- 8. 88

- 4. 88

, I C-38

I, UNIRRADIATED HEAT AFFECTED ZONE Page 2 Plant: VOGTLE 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input CVN

- 80. 00

- 60. 00

- 60. 00

- 60. 00

40. 00
40. 00
40. 00

-20. 00

-20. 00

-20. 00

.00 00

.00

40. 00
40. 00
40. 00
80. 00
80. 00
80. 00 120. 00 120. 00 120. 00 220. 00 220. 00 220. 00
27. 00
45. 00
60. 00
42. 00
29. 00 58.00 I1 9. 0 0 112. 00
60. 00
79. 00 76.00 94.00 108. 00 125. 00
95. 00 103. 00 147. 00 113. 00
93. 00 136.00 125. 00 136. 00 140. 00 126. 00 140. 00 Computed CVN
33. 88
46. 73 46.73
46. 73 61.76 61.76 61.76 77.57
77. 57 77.57 92.48 92.48 92.48 114. 98 114.98 114. 98 127. 00 127.00 127.00 132. 37 132. 37 132. 37 135. 65 135. 65 135.65 Differential

-6.

88

1. 73 13.27

-4. 73

-32. 76

-3.76

57. 24
34. 43

-17. 57 1.43

- 16.48 1.52 15.52 10.02

19. 98

-11. 98

20. 00
14. 00
34. 00
3. 63
7. 37
3. 63
4. 35
9. 65
4. 35 Correlation Coefficient =.938 C-39

CAPSULE X HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:23 AM Page 1 Coefficients of Curve 2 A = 64.6 B = 62.4 C = 118.02 TO = -2.68 D = O.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Upper Shelf Energy=l 27.0(Fixed)

Lower Shelf Energy=2.2(Fixed)

Temp@30 ft-lbs=-76.4 Deg F Temp@50 ft-lbs=-30.8 Deg F Plant: VOGTLE 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence:

n/cmA2 300 250 W

200 El 150 C) 0 r-z

> 100 50 50 0

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Deg F Charpy V-Notch Data Temperature

- 150. 00

-I100.

00

-75. 00

-50. 00

-50. 00

-25. 00

25. 00
75. 00 100. 00 Input CVN 3.00 14.00
23. 00
58. 00
41. 00
59. 00 6S. 00 109. 00 112. 00 Computed CVN
11. 70
22. 32
30. 52
40. 84
40. 84 52.94 78.97 100. 62 108.37

-8. 70

- 8. 32

-7.

52

17. 16

.16

6. 06

- 10. 97

8. 38
3. 63 Differential C-40

CAPSULE X HEAT AFFECTED 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 100. 00 150. 00 200. 00 200. 00 225. 00 225. 00 106. 00 102. 00 93.00 122. 00 172. 00 123. 00 Computed CVN 108.37 118.27 123. 10 123. 10 124.42 124. 42 Differential

- 2. 37

- 16. 27

- 30. 10

- 1. 10

47. 58

- 1.42 Correlation Coefficient =.932 C-41

UNIRRADIATED HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:37 AM Page 1 Coefficients of Curve I A = 39.56 B = 39.56 C = 70.03 TO = -41.62 D = O.O0E+OO 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: VOGTLE I Material: SAW Heat: B8805-1 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 200 150 u,

E

'0.

a 100 50 0

-9

-300 0

300 Temperature in Deg F 600 Charpy V-Notch Data Temperature

- 180. 00

- 180. 00

-140. 00

-140. 00

-110. 00

-110.00

-110. 00

-80. 00

- 80. 00 Input LE.

1. 00
1. 00
2. 00
3. 00
9. 00 5.00
9. 00 15.00 18.00 Computed L.E.

Differential

1. 49
1. 49 4.49 4.49
9. 83
9. 83
9. 83 19.81 19.81

-.49

-. 49

-2. 49

-1. 49

-. 83

-4. 83

-. 83

-4.

81

- 1. 81 C-42

I 1 -

UNIRRADIATED HEAT AFFECTED ZONE Page 2 Plant: VOGTLE 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input L.E.

