ML031060076

Submittal of Test Results for Withdrawal of Surveillance Capsule X from Reactor Vessel

ML031060076
Person / Time
Site:	Wolf Creek
Issue date:	04/08/2003
From:	Harris K Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 03-0041 WCAP-16028, Rev 0
Download: ML031060076 (168)

Text

- WELF CREEK

'NUCLEAR OPERATING CORPORATION Karl A. (Tony) Harris Manager Regulatory Affairs APR - 8 2003 RA 03-0041 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Submittal of Test Results for the Withdrawal of Surveillance Capsule X from the Reactor Vessel Gentlemen:

Surveillance capsule X was withdrawn from the Wolf Creek Generating Station reactor vessel on April 12, 2002, at the end of Refuel 12. The capsule was withdrawn to determine the vessel integrity after being subjected to neutron radiation exposure equivalent to the peak end-of-life (extended) fluence on the inside surface of the vessel. Capsule X reached the peak vessel surface fluence equivalent to 54 effective full power years (EFPY) after an actual exposure of 13.83 EFPY, since the lead factor for the capsule is 4.3.

Appendix H to 10 CFR 50 requires that a report be submitted to the Nuclear Regulatory Commission for each capsule withdrawn. The report must describe the capsule and the test results for the capsule. The enclosure provides Westinghouse report WCAP-16028 Revision 0 for the analysis of capsule X.

There are no commitments contained in this correspondence. If you have any questions concerning this matter, please contact me at (620) 3644038, or Ms. Jennifer Yunk at (620) 3644272.

Very truly yours, Karl A. (Tony) Harris KAH/rIg Enclosure cc: J. N. Donohew (NRC), w/e -- Y0 0 D. N. Graves (NRC), w/e E. W. Merschoff (NRC), w/e Senior Resident Inspector (NRC), w/e RO Box 411 / Burlington, KS 66839/ Phone. (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET

Westinghouse Non-Proprietary Class 3 WCAP-1 6028 March 2003 Revision 0 Analysis of Capsule X from Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Program I Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16028, Revision 0 Analysis of Capsule X from the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Program T.J. Laubham J. Conermann R.J. HagIer March 2003 Approved:' Lk )

J.A. Gresham, Manager Engineering & Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

©2003 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST OF TABLES .................. iv LIST OF FIGURES .................. vi PREFACE.. ................. viii EXECUTIVE

SUMMARY

.................. ix 1

SUMMARY

OF RESULTS .1-1 2 INTRODUCTION .2-1 3 BACKGROUND .3-1 4 DESCRIPTION OF PROGRAM .4-1 5 TESTING OF SPECIMENS FROM CAPSULE X .. 5-1 5.1 OVERVIEW .5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS. . . . 5-3 5.3 TENSILE TEST RESULTS................................................ .................... .............. 5-5 5.4 1/2T COMPACT TENSION AND BEND BAR SPECIMEN TESTS .5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY . .6-1

6.1 INTRODUCTION

.6-1 6.2 DISCRETE ORDINATES ANALYSIS .6-2 6.3 NEUTRON DOSIMETRY .6-5 6.4 CALCULATIONAL UNCERTAINTIES .6-6 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .7-1 8 REFERENCES .8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS CREDIBILITY .......................... A-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS ........................... B-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD ................................... C-0 APPENDIX D WOLF CREEK SURVEILLANCE PROGRAM CREDIBILITY EVALUATION.. D-0

iv LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the Wolf Creek Reactor Vessel Surveillance Materials (Unirradiated) .............................................................. 4-3 Table 4-2 Heat Treatment History of the Wolf Creek Reactor Vessel Surveillance Materials ....... 4-4 Table 4-3 Chemical Composition (wt%) of four Charpy Specimens from Wolf Creek Capsule X 4-5 Table 4-4 Chemical Results from Low Alloy Steel NIST Certified Reference Standards (wt%) ... 4-6 Table 4-5 Chemical Results from Low Alloy Steel NIST Certified Reference Standards (wt%) ... 4-7 Table 5-1 Charpy V-Notch Data for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x 10'9 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation) .5-6 Table 5-2 Charpy V-Notch Data for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x 10'9 n/cm 2 (E > 1.0 MeV) (Transverse Orientation) .5-7 Table 5-3 Charpy V-notch Data for the Wolf Creek Surveillance Weld Material Irradiated to a Fluence of 3.49 x 1019 n/cm 2 (E> 1.0 MeV) .5-8 Table 5-4 Charpy V-notch Data for the Wolf Creek Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of 3.49 x 10'9 n/cm 2 (E> 1.0 MeV) .5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x 10'9 n/cm 2 (E> 1.0 MeV)

(Longitudinal Orientation) . 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x I 0'9 n/cm 2 (E> 1.0 MeV)

(Transverse Orientation) . 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Wolf Creek Surveillance Weld Metal Irradiated to a Fluence of 3.49 x 1019 n/cm2 (E> 1.0 MeV) . 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Wolf Creek Heat-Affected-Zone (HAZ) Irradiated to a Fluence of 3.49 x 1019 n/cm2 (E> 1.0MeV) . 5-13 Table 5-9 Effect of Irradiation to 3.49 x 1019 n/cm 2 (E> 1.0 MeV) on the Notch Toughness Properties of the Wolf Creek Reactor Vessel Surveillance Materials . 5-14 Table 5-10 Comparison of the Wolf Creek Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions . 5-15

V LIST OF TABLES (Cont.)

Table 5-11 Tensile Properties of the Wolf Creek Capsule XReactor Vessel Surveillance Materials Irradiated to 3.49 x 10'9 n/cm 2 (E> 1.0MeV) ..................................... ..................... 5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center . 6-12 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface . 6-16 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall . 6-20 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) Within The Reactor Vessel Wall . 6-20 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Wolf Creek ....................................... 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 Wolf Creek Reactor Vessel ..................... 4-8 Figure 4-2 Capsule X Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters ........... 4-9 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) . 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) . 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) . 5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) . 5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) . 5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) . 5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Weld Metal . 5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Weld Metal . 5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Weld Metal . 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Heat-Affected-Zone Material . 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Heat-Affected-Zone Material . 5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Heat-Affected-Zone Material . 5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) . 5-29

vii LIST OF FIGURES (Cont.)

Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) ...................................................

5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Weld Metal

...... 5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Heat-Affected-Zone Metal .................

5-32 Figure 5-17 Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) ..................

5-33 Figure 5-18 Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation).

5-34 Figure 5-19 Tensile Properties for Wolf Creek Reactor Vessel Weld Metal .5-35 Figure 5-20 Fractured Tensile Specimens from Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation) .......................................

5-36 Figure 5-21 Fractured Tensile Specimens from Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation) .5-37 Figure 5-22 Fractured Tensile Specimens from Wolf Creek Reactor Vessel Weld Metal .5-38 Figure 5-23 Engineering Stress-Strain Curves for Lower Shell Plate R2508-3 Tensile Specimens AL-10, AL-I l and AL-12 (Longitudinal Orientation) ...................

5-39 Figure 5-24 Engineering Stress-Strain Curves for Lower Shell Plate R2508-3 Tensile Specimens AT-10, AT-Il and AT-12 (Transverse Orientation) .........................

5-40 Figure 5-25 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens AW-10, AW-Il and AW-12 .5-41 Figure 6-1 Wolf Creek rO Reactor Geometry at the Core Midplane .6-8 Figure 6-2 Wolf Creek rz Reactor Geometry .6-11

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

Reviewer:

Sections I through 5, 7, 8, Appendices B, C and D A.R. Rawluszki Section 6 and Appendix A D. M. Chapman

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule X from Wolf Creek Capsule X was removed at 13.8 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI data-base. Capsule X received a fluence of 3.49 x 1019 n/cm2 after irradiation to 13.8 EFPY. The peak clad/base metal interface vessel fluence after 13.8 EFPY of plant operation was 8.1 x 10 n/cm 2 .

This evaluation lead to the following conclusions: 1)The measured 30 ft-lb shift in transition temperature values of the lower shell plate R2508-3 contained in capsule X (longitudinal & transverse) is less than the Regulatory Guide 1.99, Revision 2 13], predictions. 2) The measured 30 ft-lb shift in transition temperature values of the weld metal contained in capsule X is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2. 3) The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Wolf Creek surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions. 4) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the extended life of the vessel (54 EFPY) as required by IOCFR50, Appendix G 12]. 5) The Wolf Creek surveillance data was found to be credible. This evaluation can be found in Appendix D.

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

1-1 1

SUMMARY

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

The Charpy V-notch data presented in WCAP-15078, Rev. 1[3] were based on a re-plot of all capsule data from WCAP-1001514 1,WCAP-11553Esl and WCAP-13365, Rev. 1[6J using CVGRAPH, Version 4.1, which is a symmetric hyperbolic tangent curve-fitting program. The results presented are only for the Capsule X test results, which are also based on using CVGRAPH, Version 4.1. This report also shows the composite plots that show the results from the previous capsule. Appendix C presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data

• Capsule X received an average fast neutron fluence (E> 1.0 MeV) of 3.49 x 10'9 n/cm2 after 13.8 effective full power years (EFPY) of plant operation.

• Irradiation of the reactor vessel lower shell plate R2508-3 (heat number C4935-2) Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 36.11 0 F and an irradiated 50 ft-lb transition temperature of 67.750 F. This results in a 30 ft-lb transition temperature increase of 61.061F and a 50 ft-lb transition temperature increase of 67.640 F for the longitudinal oriented specimens. See Table 5-9.

• Irradiation of the reactor vessel lower shell plate R2508-3 (heat number C4935-2) Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 55.970 F and an irradiated 50 ft-lb transition temperature of 97.047F. This results in a 30 ft-lb transition temperature increase of 53.961F and a 50 ft-lb transition temperature increase of 62.71 °F for the longitudinal oriented specimens. See Table 5-9.

Irradiation of the weld metal (heat number 90146) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 10.66°F and an irradiated 50 ft-lb transition temperature of 54.73°F. This results in a 30 ft-lb transition temperature increase of 68.36°F and a 50 ft-lb transition temperature increase of75.38°F. See Table 5-9.

• Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-74.34°F and an irradiated 50 ft-lb transition temperature of -52.92°F. This results in a 30 ft-lb transition temperature increase of 69.66°F and a 50 ft-lb transition temperature increase of 61.08°F. See Table 5-9.

• The average upper shelf energy of the lower shell plate R2508-3 (longitudinal orientation) resulted in an average energy decrease of 6 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 142 ft-lb for the longitudinal oriented specimens. See Table 5-9.

Summary of Results

1-2

• The average upper shelf energy of the lower shell plate R2508-3 (transverse orientation) resulted in no energy decrease after irradiation. This results m an irradiated average upper shelf energy of 94 ft-lb for the longitudinal oriented specimens. See Table 5-9.

• The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 7 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 93 ft-lb for the weld metal specimens. See Table 5-9.

• The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 26 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 135 ft-lb for the weld HAZ metal. See Table 5-9.

• A comparison, as presented in Table 5-10, of the Wolf Creek reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2Y1] predictions led to the following conclusions:

- The measured 30 ft-lb shift in transition temperature values of the lower shell plate R2508-3 contained in capsule X (longitudinal & transverse) are less than the Regulatory Guide 1.99, Revision 2, predictions.

- The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Wolf Creek surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions.

• All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the extended life of the vessel (54 EFPY) as required by 10CFR50, Appendix G 1

• The calculated end-of-license (54 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the Wolf Creek reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i e., Equation #3 in the guide) are as follows:

Calculated: Vessel inner radius* = 3.51 x 1019 n/cm2 Vessel 1/4 thickness = 2.09 x 10'9n/cm 2 Vessel 3/4 thickness = 7.42 x 1018 n/cm 2

• Clad/base metal interface. (From Table 6-2)

Summary of Results

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule X, the fourth capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-10015, "Kansas Gas and Electric Company Wolf Creek Generating Station Unit No. I Reactor Vessel Radiation Surveillance Program" 14 1. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-79, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." Capsule X was removed from the reactor after 13.8 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing of and the post-irradiation data obtained from surveillance capsule X

removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor vessel and discusses the analysis of the data.

Introduction

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

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

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208171) or the temperature 60IF less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (Kic curve) which appears in Appendix G to the ASME Code!Sl. The KI, curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KI, curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Wolf Creek reactor vessel radiation surveillance programs, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RT=T initial + M + ARTNDT) is used to index the material to the KIc curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

Background

4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Wolf Creek reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core Capsule X was removed after 13.8 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2T-CT fracture mechanics specimens made from lower shell plate R2508-3 (heat number C4935-2) and submerged arc weld metal representative of all the reactor vessel beltline region weld seams. In addition, this capsule contained Charpy V-notch specimens from the weld Heat-Affected-Zone (HAZ) metal of plate R2508-3.

Test material obtained from the Lower Shell Plate (after thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the 1/4 thickness location of the plate after performing a simulated post-weld stress-relieved weldment joining lower shell plate R2508-1 and adjacent lower shell plate R2508-3. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of the lower shell plate R2508-3 Charpy V-notch impact specimens from lower shell plate R2508-3 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major working direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major working direction).

The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular to the weld direction. The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction Tensile specimens from lower shell plate R2508-3 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented with the long dimension of the specimen perpendicular to the weld direction Compact tension test specimens from lower shell plate R2508-3 were machined in the longitudinal and transverse orientations. Compact tension test specimens from the weld metal were machined perpendicular to the weld direction with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399.

The chemical composition and heat treatment of the unirradiated surveillance materials are presented in Tables 4-1 and 4-2, respectively. The data in Table 4-1 and 4-2 was obtained from the unirradiated surveillance program report, WCAP-10015, Appendix A. Contained m Table 4-3 are the results of the chemical analysis performed on four Charpy specimens from Capsule X The results of the NBS certified standards are presented in Tables 4-4 and 4-5.

Description of Program

4-2 Capsule X contained dosimeter wires of pure iron, copper, nickel, and aluminum 0.15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np2') and uranium (U238) were placed m the capsule to measure the integrated flux at specific neutron energy levels.

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

2.5% Ag, 97.5% Pb Melting Point: 579TF (3041C) 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590'F (310 0 C)

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

Description of Program

4-3 Table 4-1 Chemical Composition (wt%) of the Wolf Creek Reactor Vessel Surveillance Materials (Unirradiated)(Y)

Element Lower Shell Plate Weld Metal (b)

R2508-3 C 0.20 0.11 Mn 1.45 1.46 P 0008 0 005 S 0 010 0.011 Si 0.20 048 Ni 0.62 0 09 Mo 0.55 0.56 Cr 0.05 0.09 Cu 0.07 0.04 Al 0.032 0 009 Co 0.014 0.010 Pb <0 001 <0.001 W <0.01 <0.01 Ti <0.01 <0.01 Zr <0.001 <0.001 V 0 003 0.005 Sn 0.002 0 003 As 0.007 0.004 Cb <0.01 <0.01 N2 0.007 0 006 B <0 001 <0.001 Notes:

(a) Data obtained from WCAP-10015 and duplicated herein for completeness.

(b) Weld wire Type B4, Heat Number 90146, Flux Type Linde 124, and Flux Lot Number 1061.

Surveillance weldment is from a weld between the lower shell plates R2508-3 and R2508- 1 and is identical to the intermediate to lower shell circumferential weld seam. In addition, this weld is made of the same weld wire heat as the longitudinal weld seams.

Description of Program

4-4 Table 4-2 Heat Treatment History of the Wolf Creek Reactor Vessel Surveillance Maierial cm raur - ea  ! Poit~

Lower Shell Plate Austemutized @ 4 hrs. Water-Quench 1600 +/- 25 R2508-3 Tempered @ 4 hrs Air-cooled 1225 +/-25 Stress Relieved(b) @ 8 hrs. 30 min. Furnace Cooled 1150 +/-50 Weld Metal (heat # 90146) Stress Relieved(b) @ 10 hrs. 15 min. Furnace Cooled 1150 +/-50 x1 1NU1rs:

(a) This table was taken from WCAP-10015t 4 ].

(b) The stress relief heat treatment received by the surveillance test plate and weldment have been simulated.

Description of Program

4-5 Table 4-3 Chemical Composition ( ) of four Charpy Specimens from Wolf Creek Capsule X Concentration in Weight Percent Weld Metal Specimens -BaseMetal

,........... : , , ,pecimen

.. S ElAW-59 -55 AW-54 AT-54 Al 0.016 0.008 0.008 0.015 Co 0.0144 0.01 0.01 0.0144 Cr 0.0681 0.116 0.112 0.0687 Cu 0.0511 0.05 0.0411 0.0747 Fe 95.0 96.8 94.5 95.6 Mn 1.34 1.45 1.41 1.34 Mo 0.502 0.545 0.527 0.511 Ni 0.589 0.112 0.108 0.591 P 0.0152 0.017 0.0142 0.0145 Si 0.189 0.326 0.310 0.179 Sn 0.004 0.004 0.003 0.003 Ti 0.006 0.006 0.004 0.004 V 0.008 0.008 0.007 0.006 Zr <0.00001 <0.00001 <0.00001 <0.00001 C 0.12 0.12 0.12 0.22 S 0.013 0.013 0.013 0.013 Description of Program

4-6 Tble44 4 C' Results fom L Sti NIST Certified Reference7

~wAlly

,,,3i',><,, ,$ 1.0 MeV) in 13.8 EFPY of operation, are presented in Tables 5-1 through 5-11 and are compared with unirradiated results 4' as shown in Figures 5-1 through 5-12.

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

Irradiation of the reactor vessel lower shell plate R2508-3 (heat number C4935-2) Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 36.1 10F and an irradiated 50 ft-lb transition temperature of 67.750 F. This results in a 30 ft-lb transition temperature increase of 61.06'F and a 50 ft-lb transition temperature increase of 67.640 F for the longitudinal oriented specimens.

Irradiation of the reactor vessel lower shell plate R2508-3 (heat number C4935-2) Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 55.970 F and an irradiated 50 ft-lb transition temperature of 97.04'F. This results in a 30 ft-lb transition temperature increase of 53.960 F and a 50 ft-lb transition temperature increase of 62.7 10F for the longitudinal oriented specimens.

Irradiation of the weld metal (heat number 90146) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 10.660 F and an irradiated 50 ft-lb transition temperature of 54.730 F.

This results in a 30 ft-lb transition temperature increase of 68.360 F and a 50 ft-lb transition temperature increase of 75.38 0 F Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-74.340 F and an irradiated 50 ft-lb transition temperature of -52.920 F. This results in a 30 ft-lb transition temperature increase of 69.66 0 F and a 50 ft-lb transition temperature increase of 61.080 F.

The average upper shelf energy of the lower shell plate R2508-3 (longitudinal orientation) resulted in an average energy decrease of 6 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 142 ft-lb for the longitudinal oriented specimens.

The average upper shelf energy of the lower shell plate R2508-3 (transverse orientation) resulted in no energy decrease after irradiation. This results in an irradiated average upper shelf energy of 94 ft-lb for the longitudinal oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 7 ft-lb after irradiation This results in an irradiated average upper shelf energy of 93 ft-lb for the weld metal specimens.

Testing of Specimens from Capsule X

It 5-4 The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average decrease of 26 ft-lb after irradiation. This results energy in an irradiated average upper shelf energy of for the weld HAZ metal. 135 ft-lb A comparison, as presented in Table 5-10, of the Wolf Creek reactor vessel surveillance material results with the Regulatory Guide 1.99, Revision test 21'1 predictions led to the following conclusions:

- The measured 30 ft-lb shift in transition temperature values of the lower shell plate R2508-3 contained in capsule X (longitudinal & transverse) are less than the Regulatory Guide 1.99, Revision 2, predictions.

- The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule X is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

- The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules X contained in the Wolf Creek surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout extended life of the vessel (54 EFPY) as required the by IOCFR50, Appendix G t2k The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule materials is shown in Figures 5-13 through X 5-16 and shows an increasingly ductile or tougher with increasing test temperature. appearance The load-time records for individual instrumented Charpy specimen tests are shown in Appendix B.

