Enclosure 1 Vogtle Electric Generating Plant WCAP-16278-NP ... · * Capsule X received an average...

Enclosure 1

Vogtle Electric Generating PlantWCAP-16278-NP, Rev. 0

Westinghouse Non-Proprietary Class 3

WCAP-1 6278-NPRevision 0

July 2004

Analysis of Capsule X from the SouthernNuclear Operating Company, Vogtle Unit 1Reactor Vessel Radiation SurveillanceProgram

*Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3

WCAP-16278-NP, Revision 0

Analysis of Capsule X from the Southern Nuclear OperatingCompany, Vogtle Unit 1 Reactor Vessel Radiation

Surveillance Program

KG KnightR. J. Hagler

J. Conermann

July 2004

Approved:~ 4~4.J. Gh'gurovich, ManagerReactor Component Design & Analysis

Westinghouse Electric Company LLCEnergy SystemsP.O. Box 355

Pittsburgh, PA 15230-0355

02004 Westinghouse Electric Company LLCAll 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 .. 5-15.1 OVERVIEW .5-15.2 CHARPY V-NOTCH IMPACT TEST RESULTS .5-35.3 TENSILE TEST RESULTS .5-55.4 1/2T COMPACT TENSION SPECIMEN TESTS .5-5

6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY . .6-16.1 INTRODUCTION .6-6.2 DISCRETE ORDINATES ANALYSIS .6-26.3 NEUTRON DOSIMETRY .6-56.4 CALCULATIONAL UNCERTAINTIES .6-6

7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE .7-1

8 REFERENCES ......... 8-1

APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ONNEUTRON 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 SYMMETRICHYPERBOLIC TANGENT CURVE-FITTING METHOD .C-0

APPENDIX D VOGTLE UNIT I SURVEILLANCE PROGRAM CREDIBILITYEVALUATION . D-0

iv

LIST OF TABLES

Table 4-1 Chemical Composition (wt %) of the Vogtle Unit I Reactor VesselSurveillance Materials (Unirradiated) ..................................................... 4-3

Table 4-2 Heat Treatment History of the Vogtle Unit 1 Reactor VesselSurveillance Materials .............. 4-4

Table 5-1 Charpy V-Notch Data for the Vogtle Unit I Intermediate Shell Plate B8805-3Irradiated 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 B8805-3Irradiated 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 MaterialIrradiated 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/cm 2 (E> 1.0 MeV) ............................... 5-9

Table 5-5 Instrumented Charpy Impact Test Results for the Vogtle Unit 1 IntermediateShell 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 IntermediateShell 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 SurveillanceWeld 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 NotchToughness Properties of the Vogtle Unit I Reactor Vessel SurveillanceMaterials .... 5-14

Table 5-10 Comparison of the Vogtle Unit 1 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide1.99, 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 SurveillanceMaterials 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 TheSurveillance Capsule Center .................. 6-12

Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and IntegratedExposures at the Reactor Vessel Clad/Base Metal Interface ....................................... 6-16

Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within TheReactor Vessel Wall ............ 6-20

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

Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn fromAlvin 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 ReactorVessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) . 5-17

Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1Reactor Vessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) . 5-18

Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I ReactorVessel Intermediate Shell Plate B8805-3 (Longitudinal Orientation) . 5-19

Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit I ReactorVessel Intermediate Shell Plate B8805-3 (Transverse Orientation) . 5-20

Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1Reactor Vessel Intermediate Shell Plate B8805-3 (Transverse Orientation) . 5-21

Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I ReactorVessel Intermediate Shell Plate B8805-3 (Transverse Orientation) . 5-22

Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 ReactorVessel Weld Metal . 5-23

Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit IReactor Vessel Weld Metal . 5-24

Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I ReactorVessel Weld Metal . 5-25

Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Vogtle Unit 1 ReactorVessel Heat-Affected-Zone Material . 5-26

Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Vogtle Unit 1Reactor Vessel Heat-Affected-Zone Material . 5-27

Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Vogtle Unit I ReactorVessel Heat-Affected-Zone Material . 5-28

Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit I Reactor VesselIntermediate 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 ReactorVessel Intermediate Shell Plate B8805-3 (Transverse Orientation) .............................. 5-30

Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit I ReactorVessel Weld Metal .......... 5-31

Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Vogtle Unit I ReactorVessel Heat-Affected-Zone Metal ..................... 5-32

Figure 5-17 Tensile Properties for Vogtle Unit 1 Reactor Vessel Lower ShellPlate B8805-3 (Longitudinal Orientation) ........................... 5-33

Figure 5-18 Tensile Properties for Vogtle Unit 1 Reactor Vessel Lower ShellPlate 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 VesselIntermediate Shell Plate B8805-3 (Longitudinal Orientation) ...................................... 5-36

Figure 5-21 Fractured Tensile Specimens from Vogtle Unit 1 Reactor VesselIntermediate 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 PlateB8805-3, 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 PlateB8805-3, 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 VogtleUnit 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 neutrontransport and dosimetry cross-section libraries was derived from the ENDF/B-VI data-base. Capsule Xreceived a fluence of 3.53 x IO"1 n/cm2 (E > 1.0 MeV) after irradiation to 14.33 EFPY. The peakclad/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 shelfenergy for all the surveillance materials of Capsules X contained in the Vogtle Unit I surveillanceprogram are less than the Regulatory Guide 1.99, Revision 2 predictions. 2) All beltline materials exhibita more than adequate upper shelf energy level for continued safe plant operation and are predicted tomaintain an upper shelf energy greater than 50 ft-lb throughout the current license (36 EFPY) and apotential license renewal date of 54 EFPY as required by 1 OCFR50, Appendix G [2]. 3) The Vogtle UnitI 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-notchdata 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 capsuleremoved 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 fromWCAP-1101113 ] and STC Report (Capsule X)171. The Charpy plots were developed usingCVGRAPH Version 5.0.2, which is a symmetric hyperbolic tangent curve-fitting program. Theirradiated capsule data from WCAP-1225614 3 (Capsule U) and WCAP-13931, Rev. 11[5 (CapsuleY) are documented in Appendix C of WCAP-1506716 1 (Capsule V). The results presented inSection 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 after14.33 effective full power years (EFPY) of plant operation.

* Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, orientedwith the longitudinal axis of the specimen parallel to the major working direction (longitudinalorientation), resulted in an irradiated 30 ft-lb transition temperature of 81.61F and an irradiated50 ft-lb transition temperature of 121.7 0F. This results in a 30 ft-lb transition temperatureincrease of 96.50F and a 50 ft-lb transition temperature increase of 99.81F for the longitudinaloriented specimens.

* Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, orientedwith 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 anirradiated 50 ft-lb transition temperature of 135.27F. This results in a 30 ft-lb transitiontemperature increase of 60.81F and a 50 ft-lb transition temperature increase of 72.71F for thetransverse oriented specimens.

* Irradiation of the weld metal (heat number 83653) Charpy specimens resulted in an irradiated30 ft-lb transition temperature of-3.8°F and an irradiated 50 ft-lb transition temperature of12.8°F. This results in a 30 ft-lb transition temperature increase of 53.4°F and a 50 ft-lbtransition temperature increase of 43.1 'F.

* Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in anirradiated 30 ft-lb transition temperature of-76.4°F and an irradiated 50 ft-lb transitiontemperature of -30.8°F. This results in a 30 ft-lb transition temperature increase of 10.6°F and a50 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 irradiatedaverage 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 irradiatedaverage 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 averageenergy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper shelfenergy of 141 ft-lb for the weld metal specimens.

* The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an averageenergy decrease of 9 ft-lb after irradiation. This results in an irradiated average upper shelfenergy 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 Ireactor vessel surveillance materials is presented in Table 5-10.

* All beltline materials exhibit a more than adequate upper shelf energy level for continued safeplant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lbthroughout 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 plateis 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 coremid-plane for the Vogtle Unit I reactor vessel using the Regulatory Guide 1.99, Revision 2attenuation 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 thereactor in the continuing surveillance program, which monitors the effects of neutron irradiation on theSouthern Nuclear Operating Company, Vogtle Unit I reactor pressure vessel materials under actualoperating conditions.

The surveillance program for the Southern Nuclear Operating Company Vogtle Unit 1 reactor pressurevessel materials was designed and recommended by the Westinghouse Electric Corporation. Adescription of the surveillance program and the pre-irradiation mechanical properties of the reactor vesselmaterials are presented in WCAP-1I01 1, "Georgia Power Company Alvin NV. Vogtle Unit No. I ReactorVessel Radiation Surveillance Program" 131. The surveillance program was planned to cover the 40-yeardesign life of the reactor pressure vessel and was based on ASTM El 85-82, "Standard RecommendedPractice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Reactor Vessels."l2 01 CapsuleX was removed from the reactor after 14.33 EFPY of exposure and shipped to the Westinghouse Scienceand Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of theCharpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing of and the post-irradiation data obtained from surveillance Capsule Xremoved from the Southern Nuclear Operating Company Vogtle Unit I reactor vessel and discusses theanalysis 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 resistfracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region ofthe reactor pressure vessel is the most critical region of the vessel because it is subjected to significantfast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties oflow alloy, ferritic pressure vessel steels such as SA533 Grade B Class I (base material of the Vogtle UnitI reactor pressure vessel beltline) are well documented in the literature. Generally, low alloy ferriticmaterials show an increase in hardness and tensile properties and a decrease in ductility and toughnessduring high-energy irradiation.

A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture ToughnessCriteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and PressureVessel Code 1[0]. The method uses fracture mechanics concepts and is based on the reference nil-ductilitytransition temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT perASTM E-2081 9]) or the temperature 601F less than the 50 ft-lb (and 35-mil lateral expansion) temperatureas determined from Charpy specimens oriented perpendicular (transverse) to the major working directionof the plate. The RTNDT of a given material is used to index that material to a reference stress intensityfactor curve (Kk, curve) which appears in Appendix G to the ASME Coderl']. The Ki, curve is a lowerbound of static fracture toughness results obtained from several heats of pressure vessel steel. When agiven material is indexed to the Kic curve, allowable stress intensity factors can be obtained for thismaterial as a function of temperature. Allowable operating limits can then be determined using theseallowable stress intensity factors.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effectsof radiation on the reactor vessel material properties. The changes in mechanical properties of a givenreactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillanceprogram, such as the Vogtle Unit I reactor vessel radiation surveillance programs1, in which asurveillance capsule is periodically removed from the operating nuclear reactor and the encapsulatedspecimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due toirradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust theRTNDT (ART) for radiation embrittlement. This ART (RTNDT initial + M + ARTNDT) is used to index thematerial to the KI, curve and, in turn, to set operating limits for the nuclear power plant that take intoaccount 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 reactorpressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plantstart-up. The six capsules were positioned in the reactor vessel between the neutron pads and the vesselwall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of thecore. The capsules contain specimens made from Intermediate Shell Plate B8805-3, and weld metalfabricated with 3/16-inch Mil B4 weld filler wire, Heat Number 83653 Linde Type 0091 flux, LotNumber 3536, which is identical to that used in the actual fabrication of the intermediate to lower shellgirth weld and all longitudinal weld seams of both the intermediate and lower shell plates of the pressurevessel.

Capsule X was removed after 14.33 effective full power years (EFPY) of plant operation. This capsulecontained Charpy V-notch, tensile, and 1/2T-CT fracture mechanics specimens made from IntermediateShell Plate B8805-3 and submerged arc weld metal representative of the intermediate shell longitudinalweld 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 formingof the plate) was taken at least one plate thickness from the quenched edges of the plate. All testspecimens were machined from the 1/4 and 3/4 thickness locations of the plate after performing a simulatedpost-weld stress-relieved treatment on the test material. Test specimens were also removed from weldand heat-affected-zone metal of a stress-relieved weldment joining intermediate shell plate B8805-1 andadjacent 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 thelongitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction) and alsoin the transverse orientation (longitudinal axis of the specimen perpendicular to the major rollingdirection). The core region-weld Charpy impact specimens were machined from the weldment such thatthe long dimension of each Charpy specimen was perpendicular to the weld direction. The notch of theweld metal Charpy specimens was machined such that the direction of crack propagation in the specimenwas in the welding direction.

Tensile specimens from Intermediate Shell Plate B8805-3 were machined in both the longitudinal andtransverse orientations. Tensile specimens from the weld metal were oriented with the long dimension ofthe specimen perpendicular to the weld direction.

Compact tension test specimens from Intermediate Shell Plate B8805-3 were machined in thelongitudinal and transverse orientations. Compact tension test specimens from the weld metal weremachined perpendicular to the weld direction with the notch oriented in the direction of welding. Allspecimens 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 inTables 4-1 and 4-2, respectively. The data in Table 4-1 and 4-2 was obtained from the unirradiatedsurveillance program report, WCAP-110 113,Appendix A.

Capsule X contained dosimeter wires of pure iron, copper, nickel, and aluminum-0.15 weight percentcobalt (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 neutronenergy levels.

The capsule contained thermal monitors made from two low-melting-point eutectic alloys and sealed inPyrex tubes. These thermal monitors were used to define the maximum temperature attained by the testspecimens during irradiation. The composition of the two eutectic alloys and their melting points are asfollows:

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 inCapsule X is shown in Figure 4-2.


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 WestinghouseEngineering 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.


4-4

Table 4-2

Heat Treatment History of the Vogtle Unit 1 Reactor Vessel Surveillance Materials(')

|Material Temperature (0f) Time Coolant

Intermediate Shell Plate Austenitized @ 4 hrs. Water-Quench1600 ± 25

B8805-3 Tempered @ 4 hrs. Air-cooled1225 * 25

Stress Relieved @ 17.5 hrs. Furnace Cooledl_ 1150 ± 50

Weld Metal (heat # 83653) Post Weld Stress Relieved 12.75 hrs. Furnace Cooledl @ 1150*50

Notes:(a) This table was taken from WCAP-I 101 1I'.


4-5

0O

CAPSULE U (58.5 )

- V (a1)

- WX.