- 80. 00

-60. 00

-60. 00

-60. 00

40. 00

-40. 00

-40. 00

-20. 00

-20.00

-20. 00 00 00 00

40. 00
40. 00 40.00
80. 00
80. 00
80. 00 120.00 120. 00 120. 00 220.00 220. 00 220. 00
19. 00
28. 00
40. 00
29. 00
20. 00
36. 00
66. 00
67. 00 44.00
51. 00
49. 00 61.00
70. 00
74. 00 64.00 64.00
81. 00
74. 00
70. 00
79. 00
77. 00 78.00
84. 00
81. 00
87. 00 Computed L.E.
19. 81
29. 40
29. 40
29. 40
40. 47
40. 47
40. 47 51.39 51.39 51.39
60. 64
60. 64
60. 64
72. 10
72. 10
72. 10 76.73 76.73 76.73 78.34
78. 34
78. 34 79.07
79. 07
79. 07 Differential

-.81

-1.40

10. 60

-. 40

- 20. 47

-4.47 25.53 15.61

-7.39

-. 39

11. 64

.36 9.36

1. 90

- 8. 10

-8. 10

4. 27

-2.73

-6. 73

.66 34 4.93 1.93

7. 93 Correlation Coefficient =.964 C-43

CAPSULE X HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:37 AM Page 1 Coefficients of Curve 2 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.=.O(Fixed)

Temp. @L.E. 35 mils=-6.4 Deg F Plant: VOGTLE I Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence:

n/cmA2 200 150 C0 100 r-0 50 0

0 '

-- a' ------

o

-II

-300 0

300 600 Temperature in Deg F Charpy V-Notch Data Temperature Input L.E.

- 150. 00

00. 00

- 7 5. 00

- 5 0. 00

-50. 00

- 25. 00

25. 00
75. 00 1 00. 0 0

. 00

1. 00
7. 00
28. 00 1S. 00
30. 00
43. 00
58. 00
69. 00 Computed L.E.
2. 38
7. 13 11.83
18. 69
18. 69
27. 60
47. 03
60. 15
63. 69 Differential

-2.

38

- 6. 13

- 4. 8 3 9.31

-.69 2.40

-4. 03

- 2. 15 5.31 C-44

I, CAPSULE X HEAT AFFECTED ZONE Page 2 Plant: VOGTLE 1 Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence:

n/cmA2 Charpy V-Notch Data Temperature I 00. 00 150.00 200. 00 200. 00 225. 00 225. 00 Input L.E.

60.00

70. 00 62.00
72. 00 7 1. 00
67. 00 Computed L.E.
63. 69 67.11 68.25 68.25 68.48 68.48-Differential

- 3.69

2. 89

-6. 25 3.75 2.52

- 1. 48 Correlation Coefficient =.986 C45

UNIRRADIATED HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:30 AM Page 1 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: VOGTLE I Material: SAW Heat: B8805-1 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 125 100 0

U) 1.O r-0.a)

QL 75 50

.1 000 0

0 0

0 00 0

01 0

-I 25 0

-300

-200

-100 0

100 200 300 Temperature in Deg F 400 500 600 Charpy V-Notch Data Temperature

- I so. 00 so. 00

-140. 00

-140. 00

- 11. 00

-110. 00

-110. 00

- 80. 00

- 80. 00 Input Percent Shear

.00

.00

.00

.00

5. 00
5. 00
5. 00
10. 00
25. 00 Computed Percent Shear 1.57 1.57 4.41 4.41 9.29 9.29 9.29 18.52
18. 52