The Charpy V-notch data presented in WCAP-15078, Rev. I "' were based on a re-plot of all capsule from WCAP-1001514 1 , WCAP-1 1553"5' and WCAP-13365, data Rev. 1[61 using CVGRAPH, Version 4.1, which is a symmetric hyperbolic tangent curve-fitting program. The results presented are only for the X test results, which are also based on using Capsule CVGRAPH, Version 4.1. This report also shows composite plots that show the results from the the previous capsule. Appendix C presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data.

Testing of Specimens from Capsule X

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule X irradiated to 3.49 x 1O'9 n/cm2 (E> 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated resultsE41 as shown in Figures 5-17 and 5-19.

The results of the tensile tests performed on the lower Shell Plate R2508-3 (longitudinal orientation) indicated that irradiation to 3.49 x 109 n/cm2 (E> 1.0 MeV) caused approximately a 9 to 11 ksi increase in the 0.2 percent offset yield strength and approximately a 8 ksi increase in the ultimate tensile strength when compared to unirradiated dataE4 3. See Figure 5-17.

The results of the tensile tests performed on the lower Shell Plate R2508-3 (Transverse orientation) indicated that irradiation to 3.49 x 1O'9 n/cm 2 (E> 1.0 MeV) caused approximately a 7 to 11 ksi increase in the 0.2 percent offset yield strength and approximately a 9 to 10 ksi increase in the ultimate tensile strength when compared to unirradiated dataE43. See Figure 5-18.

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 3.49 x 1019 n/cm 2 (E> 1.0 MeV) caused approximately a 2 to 9 ksi increase in the 0.2 percent offset yield strength and approximately a 1 to 7 ksi increase in the ultimate tensile strength when compared to unirradiated data141. See Figure 5-19.

The fractured tensile specimens for the Lower Shell Plate R2508-3 material are shown in Figures 5-20 and 5-21, while the fractured tensile specimens for the surveillance weld metal are shown in Figure 5-22. The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-25.

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

Testing of Specimens from Capsule X

II 5-6 Table 5-1 Charpy V-notch Data for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x lO' 9 n/cm 2 (E> 1.0 MeV) (Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF 0C ft-lbs Joules mils nun  %

AL48 -50 -46 2 3 0 0.00 2 AL57 0 -18 13 18 5 0.13 5 AL53 25 -4 21 28 12 0.30 10 AL52 40 4 37 50 24 0.61 15 AL56 50 10 53 72 33 0.84 20 AL55 75 24 43 58 29 0.74 30 AL59 110 43 74 100 47 1.19 50 AL50 135 57 108 146 67 1.70 65 AL60 150 66 133 180 74 1.88 90 AL49 175 79 100 136 64 1.63 75 AL51 190 88 122 165 67 1.70 85 AL46 225 107 150 203 71 1.80 100 AL54 250 121 146 198 75 1.91 100 AL58 275 135 135 183 75 1.91 100 AL47 300 149 137 186 75 1.91 100 Testing of Specimens from Capsule X

5-7 Table 5-2 Charpy V-notch Data for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x 10' 9 n/cm 2 (E> 1.0 MeV) (Transverse Orientation)

Sample Temperature J Impact Energy Lateral Expansion Shear Number OF J 0 C ft-lbs joules mils  %

AT46 -75 -59 5 7 0 0.00 2 AT50 -25 -32 11 15 4 0.10 5 AT60 15 -9 15 20 8 0.20 15 AT56 50 10 30 41 20 0.51 25 AT54 75 24 41 56 29 0.74 40 AT53 100 38 52 71 36 0.91 45 AT59 125 52 55 75 38 0.97 55 AT58 150 66 67 91 50 1.27 65 AT48 175 79 98 133 57 1.45 90 AT51 175 79 79 107 53 1.35 80 AT57 200 93 88 119 51 1.30 95 AT52 225 107 91 123 69 1.75 100 AT49 250 121 93 126 60 1.52 100 AT47 275 135 101 137 66 1.68 100 AT55 300 149 96 130 61 1.55 100 Testing of Specimens from Capsule X

II 5-8 Table 5-3 Charpy V-notch Data for the Wolf Creek Surveillance Weld Metal Irradiated to a Fluence of 3.49 x 1019 n/cm 2 (E> 1.0 MeV)

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

AW47 -75 -59 4 5 0 0.00 10 AW52 -35 -37 13 18 5 0.13 20 AW53 0 -18 23 31 15 0.38 45 AW51 25 -4 37 50 25 0.64 50 AW58 50 10 53 72 35 0.89 60 AW46 75 24 64 87 45 1.14 80 AW59 100 38 68 92 51 1.30 85 AW57 125 52 76 103 55 1.40 95 AW55 125 52 78 106 53 1.35 95 AW48 150 66 67 91 54 1.37 85 AW60 160 71 89 121 62 1.57 95 AW56 200 93 84 114 54 1.37 99 AW50 225 107 102 138 64 1.63 100 AW49 250 121 94 127 69 1.75 100 AW54 250 121 96 130 64 1.63 100 Testing of Specimens from Capsule X

5-9 Table 5-4 Charpy V-notch Data for the Wolf Creek Heat-Affected-Zone (HAZ) Material Irradiated to a Fluence of 3.49 x 1019 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF 0C Ft-lbs Joules mils mm  %

AH53 -175 -115 6 8 0 0.00 0 AH51 -100 -73 15 20 5 0.13 2 AH50 -75 -59 34 46 12 0.30 5 AH49 -50 -46 88 119 42 1.07 30 AH58 -50 -46 29 39 12 0.30 15 AH52 -25 -32 50 68 21 0.53 20 AH47 0 -18 152 206 70 1.78 100 AH55 0 -18 92 125 44 1.12 50 AH59 25 4 98 133 52 1.32 65 AH48 35 2 132 179 70 1.78 75 AH54 50 10 140 190 70 1.78 100 AH60 100 38 146 198 70 1.78 100 AH46 150 66 140 190 71 1.80 100 AH57200 93 129 175 74 1.88 100 AI-156 200 93 124 168 72 1.83 100 Testing of Specimens from Capsule X

5-10 Table 5,-5 Instrumented Charpy Impact Test Results for the Wolf Creek Lower Shell PlateR2S08^3 Irradiated to a Fluence of 3.49 x 1O'9 n/cm2 (E>1t0 MeY). (Longitudinal Orientation) __,',__

Churp Normalized Energie Enry(lt/n)Yield' 'im n Test .Load FastY Timeto MaxMaL "-,Max., Fracto< ?(Arrest Sample Temp. ,; Charpy Max. Yield,,:, Flow,,

Prop,, PPY 3Yield tGo' 3 P M , IM LoadPr 0o. (F) Lo d P Stress,; Stress E/kA, Qb' b' (msec) (Ib) (mssec)', (Ib , "', .ksi) k,s(Ib)

AL48 -50 2 16 8 8 1068 0.10 1068 0.10 1068 0 36 36 AL57 0 13 105 54 51 3868 0.20 3868 0.20 3868 0 129 129 AL53 25 21 169 58 111 3079 0.14 3810 0.21 3783 41 103 115 AL52 40 37 298 223 75 3167 0.14 4231 0.53 4212 183 105 123 AL56 50 53 427 297 130 3078 0.14 4212 0.67 4058 0 102 121 AL55 75 43 346 205 141 3110 0.14 4153 0.50 4141 854 104 121 AL59 110 74 596 292 305 2952 0.14 4040 0.68 3803 1752 98 116 AL50 135 108 870 366 505 3045 0.15 4137 0.83 3240 1054 101 120 AL60 150 133 1072 292 779 2934 0.14 4116 0.69 1461 689 98 117 AL49 175 100 806 287 519 2869 0.14 4066 0.68 3786 1041 96 115 AL51 190 122 983 293 690 2885 0.14 4069 0.70 2687 1779 96 116 AL46 225 150 1209 353 856 2825 0.14 4059 0.82 N/A N/A 94 115 AL54 250 146 1176 349 828 2788 0.14 4007 0.82 N/A N/A 93 113 AL58 275 135 1088 279 809 2631 0.15 3911 0.70 N/A N/A 88 109 AL47 300 137 1104 276 828 2714 0.14 3888 0.69 N/A N/A 90 110 Testing of Specimens from Capsule X

5-11

-.Table 5-6 Instrumented Charpy Impact Test Results for the Wolf Creek Lower Shell Plate R2508-3 Irradiated to a Fluence of 3.49 x 109 nfcm2 (E>1.0 MeV) (Transverse Orientation) >___>__

chiarpy No tlized EftergiesYedTm t as.