- W (121.5")

REACTORVESSEL

IOos

PL ANVEW

VESSELWALL

CAPSULE

.CORE

NELTRMON PAD

OORE BAIM

ELEVATKON VIEW

Figure 4-1 Arrangement of Surveillance Capsules in the Vogtle Unit 1 Reactor Vessel


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-.15zcoI i Al a

n. . -. ISICO (

~I IIII Iem579.f

olt TOR

Spacer Tensile Compact Compact Charp Charpy Charpy CompactBlock IZ A 12 7 AW60 Al 160 AW57 AH57 FAs [ T]

[WI I AW16 AW15 AW14 AW59 A1159 AW56 Arl56 I [AW3 IA5 AL16 ALISW 1 | AW58 _Al158 AW 5 25 _Al155 AW52 _ AH52

TOP OF VESSEL " CENTER

Np237

Compact Charpy Charpy Charpy Charpy

AW51 I AH51 AW48 | A148AL14 AL13 AW50 I AW47 N A1447

. j AW49 1 A1149 I AW46 A1146CENTER - - \

|ATr60 | AL60 ||AT57 I I AL57 I|ATS9 ||AL59 | AT56 I AL56

| T5 |[AL58 | 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

|AT54 AL54AT53 | l.53

- A~T52 ||Al,52|

|AT51 ||AL51 |1'\ |AL48 | | ||AT 12||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


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 specimenswas performed in the Remote Metallographic Facility (RMF) at the Westinghouse Research andTechnology Park. Testing was performed in accordance with I OCFR5O, Appendices G and Hid, ASTMSpecification El 85-821I", and Westinghouse Procedure RMF 84021't2 Revision 2 as modified byWestinghouse 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]. Thespecimens and spacer blocks were carefully removed, inspected for identification number, and checkedagainst 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 nomelting of either type of thermal monitor. Based on this examination, the maximum temperature towhich 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 8103on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumentedwith an Instron Dynatup Impulse instrumentation system, feeding information into an IBM compatiblecomputer. With this system, load-time and energy-time signals can be recorded in addition to thestandard 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 wasobserved. The load at which fast fracture was initiated is identified as the fast fracture load (PF). If thefast 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 acrack in the specimen. Therefore, the propagation energy for the crack (Ep) is the difference between thetotal 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 thespecimen measured parallel to the notch; W = height of the specimen, measured perpendicularly to thenotch; 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 Charpyspecimen 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 ofthe 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 ofthe 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 incompliance with ASTM E23-02a 15] and A370-97a '6]. The lateral expansion was measured using a dialgage 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) perASTM 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 determineddirectly from the load-extension curve. The yield strength, ultimate strength and fracture strength werecalculated using the original cross-sectional area. The final diameter and final gage length weredetermined from post-fracture photographs. The fracture area used to calculate the fracture stress (truestress at fracture) and percent reduction in area were computed using the final diameter measurement.


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 CapsuleX, which received a fluence of 3.53 x 10'9 n/cm2(E> 1.0 MeV) in 14.33 EFPY of operation, are presentedin Tables 5-1 through 5-8 and are compared with unirradiated results 41 as shown in Figures 5-1 through5-12.

The transition temperature increases and upper shelf energy decreases for the Capsule X materials aresummarized 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-fittingroutine. 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 after14.33 effective full power years (EFPY) of plant operation.

* Irradiation of the reactor vessel intermediate shell plate B8805-3 Charpy specimens, orientedwith the longitudinal axis of the specimen parallel to the major working direction (longitudinalorientation), resulted in an irradiated 30 ft-lb transition temperature of 81.61F and an irradiated50 ft-lb transition temperature of 121.7 0F. This results in a 30 ft-lb transition temperatureincrease of 96.50F and a 50 ft-lb transition temperature increase of 99.80 F for the longitudinaloriented specimens.

* Irradiation of the reactor vessel Intermediate Shell Plate B8805-3 Charpy specimens, orientedwith 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 anirradiated 50 ft-lb transition temperature of 135.20F. This results in a 30 ft-lb transitiontemperature increase of 60.81F and a 50 ft-lb transition temperature increase of 72.71F for thetransverse oriented specimens.

* Irradiation of the weld metal (heat number 83653) Charpy specimens resulted in an irradiated30 ft-lb transition temperature of-3.80F and an irradiated 50 ft-lb transition temperature of12.8°F. This results in a 30 ft-lb transition temperature increase of 53.4°F and a 50 ft-lbtransition temperature increase of 43.1 'F.

* Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in anirradiated 30 ft-lb transition temperature of-76.4°F and an irradiated 50 ft-lb transitiontemperature of -30.8°F. This results in a 30 ft-lb transition temperature increase of 10.6°F and a50 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 irradiatedaverage upper shelf energy of 109 ft-lb for the longitudinal oriented specimens.


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 irradiatedaverage 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 averageenergy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper shelfenergy of 141 ft-lb for the weld metal specimens.

* The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an averageenergy decrease of 9 ft-lb after irradiation. This results in an irradiated average upper shelfenergy 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 Ireactor vessel surveillance materials is presented in Table 5-10.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plantoperation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end ofthe 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 inFigures 13 through 16. The fractures show an increasingly ductile or tougher appearance with increasingtest temperature. Load-time records for the individual instrumented Charpy specimens are contained inAppendix A.


5-5

5.3 TENSILE TEST RESULTS

The results of the tensile tests performed on the various materials contained in Capsule X irradiated to3.53 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 (longitudinalorientation) indicated that irradiation to 3.53 x 1019 n/cm2 (E> 1.0 MeV) caused approximately a I to 10ksi increase in the 0.2 percent offset yield strength and approximately a 3 to 10 ksi increase in theultimate 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 inthe 0.2 percent offset yield strength and approximately a 2 to 5 ksi increase in the ultimate tensilestrength 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 to3.53 x 1 lO'9 n/cm2 (E> 1.0 MeV) caused approximately a 2 to 4 ksi increase in the 0.2 percent offset yieldstrength and approximately a 2 to 6 ksi increase in the ultimate tensile strength when compared tounirradiated 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 andare being stored at the Westinghouse Research and Technology Park Hot Cell facility.


11-

5-6

Table 5-1

Charpy V-notch Data for the Vogtle Unit 1 Intermediate Shell Plate B8805-3 Irradiated to aFluence 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


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 -1-- II- - n tx--%n_ +^ ̂

rAueUMA V1 0.0 AA 1. LU WIL1 kJs 1.U .V IV) V a1I4fVers VIUiiaLIUU)


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 | 89 50 1.27 95

AT57 250 121 95 129 66 1.68 100

ATS9 275 135 98 133 | 66 1.68 100


5-8

Table 5-3

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


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


5-9

Table 5-4

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


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


5-10

Table 5-5

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

Charpy Nornalized Energies TimeTest Energy (ft-lb/in2 ) Yield to Time Fast YiCId

TeSt Load Yield Max. to Max. Fract. Arrest Stress FlowSample 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


5-11

Table 5-6

Instrumented Charpy Impact Test Results for the Vogtle Unit 1 Intermediate Shell Plate B8805-3Irradiated to a Fluence of 3.53 x 1019 n/cm2 (E>1.0 MeV) (Transverse Orientation)

Charpy Normalized Energies TimeCharpy )Yild toTime Fast Yield

Energy (ft-lb/in2) Yield toTest 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 MetalIrradiated to a Fluence of 3.53 x 10i9 n/cm2 (E>1.0 MeV)

Normalized Energies TimeTest Energy (ft-lb/in2) Yield to Time Fast

Test Load Yield Max. to Max. Fract. Arrest Yield FlowSample 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


5-13

Table 5-8

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

Normalized Energies TimeCEnery (ft-lb/in2) Yield to Time Fast Arrest

Test Egy Load Yield Max. to Max. Fract. Load Yield FlowSample 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


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 SurveillanceMaterials

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 -13.0Shell PlateB8805-3 (Long.)

Intermediate 17.1 77.9 60.8 55.0 146.6 91.6 62.5 135.2 72.7 96.0 93.0 -3.0Shell Plate1B8805-3(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 andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions

30 ft-lb Transition Upper Shelf EnergyTemperature 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.


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 ReductionNumber 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 65B 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 62B 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


5-17

INTERMEDIATE SHELL PLATE B8805-3 (LONGITUDINAL ORIENTATION)

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

Curve12

300 -

250 -

thI 200 -

00

U-

E150 -a,

z>100-

50 -

-300

PlantVOGTLE IVOGTLE I

CapsuleUNIRR

x

MaterialSA533B ISA533B I

Ori. Heat #LT C0623-1LT C0623-1

-200 -100 0 100 200 300 400 500 600

Temperature in Deg F0 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 VesselIntermediate Shell Plate B8805-3 (Longitudinal Orientation)


_1

5-18


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

Curve

2

PlantVOGTLE IVOGTLE 1

CapsuleUNIRR

X

MaterialSA533B ISA533BI

Ori.LTLT

Heat #C0623-1C0623-1

200

150A

E

.o

2 100

5.50

0-300 0 300

Temperature in Deg Fo Set 2

600

o Set I

Results

Curve

2


.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 VesselIntermediate Shell Plate B8805-3 (Longitudinal Orientation)


5-19


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

Curve12

PlantVQGTLE IVOGTLE I

CapsuleUNIRR

x

MaterialSA533BISA533B1

Ori.LTLT

Heat #C0623-1C0623-1

Da)

'EU,

aIL

125-

100 -

75-

50-

25

-300 -200 -1 C

0 Set I

0O 0 100 200 300

Temperature in Deg F400 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 VesselIntermediate Shell Plate B8805-3 (Longitudinal Orientation)


'I

5-20

INTERIEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)


Curve

2


CapsuleUNIRR

x

NllaterialSA533B ISA533B I

Ori.TLTL

Hleat #C0623-1C0623-1

300

250

o

8, 20000

LL

Im 150C,

z> 100C.

50 -_

00-300 -200 -100 0 100 200 300 400 500 600

Temperature in Deg Fo Set 2o Set1

Results

Fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50Curve

2

2.2 96.0

2.2 93.0

.0

- 3. 0

17. 1

77. 9

.0

60. 8

62. 5135. 2

.0

72.7

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


5-21

INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)


Curve12


CapsuleUNIRR

X

MaterialSA533B ISA533B I

Ori.TLTL

Heat #C0623-1C0623-1

200

150

u,=

E

100C

050

0 -

-300 0 300Temperature 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 VesselIntermediate Shell Plate B8805-3 (Transverse Orientation)


_- .

5-22

INTERMEDIATE SHELL PLATE B8805-3 (TRANSVERSE ORIENTATION)

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

CurveI2

PlantVOGTLE IVOGTLE 1

CapsuleUNIRR

x

MaterialSA533B 1SA533BI

Ori.TLTL

Ifeat #C0623-1C0623-1

Ma)

ci

UnI-

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 VesselIntermediate Shell Plate B8805-3 (Transverse Orientation)


5-23

SURVEIALANCE PROGRAM WNIELD METAL

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:56 AMData Set(s) Plotted

Curve12


Capsule.UNIRRx

MaterialSAWSAW

Ori.NANA

Heat #*IRE:83653NNIRE:83653

300

250

:, 200-00U-

E 150c)C1

zZ; 1 00

50-_

0x-300 -200 -100 0 100 200 300 400 500 600

Temperature in Deg Fc 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


.... L____i1.

5-24

SURVEILLANCE PROGRAM WELD METAL


CurveI2


CapsuleUNIRRx

MaterialSAWSAW

Ori.NANA

Heat #WIRE:83653WIRE:83653

200

150on

0U,

a 100

I-I

2

50

o --300 0 300

Temperature in Deg F0 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 VesselWeld Metal


5-25

SURVEILLANCE PROGRAM WELD METAL

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03123/2004 11:01 AMData Set(s) Plotted

Plant Capsule Material Ori. Heat #VOGTLE I UNIRR SAW NA WIRE:83653VOGTLE I X SAW NA WIRE:83653

Curve12

125

100

C,

a)*0

CL

75

50

25

o -

-300 -200 -100 0 100 200 300 400 500Temperature 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 VesselWeld Metal


'I

5-26

SURVEILLANCE PROGRAM HEAT AFFECTED ZONE


Plant Capsule Material Ori. Heat #VOGTLE 1 UNIRR SAW NA B8805-1VOGTLE 1 X SAW NA B8805-1

Curve12

300 1250 --

-0 200 -00o I

m 150CD

ujz

>100

50 -_ -

-300 -200 -100

0 Set 1

0 100 200 300 400 500 600

Temperature in Deg Fo 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 VesselHeat-Affected-Zone Material


5-27

SURVEILLANCE HEAT AFFECTED ZONE

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03123/2004 10:39 AMData Set(s) Plotted

Plant Capsule Material Ori. Heat #VOGTLE 1 UNIRR SAW NA B8805-lVOGTLE 1 X SAW NA B8805-1

Curve12

200

150

r-tn0U,

a 100

a0

50

o 4-

-300 0 300

Temperature in Deq F600

0 Seti 0 Set2

Curve

2


.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 VesselHeat-Affected-Zone Material


____ ___________11

5-28

SURVEILLANCE PROGRAM HEAT AFFECTED ZONE

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:31 ANAData Set(s) Plotted

Curvel2


CapsuleUNIRR

x

MaterialSAWSAW

Ori.NANA

Heat #B8805-1B8805-1

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 @50Curve

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 VesselHeat-Affected-Zone Material


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 VesselIntermediate Shell Plate 1B8805-3 (Longitudinal Orientation)


5-30

AT49. -250F AT55 750 F

AT56.2000 F

AF53,200OF AT46,225TF AT47,2250 F AT57,2500 F AT59,275TF

Figure 5-14 Clharpy Impact Specimen Fracture Surfaces for Vogtle Unit 1 Reactor VesselIntermediate Shell Plate B8805.3 (Transverse Orientation)


5-31

AWN1 °OT AWA7 I lnOl -1-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 VeldMetal

5°F


5-32

AH58. -5U0F AH59. -500 F

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


5-33

120- ULTIMATEYIELD STRENGTH

100 -

-80 -

en 60 - 0.2% YIELD STRENGTH H

20-

0 100 200 300 400 500 600

TEMPERATURE( F)

Legend: A and o are UnirradiatedA 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 AREAA 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)


Il

5-34

120-

100 -ULTIMATE MELD STRENGTH

.;;80

'o 60Mt 0.2% YIELD STRENGTHco 40 -

20 -

0-4- II

0 100 200 300 400 500 600

TEMPERATURE( F)