-1.57

- 1.57

-4. 41

-4. 41

-4. 29

-4. 29

-4. 29

-8. 52 6.48 Differential

.i C:46

  • 1 C.46

UNIRRADIATED HEAT AFFECTED ZONE Page 2 Plant: VOGTLE I Material: SAW Heat: B8805-1 Orientation: NA Capsule: UNIRR Fluence:

n/cmA2 Charpy V-Notch Data Temperature Input Percent Shear

- 80.00

- 60. 00

- 60. 00

- 60.00

-40.00

- 40. 00

-40. 00

- 20. 00

-20.00

-20.00

.00

.00

.00

40. 00 40.00 40.00 80.00 80.00 80.00 120.00 120. 00 120.00 220.00 220.00 220. 00 18.00
20. 00 34.00
20. 00
25. 00 30.00
80. 00
80. 00 35.00
65. 00
65. 00
60. 00
56. 00 100. 00
65. 00
75. 00 I 1 0. 00
90. 00
90. 00 100. 00 100. 00 100. 00 100. 00 I 1 0. 00 100. 00 Computed Percent Shear 18.52
27. 88
27. 88
27. 88
39. 68
39. 68
39. 68
52. 82
52. 82
52. 82 65.58 65.58 65.58
84. 65
84. 65
84. 65
94. 11
94. 11
94. 11
97. 88
97. 88
97. 88
99. 85
99. 85
99. 85 Differential

-. 52

- 7. 88

6. 12

-7.

88

- 14. 68

-9.

68

40. 32
27. 18

- 17. 82

12. 18

-. 58

-5.58

9. 58
15. 35
19. 65
9. 65
5. 89

-4. 11

-4.

11

2. 12
2. 12
2. 12

.15

.15

.15 Correlation Coefficient =.954 C-47

CAPSULE X HEAT AFFECTED ZONE CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:30 AM Page I Coefficients of Curve 2 A = 50. B = 50. C = 82.03 TO = -7.57 D = 0.OOE+00 Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -7.5 Plant: VOGTLE I Material: SAW Heat: B8805-1 Orientation: NA Capsule: X Fluence:

nlcmA2 lI 125 100 a)

CO, s

a) i-0) 75 50 25 0 *-l

-300

-200

-100 0

100 200 300 Temperature in Deg F 400 500 600 Charpy V-Notch Data Temperature Input Percent Shear

- 150. 00

- 100. 00

-75. 00

-50. 00

-50. 00

- 25. 00 25.00

75. 00 I 00. 00
2. 00
5. 00
15. 00
30. 00
25. 00
40. 00
70. 00
90. 00
90. 00 Computed Percent Shear 3.01 9.50
16. 19
26. 22
26. 22 39.53
68. 87 88.22 93.23 Differential

-I. 01

-4.

50

-1.

19 3.78

- 1. 22

.47 1.13 1.78

-3. 23 C-48

' t

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 1 00. 00 150.00 200. 00 200. 00 225. 00 225. 00

90. 00 95.00 1 00. 00 1 00. 00 1 00. 00 1 00. 00 Computed Percent Shear
93. 23 97.90
99. 37
99. 37
99. 66
99. 66 Differential

- 3. 2 3

-2. 90

.63

. 63

.34

.34 Correlation Coefficient =.998 C-49

D-O APPENDIX D VOGTLE UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D

D-l INTRODUCTION:

Regulatory Guide 1.99, Revision 2, 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. Position C.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 Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there has been four surveillance capsules removed from the Vogtle Unit I reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Vogtle Unit I reactor vessel surveillance data and determine if the Indian Point Unit I surveillance data is credible.

EVALUATION:

Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The Vogtle Unit I reactor vessel consists of the following beltline region materials:

Intermediate Shell Plates B8805-1, 2, and 3 (Heat No.'C0613-1,-2 and C0623-1)

Lower Shell Plates B8606-1, 2, and 3 (HeatNo. C2146-1, -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)

At the time when the Vogtle Unit 1 surveillance program material was selected it was believed that copper and phosphorus were the elements most important to embrittlement of the reactor vessel steels.

The 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, the intermediate shell plate B8805-3 had approximately the same copper and phosphorus content as the other beltline plate materials. Therefore, Appendix D

D-2 based on the highest initial RTNDT and one of the lowest USE, the intermediate shell plate B8606-1 was chosen for the surveillance program.

The weld material in the Vogtle Unit 1 surveillance program was made of weld wire (Heat #83653), the same as all the beltline weld seams, thus it was chosen as the surveillance weld material.