Tet <Energy (ft-lb/in)t Fract. Arrest o samp T5eNmp. >- Charpy. Mas> ProP> Py Yield tGy Load PM load Pi Load PA,- Stres. Stress No  : (, ) (f-Ib) E.. EM/A E5 /A (lb> (msec) (Ib >.yec):>.> 5 . (Ii,>>.> >> (ksi) (ksi)

AT46 -75 5 40 19 21 2391 0.13 2391 0.13 2391 0 80 80 AT50 -25 11 89 49 40 3839 0.19 3839 0.19 3839 0 128 128 AT60 15 15 121 58 63 3814 0.21 3814 0.21 3814 94 127 127 AT56 50 30 242 151 91 2922 0.14 3909 0.40 3877 263 97 114 ATS4 75 41 330 159 171 3086 0.14 4001 0.41 3991 1021 103 118 AT53 100 52 419 209 209 2979 0.14 4066 0.52 4035 1530 99 117 AT59 125 55 443 200 243 2888 0.14 3873 0.51 3823 1335 96 113 AT58 150 67 540 197 343 2861 0.14 3864 0.51 3499 1191 95 112 AT48 175 98 790 288 501 2740 0.15 4089 0.70 3353 2066 91 114 AT51 175 79 637 273 363 2890 0.14 3906 0.66 3674 1106 96 113 ATS7 200 88 709 276 433 2833 0.14 3947 0.67 3552 2362 94 113 AT52 225 91 733 268 465 2718 0.14 3865 0.67 N/A N/A 90 110 AT49 250 93 749 264 485 2814 0.15 3948 0.66 N/A N/A 94 113 AT47 275 101 814 287 527 2776 0.15 3999 0.70 N/A N/A 92 113 ATSS 300 96 774 258 515 2717 0.14 3842 0.65 N/A N/A 90 109 Testing of Specimens from Capsule X

5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Wolf Creek Surveillanse Weld Metal Irradiated to a Fluence of 3 49 x IO1 ni/crn (E>Io. MeY)

CharpY.. Normalized Energies Energy:ftbi) ed Time to I Fast p

l'Test Te Charpy Max." Prp lTim to M M I jLoadPF A YLiad Yeld Flow Yield t y PGYrop. Load PM t L PA Stress cry Stress AW47 -75 4 32 14 18 1800 0.12 1800 0.12 1800 0 60 60 AW52 -35 13 105 42 63 3814 0.17 3814 0.17 3814 416 127 127 AW53 0 23 185 33 152 3201 0.16 3201 0.16 3201 1941 107 107 AW51 25 37 298 137 161 3104 0.14 4062 0.36 4000 1984 103 119 AW58 50 53 427 212 215 3286 0.15 4242 0.50 4177 1597 109 125 AW46 75 64 516 219 297 3344 0.14 4316 0.50 4188 2141 I11 128 AW59 100 68 548 216 332 3151 0.14 4121 0.52 3736 2053 105 121 AW57 125 76 612 219 394 3070 0.14 4135 0.52 3812 2294 102 120 AW55 125 78 628 215 414 3174 0.15 4169 0.52 3486 2501 106 122 AW48 150 67 540 199 341 2934 0.14 3879 0.50 3693 2308 98 113 AW60 160 89 717 281 436 3027 0.14 4081 0.65 3389 2313 101 118 AW56 200 84 677 207 470 3000 0.14 4012 0.52 2679 2204 100 117 AW50 225 102 822 299 523 3098 0.15 4181 0.67 N/A N/A 103 121 AW49 250 94 757 276 482 2941 0.14 3980 0.66 N/A N/A 98 115 AW54 250 96 774 292 482 3009 0.17 4034 0.70 N/A N/A 100 117 Testing of Specimen' (frnm r-incu, v-

-- , -- -1. -- --- -

5-13

[Table 5-8 Instrumented Charpy, Impact Test Results for the Wolf Creek Heat-Affected-Zone (HAZy Metal 1'd;Sh~d f*~ - I,6 .fl is, iii'-,.i n Itxi,,

hP,. Normlized .nr..e.

') sa: emerge~o>>ss3?>Tet  ;  ?'T~ t o fv~ ls Mal.toC, Tii" Vinci i Fa': tl!';t,, Arrest Yield v  ; . Flow,

• Charpy MaL PrPop, GYI I Yield tGV Lad P. M I AS ad a PF LoaI Stress ay Stres (3F) ( llNo.

lb) E3WA EA , jA Qb) , P *(b), . i) *ksi3(k~i)

AH53 -175 6 48 23 25 2931 0.14 2931 0.14 2931 N/A 98 98 AH51 -100 15 121 70 51 4991 0.21 4991 0.21 4991 N/A 166 166 AH50 -75 34 274 224 50 3892 0.15 4782 0.47 4749 N/A 130 144 AH49 -50 88 709 344 365 3874 0.16 4764 0.68 4082 N/A 129 144 AH58 -50 29 234 176 58 3648 0.15 4483 0.40 4452 N/A 121 135 AH52 -25 50 403 329 74 3701 0.15 4675 0.66 4605 N/A 123 139 AH47 0 152 1225 329 895 3583 0.15 4569 0.68 N/A N/A 119 136 AH55 0 92 741 322 419 3510 0.15 4559 0.67 3925 1008 117 134 AH59 25 98 790 321 468 3475 0.15 4495 0.68 4039 1337 116 133 AH48 35 132 1064 326 738 3499 0.15 4499 0.69 2002 404 117 133 AH54 50 140 1128 324 804 3395 0.15 4468 0.70 N/A N/A 113 131 AH60 100 146 1176 317 860 3383 0.15 4430 0.68 N/A N/A 113 130 AH46 150 140 1128 318 810 3203 0.15 4341 0.70 N/A N/A 107 126 AH57 200 129 1039 302 737 3074 0.14 4252 0.69 N/A N/A 102 122 AH56 200 124 999 304 695 3058 0.15 4212 0.70 N/A N/A 102 121 Testing of Specimens from Capsule X

5-14

.Table 5-9> Effect of Irradiation to 3.49 s 19 nicn 2 (E>1.0 MeV) on the Capsule V Notch Toughness Propertiesof the Wolf Creek Reactor Vessel s

,SurvFeillance Mate'r's().

' . " , '"..y ,,,,',

Averag 30Q (t-lbAverage ft-lb) Avcrage 3 mI l er 1° AveragecS50 ft.Ibj ft-lb(, Energ Absorption~a Matenal Transition Temperature (,) Expansion Temperature (0 5) Transition Temperature at Full Shear (ft-Jb)

'5l:5555,5,',

UnrdaeIid 'eATi: Unirradiated Irradiated, AT, I nin-adiated Irradiated, AT Umaiae rradiated E Unirradiated,~ n aed>L urrd rnfad ;l teteM Lower Shell Plate -24.95 36.11 61.06 -0.4 72.95 73.36 0.11 67.75 67.64 148 142 -6 R2508-3 (Long.)

Lower Shell Platc 2.0 55.97 53.97 25.44 102.4 76.95 34.32 97.04 62.71 94 95 +1 R2508-3 (Trans.)

Weld Metal -57.69 10.66 68.36 -27.06 53.49 80.56 -20.64 54.73 75.38 100 93 -7 (Heat # 90146)

HAZMetal -144.01 -74.34 69.66 -89.86 -27.62 62.24 -114 -52.92 61.08 161 135 -26

a. "Average" is defined as the value read from the curve fit through thc 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 5-1 1).

Testing of Specimens from Capsule X

5-15 Table 5-10 Comparison of theiWolf Creek Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 1...... 30 ft-lb Transition TUpper Shelf Energy e,,,._jTemperature Shift' Decrease Material 'Capsule FluencyPredicted 1 Measured -'Predicted Measured:

E 0l.MeV) ____

Lower Shell Plate U 0.316 34.88 36.46 14.5 2 R2508-3 Y 1.19 53.55 16.03 20 11 V 2.22 62.22 52.03 23 13 (Longitudinal) X 3.49 67.83 61.06 25 4 Lower Shell Plate U 0.316 34.88 23.79 14.5 0 R2508-3 Y 1.19 53.55 35.39 20 0 V 2.22 62.22 54.53 23 6 (Transverse) X 3.49 67.83 53.96 25 0 Surveillance U 0.316 33.24 27.21 16 8 Program Y 1.19 51.03 45.09 22 6 Weld Metal V 2.22 59.29 46.33 25 11 X 3.49 64.64 68.36 28 7 Heat Affected Zone U 0.316 --- 58.41 --- 13 Material Y 1.19 --- 12.98 --- 0 V 2.22 --- 55.91 --- 0 X 3.49 --- 69.66 --- 16 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 4.1 (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.

Testing of Specimens from Capsule X

5-16 Table 5-11 Tensile Properties of the Wolf Creek Capsule X Reactor Vessel Surveillance Material Trrafllntedi hr AQ 49 x l 9

,,,-,l2 {. s 1 n .

Material Sample Test 0.2% Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp. Yield Strength Load Stress (ksi) Strength Elongation Elongation in Area (OF) Strength (ksi) (kip) (ksi) (%) (%) (%)

__ (ksi)

Lower Shell Plate AL-10 75 70.0 89.3 2.75 176.6 56.0 13.0 28.2 68 R2508-3 (Long.)

AL- 1 300 63.7 81.5 2.65 175.1 54.0 11.0 23.7 69 Al- 550 60.9 85.5 2.76 149.9 56.2 10.9 21.6 62 Lower Shell Plate AT-I0 75 71.3 90.2 3.10 168.4 63.2 12.8 25.3 62 R2508-3 (Trans.)

AT-II 300 63.7 82.3 T 2.80 140.7 57.0 11.3 20.8 59 AT- 12 550 58.6 86.3 0.35 15.9 7.1 11.3 18.1 55 WeId Meal AW-IO 75 81.0 96.4 3.15 180.7 64.2 11.5 25.1 64 AW- I I T 300 70.8 J 86.0 f 2.81 T 165.8 J 57.2 9.5 21.1 65 AW-12 550 70.8 f 92.8 j 3.20 178.7 65.2 11.0 22.2 64 Testing of Specimens from Capsule X

5-17 LOWER SHELL PLATE R2508-3 (LONGITUDINAL)

CVGRAPH 41 Hyperbolic Tangent Curve Pnnted at 0942j0 on 01-10-203 Plesults Curve Fluence MSE d-LSE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1 0 219 0 148 0 -2495 0 11 0 2 0 219 0 145 -3 11.51 36846 34B5 34.73 3 0 219 0 131 -17 -91 1603 3154 31,43 4 0 219 0 129 -19 271B 5Z03 4698 48.86 5 0 2.19 0 142 -6 3611 61.06 67.75 67.64 crw250 -

1.0 4-,F ___

I 0__ 10 -3000 20-0 0- 0 0

~177

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Leged i I 0-~ 2 0----- a0 ~ 4 -~ 5 Data Setgs) Plotted Curve Plant Capsule Material OriL Heat#

I WCI UNIRR PLATE SA533BI LT C4935-2 2 C!C U PLATE SA533BI LT C4935-2 3 WfCI Y PLATE SA533BI LT C4935-2 4 W'I V PLATE SA533BI LT C4935-2 5 iC! X PLATE SA533BI LT C4935-2 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

I' 5-18 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-19 LOWER SHELL PLATE R-2508-3 (LONGITUDINAL)