Legend: A and o are UnirradiatedAand * are Irradiated to 3.53 x IlO' n/cm 2 (E > 1.0 MeV)

~~~~~~-- ----- --

70REDUCTION IN AREA

60 - 4

-; 50-

> 40- 30 TOTAL ELONGATION

Q0a 20 .-o

10UNIFORM 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)


5-35

100ULTIMATE YIELD STRENGTH

80 -

^. 60 - 0.2% YIELD STRENGTHC,,ww 40-

20 -

0- .I ,

0 100 200 300 400 500 600

TEMPERATURE( F)

Legend: A and o are UnirradiatedA 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 A10 -A

0 100 200 300 400 500 600

TEMPERATURE ( F)

Figure 5-19 Tensile Properties for Vogtle Unit 1 Reactor Vessel Weld Metal


5-36

Specimen AL10 Tested at 750F

Specimen ALl1 Tested at 300OF

Specimen AL12 Tested at 5500 F

Figure 5-20 Fractured Tensile Specimens from Vogtle Unit 1 Reactor Vessel Intermediate SliclPlate B8805-3 (Longitudinal Orientation)


5-37

Specimen ATIO Tested at 750 F

-.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 ShellPlate B8805-3 (Transverse Orientation)


5-38

Specimen AWIO Tested at 750 F

Specimen AWl 1 Tested at 300OF

Specimen AW12 Tested at 5500 F

Figure 5-22 Fractured Tensile Specimens from Vogtle Unit I Reactor Vessel Weld Metal


5-39

VOGTLE UNIT #1'X CAPSULE

100 I

Uso

a, 60 -U)

U)40 -

20

AL-1075F

00 0.05 0.1 0.15

STRAIN. INIIN

0.2 0.25 0.3

VOGTLE UNIT #1X CAPSULE

100

So

so

70

X 60

U)soUOw

U) 40

30

20

10

0

AL-11300 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)


540

VOGTLE UNIT # 1'X' CAPSULE

100

90

80

70

X 60

n 50

to 40

30

20

10

AL-12550 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 PlateB8805-3, Capsule X, Tensile Specimen AL-12 (Longitudinal Orientation)


541

VOGTLE UNIT #1X CAPSULE

100

90

so1

70

s 60

U, 50*M'U

U, 40

30

20

10

0-

_ _

AT-1075 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 # IAX CAPSULE

AT-11300 F

C 0.05 0.1 0.15STRAIN. 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)


5-42

VOGTLE UNIT # 1X CAPSULE

100

90

80

70

W 60to0 sU) 50LU

(D40

30

20

10

AT-12550 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 PlateB8805-3, Capsule X, Tensile Specimen AT-12 (Transverse Orientation)


5-43

VOGTLE UNIT #1X' CAPSULE

100*goo90-

80-

70

X 60

"' 50

co 40

30

20

10

AW-1075 F

00 0.05 0.1 0.15

STRAIN. ININ

02 0.25 0.3

100

90

eo70 -

s 60

Cl,

,,, 40

30

20

10

0

VOGTLE UNIT #1X' CAPSULE

AW-il300 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 SpecimensAW-10, and AWV-l1


_________1

5-44

VOGTLE UNIT #1"X'CAPSULE

100-

90

80

70

1 60'

U) 50

co 40-

30-

20-

10-

0-

AW-12550 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 SpecimenANV- 12


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 Ireactor to determine the neutron radiation environment within the reactor pressure vessel andsurveillance 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 specificbasis. An evaluation of the most recent dosimetry sensor set from Capsule X, withdrawn at the end of theeleventh plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable tothe 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 withthe analytical predictions served to validate the plant specific neutron transport calculations. Thesevalidated calculations subsequently formed the basis for providing projections of the neutron exposure ofthe 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 theneutron exposure of the material has traditionally been accepted for the development of damage trendcurves as well as for the implementation of trend curve data to assess the condition of the vessel. Inrecent years, however, it has been suggested that an exposure model that accounts for differences inneutron energy spectra between surveillance capsule locations and positions within the vessel wall couldlead to an improvement in the uncertainties associated with damage trend curves and improved accuracyin the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damagefunction for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light-WVaterReactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along withfluence (E > 1.0 MeV) to provide a database for future reference. The energy dependent dpa function tobe used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing NeutronExposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." The application of thedpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vesselwall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement ofReactor Vessel Materials."

All of the calculations and dosimetry evaluations described in this section and in Appendix A were basedon the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latestavailable calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologiesfollow the guidance and meet the requirements of Regulatory Guide .1.190, "Calculational and DosimetryMethods for Determining Pressure Vessel Neutron Fluence."122 Additionally, the methods used todevelop the calculated pressure vessel fluence are consistent with the NRC approved methodologydescribed in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating SystemSetpoints 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 thereactor 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 followingoctant symmetric locations represented in Figure 6-1: 29° from the core cardinal axes (for the 610 and2410 dual surveillance capsule holder locations) and 31.5° from the core cardinal axes (for the 121.50 and301.5° single surveillance capsule holder locations, and for the 58.5° and the 238.50 dual surveillancecapsule holder locations). The stainless steel specimen containers are 1.182-inch by 1-inch and areapproximately 56 inches in height. The containers are positioned axially such that the test specimens arecentered 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 andthe neutron energy spectrum in the water annulus between the neutron pads and the reactor vessel. Inorder to determine the neutron environment at the test specimen location, the capsules themselves mustbe included in the analytical model.

In performing the fast neutron exposure evaluations for the Alvin W. Vogtle Unit I reactor vessel andsurveillance capsules, a series of fuel cycle specific forward transport calculations were carried out usingthe 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 transportsolution in r,0 geometry, 4(rz) is the two-dimensional solution for a cylindrical reactor model using theactual axial core power distribution, and +(r) is the one-dimensional solution for a cylindrical reactormodel using the same source per unit height as that used in the r,0 two-dimensional calculation. Thissynthesis 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 utilizedsince, with the exception of the neutron pads, the reactor is octant symmetric. These r,0 models includethe core, the reactor internals, the neutron pads - including explicit representations of octants notcontaining surveillance capsules and octants with surveillance capsules at 290 and 31.5°, the pressurevessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biologicalshield wall. These models formed the basis for the calculated results and enabled making comparisons tothe surveillance capsule dosimetry evaluations. In developing these analytical models, nominal designdimensions were employed for the various structural components. Likewise, water temperatures, andhence, coolant densities in the reactor core and downcomer regions of the reactor were taken to berepresentative of full power operating conditions. The coolant densities were treated on a fuel cyclespecific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, andmiscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric meshdescription of the r,0 reactor models consisted of 183 radial by 99 azimuthal intervals. Mesh sizes were


6-3

chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis. Thepoint-wise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of0.001.

The rz model used for the Alvin W. Vogtle Unit 1 calculations is shown in Figure 6-2 and extendsradially from the centerline of the reactor core out to a location interior to the primary biological shieldand over an axial span from an elevation below the lower core plate to above the upper core plate. As inthe case of the r,0 models, nominal design dimensions and full power coolant densities were employed inthe calculations. In this case, the homogenous core region was treated as an equivalent cylinder with avolume equal to that of the active core zone. The stainless steel former plates located between the corebaffle and core barrel regions were also explicitly included in the model. The rz geometric meshdescription of these reactor models consisted of 153 radial by 188 axial intervals. As in the case of ther,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations wasachieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in ther,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 radialmesh 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 SouthernNuclear Co and the Nuclear Fuels Division of Westinghouse. for each of the first twelve fuel cycles atAlvin \V. Vogtle Unit 1. Specifically, the data utilized included cycle dependent fuel assembly initialenrichments, burnups, and axial power distributions. This information was used to develop spatial andenergy dependent core source distributions averaged over each individual fuel cycle. Therefore, theresults 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 neutronexposure for each fuel cycle. In constructing these core source distributions, the energy distribution ofthe source was based on an appropriate fission split for uranium and plutonium isotopes based on theinitial enrichment and burnup history of individual fuel assemblies. From these assembly dependentfission splits, composite values of energy release per fission, neutron yield per fission, and fissionspectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discreteordinates code Version 3.1[241 and the BUGLE-96 cross-section library.E251 The BUGLE-96 libraryprovides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for lightwater reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5legendre 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, weretreated 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 surveillancecapsule positions, i.e., for the 29° dual capsule, 31.50 dual capsule, and 31.50 single capsule. Theseresults, representative of the axial midplane of the active core, establish the calculated exposure of the


- - _ _ _ _ _ _ _ I L

64

surveillance capsules withdrawn to date as well as projected into the future. Similar information isprovided in Table 6-2 for the reactor vessel inner radius at four azimuthal locations. The vessel datagiven in Table 6-2 were taken at the clad/base metal interface, and thus, represent maximum calculatedexposure 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 datatabulations include both plant and fuel cycle specific calculated neutron exposures at the end of theeleventh fuel cycle as well as future projections to 20, 24, 32, 40, 48, and 54 EFPY. The calculations forCycle 5 account for an uprate from 3411 MWt to 3565 MWt. The projections were based on theassumption that the core power distributions and associated plant operating characteristics from Cycle 12were representative of future plant operation. The future projections are also based on the current reactorpower level of 3565 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given inTables 6-3 and 64, respectively. The data, based on the cumulative integrated exposures from Cycles Ithrough 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 atthe 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 theplant and fuel cycle specific neutron transport calculations performed for the Alvin W. Vogtle Unit Ireactor.

Updated lead factors for the Alvin NV. Vogtle Unit I surveillance capsules are provided in Table 6-6. Thecapsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric centerof the surveillance capsule to the corresponding maximum calculated fluence at the pressure vesselclad/base metal interface. In Table 6-6, the lead factors for capsules that have been withdrawn from thereactor (U, Y, V and X) were based on the calculated fluence values for the irradiation periodcorresponding to the time of withdrawal for the individual capsules. For the capsules remaining in thereactor (W and Z), the lead factor corresponds to the calculated fluence values at the end of Cycle 11, thelast completed fuel cycle for Alvin NV. Vogtle Unit 1.


6-5

63 NEUTRON DOSIMETRY

The validity of the calculated neutron exposures previously reported in Section 6.2 is demonstrated by adirect comparison against the measured sensor reaction rates and via a least squares evaluation performedfor each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merelyserves to validate the calculated results, only the direct comparison of measured-to-calculated results forthe most recent surveillance capsule removed from service is provided in this section of the report. Forcompleteness, the assessment of all measured dosimetry removed to date, based on both direct and leastsquares evaluation comparisons, is documented in Appendix A.

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

Reaction Rates (rpslatom) M/CReaction Measured Calculated Ratio

6 3Cu(n,a)60Co 4.62E-1 7 4.28E-1 7 1.085 4 Fe(n,p) 54Mn 4.69E-1 5 4.70E-15 1.00SSNi(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)137Cs (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 from0.98 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 ofcomparisons given in Appendix A for all measured dosimetry removed to date from the Alvin W. VogtleUnit 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.


~- - ___________ _Ij-

6-6

6.4 CALCULATIONAL UNCERTAINTIES

The uncertainty associated with the calculated neutron exposure of the Alvin NV. Vogtle Unit 1surveillance capsule and reactor pressure vessel is based on the recommended approach provided inRegulatory Guide 1.190. In particular, the qualification of the methodology was carried out in thefollowing 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 measurementsfrom the H. B. Robinson power reactor benchmark experiment.

3 - An analytical sensitivity study addressing the uncertainty components resulting fromimportant input parameters applicable to the plant specific transport calculations used in theneutron exposure assessments.