Hence, Criterion 1 is met for the Vogtle Unit I 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 upper shelf energy 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 upper shelf energy 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 280F for welds and 170F 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 upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

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 281F for welds and less than 1 71F for the plate.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998. At this meeting the NRC presented five cases. Of the five cases Case I ("Surveillance data available from plant but no other source") most closely represents the situation listed above for Vogtle Unit 1 surveillance weld metal and plate materials.

Appendix D

D-3 TABLE D-I Calculation of Chemistry Factors using Vogtle Unit I Surveillance Capsule Data Material Capsule Capsule fza)

FF(b)

ARTNDT()

l FF*ARTNDT FF 2 Lower Shell U

0.334 0.698 13.56 9.5 0.487 Plate B8805-3 Y

1.16 1.041 31.94 33.3 1.084 (Longitudinal)

V 1.97 1.185 42.66 50.6 1.404 X

3.53 1.329 96.50 128.2 1.766 Lower Shell U

0.334 0.698 O(d) 0.0 0.487 Plate B8805-3 Y

1.16 1.041 15.19 15.8 1.084 (Transverse)

V 1.97 1.185 33.79 40.0 1.404 X

3.53 1.329 60.80 80.8 1.766 SUM:

358.2 9.482 CFjssos53 = X(FF

( FF2) = (358.2)

(9.482) = 37.80 F Surveillance Neld U

0.334 0.698 24.98 17.4 0.487 Material Y

1.16 1.041 7.70 8.0 1.084 V

1.97 1.185 O(d) 0.0 1.404 X

3.53 1.329 53.40 71.0 1.766 SUM:

96.4 4.741 CF Surv. Weld = X (FF

  • RThijT)
  • 7( FF2) = (96.4) * (4.741) = 20.3OF Notes:

(a) f = fluence. Calculated fluence from Section 6 of this report, [x 1019 n/cm2, E > 1.0 MeV].

(b)

FF = fluence factor = f.u 0 log f)

(c)

ARTNDT values are the measured 30 ft-lb shift values taken from Appendix C, herein [TF].

(d)

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

Appendix D

.1

it D4 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-2:

Vogtle Unit I Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials.

Scatter

<17'F (Base Material Capsue F

FF Measured Predicted Metals)

Maeil Cpue(SIlopebest fit)

ARTNDT ARTNDT ARTNDT Meas (OF)

<280F (Weld)

Intermediate Shell U

37.8 0.698 13.56 26.38

-12.82 Yes Plate B8805-3 Y

37.8 1.041 31.94 39.35

-7.41 Yes (Longitudinal)

V 37.8 1.185 42.66 44.79

-2.13 Yes X

37.8 1.329 96.50 50.24 46.26 No Intermediate Shell U

37.8 0.698 0.0 26.38

-26.38 No Plate B8805-3 Y

37.8 1.041 15.19 39.35

-24.16 No (Transverse)

V 37.8 1.185 33.79 44.79

-11 Yes X

37.8 1.329 60.80 50.24 10.56 Yes U

20.3 0.698 24.98 14.17 10.81 Yes Vessel Beltline Y

20.3 1.041 7.70 21.13

-13.43 Yes Welds (Heat # 83653)

V 20.3 1.185 0.0 24.06

-24.06 Yes X

20.3 1.329 53.40 26.98 26.42 Yes Table D-2 indicates that 3 of 8 data point falls outside the +/- I a of 170F scatter band for the Intermediate Shell Plate B8805-3 surveillance data. Therefore, the surveillance plate data is deemed "not credible".

No data points fall outside the +/- 1 a of 28 0F scatter band for the surveillance weld data. Therefore, the weld data is deemed credible per the third criterion.

Appendix D

D-5 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 +/- 251F.

The capsule specimens are located in the reactor between the core barrel 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 257F. Hence, this criterion is met.

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 I surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to the Vogtle Unit I surveillance program.

CONCLUSION:

Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B and 10 CFR 50.61, the Vogtle Unit I surveillance plate is not credible but the weld data is credible.

Appendix D