CYGRAPH 41 Hyperbolhc Tangent Curve Printed at 095013 on 01-10-20)03 Results Curve Fluence To 50z Shear d-T o 50%. Shear I 0 3843 0 2 0 712 M76 3 0 5484 16.4 4 0 9022 52M8 5 0 10664 6a2 Cu a)3

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Cune legend 120- 30 4^5 Data Set(s) Plotted Curve Plant Capsule llaterial Ori Heaqti 1 llCI UNIRR PLATE SA533BI LT C49352 2 WfC[ U PLATE SA533I LT C49352 3 lfCI Y PLATE SA533BI LT C4935-2 4 lfC1 V PLATE SA533BI LT C49352 S lfC1 X PLATE SA533BI LT

_. C4935 2 aww Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5.20 LOWER SHELL PLATE R2508-3 (TRANSVERSE)

CVGRAPH 41 Hyperbolc Tangent Curve Printed at 100356 on 01-10-2003 Resulb Curve Fluence LSE d-ISE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1 n .^ ^

I U 4..W U 94 U 2 0 34.32 0 2 0 219 0 96 2 25B 23.79 59.55 2523 3 0 219 0 94 0 3739 35.39 8149 4716 4 0 219 0 88 -6 5654 54.53 90.59 5627 5 0 219 0 95 1 55X97 53.96 97.04 62.71 C1 7

a)

-30 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10 20--- 30 4 ^- 5 Data Set(s) Plotted Curve Plant Capulle Matenal r---n- HtIPf Ori Tlva

• vtI~

I OC! UNIRR PLATE SA533BI TL C4935-2 2 TCI U PLATE SA533BI TL C4935-2 3 TO Y PLATE SA533BI TL C4935-2 4 WT V PLATE SA533B1 TL C4935-2 5 TI X PLATE SA533BI TL C4935-2 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-21 LOWER SHELL PLATE R2508-3 (TRANSVERSE)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at IW)723 on 01-10-2C03 Resilts Curve Fluence USE d-USE T o LE35 d-T o LE35 I ° 605 0 2544 0 2 0 72B6 481 357 1025 3 0 7548 743 67.84 42.39 4 0 61.41 -663 9a79 68.34 5 0 64.44 -a6 1024 7695 1507_

~1007-CO 50

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend I O- 2 0-- 30 4 - 5 -

Data Set(s) Plotted Curve Plant Capsule Material Ori Heatf I IC UNIRR PLATE SA533BI TL C4935-2 2 lCI U PLATE SA533BI TL C4935-2 3 W!a Y PLATE SA533BI TL C4935-2 4 WC! V PLATE SA533BI TL C4935-2 5 WC! x PLATE SA533BI TL C4935-2 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-22 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-23 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 41 Hyperbolic Tangent Curve Printed at 10182 on 01-10-2003 Reults Curve Fluence BSE d-ISE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 I 0 219 0 100 0 -57.69 0 -20.64 0 2 0 219 0 92 -8 -3047 2721 644 27.09 3 0 219 0 94 -6 -1259 45.09 2082 41.47 4 0 2.19 0 89 -Ll -1136 46.33 3179 5Z44 5 0 219 0 93 -7 10.66 68.36 5473 7538 a) z 0--

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend Io - 2 0-- 30 4 - 5 V.-

Data Set(s) Plotted Curve Plant Capsule Material Ori Heat#

I WlC UNIER WD WIRE HEAT NO.90146 2 WC] U van WIRE HEAT NO.90146 3 WC1 Y Wm WIRE HEAT NO.90146 4 WC1 V Wma WMRE HEAT NO.90146 5 lCI x Wm WIRE HEAT NO.90146 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Weld Mletal Testing of Specimens from Capsule X

II 5-24 SURVEILLANCE PROGRAM WELD METAL CYGRAPH 41 Hyperbolic Tangent Curve Printed at 102325 on 01-10-2003 Reults Curve Fluence USE d-USE To LE35 d-T o LE35 0 7526 0 -Z7.06 0 2 0 7722 L96 -13.04 14.02 3 0 7006 -517 17.96 45.03 4 0 6722 403 4552 7259 5 0 E.57 -1269 53.49 8056 4-4

-300 -200 -100 0 100 200 300 40 500 6 Temperature in Degrees F Curve Legend I - 20---- 30 4 5 Data Set(s) Plotted Curve Plant Capsule Material Ori Heatf I WCI UNIRR WELD WMRE HEAT NOS90146 2 WCl U WELD WIRE HEAT NO.90146 3 WCI Y WELD WIRE HEAT NO.90146 4 hCI V WELD WIRE HEAT N090146 5 WCI X WED WIRE HEAT N090146 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Wolf Creek Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-25 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 41 Hyperbolic Tangent Curve Printed at 102909 on 01-10-2003 ReuIts Curve Fluence T o 50z. Shear d-T o 50?/ Shear I 0 -7394 0 2 0 20.94 94B9 3 0 55.03 4 0 2076 9471 5 0 2109 9504 U) a)

C) a)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1 - 2 C--- 30 4^ - 5 Data Set(s) Plotted Curve Plant Capsule Material Or. Heat#

I UC UNIRR WIRE HEAT NO.90146 2 TOC U WED WIRE HEAT N090146 3 WCI Y WELD TIRE HEAT NO.90146 4 WTOI V WmL WIRE HEAT NO.90146 5 KCI X WmL WIRE HEAT NO.90146 Figure 5-9 . Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-26 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Wolf Creek Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule X

5-27 HEAT AFFECTED ZONE CVGRAPH 41 Hyperbolic Tangent Curve Printed at 10(3855 on 01-10-2003 RBults Curve nuence USE d-USE T o LE35 d-T o LE35 I 0 8466 0 -896 0 2 0 6726 26 -54.79 35.07 3 0 97.96 1329 -6051 29.35 4 0 6114 -16.51 -436 4626 5 0 72.49 -l116 -2762 6224 U)

P--

Ct

4

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend I o- 20-- 30 - 4- 5 Data Set(s) Plotted Curve Plant Capsule Material Ori HeatR I OCI UNIRR HEAT AFFD ZONE WIRE HEAT NO.S0146 2 ITCl U HEAT AFFD ZONE WIRE HEAT NO.00146 3 WC1 Y HEAT AFFD ZONE TIlRE HEAT N.090146 4 IC1 V HEAT AFFD ZONE TIE HEAT NO.00146 5 WC1 X HEAT AFYD ZONE TME HEAT NO.00146 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Volf Creek Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule X

5-28 HEAT AFFECTED ZONE CVGRAPH 41 Hyperbolic Tangent Curve Pnnted at 1a4139 on 01-10-2003 Results Curve Fluence T o 50z. Shear .I-T A 5rdx .%-<

T 0 50z. Shr I 0 -77.81 0 2 0 -20.47 57.34 3 0 -47.81 30 4 0 -30.46 47.34 5 0 _m2s 65.62 C.

U)

C.)

I

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend I0 2 0-- 30 4^ 5 ~

Data Set(s) Plotted Curve Plant Capsule Ilaterial Material Ori. Ueat On llptI I WC1 UNIRR HEAT AFFD ZONE WERE HEAT NO.90146 2 lC! U HEAT AFFD ZONE lIRE HEAT NO.S0146 3 lCI Y HEAT AFF'D ZONE WME HEAT NO.90146 4 lCI V HEAT AFFD ZONE WIRE HEAT NO090146 5 liCI X HEAT AFFD ZONE WIRE HEAT NO.S0146 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Wolf Creek Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule X

5-29 AL48,-50F AL57, 00 F AL53, 25 0 F AL52, 40 0 F AL56, 500 F AL55, 75 0F AL59, 110F AL50, 135 0F AL60, 150 0F AL49, 175 0F AL51, 190 0F AL46, 225-F ALS4, 250 0F AL58, 275 0F AL47, 300 0F Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-30 AT46,-75 0 F AT50,-25 0 F AT60,15-F AT56, 50 0 F AT54, 75 0 F AT53, 100 0F AT59, 125 0F AT58, 150 0F AT48, 175 0F AT51, 175 0F AT57,200 0F AT52, 225 0F AT49,250 0F AT47,275 0F AT55, 300 0F Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-31 AW47, -75 0 F AW52, -350 F AW53, 0F AW51, 250 F AW58, 500 F AW46, 750 F AW59, 1000 F AW57, 125 0 F AW55, 125TF AW48, 150TF AW60, 160TF AW56, 2000 F AW50,2250 F AW49, 250 0 F AW54, 2500 F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-32 AH53, -1750 F AH51, -1000 F AH50,-750 F AH49,-50 0 F AH58,-50 0 F AH52, -25 0 F AH47, 0F AH55, 00F AH59, 257F AH48, 35F AH54, 50 0F AH60, 100 0F AH46, 150TF AH57,200TF AH56, 200 0F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Heat-Affected-Zone Metal Testing of Specimens from Capsule X

5-33 (O C) 0 50 100 150 200 250 300 120 I I I I I I l_ 800 110 100 700 ULTIMATE TENSILE STRENGTH Cn 90 -A

-A A = A_ 600 Cs l 80 La

- 70 -_ AA 500 C-,

60 0 0 0 400 0 -

50 0.2% YIELD STRENGTH 300 40 LEGEND:

A 0 UNIRRADIATED 19 2 0 A s IRRADIATED TO A FLUENCE OF 3.49 X 10 nlcm (E>1.OMeV) AT 550 F 80 70 as 60

>- 50 1-

"_. 40 3 30 20 10 0

0 100 200 300 400 500 600 TEMPERATURE (OF)

Figure 5-17 Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-34 (O C) 0 50 100 150 200 250 300 120 I I I I I I I_ 800 110 100 700 I& ULTIMATE TENSILE STRENGTH 90 2 A- 600 C-1 80 2 _

La I-~ 70 500 60 0 0 0

%Y S2 400 50 0 2% YIELD STRENGTH 300 40 LEGEND:

A 0 UNIRRADIATED 19 2 0

• . IRRADIATED TO A FLUENCE OF 3.49 X 10 n/cm (E>1.0MeV) AT 550 F 80 70 - 2 REDUCTION INAREA 60 A I-1

_ A A

50 W-4

-J 0-4 40 C-,

30 _ o TOTAL ELONGATION 2 0

10 UNIFORM ELONGATION 0 I I I I I 0 100 200 300 400 500 600 TEMPERATURE (OF)

Figure 5-18 Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-35 (0C) 0 50 100 150 200 250 300 120 I I I I I I I 800 110 ULTIMATE TENSILE STRENGTH 100 700 A 't It 90 600 -d 80 0l-Lin C-,

70 500 60 - 0.2% YIELD STRENGTH 2/

400 50 300 40 LEGEND:

A 0 UNIRRADIATED 19 2 0 As IRRADIATED TO A FLUENCE OF 3.49 X 10 nlcm (E>1.OMeV) AT 550 F 80 REDUCTION INAREA 70 60 I-1 t-4 50 0-J-

-j I-- 40 30 TOTAL ELONGATION 2 20 10 UNIFORM ELONGATION 0 *I I I I I 0 100 200 300 400 500 600 TEMPERATURE (OF)

Figure 5-19 Tensile Properties for Wolf Creek Reactor Vessel Weld Metal Testing of Specimens from Capsule X

Il 5-36 Specimen ALI0 Tested at 75 0 F Specimen AL 1I Tested at 3 00 0 F Specimen AL 12 Tested at 550F Figure 5-20 Fractured Tensile Specimens from NVolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-37 Specimen ATI 0 Tested at 750 F Specimen AT 1 Tested at 3000 F Specimen AT12 Tested at 550OF Figure 5-21 Fractured Tensile Specimens from Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)

Testing of Specimens from Capsule X

5-38 Specimen AWIO Tested at 75 0 F Specimen AW 1I Tested at 300'F Specimen AW12 Tested at 550 0 F Figure 5-22 Fractured Tensile Specimens from Wolf Creek Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-39 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.X-100 90 80 70 y 60 to I 40 30 AL-10 75 F 20 10 0

0 0.05 01 0.15 0.2 025 03 STRAIN, INAN STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.X' 100 90 80 70 y 60 co50 Uj 40 AL-11 30 300 F 20 10 0

0 005 01 0.15 0.2 025 0.3 STRAIN, INAN Figure 5-23 Engineering Stress-Strain Curves for Wolf Creek Lower Shell Plate R2508-3 Tensile Specimens AL-10, AL-11 and AL-12 (Longitudinal Orientation)

Testing of Specimens from Capsule X

5-40 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.-X" 100' 90 80-70-Se 60-LO 50-w o 40 30 AL-12 20 550 F 10 0

0 005 01 0 15 02 025 03 STRAIN, INAN Figure 5-23 Continued Testing of Specimens from Capsule X

5-41 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE *X-100 90 80 70 e 60 U,

ED 50 40 AT-10 30 75 F 20 10 0

0 005 01 015 02 025 03 STRAIN, INAN STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE X" 100 -

90 -

80 -

70 -

i 60 -

(6 co 50 -

I 40 -

AT-1 1 30 - 300 F 20 -

10 -

0-0 0.05 0.1 0.15 02 0.25 03 STRAIN, IN/IN Figure 5-24 Engineering Stress-Strain Curves for Wolf Creek Lower Shell Plate R2508-3 Tensile Specimens AT-10, AT-11 and AT-12 (Longitudinal Orientation)

Testing of Specimens from Capsule X

.1 5-42 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE"X" 100 90 80 70 u4 60 w

CO50 40 AT-12 30 550 F 20 10 0

0 005 0.1 0.15 0.2 0.25 0.3 STRAIN, INAN Figure 5-24 - Continued Testing of Specimens from Capsule X

543 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE *X' 100 90 80 70

~4to 60 to 50 w

,,, 40 AW-10 30 75 F 20 10 0

0 005 0.1 0.15 0.2 025 03 STRAIN, ININ STRESS-STRAIN CURVE BEAVER VALLEY UNIT 2 W CAPSULE 100 90 80 70

$ 60 t) 50 400 30 AW-11 300 F 20 10 0

0 005 01 0.15 02 025 03 STRAIN, ININ Figure 5-25 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens AW-10, AW-11 and AW-12 Testing of Specimens from Capsule X

5-44 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE X' 100 90 80 70 9 60 to 50 40 AW-12 30 550 F 20 10 0

0 0 05 01 0.15 0.2 0.25 03 STRAIN, ININ Figure 5 Continued Testing of Specimens from Capsule X

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates S,, transport analysis performed for the Wolf Creek reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules.

In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule X, withdrawn at the end of the twelfth plant operating cycle, is provided. In addition, in order to provide a complete measurement database applicable to Wolf Creek, results from prior in-vessel irradiations are included in Appendix A to this report. The data included in Appendix A were previously documented in Reference 3. Comparisons of the results from these dosimetry evaluations with the analytical predictions served to validate the plant specific neutron transport calculations. These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY). These projections also account for a plant uprating, from 3411 MWt to 3565 MWt, which occurred during and post the seventh operating cycle.

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

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

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDFIB-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance and meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," 120 1.Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996211. The specific calculational methods applied are also consistent with those described in WCAP-15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology."22]

Radiation Analysis and Neutron Dosimetry

.1 6-2 methods applied are also consistent with those described in WCAP-15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology."1 2 2' 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the Wolf Creek reactor geometry at the core midplane is shown in Figures 6-1 a-c.

Six irradiation capsules attached to the neutron pads are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 610 and 241 (290 from the core cardinal axes) and 58.5°, 121.50, 238.50, and 301.5° (31.50 from the core cardinal axes) as shown in Figure 4-1. The stainless steel specimen containers are 1.182-inch by 1-inch and approximately 56-inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.

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

The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pads and the reactor vessel.

In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

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

0(r,0, z) = s(r, 6)

• 0(r, z) 0(r) where P(rO,z) is the synthesized three-dimensional neutron flux distribution, P(rO) is the transport solution in r,9 geometry, 4(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and ¢(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the rO two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Wolf Creek.

For the Wolf Creek transport calculations, the r,6 models depicted in Figures 6-la-c were utilized since the reactor is octant symmetric. This rO model includes the core, the reactor internals, the neutron pad --

including explicit representations of the surveillance capsules at 290 and 31.5°, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing this analytical model set, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh Radiation Analysis and Neutron Dosimetry

6-3 description of the re reactor model consisted of 183 radial by 99 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the rO calculations was set at a value of 0.001.

The rz model used for the Wolf Creek calculations that is shown in Figure 6-2 extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation approximately 1.5-feet below the active fuel to approximately 2.5-feet above the active fuel. As in the case of the re model, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The rz geometric mesh description of the reactor model consisted of 153 radial by 107 axial intervals. As in the case of the rO calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the rz calculations was also set at a value of 0.001.

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

The core power distributions used in the plant specific transport analysis were taken from the appropriate Wolf Creek fuel cycle design reports. The data extracted from the design reports represented cycle dependent fuel assembly enrichments, burnups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.11231 and the BUGLE-96 cross-section library12 41 . The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion and angular discretization was modeled with an S16 order of angular quadrature for the r and rz models while an S8 order of angular quadrature was used in the r,0 models. Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

Radiation Analysis and Neutron Dosimetry

Il 6-4 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 two azimuthally symmetric surveillance capsule positions (290 and 31.50). Also note that Table 6-1 presents calculated exposure rates and integrated exposures for Capsule X, which was irradiated at a 31.50 location during Cycles 1 through 12 until it was removed from service.

These results, representative of the axial midplane of the active core, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future.

Similar information is provided in Table 6-2 for the reactor vessel inner radius. The vessel data given in Table 6-2 are representative of the axial location of the maximum neutron exposure at each of the four azimuthal locations. It is also important to note that the data for the vessel inner radius were taken at the clad/base metal interface, and thus, represent the maximum calculated exposure levels of the vessel forgings and welds.

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Tables 6-1 and 6-2. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the twelfth operating fuel cycle as well as projections to 15.53, 20, 24, 32, 40, 48, and 54 effective full power years (EFPY). The projections were based on the assumption that the radial power distribution from fuel cycle 12 was representative of future plant operation (excluding cycle 13 projections). All remaining core parameters were obtained by utilizing cycle 12 (excluding cycle 13 projections). The future projections are also based on the current reactor power level of 3565 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-3 and 6-4, respectively. The data, based on the cumulative integrated exposures from Cycles 1 through 12, are presented on a relative basis for each exposure parameter at several azimuthal locations. Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-3 and 6-4.

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

Updated lead factors for the Wolf Creek surveillance capsules are provided in Table 6-6. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-6, the lead factors for capsules that have been withdrawn from the reactor (U, Y, V, and X) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (W and Z), the lead factors correspond to the calculated fluence values at the end of cycle 12 operations.

Radiation Analysis and Neutron Dosimetry

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

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule X that was withdrawn from Wolf Creek at the end of the twelfth fuel cycle, is summarized below.

Reaction'Rates (rps/atoin) > MC Reaction Measued Calculated Ratio 63Cu(n,a)6 Co 4.65E-17 4.34E-17 1.07

' 4Fe(n,p)54Mn 4.72E-15 4.79E-15 0.99 58Ni(n,p)"Co 6.49E-15 6.71E-15 0.97 23 SU(np)l 7Cs (Cd) 3.01E-14 2.56E-14 1.18 2 7Np(n,f)'3 7 Cs (Cd) 2.56E-13 2.50E-13 1.02 Average: 1.05

% Standard Deviation: 8.0 The measured-to-calculated (MIC) reaction rate ratios for the Capsule X threshold reactions range from 0.97 to 1.18, and the average MWC ratio is 1.05 +/- 8.0% (la). This direct comparison falls well within the

+/- 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the Wolf Creek reactor.

As a result, these comparisons validate the current analytical results described in Section 6.2 which are deemed applicable for Wolf Creek.

Radiation Analysis and Neutron Dosimetry

6-6 6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Wolf Creek surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodologywas carned out in the following four stages-1 - Comparison of calculations with benchmark measurements from the Pool Cntical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2 - Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

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

4 - Comparisons of the plant specific calculations with all available dosimetry results from the Wolf Creek surveillance program.