4 - Comparisons of the plant specific calculations with all available dosimetry results from theAlvin W. Vogtle Unit I surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basictransport 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 addressuncertainties in operational or geometric variables that impact power reactor calculations. The secondphase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areasthat are primarily methods related and would tend to apply generically to all fast neutron exposureevaluations. The third phase of the qualification (analytical sensitivity study) identified the potentialuncertainties introduced into the overall evaluation due to calculational methods approximations as wellas to a lack of knowledge relative to various plant specific input parameters. The overall calculationaluncertainty applicable to the Alvin W. Vogtle Unit I analysis was established from results of these threephases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Alvin W. Vogtle Unit Imeasurements) was used solely to demonstrate the validity of the transport calculations and to confirmthe uncertainty estimates associated with the analytical results. The comparison was used only as a checkand was not used in any way to modify the calculated surveillance capsule and pressure vessel neutronexposures previously described in Section 6.2. As such, the validation of the Alvin W. Vogtle Unit Ianalytical 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 methodologyqualification. Additional information pertinent to these evaluations is provided in Reference 2.


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 theanalytical results.

The plant specific measurement comparisons described in Appendix A support these uncertaintyassessments for Alvin W. Vogtle Unit 1.


Ij-

6-8

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

240

180

._2

xC):1 20

60

0

0 75 150 225 300

R Axis (cm)


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)


~- - - _________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)


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)


- -

6-12

Table 6-1

Calculated Neutron Exposure Rates and Integrated ExposuresAt The Surveillance Capsule Center

Cumulative Cumulative Neutron Flux (E > 1.0 MeV)Cycle Irradiation Irradiation In/cm2-sl

Length Time Time Dual Dual SingleCycle [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+102 3.58E+07 7.19E+07 2.28 7.55E+10 8.09E+10 8.01E+103 4.19E+07 1.14E+08 3.61 7.78E+10 8.58E+10 8.51E+104 3.92E+07 1.53E+08 4.85 6.32E+10 6.84E+10 6.77E+105 4.19E+07 1.95E+08 6.17 6.69E+10 7.24E+10 7.18E+106 4.24E+07 2.37E+08 7.52 6.82E+10 7.33E+10 7.25E+107 3.98E+07 2.77E+08 8.78 6.11E+10 6.61E+10 6.55E+108 4.22E+07 3.19E+08 10.11 6.14E+10 6.75E+10 6.69E+109 4.61 E+07 3.65E+08 11.57 7.53E+10 8.74E+10 8.68E+1010 4.28E+07 4.08E+08 12.93 7.74E+10 8.49E+10 8.41 E+ 1011 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+10Future 1.26E+08 7.57E+08 24.00 6.77E+10 7.34E+10 7.27E+10Future 2.53E+08 I.01E+09 I 32.00 6.77E+10 7.34E+10 7.27E+10Future 2.53E+08 1.26E+09 40.00 6.77E+10 7.34E+10 7.27E+10Future 2.53E+08 1.51E+09 48.00 6.77E+10 7.34E+10 7.27E+10Future 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 atthe core midplane.


6-13

Table 6-1 cont'd

Calculated Neutron Exposure Rates And Integrated ExposuresAt The Surveillance Capsule Center

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)Cycle Irradiation Irradiation ln/cm2l

Length Time Time Dual Dual SingleCycle 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+182 3.58E+07 7.19E+07 2.28 5.82E+18 6.23E+18 6.17E+183 4.19E+07 1.14E+08 3.61 9.08E+18 9.83E+18 9.74E+184 3.92E+07 1.53E+08 4.85 1.16E+19 1.25E+19 1.24E+195 4.19E+07 1.95E+08 6.17 1.44E+19 1.56E+19 1.54E+196 4.24E+07 2.37E+08 7.52 1.73E+19 1.87E+19 1.85E+197 3.98E+07 2.77E+08 8.78 1.97E+19 2.13E+19 2.11E+198 4.22E+07 3.19E+08 10.11 2.23E+19 2.41E+19 2.39E+199 4.61E+07 3.65E+08 11.57 2.57E+19 2.82E+19 2.79E+1910 4.28E+07 4.08E+08 12.93 2.91E+19 3.18E+19 3.15E+1911 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+19Future 1.26E+08 7.57E+08 24.00 5.29E+19 5.77E+19 5.72E+19Future 2.53E+08 1.0 IE+09 32.00 6.99E+19 7.63E+19 7.55E+19Future 2.53E+08 1.26E+09 40.00 8.70E+19 9.48E+19 9.39E+19Future 2.53E+08 1.51E+09 48.00 1.04E+20 1.13E+20 1.12E+20Future 1.89E+08 1.70E+09 54.00 1.17E+20 1.27E+20 1.26E+20



-- . I'-.

6-14

Table 6-1 cont'd


Cumulative Cumulative Iron Atom Displacement RateCycle Irradiation Irradiation fda/sl

Length Time Time Dual Dual SingleCycle [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-102 3.58E+07 7.19E+07 2.28 1.47E-10 1.57E-10 1.56E-103 4.19E+07 1.14E+08 3.61 1.51E-10 1.67E-10 1.65E-104 3.92E+07 1.53E+08 4.85 1.23E-10 1.33E-10 1.31E-105 4.19E+07 1.95E+08 6.17 1.30E-10 1.40E-10 1.39E-106 4.24E+07 2.37E+08 7.52 1.32E-10 1.42E-10 1.40E-107 3.98E+07 2.77E+08 8.78 1.18E-10 1.28E-10 1.27E-108 4.22E+07 3.19E+08 10.11 1.20E-10 1.32E-10 1.30E-109 4.61 E+07 3.65E+08 11.57 1.46E-10 1.70E-10 1.68E-1010 4.28E+07 4.08E+08 12.93 1.50E-1O 1.65E-10 1.63E-1011 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-10Future 1.26E+08 7.57E+08 24.00 1.31E-10 1.42E-10 1.41E-10Future 2.53E+08 1.I E+09 32.00 1.31E-10 1.42E-10 1.41 E-I0Future 2.53E+08 1.26E+09 40.00 1.31E-10 1.42E-10 1.41E-10Future 2.53E+08 1.51E+09 48.00 1.31E-10 1.42E-10 1.41 E-10Future 1.89E+08 1.70E+09 54.00 1.311E-10 1.42E-10 1.41E-10



6-15

Table 6-1 cont'd


Cumulative Cumulative Iron Atom DisplacementsCycle Irradiation Irradiation Idpal

Length Time Time Dual Dual SingleCycle 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-032 3.58E+07 7.19E+07 2.28 1.14E-02 1.22E-02 1.20E-023 4.19E+07 1.14E+08 3.61 1.77E-02 1.91E-02 1.89E-024 3.92E+07 1.53E+08 4.85 2.25E-02 2.43E-02 2.41E-025 4.19E+07 1.95E+08 6.17 2.79E-02 3.02E-02 2.99E-026 4.24E+07 2.37E+08 7.52 - 3.35E-02 3.62E-02 3.58E-027 3.98E+07 2.77E+08 8.78 3.82E-02 4.13E-02 4.09E-028 4.22E+07 3.19E+08 .10.11 4.33E-02 4.69E-02 4.64E-029 4.61E+07 3.65E+08 11.57 5.OOE-02 5.47E-02 5.41E-0210 4.28E+07 4.08E+08 12.93 5.64E-02 6.17E-02 6.11 E-0211 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-02Future 1.26E+08 7.57E+08 24.00 1.03E-01 1.12E-01 1.11 E-01Future 2.53E+08 1.01E+09 32.00 1.36E-01 1.48E-01 1.46E-01Future 2.53E+08 1.26E+09 40.00 1.69E-01 1.84E-01 1.82E-01Future 2.53E+08 1.51 E+09 48.00 2.02E-01 2.20E-01 2.18E-01Future 1.89E+08 1.70E+09 54.00 2.27E-01 2.47E-01 2.44E-01



__ _ _ _ _U-

6-16

Table 6-2

Calculated Azimuthal Variation of Maximum Exposure Ratesand 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 ICycle 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+102 3.58E+07 7.19E+07 2.28 9.98E+09 1.57E+10 1.89E+10 1.80E+103 4.19E+07 1.14E+08 3.61 9.15SE+09 I.43E+10 1.95E+10 1.94E+104 3.92E+07 1.53E+08 4.85 1.04E+10 1.41E+10 1.62E+10 1.67E+105 4.19E+07 1.95E+08 6.17 9.05E+09 1.38E+10 1.68E+10 1.67E+106 4.24E+07 2.37E+08 7.52 9.13E+09 1.45E+10 1.72E+10 1.68E+107 3.98E+07 2.77E+08 8.78 8.88E+09 L.26E+10 1.55E+10 1.52E+108 4.22E+07 3.19E+08 10.11 9.08E+09 1.39E+10 1.57E+10 1.80E+109 4.61 E+07 3.65E+08 11.57 9.54E+09 1.36E+10 1.92E+10 2.18E+1010 4.28E+07 4.08E+08 12.93 9.04E+09 1.46E+10 1.93E+10 1.94E+1011 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+10Future 1.26E+08 7.57E+08 24.00 9.92E+09 1.45E+10 1.71 E+10 1.75E+10Future 2.53E+08 1.01E+09 32.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10Future 2.53E+08 1.26E+09 40.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10Future 2.53E+08 1.51 E+09 48.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10Future 1.89E+08 1.70E+09 54.00 9.92E+09 1.45E+10 1.71E+10 1.75E+10


6-17

Table 6-2 cont'd



Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)Cycle Irradiation Irradiation InJ ml-

Length Time TimeCycle 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+172 3.58E+07 -7.19E+07 2.28 8.18E+17 1.25E+18 1.46E+18 1.45E+183 4.19E+07 1.14E+08 3.61 1.20E+18 1.85E+18 2.28E+18 2.26E+184 3.92E+07 1.53E+08 4.85 1.60E+18 2.38E+18 2.90E+18 2.90E+185 4.19E+07 1.95E+08 6.17 1.98E+18 2.96E+18 3.60E+18 3.60E+186 4.24E+07 2.37E+08 7.52 2.36E+18 3.58E+18 4.33E+18 4.31E+187 3.98E+07 2.77E+08 8.78 2.71E+18 4.08E+18 4.94E+18 4.911E+188 4.22E+07 3.19E+08 10.11 3.1 OE+1 8 4.66E+18 5.60E+1 8 5.67E+189 4.61E+07 3.65E+08 11.57 3.54E+18 5.29E+18 6.49E+18 6.67E+1810 4.28E+07 4.08E+08 12.93 3.92E+18 5.911E+18 7.311E+18 7.50E+1811 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+19Future 1.26E+08 7.57E+08 24.00 7.38E+18 1.1OE+19 1.33E+19 1.37E+19Future 2.53E+08 1.01E+09 32.00 9.89E+18 1.47E+19 1.77E+19 1.81E+19Future 2.53E+08 1.26E+09 40.00 1.24E+19 1.83E+19 2.20E+19 2.25E+19Future 2.53E+08 1.51E+09 48.00 1.49E+19 2.20E+19 2.63E+19 2.70E+19Future 1.89E+08 1.70E+09 54.00 1.68E+19 2.48E+19 2.95E+19 3.03E+19


- ll U-

6-18

Table 6-2 cont'd

Calculated Azimuthal Variation of Fast Neutron Exposure Ratesand Iron Atom Displacement Rates At The Reactor Vessel


Cumulative Cumulative Iron Atom Displacement RateCycle Irradiation Irradiation Id alsl

Length Time TimeCycle 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 12 3.58E+07 7.19E+07 2.28 1.55E-11 2.42E-11 2.91E-11 2.85E-I I3 4.19E+07 1.14E+08 3.61 1.42E-11 2.20E-11 3.01E-1 I 3.07E- I4 3.92E+07 1.53E+08 4.85 1.61E-11 2.17E-11 2.51E-11 2.64E-1 15 4.19E+07 1.95E+08 6.17 1.41E-11 2.13E-11 2.59E-11 2.64E-I I6 4.24E+07 2.37E+08 7.52 1.42E-11 2.23E-11 2.66E-11 2.66E-II7 3.98E+07 2.77E+08 8.78 1.38E-11 1.95E-1I 2.39E-1I 2.40E-I I8 4.22E+07 3.19E+08 10.11 1.41E-11 2.14E-11 2.42E-1I 2.84E-I19 4.61E+07 3.65E+08 11.57 1.48E-11 2.09E-11 2.96E-11 3.44E-I I10 4.28E+07 4.08E+08 12.93 1.41E-11 2.25E-11 2.98E-11 3.06E-I II1 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 IFuture 1.26E+08 7.57E+08 24.00 1.54E-1 I 2.24E- I 2.64E- I 2.76E-1 IFuture 2.53E+08 1.O1E+09 32.00 1.54E-11 2.24E-1I 2.64E-11 2.76E-1I IFuture 2.53E+08 1.26E+09 40.00 1.54E-11 2.24E-11 2.64E-1I 2.76E-I IFuture 2.53E+08 1.51E+09 48.00 1.54E-11 2.24E-1I 2.64E-11 2.76E-I1 IFuture 1.89E+08 1.70E+09 54.00 1.54E-11 2.24E-11 2.64E-11 2.76E-I I


6-19

Table 6-2 cont'd



Cumulative Cumulative Iron Atom DisplacementsCycle Irradiation Irradiation Id pal

Length Time TimeCycle 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-032 3.58E+07 7.19E+07 2.28 1.27E-03 1.92E-03 2.25E-03 2.29E-033 4.19E+07 1.14E+08 3.61 1.87E-03 2.84E-03 3.51 E-03 3.58E-034 3.92E+07 1.53E+08 4.85 2.48E-03 3.66E-03 4.46E-03 4.58E-035 4.19E+07 1.95E+08 6.17 3.07E-03 4.55E-03 5.55E-03 5.68E-036 4.24E+07 2.37E+08 7.52 3.67E-03 5.50E-03 6.67E-03 6.811E-037 3.98E+07 2.77E+08 8.78 4.22E-03 6.27E-03 7.62E-03 7.76E-038 4.22E+07 3.19E+08 10.11 4.81 E-03 7.17E-03 8.64E-03 8.96E-039 4.61 E+07 3.65E+08 11.57 5.50E-03 8.14E-03 I.OOE-02 1.05E-0210 4.28E+07 4.08E+08 12.93 6.1OE-03 9.10E-03 1.13E-02 1.19E-02I 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-02Future 1.26E+08 7.57E+08 24.00 1.15E-02 1.69E-02 2.06E-02 2.17E-02Future 2.53E+08 1.01E+09 32.00 1.54E-02 2.26E-02 2.72E-02 2.86E-02Future 2.53E+08 1.26E+09 40.00 1.93E-02 2.82E-02 3.39E-02 3.56E-02Future 2.53E+08 1.5 1 E+09 48.00 2.32E-02 3.39E-02 4.06E-02 4.26E-02Future 1.89E+08 1.70E+09 54.00 2.61E-02 3.81 E-02 4.56E-02 4.78E-02


-_ _ _ _ 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 1.000225.59 0.571 0.567 0.561 0.557231.06 0.282 0.277 0.272 0.269236.54 0.134 0.130 0.127 0.125242.01 0.064 0.059 0.057 0.056

Note: Base Metal Inner Radius = 220.11 cmBase Metal I/4T = 225.59 cmBase Metal 1/2T = 231.06 cmBase Metal 3/4T = 236.54 cmBase 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 1.000225.59 0.642 0.637 0.635 0.645231.06 0.390 0.382 0.381 0.392236.54 0.237 0.227 0.227 0.235242.01 0.142 0.128 0.127 0.130

Note: Base Metal Inner Radius = 220.11 cmBase Metal 1/4T = 225.59 cmBase Metal 1/2T = 231.06 cmBase Metal 3/4T = 236.54 cmBase Metal Outer Radius = 242.01 cm


6-21

Table 6-5

Calculated Fast Neutron Exposure of Surveillance CapsulesWithdrawn from Alvin W. Vogtle Unit 1

Irradiation Time Fluence (E > 1.0 MeV) Iron DisplacementsCapsule fEFPY] I In/cm2 1 I fdpal

U 1.14 3.34E+18 6.52E-03Y 4.85 1.16E+19 2.25E-02V 8.78 1.97E+19 3.82E-02X 14.33 3.53E+19 6.85E-02

Table 6-6

Calculated Surveillance Capsule Lead Factors

Capsule IDAnd Location Status Lead FactorU(31.50 Dual) Withdrawn EOC 1 4.15Y (29.00 Dual) Withdrawn EOC 4 3.99V (29.00 Dual) Withdrawn EOC 7 3.98X (31.5° Dual) Withdrawn EOC 11 4.21

W (31.5° Single) In Reactor 4.17Z (31.5° Single) In Reactor 4.17

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


7-1

7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE

The following surveillance capsule removal schedule meets the requirements of ASTM El 85-82 and isrecommended for future capsules to be removed from the Vogtle Unit I reactor vessel. Thisrecommended 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, andAppendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear RegulatoryCommission, Washington, D.C.

3. WCAP-1 1011, "Georgia Power Company Alvin W. Vogtle Unit No. I Reactor Vessel RadiationSurveillance Program", L.R. Singer, February 1986.

4. WCAP-12256, "Analysis of Capsule U from the Georgia Power Company Vogtle Unit 1 ReactorVessel 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 IReactor Vessel Radiation Surveillance Program", M. J. Malone, et. al., August 1995.

6. WCAP-1 5067, "Analysis ofCapsule Vfrom Southern Nuclear Vogtle Electric Generating Plant UnitI 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-DuctilityTransition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society forTesting and Materials, Philadelphia, PA.

10. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G. Fracture Toughness Criteriafor Protection Against Failure

11. ASTM El 85-82, Standard Practice/or Conducting Surveillance Testsfor Light-Water CooledNuclear Power Reactor Vessels, in ASTM Standards, Section 3, American Society for Testing andMaterials, 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, inASTM 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 MetallicMfaterials, 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 ASTMStandards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

20. ASTM El 85-82, Standard Recommended Practicefor Conducting Surveillance TestsforLight- WaterCooled Nuclear Reactor Vessels.

21. WCAP-14370, Use of the Hyperbolic Tangent Function for Fitting Transition TemperatureToughness Data, T. R. Mager, et al, May 1995.

22. Regulatory Guide RG-1.190, Calculational and Dosimetry Methods for Determining Pressure VesselNeutron 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 MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

24. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two- and Three-DimensionalDiscrete Ordinates Neutron/Photon Transport Code System, August 1996.

25. RSIC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray GroupCross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel DosimetryApplications," March 1996.

References

A-O

APPENDIX A

VALIDATION OF THE RADIATION TRANSPORTMODELS 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 forall 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 dosimetryevaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods forDetermining Pressure Vessel Neutron Fluence.I'[A I One of the main purposes for presenting thismaterial is to demonstrate that the overall measurements agree with the calculated and least squaresadjusted values to within + 20% as specified by Regulatory Guide 1.190, thus serving to validate thecalculated neutron exposures previously reported in Section 6.2 of this report. This information may alsobe useful in the future, in particular, as least squares adjustment techniques become accepted in theregulatory 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 ofthe Alvin X'. Vogtle Unit I Reactor Vessel Materials Surveillance Program are presented. The capsuledesignation, location within the reactor, and time of withdrawal of each of these dosimetry sets were asfollows:

Azimuthal Withdrawal IrradiationCapsule 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 azimuthalangle 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 aresummarized 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 theradial, azimuthal, and axial center of the material test specimen array, gradient corrections were notrequired for these reaction rates. Pertinent physical and nuclear characteristics of the passive neutronsensors are listed in Table A-1.

The use of passive monitors such as those listed above does not yield a direct measure of the energydependent neutron flux at the point of interest. Rather, the activation or fission process is a measure ofthe integrated effect that the time and energy dependent neutron flux has on the target material over thecourse of the irradiation period. An accurate assessment of the average neutron flux level incident on thevarious monitors may be derived from the activation measurements only if the irradiation parameters arewell 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 documentedin References A-2, A-5, and A-6. The radiometric counting of the sensors from Capsule X was carriedout by Pace Analytical Services, Inc., located at the Westinghouse Waltz Mill Site. In all cases, theradiometric 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 gammaspectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performedby direct counting of each of the individual samples. In the case of the uranium and neptunium fissionsensors, the analyses were carried out by direct counting preceded by dissolution and chemical separationof cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules U, Y, V, and Xwas based on the monthly power generation of Alvin W. Vogtle Unit I from initial reactor criticalitythrough the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillancecapsules, the half-lives of the product isotopes are long enough that a monthly histogram describingreactor operation has proven to be an adequate representation for use in radioactive decay corrections forthe 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 operatinghistory of the reactor, reaction rates referenced to full-power operation were determined from thefollowing 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 corepower 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 averagej(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 irradiationperiod.

Appendix A

A-4

In the equation describing the reaction rate calculation, the ratio [Pj]/[P,,f] accounts for month-by-monthvariation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. Theratio Cj, which was calculated for each fuel cycle using the transport methodology discussed inSection 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced bychanges in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cjis normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing lowleakage fuel management, the additional Cj term should be employed. The impact of changing flux levelsfor constant power operation can be quite significant for sensor sets that have been irradiated for manycycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or forsensor 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 therespective 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 impuritiesin 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 inducedfission reactions that occurred over the course of the capsule irradiations. The correction factors appliedto the Alvin W. Vogtle Unit I fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V Capsule X23SU 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 neptuniumfission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules U, Y, V, and X are given in Table A-4. InTable A-4, the measured specific activities, decay corrected saturated specific activities, and computedreaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensorreaction rates are listed both with and without the applied corrections for 28U impurities, plutoniumbuild-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 thecorresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum withassociated uncertainties. Best Estimates for key exposure parameters such as ¢(E > 1.0 MeV) or dpa/salong with their uncertainties are then easily obtained from the adjusted spectrum. In general, the leastsquares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measuredsensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrumwithin their respective uncertainties. For example,

Ri + 8R, =Z(o, 8 + Ar )(Og -+60g

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

For the least squares evaluation of the Alvin NV. Vogtle Unit I surveillance capsule dosimetry, the FERRETcode!A-3] was employed to combine the results of the plant specific neutron transport calculations andsensor 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 withdrawnto 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 resultsof plant specific neutron transport calculations described in Section 6.2 of this report. The sensorreaction rates were derived from the measured specific activities using the procedures described inSection A.I.I. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRMLdosimetry cross-section library[A4]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations byASTM Standard E1018, "Applicationof ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculatedneutron 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 theAlvin 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 basicmeasurement process, irradiation history corrections, and corrections for competing reactions. A highlevel of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures thatconform 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 countingand data evaluation procedures were assigned the following net uncertainties for input to the leastsquares evaluation:

Reaction Uncertainty6 3Cu(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 SNLRMLlibrary. This data library provides reaction cross-sections and associated uncertainties, includingcovariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are providedin a fine multigroup structure for use in least squares adjustment applications. These cross-sections werecompiled from the most recent cross-section evaluations and they have been tested with respect to theiraccuracy and consistency for least squares evaluations. Further, the library has been empirically testedfor use in fission spectra determination as well as in the fluence and energy characterization of 14 MeVneutron sources.

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

Appendix A

A-7

Reaction Uncertainty63Cu(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 withthe sensor sets used in LWR irradiations.

Calculated Neutron Spectrum

The neutron spectra input to the least squares adjustment procedure were obtained directly from theresults of plant specific transport calculations for each surveillance capsule irradiation period andlocation. The spectrum for each capsule was input in an absolute sense (rather than as simply a relativespectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data weretreated equally with the measurements.

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

Mgg* =R + R* Rg, Pgg.

where R., specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg andRg. specify additional random groupwise uncertainties that are correlated with a correlation matrix givenby:

Pg, [1- aks, + 9 e-

where

H=(g-g')2H=272

The first term in the correlation matrix equation specifies purely random uncertainties, while the secondterm describes the short-range correlations over a group range y (0 specifies the strength of the latterterm). 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 calculatedspectra 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 surveillancecapsules withdrawn to date are provided in Tables A-S and A-6. In Table A-5, measured, calculated, andbest-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulationare 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 energyspectra to the measured reaction rates both before and after adjustment. In Table A-6, comparison of thecalculated and best estimate values of neutron flux (E > 1.0 MeV) and iron atom displacement rate aretabulated 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 calculatedspectra 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 thatthe use of the least squares evaluation results in a reduction in the uncertainties associated with theexposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that theuncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atomdisplacements 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 exposureparameters have been reduced to 6% for neutron flux (E > 1.0 MeV) and 8% for iron atom displacementrate. 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 sensorreaction rates are compared directly with the corresponding measurements. These threshold reaction ratecomparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculatedenergy spectra. In Table A-8, calculations of fast neutron exposure rates in terms of (E > 1.0 MeV) anddpa/s are compared with the best estimate results obtained from the least squares evaluation of thecapsule dosimetry results. These two levels of comparison yield consistent and similar results with allmeasurement-to-calculation comparisons falling well within the 20% limits specified as the acceptancecriteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/Ccomparisons for fast neutron reactions range from 0.88 to 1.14 for the 20 samples included in the dataset. The overall average M/C ratio for the entire set of Alvin NV. Vogtle Unit I data is 1.02 with anassociated standard deviation of 7.3%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the correspondingBEIC comparisons for the capsule data sets range from 0.95 to 1.04 for neutron flux (E > 1.0 MeV) andfrom 0.96 to 1.04 for iron atom displacement rate. The overall average BE/C ratios for neutronflux (E > 1.0 MeV) and iron atom displacement rate are 0.99 with a standard deviation of 3.5% and 1.00with a standard deviation of 3.3%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided inSection 6.2 of this report are validated for use in the assessment of the condition of the materialscomprising 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 FissionMaterial 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 dNickel "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 6.02Neptunium-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 theAlvin W. Vogtle Unit I surveillance capsules, approximately 90% of the sensor responseis due to neutrons in the energy range specified with approximately 5% of the totalresponse due to neutrons with energies below the lower limit and 5% of the totalresponse due to neutrons with energies above the upper limit.

Appendix A

A-li

Table A-2

Monthly Thermal Generation During The First Eleven Fuel CyclesOf The Alvin W. Vogtle Unit I Reactor

(Reactor power of 3411 MWt from startup through Cycle 4 (3/13/93) and3565 MWt from Cycle 5 (4/27/93) through the End of Cycle 11)

Thermal Thermal ThermalGeneration Generation Generation

Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr)1987 3 68766 1990 3 .0 1993 3 8497521987 4 797491 1990 4 591136 1993 4 1667501987 5 1044332 1990 5 2311713 1993 5 24015021987 6 759746 1990 6 2299026 1993 6 25644371987 7 1835718 1990 7 2196834 1993 7 24991301987 8 2509822 1990 8 2512580 1993 8 26459701987 9 2452829 1990 9 2452206 1993 9 25601401987 10 707673 1990 10 2534258 1993 10 26499621987 11 1927388 1990 11 2428733 1993 11 25582331987 12 2467702 1990 12 1692955 1993 12 26460461988 1 1365280 1991 1 2534837 1994 1 26397581988 2 1387377 1991 2 2260779 1994 2 21566171988 3 2456340 1991 3 2495386 1994 3 25812091988 4 1907244 1991 4 2449552 1994 4 25573271988 5 2531355 1991 5 2533685 1994 5 25541731988 6 2444967 1991 6 2449889 1994 6 25613791988 7 2220349 1991 7 2534501 1994 7 26469041988 8 2415264 1991 8 2483204 1994 8 24489461988 9 2370737 1991 9 969976 1994 9 6299271988 10 483956 1991 10 0 1994 10 10997011988 11 52233 1991 11 215953 1994 11 25644651988 12 2135007 1991 12 2466013 1994 12 26316521989 1 1771903 1992 1 2534684 2001 1 26503771989 2 1905573 1992 2 2371364 2001 2 21306211989 3 2533004 1992 3 2528590 1995 3 26501821989 4 2380073 1992 4 2239948 1995 4 25618021989 5 2264902 1992 5 1866712 1995 5 26308211989 6 2452382 1992 6 2452840 1995 6 25649441989 7 2443387 1992 7 2534681 1995 7 23817191989 8 2286024 1992 8 2535008 1995 8 26508441989 9 2450229 1992 9 2188889 1995 9 25199611989 10 2142954 1992 10 2538900 1995 10 26515201989 11 2391716 1992 11 2454211 1995 11 25649131989 12 2535607 1992 12 2536190 1995 12 26507291990 1 2374089 1993 1 2536730 1996 1 26506081990 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 CyclesOf The Alvin W. Vogtle Unit 1 Reactor

(Reactor power of 3411 MWt from startup through Cycle 4 (3/13/93) and3565 MWt from Cycle 5 (4/27/93) through the End of Cycle 11)