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

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

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

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 22.

Radiation Analysis and Neutron Dosimetry

6-7 Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

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

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

The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for Wolf Creek.

Radiation Analysis and Neutron Dosimetry

--- - - 11 6-8 Figure 6-la Wolf Creek r.O Reactor Geometry at the Core Midplane Wolf Creek Unit 1 R-T No Capsule Present 12.5 Degree DDORT Geometry

-c R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-9 Figure 6-lb Wolf Creek r,0 Reactor Geometry at the Core Midplane Wolf Creek Unit 1 R-T Single Capsule Present 20.0 Degree DORT Geometry

.Z I

. C C,

R Axis (cm)

Radiation Analysis and Neutron Dosimetry

IE 6-10 Figure 6- I c Wolf Creek rO Reactor Geometry at the Core Midplane Wolf Creek Unit 1 R-TDuaICapsule Present 22.5 Degree DORT Geometry 91 an

, O R Axis (cm)

Radiation Analysis and Neutron Dosimetry

6-11 Figure 6-2 Wolf Creek r,z Reactor Geometry Wolf Creek Unit 1 R-Z DORT Geometry R Axis (cm)

Radiation Analysis and Neutron Dosimetry

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

Cumulative Flux Operating [E>1.0 MeV]

Time [n/cmA2-sec]

Cycle [EFPY] 29 Deg 31.5 Deg 1 1.07 8.73E+10 9.33E+10 2 1.75 9.08E+10 1.OOE+11 3 2.43 7.67E+10 8.33E+10 4 3.57 7.30E+10 8.04E+10 5 4.79 7.12E+10 7.60E+10 6 5.82 6.66E+10 7.05E+10 7 7.12 6.44E+10 6.98E+10 8 8.33 7.44E+10 7.91E+10 9 9.78 6.22E+10 7.1 OE+10 10 11.10 8.19E+10 8.64E+10 11 12.47 7.23E+10 8.25E+10 12 13.83 7.29E+10 8.29E+10 Projection 15.53 6.97E+10 7.79E+10 Projection 20.00 7.29E+10 8.29E+10 Projection 24.00 7.29E+10 8 29E+10 Projection 32.00 7.29E+10 8.29E+10 Projection 40.00 7.29E+10 8.29E+10 Projection 48.00 7.29E+10 8.29E+10 Projection 54.00 7.29E+10 8.29E+10 Note Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

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

Cumulative Fluence Operating [E>1.0 MeV]

Time [n/cmA2-sec]

Cycle [EFPY] 29 Deg 31.5 Deg 1 1.07 2.96E+18 3.16E+18 2 1.75 4.91 E+18 5.32E+18 3 2.43 6.54E+18 7.09E+18 4 3.57 9.18E+18 1.OOE+19 5 4.79 1.19E+19 1.29E+19 6 5.82 1.41 E+19 1.52E+19 7 7.12 1.67E+19 1.81 E+19 8 8.33 1.96E+19 2.11 E+19 9 9.78 2.22E+19 2.42E+19 10 11.10 2.57E+19 2.78E+19 11 12.47 2.88E+19 3.13E+19 12 13.83 3.19E+19 3.49E+19 Projection 15.53 3.58E+1 9 3.93E+1 9 Projection 20.00 4.61 E+19 5.1 0E+19 Projection 24.00 5.53E+19 6.14E+19 Projection 32.00 7.37E+1 9 8.24E+19 Projection 40.00 9.21 E+1 9 1.03E+20 Projection 48.00 1.11 E+20 1.24E+20 Projection 54.00 1.24E+20 1.40E+20 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane Radiation Analysis and Neutron Dosimetry

6-14 Table 6-1 cont'd Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENT RATES Cumulative Operating Displacement Rate Time [dpa/sec]

Cycle [EFPY] 29 Deg 31.5 Deg 1 1.07 1.71 E-10 1.82E-10 2 1.75 1.78E-10 1.97E-10 3 2.43 1.49E-10 1.62E-10 4 3.57 1.42E-10 1.56E-10 5 4.79 1.38E-10 1.48E-10 6 5.82 1.29E-10 1.37E-10 7 7.12 1.25E-10 1 35E-10 8 8.33 1.44E-10 1.53E-10 9 9.78 1.21 E-10 1.38E-10 10 11.10 1.59E-10 1.68E-10 11 12.47 1.40E-10 1.60E-10 12 13.83 1.42E-10 1.61E-10 Projection 15.53 1.35E-10 1.51 E-10 Projection 20.00 1.42E-10 1.61 E-1 0 Projection 24.00 1.42E-10 1.61 E-10 Projection 32.00 1.42E-10 1.61 E-1 0 Projection 40.00 1.42E-10 1.61E-10 Projection 48.00 1.42E-1 0 1.61 E-1 0 Projection 54.00 1.42E-10 1.61 E-1 0 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane Radiation Analysis and Neutron Dosimnetry

6-15 Table 6-1 cont'd Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENTS Cumulative Operating Displacements Time [dpa]

Cycle [EFPY] 29 Deg 31.5 Deg 1 1.07 5.79E-03 6.1 8E-03 2 1.75 9.61 E-03 1.04E-02 3 2.43 1.28E-02 1.39E-02 4 3.57 1.79E-02 1.95E-02 5 4.79 - 2.32E-02 2.52E-02 6 5.82 2.74E-02 2.96E-02 7 7.12 3.25E-02 3.52E-02 8 8.33 3.80E-02 4.1 OE-02 9 9.78 4.32E-02 4.69E-02 10 11.10 4.99E-02 5.40E-02 11 12.47 5.59E-02 6.09E-02 12 13.83 6.20E-02 6.78E-02 Projection 15.53 6.96E-02 7.63E-02 Projection 20.00 8.96E-02 9.91 E-02 Projection 24.00 1.08E-01 1.19E-01 Projection 32.00 1.43E-01 1.60E-01 Projection 40.00 1.79E-01 2.01 E-01 Projection 48.00 2.15E-01 2.42E-01 Projection 54.00 2.42E-01 2.72E-01 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry

6-16 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Maximum Pressure Vessel Flux Operating [E>1.0 MeV]

Time [n/cmA2-sec]

Cycle [EFPY] 0 Deg 15 Deg 30 Deg 45 Deg 1 1.07 1.26E+10 1.87E+10 2.17E+10 2.19E+10 2 1.75 1.39E+10 1.99E+10 2.31E+10 2.68E+10 3 2.43 1.13E+10 1.63E+10 1.91E+10 1.85E+10 4 3.57 1.22E+10 1.67E+10 1.83E+10 1.90E+10 5 4.79 1.15E+10 1.67E+10 1.77E+10 1.71 E+10 6 5.82 9.47E+09 1.59E+10 1.69E+10 1.62E+10 7 7.12 8.16E+09 1.35E+10 1.67E+10 1.63E+10 8 8.33 9.32E+09 1.65E+10 1.90E+10 1.65E+10 9 9.78 7.78E+09 1 09E+10 1.57E+10 1.72E+10 10 11.10 9.92E+09 1.56E+10 1.99E+10 1.90E+10 11 12.47 9.15E+09 1.31E+10 1.79E+10 1.98E+10 12 13.83 9.OOE+09 1.36E+10 1.81E+10 2.14E+10 Projection 15.53 9.30E+09 1.36E+10 1.73E+10 1.94E+10 Projection 20.00 9.00E+09 1.36E+1 0 1.81 E+1 0 2.14E+1 0 Projection 24.00 9.00E+09 1.36E+10 1.81E+10 2.14E+10 Projection 32.00 9.OOE+09 1.36E+10 1.81E+10 2.14E+10 Projection 40.00 9.OOE+09 1.36E+10 1.81E+10 2.14E+10 Projection 48.00 9.OOE+09 1.36E+10 1.81E+10 2.14E+10 Projection 54.00 9.OOE+09 1.36E+10 1.81E+10 2.14E+10 Radiation Analysis and Neutron Dosimetry

6-17 Table 6-2 cont'd Calculated Azimuthal Variation of Maximum Exposure Rates And Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Maximum Pressure Vessel Fluence Operating [E>1.0 MeV]

Time [n/cmA 2]

Cycle [EFPY] 0 Deg 15 Deg 30 Deg 45 Deg 1 1.07 4.26E+1 7 6.32E+17 7.36E+17 7.44E+17 2 1.75 7.11E+17 1.04E+18 1.21 E+18 1.29E+18 3 2.43 9.50E+1 7 1.39E+18 1.62E+18 1.69E+18 4 3.57 1.39E+18 1.99E+18 2.27E+18 2.37E+1 8 5 4.79 1.84E+18 2.63E+18 2.96E+18 3.03E+18 6 5.82 2.14E+18 3.14E+18 3.50E+18 3.55E+18 7 7.12 2.45E+1 8 3.66E+18 4.14E+18 4.18E+18 8 8.33 2.80E+18 4.28E+18 4.85E+18 4.79E+18 9 9.78 3.14E+18 4.74E+18 5.53E+18 5.54E+1 8 10 11.10 3.55E+1 8 5.40E+18 6.36E+18 6.33E+1 8 11 12.47 3.95E+1 8 5 96E+18 7.13E+18 7.19E+18 12 13.83 4.33E+18 6.54E+18 7.91 E+18 8.1 OE+18 Projection 15.53 4.86E+1 8 7.31 E+18 8.88E+18 9.20E+1 8 Projection 20.00 6.12E+18 9.23E+18 1.14E+19 1.22E+19 Projection 24.00 7.26E+1 8 1.09E+19 1.37E+19 1.49E+19 Projection 32.00 9.53E+1 8 1.44E+19 1.83E+19 2.03E+1 9 Projection 40.00 1.18E+19 1.78E+1 9 2.28E+1 9 2.57E+19 Projection 48.00 1.41 E+19 2.12E+19 2.74E+19 3.11E+19 Projection 54.00 1.58E+19 2.38E+19 3.08E+19 3.51 E+19 Radiation Analysis and Neutron Dosimetry

6-18 Table 6-2 cont'd Calculated Azimuthal Variation of Fast Neutron Exposure Rates And Iron Atom Displacement Rates At the Reactor Vessel Clad/Base Metal Interface Cumulative Maximum Iron Atom Displacements Operating Time [dpa/sec]