Thermal Thermal ThermalGeneration Generation Generation

Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr)1996 3 130446 1999 3 299858 2002 3 4000691996 4 648324 1999 4 2560910 2002 4 7283411996 5 2258085 1999 5 2651230 2002 5 26516041996 6 1467397 1999 6 2565431 2002 6 25640801996 7 2651000 1999 7 2651031 2002 7 26513221996 8 2651013 1999 8 2650940 2002 8 26275591996 9 2565318 1999 9 2565539 2002 9 25654441996 10 2654401 1999 10 2654884 2002 10 26548081996 11 2442399 1999 11 2565850 2002 11 20387921996 12 2648271 1999 12 2650654 2002 12 23807641997 1 2648498 2000 1 2651323 2003 1 26187181997 2 2393961 2000 2 2479813 2003 2 23946791997 3 2392019 2000 3 2650657 2003 3 26512551997 4 1086834 2000 4 2562272 2003 4 25612561997 5 2489873 2000 5 2651160 2003 5 24095391997 6 2565296 2000 6 2375260 2003 6 25604071997 7 2645858 2000 7 2649880 2003 7 26467051997 8 2650538 2000 8 2601536 2003 8 26474041997 9 503695 2000 9 1185996 2003 9 22797391997 10 677865 2000 10 10626981997 11 2565214 2000 11 25655531997 12 2652171 2000 12 25316491998 1 2650194 2001 1 26514031998 2 2395254 2001 2 23948101998 3 2650194 2001 3 26187841998 4 2563238 2001 4 25623811998 5 2547428 2001 5 26508971998 6 2563238 2001 6 25657001998 7 2648218 2001 7 26509201998 8 2632408 2001 8 22974811998 9 2563238 2001 9 25532011998 10 2650194 2001 10 26442591998 11 2561262 2001 11 25570601998 12 2640313 2001 12 26440571999 1 2648218 2002 1 26465811999 2 2211460 2002 2 2385611

Appendix A

A-13

t III,Table A-3

Calculated Cj Factors at the Surveillance Capsule CenterCore Midplane Elevation

Fuel *(E > 1.0 MeV) [n/cm2-sJC ycle _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Cycle Capsule U Capsule Y Capsule V Capsule XI 9.25E+10 8.63E+10 8.63E+10 9.25E+102 7.55E+10 7.55E+10 8.09E+103 7.78E+10 7.78E+10 8.58E+104 6.32E+10 6.32E+10 6.84E+105 6.69E+10 7.24E+106 6.82E+10 7.33E+107 6.1lE+10 6.61E+108 6.75E+109 8.74E+1 010 8.49E+1 011 7.99E+10

Average 9.25E+10 7.58E+10 7.1lE+10 7.81E+10

Fuel C,Cycle

Capsule U Capsule Y | Capsule V Capsule XI 1.00 1.14 1.21 1.192 1.00 1.06 1.043 1.03 1.10 1.104 0.84 0.89 0.885 0.94 0.936 0.96 0.947 0.86 0.858 0.869 1.1210 1.091 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 RadiallyAdjusted Adjusted

Measured Saturated Saturated ReactionActivity 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-17Middle 4.38E+04 3.34E+05 3.34E+05 5.1OE-17Bottom 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-15Middle 1.34E+06 3.20E+06 3.20E+06 5.07E- 15Bottom 1.36E+06 3.25E+06 3.25E+06 5.14E- 15Average 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-15Average 7.34E-15

-23U (n,f) 137 CS (Cd) | Middle | 1.29E+05 | 5.02E+06 | 5.02E+06 I 3.30E-14U (n,f) Cs (Cd) Including 235U, 2 9Pu, and y,fission corrections: 2.78E-14

237Np (n,f) '37Cs (Cd) I Middle | 1.24E+06 | 4.82E+07 I 4.82E+07 | 3.08E-13231Np (n,f) Cs (Cd) Including yfission correction: 3.05E-13

59Co (n,y) OCo | Top 1.03E+07 7.86E+07 7.86E+07 5.13E-12Middle | l.OIE+07 7.71E+07 | 7.71E+07 | 5.03E-12Bottom | 1.05E+07 8.01E+07 8.01E+07 j 5.23E-12Average | 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-12Average 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


Surveillance Capsule Y



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-17Middle 1.21E+05 2.89E+05 j 2.89E+05 4.41E-17Bottom 1.23E+05 2.94E+05 T 2.94E+05 4.48E-17Average _ _ 4.64E-17

1 4Fe (n,p) 54Mn Top 1.63E+06 2.87E+06 2.87E+06 4.55E-15Middle 1.47E+06 2.59E+06 2.59E+06 4.1OE-15Bottom 1.48E+06 2.60E+06 2.60E+06 4.13E-15Average 4.26E-15

58Ni (n,p) 58Co Top 8.43E+06 4.47E+07 4.47E+07 6.40E-1 5Middle 7.75E+06 4.11 E+07 4.11 E+07 5.89E-15Bottom 7.63E+06 4.05E+07 4.05E+07 5.79E-15Average 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-13Np (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-12Middle 2.35E+07 5.62E+07 5.62E+07 3.66E-12Bottom 2.34E+07 5.59E+07 5.59E+07 3.65E-12Average . 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-12Bottom 1.29E+07 3.08E+07 3.08E+07 2.01E-12Average 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



Appendix A

- _ _ __1 2

A-16

Table A-4 cont'd


Surveillance Capsule V



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-17Middle 1.55E+05 2.72E+05 2.72E+05 4.15E-17Bottom 1.55E+05 2.72E+05 2.72E+05 4.15E-17Average 4.33E-17

54Fe (n,p) 54Mn Top 1.35E+06 2.73E+06 2.73E+06 4.33E-15Middle 1.24E+06 2.51E+06 2.51 E+06 3.98E-15Bottom 1.23E+06 2.49E+06 2.49E+06 3.94E-15Average 4.08E-15

58Ni (n,p) 5 8 Co Top 4.20E+06 4.37E+07 4.37E+07 6.25E-15Middle 3.89E+06 4.04E+07 4.04E+07 5.79E-15Bottom 3.88E+06 4.03E+07 4.03E+07 5.77E-15Average 5.94E-15

238U (n,f) '37Cs (Cd) ] Middle | 8.45E+05 | 4.82E+06 I 4.82E+06 | 3.17E-1423SU (nf) '37Cs (Cd) | Including ..5U, " 9Pu, and y,fission corrections: 2.47E-14

27Np (nf) '"CS (Cd) Middle | 6.27E+06 I 3.58E+07 I 3.58E+07 2.28E-13237Np (nf) '-"Cs (Cd) Including yfission correction: 2.26E-13

59Co (ny) 6wCo Top 2.88E+07 5.05E+07 5.05E+07 3.30E-12Middle 2.88E+07 5.05E+07 5.05E+07 3.30E-12Bottom 2.87E+07 5.03E+07 5.03E+07 3.28E-12Average 3.29E-12

59Co (n,'y) 60Co (Cd) Top | 1.45E+07 2.54E+07 2.54E+07 1.66E-12Middle | 1.50E+07 2.63E+07 2.63E+07 1.72E-12Bottom | 1.53E+07 2.68E+07 2.68E+07 1.75E-12

_ Average | _ _ 1.71E-12Notes: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



Appendix A

A-17

Table A-4 cont'd


Surveillance Capsule X



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-17Middle 2.23E+05 2.89E+05 2.89E+05 4.41E-17Bottom 2.26E+05 2.93E+05 2.93E+05 4.47E-17Average . . 4.62E-17

54Fe (n, p) 54Mn Top 2.41E+06 3.15E+06 3.15E+06 4.99E-15Middle 2.20E+06 2.88E+06 2.88E+06 4.56E-15Bottom 2.19E+06 2.86E+06 2.86E+06 4.54E-15Average 4.70E-15

"Ni (n,p) 58Co Top 1.55E+07 4.69E+07 4.69E+07 6.71E-15r Middle 1.44E+07 4.35E+07 4.35E+07 6.23E-15_ Average | _ _ 6.47E-15

U (nf) "3 7Cs (Cd) I Middle | 1.63E+06 | 5.94E+06 | 5.94E+06 3.90E-142sU (nf) "Cs (Cd) Including 235U, 239Pu, and y,fission corrections: 2.86E-14

237Np (n,f) I37Cs (Cd) I Middle 1.0-9E+07 | 3.97E+07 | 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-12Middle 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-12Middle 2.56E+07 3.32E+07 3.32E+07 2.16E-12Bottom 2.60E+07 3.37E+07 3.37E+07 2.20E-12Average | 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


Appendix A

1L-

A-18

Table A-5

Comparison of Measured, Calculated, and Best EstimateReaction Rates At The Surveillance Capsule Center

Capsule U

Reaction Rate Irns/atomlBest

Reaction Measured Calculated Estimate IUC MIBE63 Cu(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 0.985Ni(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 0.97237Np(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 1.0259Co(ny)60Co (Cd) 2.70E-12 2.89E-12 2.74E-12 0.93 0.99

Capsule V

Reaction Rate I s/atomlBest

Reaction Measured Calculated Estimate MIC MIBE63Cu(n,a)6OCo 4.64E-17 4.11E-17 4.43E-17 1.13 1.0554Fe(n,p)54Mn 4.26E-15 4.52E-15 4.44E-15 0.94 0.9658Ni(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 1.12237Np(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 1.0259Co(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 EstimateReaction Rates At The Surveillance Capsule Center

Capsule Y

Reaction Rnte Irmq/atnm1Best

Reaction Measured Calculated Estimate M/C M/BE6 3Cu(n,a)"Co 4.32E-17 3.93E-17 4.17E-17 1.10 1.045Fe(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 1.09137Np(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/atomlBest

Reaction Measured Calculated Estimate MWC M/BE63Cu(n,a)60Co 4.62E-17 4.28E-17 4.51E-17 1.08 1.0254Fe(n,p)54 Mn 4.69E-15 4.70E-15 4.79E-15 1 0.985"Ni(n,p)"Co 6.47E-15 6.58E-15 6.67E-15 0.98 0.97

238U(nf)137Cs (Cd) 2.86E-14 2.50E-14 2.57E-14 -1.14 1.11137Np(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 RatesAt The Surveillance Capsule Center

R(E > 1.0 MeV) fn/cm2-sl lBest Uncertainty

Capsule ID Calculated Estimate (1a) BE/CU 9.34E+10 9.04E+10 6% 0.97Y 7.62E+10 7.25E+10 6% 0.95V 7.17E+10 7.17E+10 6% 1.00X 7.88E+10 8.11E+10 6% 1.03

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

Iron Atom Displac ment Rate [dpa/slJ Best UncertaintyCapsule ID Calculated Estimate (10 BE/C

U 1.81E-10 1.78E-10 8% 0.99Y 1.47E-10 1.40E-10 8% 0.95V 1.38E-10 1.38E-10 8% 1.01X 1.52E-10 1.52E-10 8% 1.03

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

Appendix A

A-21

Table A-7

Comparison of Measured/Calculated (M/C) Sensor Reaction RateRatios Including all Fast Neutron Threshold Reactions

M/C RatioReaction

Capsule U Capsule Y Capsule V Capsule X63Cu(n,a)60Co 1.10 1.13 1.10 1.085 4 Fe(n,p)54Mn 0.97 0.94 0.95 1.0058Ni(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 associatedstandard deviation of 7.1%.

Table A-8

Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios

BE/C RatioCapsule ID *(E > 1.0 MeV) dpa/s

U 0.98 0.99Y 0.96 0.95V 1.01 1.01X 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 PressureVessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear RegulatoryResearch, March 1995.

A-2. WCAP-12256, "Analysis of Capsule U from the Georgia Power Company Vogtle Unit I ReactorVessel Radiation Surveillance Program," May 1989.

A-3. A. Schmittroth, FERRETData Analysis Core, HEDL-TME 7940, Hanford EngineeringDevelopment Laboratory, Richland, WA, September 1979.

A-4. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-SectionCompendium", July 1994.

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

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

Appendix A

B-O'

APPENDIX B

LOAD-TIME RECORDS FOR CHARPYSPECIMEN 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 3000.000-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- 3000.000-J

2000.00

1000.00

/. %AD111111 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

0.000.00 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

0.000.00 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

0.000.0 6.000 1.00 2.00 3.00

rAne81 (ms2

AL48, 125°F

4.00 5.00

5000.00

4000.00

.7 3000.00coa0

-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

-, 3000.000 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

, 3000.0003

2000.00

0.00 1.00 2.00 3.00 4.00 5.00

Tine-I (ms)

AL46, 1750 F

6.00

Appendix B

B-6

5000.00

4000.00

,33000.00

2000.00

1000.00

0.000.00 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

0.000.00 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

-, 3000.000

2000.00

1000.00 I

0.000.00 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

0.000.o 1.00 2.00 3.00 4.00 5.00 6.00

Time-I (ms)

AL50, 2500F

Appendix B

B-8

5000.00

3000.0030M0

2000.00

1000.00

0.000.00 1.00 2.00 3.00 4.00 5.00 6.00

Time-1 (ms)

AL56, 2750 F

5000.00 .

4000.00.

.0.

.3 3000.000

-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

0.0000 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

0.000.

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- 3000.00a

-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.

ofoo0.00 1.00 2.00 3.00 4.00 5.00 6.00

rime-1 (ms)

AT51, 1500 F

5000.00

4000.00

, 3000.000

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

0.000.0

_ . . . .

0I 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' 3000.000-J

2000.00

1000.00

0.000.00

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

0.000.00 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

0.000.00 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

0.000.00 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

X3000.00

0

2000.00

1000.00

0.000.00 1.00 2.00 3.00 4.00 5.00 6.00

Tme-1 (ms)

AW57, -1000F

5000.001

4000.00.

'a'0 3000.000

2000.00

1000.00

0.00l 1OS -vP r o ,- , - | -

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

a0-A

Time-1 (Ms)

AW56, 250 F

5000.00*

4000.00

a 3000.00030

-J

2000.00

1000.00.

00 3.00

Time-i (ms)

AW52, 50F

6.00

Appendix B

B-19'

5000.00

4000.00

- 3000.0001

-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 3000.000

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.W0.00 1.00 2.00 3.00 4.00 5.00 6.00

Time-1 (Ms)

AW58, 1750F

4000.00 .

Ca 3000.000-j

2000.00

1000.00

0.000.00 1.00 2.00 3.00 4.00 5.00 6.00

Tme-1 (ms)

AW60, 2000 F

Appendix B

B-23

a

0-J

5000.00

4000.00

3000.00

2000.00

1000.00

0.00IDO 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.

IJLIFew

0.00 1.00 2.00 3.00

Twne-I (ms)

AH49, -1500 F

4.00 5.00 6H0

Appendix B

l.

B-24

5000.00

4000.00

-

X7 3000.000

-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, -1000 F

0.00 1.00 2.00 3.00 4.00 5.00

Time-I (ms)

AH48, -750 F

6.00

Appendix B

B-25

5000.00

4000.00

;, 3000.000

-J

2000.00

1000.00.

0.000.

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

a600

-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.

2000.00|

1000.00

0.00I0.00 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.

0.000.00 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

0.000.00 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

0.000.00 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 3000.000-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 3000.000

-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 XUSING 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 CharpyV-notch plots for Capsule X using CVGRAPH, Version 5.0.2. Applicable Charpy V-notch plots forCapsule U, Y, and V are included in WCAP-15067 61. The definition for Upper Shelf Energy (USE) isgiven 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 insets of three at each test temperature, the set having the highest average may be regarded asdefining the upper shelf energy."

If there are specimens tested in set of three at each temperature Westinghouse reports the set having thehighest average energy as the USE (usually unirradiated material). If the specimens were not tested insets of three at each temperature Westinghouse reports the average of all 100% shear Charpy data as theUSE. Hence, the USE values reported in Table C-1 and used to generate the Charpy V-notch curves weredetermined 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 109B8805-3 (Long.)

Intermediate Shell Plate 96 98 106 94 93B8805-3 (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 PMPage 1

Coefficients of Curve IA = 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 FPlant: VOGTLE I Material: SA533B1 Heat: C0623-1

Orientation: LT Capsule: UNIRR Fluence: n/cm^2300

250

(', 200

800

IL

M 150a)r-

z> 100

50

0-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data

Temperature Input CVN

-40. 00-40. 00-40. 00- 20. 00- 20. 00-20. 00

.00

.00

.00

1 1. 0019. 00

9. 0054. 0028. 0012.0052.0047. 0042. 00

Computed CVN

20. 1920. 1920. 1927. 7527. 7527. 7537. 3937. 3937. 39

Differenti3l

-9. 19-1. 19

1I 1. 1926. 25

. 2515. 7514.619.614.61

I C1-2

a-

UNIRRADIATED (LONGITUDNAL ORIENTATION)

Page 2Plant: VOGTLE I Material: SA533BI Heat: C0623-1

Orientation: LT Capsule: UNLR Fluence: n/cmA2

Charpy V-Notch Data


40. 0040. 0040. 0080. 0080. 0080. 00

100. 001 00. 001 00. 00120.00120.00120.00180. 00180. 00180. 00260. 00260. 00320. 00320. 00

48.0062. 0060. 0093.0064. 0070. 0084.00

107.001 10. 00100. 00116. 00109. 00126. 00115.0011 6.00129. 0012 1. 00131. 001 9.00

Computed CVN

61.4561.4561.4585. 7285. 7285. 7295.5695. 5695.56

103. 33103.33103. 33116.17116. 17] 16.17120. 90120.90121. 69121.69

Differential

- 13.45.55

- 1.457.28

-21.72- 15. 72

I 1.5611.4414.44- 3. 3312.675.679. 83

-1.17- .178. 10

.109.31

- 2.69

Correlation Coefficient = .962

(,-,3

CAPSULE X (LONGITUDINAL ORIENTATION)


Coefficients of Curve 2A = 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 FPlant: \'OGTLE I Material: SA533B] Heat: C0623-1Orientation: LT Capsule: X Fluence: n1cmA2

300

250

w, 200

, -00IL

i 1500

z> 1000

50

0 k-

-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data

Temperature

-25. 0025. 0050. 0050. 0075.00

100. 00125. 00150. 001 60. 00

Input CVN

3.008.00

30. 0013.0035.0042.0059.0048. 0066. 00

Computed CVN Differential

6. 1312. 6318.661S. 6627. 2738. 5751. 8765. 6670. 90

- 3. 13-4. 631]. 34-5. 66

7. 733.437. 13

- 17. 66-4. 90

I P-A


Page 2Plant: VOGTLE I Material: SA533BI Heat: C0623-1

Orientation: LT Capsule: X Fluence: n/cmA2

Charpy V-Notch Data

Temperature

175.00200. 00225. 00225. 00250. 00275. 00

Input CVN

75. 0075. 00

117. 00105. 00103. 00113..00

Computed CVN

78. 1888. 2695. 6395. 63

100. 63103. 87

Differential

- 3.18-13.2621.379.372.379. 13


UNIRRADIATED (LONGITUDINAL ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03119/2004 01:57 PMPage 1

Coefficients of Curve IA = 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)

[email protected]. 35 mils=18.8 DegFPlant: VOGTLE I Material: SA533B1 Heat: C0623-1

Orientation: LT Capsule: UNIRR Fluence: n/cmA2200

1502

0

2 100

'I-a)

50

0-300 0 300


600

Charpy V-Notch Data

Temperature Input L.E.

-40. 00-40. 00-40. 00-20. 00-20. 00-20. 00

. 00

. 00

.00

7.0012. 006.00

38. 0022.00

8.0039. 0034. 0033.00

Computed L.E.

15. 1015. 1015. 1020. 6920.6920. 6927. 5727. 5727. 57

Differential

- 8. 10- 3. 10-9. 1017.31

1.3112. 6911.436.435.43

C-6


Page 2Plant: 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. 0540.00 43.00 44. 05 - 1.0540.00 44.00 44.05 - .0580.00 66.00 60.43 5.5780.00 45.00 60.43 -15.4380. 00 55. 00 60.43 - 5.43

100. 00 62. 00 67.20 - 5.20100.00 69.00 67.20 1.80100. 00 77. 00 67.20 9. 80120. 00 72. 00 72.68 -. 68120. 00 81. 00 72.68 8.32120.00 74.00 72.68 1.32180. 00 84. 00 82.29 1. 71180. 00 82. 00 82.29 - . 29180. 00 80. 00 82.29 - 2.29260.00 92.00 86.27 5.73260.00 83.00 86.27 -3.27320. 00 85. 00 87.02 - 2. 02320. 00 85.00 87.02 - 2.02




Coefficients of Curve 2A = 38.24 B = 38.24 C = 104.72 TO = 139.31 D = O.OOE+0O


Temp. @L.E. 35 niils=130.5 Deg FPlant: VOGTLE I Material: SA533BI Heat: C0623-1

Orientation: LT Capsule: X Fluence: n/cmA2

200

1502

E

aa

0

2.100

50

0-300 0 300

Temperature in Deg F600

Charpy V-Notch Data


-25. 0025. 0050. 0050. 0075. 00

100. 00125. 00150.00160. 00

. 003. 00

14. 007. 00

21. 0027. 0040. 0036. 0045. 00

Computed L.E.

3. 187.75

11.7611.7617.3224. 5233. 0542. 1345. 70

Differential

-3. 18-4.75

2. 24-4. 76

3. 682. 486. 95

- 6.13- . 70

C-8


Plant: VOGTLE IOrientation: LT

Page 2Material: SA533B1Capsule: X Fluence:

Heat: C0623-1n/cmA2

Charpy V-Notch Data


175. 00200. 00225. 00225. 00250. 00275.00

50. 0051.0070. 0068. 0067.0070. 00

Computed L.E.

50. 7958.2164.0264. 0268. 2471.15

Differential

-. 79-7. 215.983. 98

- 1. 24- 1. 15


. 1. C-9



Coefficients of Curve 1A = 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: C0623-1Orientation: LT Capsule: UNIRR Fluence: n/cmA2

125

100

3.-

co

r-0i0P

75

50

25

o 4-

-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data

Temperature Input Percent Shear

-40. 00-40.00-40.00- 20. 00- 20.00- 20.00

.00

.00

.00

5.0014.009.00

27.0014.005.00

30. 0030. 0025.00

Computed Percent Shear

9. 369. 369. 36

13.7613.7613.7619. 7819. 7819. 78

Differential

-4.364. 64-. 36

13.24.24

- 8.7610.2210.22

5. 22

. i C-10

a-


Page 2Plant: VOGTLE 1 Material: SA533B1 H

Orientation: LT Capsule: UNIRR Fluence:.eat: C0623-1

n/cmA2

Charpy V-Notch Data

Temperature

40. 0040. 0040. 0080. 0080. 0080. 00

100. 00100. 00100. 00120. 00120. 00120. 00180. 00180. 00180. 00260. 00260. 00320. 00320. 00

Input Percent Shear

30.0036.0036. 0060. 0045. 0040. 0055. 0075. 0075. 0080. 0085. 0085. 00

100. 00100. 00100. 00100. 00100. 00100. 00100. 00


37. 0537. 0537.0558.4258. 4258.4268.4668.4668.4677.0377. 0377. 0392.5292.5292. 5298.6098.6099. 6299. 62

Differential

-7. 05- 1. 05- I. 05

1.58- 13.42

- 18.42- 13. 46

6.546.542.977.977.977.487.487.481.401.40

.38

.38


C-1l1



Coefficients of Curve 2A = 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: C0623-1Orientation: LT Capsule: X Fluence: nlcmA2

125

100

4-acm0

coCa)20~

75

50

25

0 4--300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data


- 25. 0025.0050. 0050. 0075. 00

100. 00125.00150. 00160. 00

2. 005. 00

15. 0025. 0025. 0030. 0040. 0045. 0050. 00


2.497. 13

11.7411. 7418.7428.5640. 9454.5859. 96

Differential

- . 49- 2. 133.26

13.266.261. 44- . 94

-9.58-9. 96

C-12

I. I


Plant: VOGTLE 1Orientation: LT

Page 2Material: SA533BI

Capsule: X Fluence:Heat: C0623-1

n/cmA2

Charpy V-Notch Data


175.00200. 00225. 00225. 00250. 00275. 00

65.0070. 00

100. 00100. 00100. 00100. 00


67.5778. 3286.2386. 2391.5794. 96

Differential

- 2. 57-8. 3213.7713.77

8. 435. 04


C-13

VNIRRADIATED (TRANSVERSE ORIENTATION)


Coefficients of Curve IA = 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 FPlant: VOGTLE I Material: SA533B1 Heat: C0623-1

Orientation: TL Capsule: UNIRR Fluence: n/cmA2300

250

-D 200900

M150a)

zi 100

50

0-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data


-40. 00-40. 00

. 00. 00. 00

40. 0040. 0040. 0080. 00

9.0016.0026.0024. 0024. 0035.0046. 0053.0055. 00

Computed CVN

13.4013. 4023. 8623. 8623. 8639. 6739. 6739. 6758. 07

. Differential

-4. 402. 602. 14

.14

.14-4.67

6. 3313. 33-3. 07

C-14

1_

UNIRRADIATED (TRANSVERSE ORIENTATION)

Page 2Plant: VOGTLE 1 Material: SA533BI H

Orientation: TL Capsule: UNIRR Fluence:Seat: C0623-1

nlcmA2

Charpy V-Notch Data

Temperature

80. 0080. 00

120.00120.00120. 00140. 00140.00180. 00180.00180. 00240. 00240. 00320. 00320. 00

Input CVN

45. 0056.0069. 0073.0071. 0084. 0079.0094. 0098. 0090. 0097. 00

102. 0097. 0096. 00

Computed CVN

58. 0758. 0773. 9973. 9973.9979. 9879.9888. 0288.0288. 0293.4393.4395. 4695. 46

Differential

-13. 07-2. 07-4. 99

- .99-2. 994.02- .985.989.981.983. 578.571.54

.54


C-15

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/19/2004 02:41 PMPage 1 ;

Coefficients of Curve 2

300

250

o,- 20000LL

E0 150a)

r-

wz> 100

50

0

A = 47.6 B = 45.4 C = 124.11 TO = 128.59 D = 0.OOE+00Equation 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-lOrientation: TL Capsule: X Fluence: n/cm^2

_ _ _ ___ _ _ _ ___ _____________,

0 ~-- -- - - - -- - - - - - - -

------------ -- - -

-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data

Temperature Input CVN Computed CVN Differential

-50. 00- 25.00

25. 0050. 0075.00

100. 00125. 00150. 00175. 00

5.007.00

14. 0025. 0033. 0042. 0047.0052. 0064.00

7.049.25

16.5922. 1629. 1337.3246.2955.3663.83

-2.04-2.25-2.59

2. 843. 874. 68

.71-3. 36

.17

*i C-16


Plant: VOGTLE 1Orientation: TL

Page 2Material: SA533B1 Heat: C0623-1Capsule: X Fluence: n/cmA2

Charpy V-Notch Data


200. 00200. 00225. 00225. 00250. 00275. 00

64. 0061.0085. 0066. 0095. 0098.00

Computed CVN

71.1871.1877. 1577. 1581. 7585. 16

Differential

-7. 18- 10. 1 8

7. 85-I 1. 15

13. 2512. 84


C-17



Coefficients of Curve IA =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 FPlant: VOGTLE I Material: SA533BI Heat: C0623-1


150

n0

2 100

1..

0

9

-300 0 300


600

Charpy V-Notch Data


-40. 00-40. 00

.00

.00

. 0040. 0040. 0040. 0080. 00

5.00II. 0019.0017. 0019. 0027. 0037. 0041. 0040. 00

Computed L.E.

14.3814.3821.5621.5621.5630.9830.9830.9842. 17

-9. 38-3. 38- 2. 5 6-4. 56-2. 56- 3. 9 8

6. 0210. 02- 2. 17

Differential

C-18

-


Page 2Plant: VOGTLE 1 Material: SA533B1 Heat: C0623-1


Charpy V-Notch Data


80. 0080. 00

120. 00120. 00120. 00140. 00140. 00180.00180. 00180. 00240. 00240. 00320. 00320. 00

37. 0044. 0053.0056. 0058. 0063. 0062.0073.0071. 0080. 00*77. 0074. 0080. 0074. 00

Computed L.E.

42. 1742. 1754. 0554. 0554. 0559.8159. 8170. 1870. 1870. 1881. 7481.7490. 1890. 18

Differential

-5. 171. 83

- 1. 051. 953. 953. 192. 192. 82

.829. 82

- 4. 74-7. 74

- 10. 18- 16. 18

Cornelation Coefficient = .966

C-19



200

0

r-B.n

CE

150

100

50

Coefficients of Curve 2A = 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)

[email protected]. 35 mils= 146.6 Deg FPlant: VOGTLE I Material: SA533BI Heat: C0623-1

Orientation: TL Capsule: X Fluence: nlcmA2

fa 1P ,

------ ------ ------

0

-300 0 300 600


Charpy V-Notch Data

Temperature

-50. 00- 25.00

25. 0050. 0075. 00

100. 00125. 00150. 00175. 00

Input L.E.

.00

.007. 00

15. 0025.0031.0032.0038.0045. 00

Computed L.E.

5.527.21

12. 0915.4419.5024. 2929. 7835. 8742.36

Differential

-5. 52-7.21-5. 09* 445.506.712.222. 132. 64

C-20

itQ


Plant: VOGTLE IOrientation: TL

Page 2Material: SA533BI Heat: C0623-1Capsule: X Fluence: n/cmA2

Charpy V-Notch Data


200. 00200. 00225. 00225. 00250. 00275. 00

43. 0044.0060. 0050. 0066. 0066. 00

Computed L.E.

49. 0349. 0355.5955.5961.7967. 44

Differential

-6. 03-5. 034.41

-5. 594.21

- 1. 44


E iC-21



Coefficients of Curve IA = 50. B = 50. C = 88.03 TO = 80.81 D = 0.OOE+00


Plant: VOGTLE I Material: SA533B1 Heat: C0623-1Orientation: 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


Charpy V-Notch Data


-40.00-40.00

. 00

. 00

. 0040. 0040. 0040. 0080. 00

9.0014.0020.0014.0025.0018.0036. 0040. 0030. 00


6. 046. 04

13.7513.7513.7528. 3528. 3528.3549.54

Differential

2.967.966. 256 2 c

.2 511. 25

- 10.357. 65

11. 65- 19.54

.i C-22


Page 2Plant: VOGTLE 1 Material: SA533B1 H

Orientation: TL Capsule: UNIRR Fluence:reat: C0623-1

n/cmA2

Charpy V-Notch Data

Temperature

80. 0080. 00

120.00120. 00120. 00140. 00140. 00180. 00180. 00180.00240. 00240. 00320. 00320. 00

Input Percent Shear

41. 0048.0065. 0070. 0070. 0090. 0080. 00

1 0 0. 0 01 00. 001 0 0. 001 00. 001 0 0. 001 00. 00100.00


49.5449.5470.9070. 9070. 9079. 3379. 3390. 5090. 5090. 5097. 3897. 3899. 5799. 57

Differential

- 8. 54-1. 54-5.90

-. 90-. 90

10. 67.67

9.509. 509. 502.622.62

.43

.43


C-23



Coefficients of Curve 2A = 50. B 50. C= 77.11 TO = 145.71 D =O.OOE+OO


Plant: VOGTLE I Material: SA533B I Heat: C0623-1Orientation: TL Capsule: X Fluence: nlcmA2

125

100

1-

s0

CO*-a

a)00)

75

50

D '.

III

II

IIIjb

II011IIJ- I

III

III AIIa

0a0I IC;1API1- --- 9

I I I

25

0-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data

Temperature

-50. 00- 25. 00

25. 0050.0075. 0000. 00

125.00150. 00175.00

Input Percent Shear

2.002.005.00

10.0020. 0025. 0035.0050.0065. 00


. 621.184. 197.71

13.7823.4036. 8852.7868. 13

Differential

1.38.82.81

2.296.221. 60

- 1.88-2.78-3. 13

' .C-24

a-


Plant: VOGTLE 1Orientation: TL

Page 2Material: SA533B1 Heat: C0623-1Capsule: X Fluence: n/cmA2

Charpy V-Notch Data


200. 00200. 00225. 00225. 00250. 00275. 00

75. 0075.00

1 00. 0095. 00

1 00. 0 01 00. 00


80. 3480. 3488. 6688. 6693. 7396. 62

Differential

-5. 34-5. 3411. 34

6. 346.273. 38


C-25

UNIRRADIATED WELD METAL

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:54 AMPage 1

Coefficients of Curve I

A = 73.6 B = 71.4 C = 73.31 TO = -5.18 D = O.OOE+00Equation 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:83653Orientation: NA Capsule: UNIRR Fluence: nlcmA2

300

250

, 2004-00IL

El 150a)a,ULiz> 100

50

0 o-300 -200 -100 0 100 200 300 400 500


Charpy V-Notch Data

600


- 120. 00- 120. 00

- 80. 00- 80. 00- 80. 00- 60. 00- 60.00- 60.00-40. 00

7. 005.00

12.0015. 0016.0010. 0011. 0024. 0058.00

Computed CVN

8. 178. 17

18.6118.6118.6128.3428. 3428. 3442. 02

Differential

-1. 17- 3.17- 6. 61-3.61-2.61

- 18. 34- 17. 34-4.3415.98

. -C-26


Page 2Plant: VOGTLE I Material: SAW Heat: WIRE:83653

Orientation: NA Capsule: UNIRR Fluence: n/cmA2

Charpy V-Notch Data


40. 0040. 00

-20. 00-20. 00-20. 00

0000

.0040. 0040. 0040. 0060. 0060. 0060. 0080. 0080.0080. 00

120. 00120.00120.00180. 00180. 00180. 00240. 00240. 00240. 00320. 00320. 00320. 00

22. 0035.0094. 0076. 0086. 0068.0095. 0060. 00

101.00101.00100. 00130. 00118. 00123. 00128. 00126. 00153. 00141.00140. 00135.00144. 00158. 00154. 00144. 00135. 00145. 00144.00143.00143. 00

Computed CVN

42.0242.0259. 3659. 3659. 3678. 6478. 6478.64

112.77112.77112. 77124. 36124. 36124.36132. 27132.27132.27140.46140.46140. 46144.09144. 09144. 09144. 82144. 82144. 82144.98144. 98144. 98

-20. 02-7. 0234. 6416.6426. 64I10. 6416. 3618.64

-11. 77-11. 77- 12.77

5.64- 6.36- 1.36-4.27-6. 2720. 73

.54-. 46

-5. 46-. 09

13.919. 91- . 82

- 9. 82.18

- 1.98- 1. 98

Differential


C-27

CAPSULE X WELD METAL

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed.on 03/23/2004 10:55 AMPage 1

Coefficients of Curve 2A = 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 FPlant: VOGTLE I Material: SAW Heat: WIRE:83653

Orientation: NA Capsule: X Fluence: n/cmA2

300

250

W

-. 200100

LL

El 1500

z> 100C.

;0E 113

------ ---------- l- ---------- l--1 ;

50

0-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data


- 100. 00-50. 00

.00I0. 0025. 0050. 0075. 00

100. 00125. 00

3.006.00

22.0042. 0085.00

1 01. 00127.00136.00112.00

Computed CVN

2. 676.49

34.0346.2668.28

104.23126. 35135. 84139. 27

Differential

.33-. 49

- 12.03-4. 2616.72- 3.23

.65

.16- 27. 27

C-28_

it-


Plant: VOGTLE IOrientation: NA

Page 2Material: SAW Heat: WIRE:83653Capsule: X Fluence: nIcmA2

Charpy V-Notch Data

Temperature

125. 00150. 001.50. 00175. 00200. 00225. 00

Input CVN

128. 00127. 00137. 00142. 00142. 00145. 00

Computed CVN

1 39. 27140.43140.43140. 81140. 94140.98

Differential

- 11.27- 13.43

- 3. 431.191. 064.02


,-. C-29



Coefficients of Curve 1A = 44.13 B = 44.13 C = 54.42 TO = -21.21 D = O.OOE+O0


Temp. @L.E. 35 mils=-32.6 Deg FPlant: VOGTLE I Material: SAW Heat: WIRE:83653

Orientation: NA Capsule: UNTRR Fluence: nlcmA2200

150u,

0I-

0

la 100

50

o 4--300 0 300


Charpy V-Notch Data

600


- 120. 00- 12 0. 00

-80. 00- 80. 00-so. 00- 60. 00- 60. 00- 60. 00-40. 00

3.002.007.00

10. 0010.007.007.00

18.0040.00

Computed L.E.

2.282. 289. 129. 129. 12

17. 1017. 1017. 1029.46

Differential

.72-. 28

-2. 12.88.88

- 10. 10-10. 10

.9010.54

C-30

ii-


Page 2Plant: VOGTLE 1 Material: SAW Heat: WIRE:83653


Charpy V-Notch Data


-40. 00-40. 00-20. 00-20. 00-20. 00

.0000

.0040.0040. 0040. 0060. 0060. 0060. 0080. 0080. 0080. 00

120. 00120. 00120. 00180.00180. 00180. 00240. 00240. 00240. 00320. 00320. 00320. 00

16.0026. 0067. 0056. 0052. 0050. 0068. 0045. 0072. 0074. 0072. 0087. 0077. 0086. 0087.0084. 00

107. 0090. 0088. 0089. 0097. 0082. 0091. 0090. 0088. 0088. 0086. 0092. 0077. 00

Computed L.E.

29.4629.4645. 1045. 1045. 1060. 5060.5060. 5079. 8379. 8379. 8384. 0084. 0084. 0086. 1686. 1686. 1687. 7687.7687. 7688.2088.2088.2088.2588.2588. 2588.2588.2588. 25

Differential

-13.46-3. 4621. 9010. 906.90

- 10. 507.50

- 15.50-7. 83-5. 83-7. 833.00

-7. 002. 00

.84-2. 1620. 84

2. 24.24

1. 248. 80

-6. 202. 801.75-. 25-. 25

-2.253.75

I11.25


C-31



Coefficients of Curve 2A = 41.02 B = 41.02 C = 44.72 TO = 33.45 D = O.OOE+00


Temp. @L.E. 35 mils=26.9 Deg FPlant: VOGTLE I Material: SAW Heat: WIRE:83653

Orientation: NA Capsule: X Fluence: n/cm^2200

C

In0

r-

L.

150

100

50

1

0

-300 0 300 600


Charpy V-Notch Data


- 100. 00-50. 00

.0010. 0025. 0050. 0075. 00

1 00. 00125. 00

. 00

. 0015. 0031. 0025. 0054. 0073. 0085. 0076.00

Computed L.E.

.211. 92

15.0121.2833. 3555.5370. 9678.0580. 69

Differential

- .21- 1. 92

- . 019.72

-8.35-1.532.046.95

-4.69

. . C-32



Orientation: NA Capsule: X Fluence: nlcmA2

Charpy V-Notch Data


1 25. 00150. 00150. 00175.00200. 00225. 00

84.0082. 0080. 0083. 0081. 0078.00

Computed L.E.

80. 6981.5981.5981. 8981.9982. 02

Differential

3.31.41

- 1.591.11- . 99

- 4. 02


C-33


CVGRAPH 5.0.2 Hyperbo]ic Tangent Curve Printed on 03/23/2004 10:59 AMPage 1

Coefficients of Curve 1A = 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:83653Orientation: 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


Charpy V-Notch Data


- 120. 00- 120. 00

- 8 0. 00-80. 00- 80. 00- 60. 00- 60. 00- 60. 00-40. 00

.00

.005.005.005.005.005.00

25.0043. 00


4. 444. 44

12.0312. 0312. 0318.9818. 9818.9828. 66

Differential

-4. 44-4. 44-7. 03-7. 03-7. 03.

- 13. 98- 13. 98

6. 0214. 34

C-34




Charpy V-Notch Data

Temperature

-40. 00-40. 00-20. 00- 20. 00- 20. 00

.00

.00

.0040. 0040. 0040. 0060. 0060. 0060. 0080.0080. 0080. 00

120.00120. 00120.00180. 001 80. 00180. 00240. 00240. 00240. 00320. 00320. 00320.00

Input Percent Shear

15.0033. 0065. 0056. 0050. 0048. 0065. 0035. 0075. 0070. 0075. 0085. 0085. 0085. 0080. 0080. 00

100. 00100. 00100. 00100. 00100. 00100. 00100. 001 00. 0 0100. 001 00. 0 0100. 00100. 00100.00


28. 6628. 6640. 7840. 7840. 7854. 1454. 1454. 1477. 6277. 6277. 6285.6185. 6185. 6191.0791.0791.0796.7796. 7796. 7799. 3499. 3499. 3499. 8799. 8799. 8799. 9899. 9899. 98

13.664. 34

24.2215.229.22

-6. 1410. 8619. 14-2. 62-7. 62-2. 62

. 61

. 61

11. 0711. 078.933.233.233. 23

666666.13.13.13020202

Differential


C-35.


CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed~on 03/23/2004 10:59 AMPage 1

Coefficients of Curve 2A 50. B = 50. C = 60.36 TO = 11.87 D = 0.OOE+00


Plant: VOGTLE 1 Material: SAW Heat: WIRE:83653Orientation: NA Capsule: X Fluence: n/cmA2

125

100

I)s

CO1..r-

e)0.

75

50

25

0 !-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data


- 100. 00-50. 00

.0010. 0025. 0050. 0075. 00

100. 00125. 00

2. 0010. 0040. 0045. 0065. 0080. 0090. 0095.0090. 00


2.4011.4040. 2948.4560.7177. 9689.0194. 8897.70

Differential

- .40- 1.40

-. 29- 3.45

4. 292.04

.99

.12-7.70

C-36

;

It


Page 2Plant: VOGTLE I Material: SAW Heat: WERE:83653


Charpy V-Notch Data


125. 00150. 00150. 00175.00200. 00225. 00

90. 0095. 0098. 00

100. 00100. 00100. 00


97.7098. 9898.9899. 5599. 8099.91

Differential

-7. 70-3. 98

- .98.45.20.09


i C-37

UNIRRADIATED HEAT AFFECTED ZONE

CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:23 AMPage I

Coefficients of Curve IA = 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 FPlant: VOGTLE I Material: SAW Heat: B8805-1


250

Ln- 200

a00

50

Im150

z>100

50

0 =

-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data


-180.00-180.00- 140. 00- 140. 00-110. 00-I I 0. 00-110. 00

- 80.00-80. 00

4. 006. 004. 005. 00

17. 0010. 0016.0025. 0029. 00

Computed CVN

5. 955.95

11. 4911.4919. 8719. 8719. 8733.8833. 88

Differential

- 1. 95.05

-7. 49- 6.49-2.87-9. 87- 3. 87- 8. 88- 4. 88

, I C-38

I,


Page 2Plant: VOGTLE 1 Material: SAW Heat: B8805-1


Charpy V-Notch Data


- 80. 00- 60. 00- 60. 00- 60. 00

40. 0040. 0040. 00

-20. 00-20. 00-20. 00

.0000

.0040. 0040. 0040. 0080. 0080. 0080. 00

120. 00120. 00120. 00220. 00220. 00220. 00

27. 0045. 0060. 0042. 0029. 0058.00

I1 9. 0 0112. 0060. 0079. 0076.0094.00

108. 00125. 0095. 00

103. 00147. 00113. 0093. 00

136.00125. 00136. 00140. 00126. 00140. 00

Computed CVN

33. 8846. 7346.7346. 7361.7661.7661.7677.5777. 5777.5792.4892.4892.48

114. 98114.98114. 98127. 00127.00127.00132. 37132. 37132. 37135. 65135. 65135.65

Differential

-6. 881. 73

13.27-4. 73

-32. 76-3.7657. 2434. 43-17. 57

1.43- 16.48

1.5215.5210.0219. 98-11. 9820. 0014. 0034. 00

3. 637. 373. 634. 359. 654. 35


C-39

CAPSULE X HEAT AFFECTED ZONE


Coefficients of Curve 2A = 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 FPlant: VOGTLE 1 Material: SAW Heat: B8805-1


250

W- 200

El 150C)0r-

z> 100

50

50

0-300 -200 -100 0 100 200 300 400 500 600


Charpy V-Notch Data

Temperature

- 150. 00-I100. 00

-75. 00-50. 00-50. 00-25. 0025. 0075. 00

100. 00

Input CVN

3.0014.0023. 0058. 0041. 0059. 006S. 00

109. 00112. 00

Computed CVN

11. 7022. 3230. 5240. 8440. 8452.9478.97

100. 62108.37

-8. 70- 8. 32-7. 5217. 16

.166. 06

- 10. 978. 383. 63

Differential

' C-40




Charpy V-Notch Data


100. 00150. 00200. 00200. 00225. 00225. 00

106. 00102. 0093.00

122. 00172. 00123. 00

Computed CVN

108.37118.27123. 10123. 10124.42124. 42

Differential

- 2. 37- 16. 27- 30. 10

- 1. 1047. 58- 1.42


C-41



Coefficients of Curve IA = 39.56 B = 39.56 C = 70.03 TO = -41.62 D = O.O0E+OO


Temp. @L.E. 35 mils=49.7 Deg FPlant: VOGTLE I Material: SAW Heat: B8805-1


150u,

E'0.

a 100

500

-9

-300 0 300


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. 001. 002. 003. 009. 005.009. 00

15.0018.00

Computed L.E. Differential

1. 491. 494.494.499. 839. 839. 83

19.8119.81

- .49-. 49

-2. 49-1. 49

-. 83-4. 83

- . 83-4. 81- 1. 81

C-42

I -- - - 1 -




Charpy V-Notch Data


- 80. 00-60. 00-60. 00-60. 00

40. 00-40. 00-40. 00

-20. 00-20.00-20. 00

000000

40. 0040. 0040.0080. 0080. 0080. 00

120.00120. 00120. 00220.00220. 00220. 00

19. 0028. 0040. 0029. 0020. 0036. 0066. 0067. 0044.0051. 0049. 0061.0070. 0074. 0064.0064.0081. 0074. 0070. 0079. 0077. 0078.0084. 0081. 0087. 00

Computed L.E.

19. 8129. 4029. 4029. 4040. 4740. 4740. 4751.3951.3951.3960. 6460. 6460. 6472. 1072. 1072. 1076.7376.7376.7378.3478. 3478. 3479.0779. 0779. 07

Differential

- .81-1.40

10. 60-. 40

- 20. 47-4.47

25.5315.61-7.39

-. 3911. 64

.369.361. 90

- 8. 10-8. 10

4. 27-2.73-6. 73

.66

344.931.937. 93


C-43



Coefficients of Curve 2A = 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 FPlant: VOGTLE I Material: SAW Heat: B8805-1


150

C0

100r-

0

50

0

0 ' - -- a' ------'--- --------------

o , ,

-II

-300 0 300 600


Charpy V-Notch Data


- 150. 00- 00. 00

- 7 5. 00- 5 0. 00-50. 00- 25. 00

25. 0075. 00

1 00. 0 0

. 001. 007. 00

28. 001S. 0030. 0043. 0058. 0069. 00

Computed L.E.

2. 387. 13

11.8318. 6918. 6927. 6047. 0360. 1563. 69

Differential

-2. 38- 6. 13- 4. 8 39.31- .692.40

-4. 03- 2. 15

5.31

C-44

I,




Charpy V-Notch Data

Temperature

I 00. 00150.00200. 00200. 00225. 00225. 00

Input L.E.

60.0070. 0062.0072. 007 1. 0067. 00

Computed L.E.

63. 6967.1168.2568.2568.4868.48-

Differential

- 3.692. 89

-6. 253.752.52

- 1. 48


C45



Coefficients of Curve IA = 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: B8805-1Orientation: NA Capsule: UNIRR Fluence: n/cmA2

125

100

0

U)1.Or-

0.a)

QL

75

50

_ .1000

0 0

0

0

00

0

01 0 - - -I -- -.-----

25

0-300 -200 -100 0 100 200 300


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

.005. 005. 005. 00

10. 0025. 00


1.571.574.414.419.299.299.29

18.5218. 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


Page 2Plant: VOGTLE I Material: SAW Heat: B8805-1


Charpy V-Notch Data


- 80.00- 60. 00- 60. 00- 60.00-40.00- 40. 00-40. 00- 20. 00-20.00-20.00

.00

.00.00

40. 0040.0040.0080.0080.0080.00

120.00120. 00120.00220.00220.00220. 00

18.0020. 0034.0020. 0025. 0030.0080. 0080. 0035.0065. 0065. 0060. 0056. 00

100. 0065. 0075. 00

I 1 0. 0090. 0090. 00

100. 00100. 00100. 00100. 00I 1 0. 00100. 00


18.5227. 8827. 8827. 8839. 6839. 6839. 6852. 8252. 8252. 8265.5865.5865.5884. 6584. 6584. 6594. 1194. 1194. 1197. 8897. 8897. 8899. 8599. 8599. 85

Differential

- . 52- 7. 88

6. 12-7. 88

- 14. 68-9. 6840. 3227. 18

- 17. 8212. 18

- . 58-5.589. 58

15. 3519. 65

9. 655. 89

-4. 11-4. 11

2. 122. 122. 12

.15

.15

.15


C-47


CVGRAPH 5.0.2 Hyperbolic Tangent Curve Printed on 03/23/2004 10:30 AMPage I

Coefficients of Curve 2

A = 50. B = 50. C = 82.03 TO = -7.57 D = 0.OOE+00Equation is A + B * [Tanh((T-To)/(C+DT))]

Temperature at 50% Shear = -7.5Plant: 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


Charpy V-Notch Data


- 150. 00- 100. 00

-75. 00-50. 00-50. 00- 25. 00

25.0075. 00

I 00. 00

2. 005. 00

15. 0030. 0025. 0040. 0070. 0090. 0090. 00


3.019.50

16. 1926. 2226. 2239.5368. 8788.2293.23

Differential

-I . 01-4. 50-1. 19

3.78- 1. 22

.471.131.78

-3. 23

C-48.

' t


Page 2Plant: VOGTLE I Material: SAW Heat: B8805-1


Charpy V-Notch Data


1 00. 00150.00200. 00200. 00225. 00225. 00

90. 0095.00

1 00. 001 00. 001 00. 001 00. 00


93. 2397.9099. 3799. 3799. 6699. 66

Differential

- 3. 2 3-2. 90

.63

. 63

.34

.34


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 forcalculating 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 methodfor calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vesselbeltline materials using surveillance capsule data. The methods of Position C.2 can only be applied whentwo 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 usethese surveillance data sets, they must be shown to be credible. In accordance with the discussion ofRegulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance datato 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 Isurveillance data is credible.

EVALUATION:

Criterion 1: Materials in the capsules should be those judged most likely to be controlling withregard 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 thatcopper 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 ofall plate materials in the beltline region. In addition, the intermediate shell plate B8805-3 hadapproximately 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 waschosen for the surveillance program.

The weld material in the Vogtle Unit 1 surveillance program was made of weld wire (Heat #83653), thesame 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 andunirradiated conditions should be small enough to permit the determination of the 30ft-lb temperature and upper shelf energy unambiguously.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permitthe determination of the 30 ft-lb temperature and the upper shelf energy of the Vogtle Unit 1 surveillancematerials 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 ofARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1normally should be less than 280 F for welds and 170F for base metal. Even if thefluence range is large (two or more orders of magnitude), the scatter should notexceed twice those values. Even if the data fail this criterion for use in shiftcalculations, they may be credible for determining decrease in upper shelf energy ifthe upper shelf can be clearly determined, following the definition given in ASTME185-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilizedto determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about thisline 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 RegulatoryGuide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will befollowed. The NRC methods were presented to industry at a meeting held by the NRC on February 12and 13, 1998. At this meeting the NRC presented five cases. Of the five cases Case I ("Surveillancedata available from plant but no other source") most closely represents the situation listed above forVogtle Unit 1 surveillance weld metal and plate materials.

Appendix D

D-3

TABLE D-ICalculation of Chemistry Factors using Vogtle Unit I Surveillance Capsule Data

Material Capsule Capsule fza) FF(b) ARTNDT() | 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 * RTNDT) e ( 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/cm 2, 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 inRegulatory 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 (BaseMaterial Capsue F FF Measured Predicted Metals)

Maeil Cpue(SIlopebest fit) ARTNDT ARTNDT ARTNDT Meas(OF) <280 F (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 YesWelds

(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 IntermediateShell 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, theweld 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 thevessel 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 arepositioned 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 thereactor vessel wall and the specimens experience equivalent operating conditions such that thetemperatures 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 fallwithin the scatter band of the database for that material.

The Vogtle Unit I surveillance program does not contain correlation monitor material. Therefore, thiscriterion 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 and10 CFR 50.61, the Vogtle Unit I surveillance plate is not credible but the weld data is credible.

Appendix D

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