Cycle [EFPY] 0 Deg 15 Deg 30 Deg 45 Deg 1 1.07 1.95E-11 2.87E-1 1 3.35E-1 1 3.47E-1 1 2 1.75 2.16E-1 1 3.06E-1 1 3.56E-1 1 4.22E-1 1 3 2.43 1.75E-11 2.51 E-11 2.95E-1 1 2.93E-1 1 4 3.57 1.90E-1 1 2.58E-1 1 2.82E-1 1 3.01 E-1 1 5 4.79 1.79E-11 2.56E-1 1 2.73E-1 1 2.70E-1 1 6 5.82 1.48E-11 2.44E-1 1 2.61 E-11 2.57E-1 1 7 7.12 1.27E-11 2.08E-11 2.58E-1 1 2.57E-1 1 8 8.33 1.46E-1 1 2.54E-1 1 2.93E-1 1 2.61 E-11 9 9.78 1.21 E-11 1.68E-1 1 2.43E-1 1 2.73E-1 1 10 11.10 1.55E-11 2.40E-1 1 3.06E-1 1 3.OOE-1 1 11 12.47 1.43E-1 1 2.02E-1 1 2.77E-1 1 3.13E-1 1 12 13.83 1.40E-1 1 2.1 OE-11 2.79E-1 1 3.37E-11 Projection 15.53 1.45E-1 1 2.1 OE-1 1 2.67E-1 1 3.07E-1 1 Projection 20.00 1.40E-11 2.1OE-1 1 2.79E-1 1 3.37E-1 1 Projection 24.00 1.40E-1 1 2.1 OE-11 2.79E-1 1 3.37E-11 Projection 32 00 1.40E-11 2.1OE-1 1 2.79E-1 1 3.37E-1 1 Projection 40 00 1.40E-11 2.1 OE-11 2.79E-1 1 3.37E-1 1 Projection 48.00 1.40E-1 1 2.10E-1 1 2 79E-11 3.37E-1 1 Projection 54.00 1.40E-1 1 2.10E-1 1 2.79E-1 1 3 37E-11 Radiation Analysis and Neutron Dosimetry

6-19 Table 6-2 cont'd Calculated Azimuthal Variation of Maximum Exposure Rates And Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cumulative Maximum Iron Atom Displacements Operating Time [dpa]

Cycle [EFPY] 0 Deg 15 Deg 30 Deg 45 Deg 1 1.07 6.62E-04 9.74E-04 1.14E-03 1.18E-03 2 1.75 1.11 E-03 1.60E-03 1.87E-03 2.04E-03 3 2.43 1.48E-03 2.14E-03 2.49E-03 2.67E-03 4 3.57 2.16E-03 3.07E-03 3.51 E-03 3.75E-03 5 4.79 2.85E-03 4.05E-03 4.56E-03 4.79E-03 6 5.82 3.33E-03 4.84E-03 5.41 E-03 5.62E-03 7 7.12 3.82E-03 5.64E-03 6.40E-03 6.61 E-03 8 8.33 4.36E-03 6.59E-03 7.50E-03 7.59E-03 9 9.78 4.89E-03 7.31 E-03 8.54E-03 8.76E-03 10 11.10 5.53E-03 8.32E-03 9.82E-03 1.OOE-02 11 12.47 6.15E-03 9.19E-03 1.1 OE-02 1.14E-02 12 13.83 6.75E-03 1.01 E-02 1.22E-02 1.28E-02 Projection 15.53 7.57E-03 1.1 3E-02 1.37E-02 1.45E-02 Projection 20.00 9.55E-03 1.42E-02 1.77E-02 1.93E-02 Projection 24.00 1.13E-02 1.69E-02 2.12E-02 2.36E-02 Projection 32.00 1.49E-02 2.22E-02 2.82E-02 3.21 E-02 Projection 40.00 1.84E-02 2.75E-02 3.53E-02 4.06E-02 Projection 48.00 2.20E-02 3.28E-02 4.23E-02 4.91 E-02 Projection 54.00 2.46E-02 3.68E-02 4.76E-02 5.55E-02 Radiation Analysis and Neutron Dosimetry

'I 6-20 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1 0 MeV)

Within The Reactor Vessel Wall RAD1IUS AZIMUTHALANGLE (cm) 00 150 300 45 l 220.35 1.00 1.00 1.00 1.00 225.87 0.56 0.56 0 55 0.550 231.39 0.28 0.27 0.26 0.26 236.90 0.13 0.13 0.12 0 12 242.42 0.06 0.06 0 06 0.05 Note: Base Metal Inner Radius = 220.35 cm Base Metal 1/4T = 225.87 cm Base Metal 1/2T = 231.39 cm Base Metal 3/4T = 236.90 cm Base Metal Outer Radius = 242.42 cm Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTHALANGLE (cm) .0° 150 300 450 220.35 1.00 1.00 1.00 1.00 225.87 0.64 0 63 0 63 0.64 231.39 0.39 0.38 0.37 0.39 236.90 0.23 0.22 0.22 0.23 242.42 0.14 0.13 0.12 0.13 Note: Base Metal Inner Radius = 220.35 cm Base Metal 1/4T = 225.87 cm Base Metal 1/2T = 231.39 cm Base Metal 3/4T = 236.90 cm Base Metal Outer Radius = 242.42 cm Radiation Analysis and Neutron Dosimetry

6-21 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Wolf Creek Capsule A Irradiation Time Fluence D(E.O MeY) Iron Displacements

,EPY]

1 .s .Xfnn IK A [

f1 dpal U 1.07 3.16E+18 6.18E-03 Y 4.79 1.19E+19 2.32E-02 V 9.78 2.22E+19 4.32E-02 X 13.83 3.49E+19 6 78E-02 Table 6-6 Calculated Surveillance Capsule Lead Factors CA le ID

-And Location StatusLead Factor -

U (31.50) Withdrawn EOC 1 (for analysis) 4.25 Y (290) Withdrawn EOC 5 (for analysis) 3.93 V (290) Withdrawn EOC 9 (for analysis) 4.02 X (31.50) Withdrawn EOC 12 (for analysis) 4.30 W (31.5°) In Reactor 4.11 Z (31.50 ) In Reactor 4.11 Notes (1) Capsules U, Y,V,and X were contained in dual capsule holders, while Capsules W and Z are being irradiated in single capsule holders.

(2) Lead factors for capsules remaining in the reactor are based on exposure calculations through Cycle 12 operations for the single capsule holders.

Radiation Analysis and Neutron Dosimetry

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM El 85-82 and is recommended for future capsules to be removed from the Wolf Creek reactor vessel. This recommended removal schedule is applicable to 32 EFPY of operation.

Table 7 Recommended Suellance Capsule With Schedule CapsCap Cpsul Location edFacor( Wit wal EFPYV Fluence(Wcr(a 2

)

U 58.50 4.25 1.07 3.16 x loll (c)

Y 2410 3.93 4.79 1.19 x 1019 (c)

V 60.10 4.02 9.78 2.22 x 10'9 (c)

X 238.50 4.30 13.83 3.49 x 1019 (c)

W 121.50 4.11 Standby (d)

Z 301.50 4.11 Standby (d)

Notes (a) Updated in Capsule X dosimetry analysis (b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) The standby capsules have already reached a peak vessel surface fluence, equivalent to 54 EFPY.

They will reach two times this fluence at 26.8 EFPY. Thus, it is recommended that the standby capsules be removed and placed in storage, as recommended in NUREG-1 801, to preserve meaningful metallurgical data 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, I OCFR50, Appendix G. Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.

3. WCAP-15078, Revision I, Analysis of Capsule Vfrom the W1lolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program, E. Terek, et. al., dated September 1998.

4. WCAP-10015, Kansas Gas and Electric Company Wolf Creek Generation Station Unit No. I Reactor Vessel Radiation Surveillance Program,L.R. Singer, dated June 1982.

5. WCAP-1 1553,Analysis of Capsule Ufroin the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program,S.E. Yanichko, et. al., dated August 1987.

6. WCAP-13365, Revision 1, Analysis of Capsule Yfrom the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program, J.M. Chicots, et. al., dated April 1993.

7. ASTM E208, StandardTest Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of FerriticSteels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

8. Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G. FractureToughness Criteria for ProtectionAgainst Failure

9. ASTM El 85-82, StandardPracticefor Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.

10. Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 2.

11. Procedure RMF 8102, Tensile Testing, Revision 1.

12. Procedure RMF 8103, Charpy Impact Testing, Revision 1.

13. ASTM E23-98, Standard Test Methodfor Notched BarImpact Testing of Metallic Materials,ASTM, 1998.

14. ASTM A370-97a, StandardTest Methods and Definitionsfor Mechanical Testing of Steel Products, ASTM, 1997.

References

Il 8-2 16 ASTM E21-92 (1998), StandardTestMethodsforElevated Temperature Tension Tests ofMetallzc Materials,ASTM, 1998 17 ASTM E83-93, StandardPracticefor Verification and Classification of Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

18. ASTM E 185-79, StandardPracticefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels

19. WCAP-143 70, Use of the Hyperbolic Tangent Functionfor Fitting Transition Temperature Toughness Data, T. R. Mager, et al, May 1995.

20. Regulatory Guide RG- 1.190, Calculatonaland DosimetryMethods for DeterminingPressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

21 WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold OverpressureMitigating System Setpoints andRCSHeatup and Cooldown Limit Curves, January 1996.

22 WCAP- 15557, Revision 0, Qualificationof the Westinghouse Pressure Vessel Neutron Fluence EvaluationMethodology, August 2000.

23. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two-and Three-Dimensional Discrete OrdinatesNeutron/PhotonTransport Code System, August 1996.

24. RSIC Data Library Collection DLC-1 85, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

References

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

A-I A.1 Neutron Dosimetry Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Wolf Creek are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (Reference A-I). One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report. This information may also be useful in the future, in particular, as least squares adjustment techniques become accepted in the regulatory environment.

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

Capsule ID Azimuthal Withdrawal Irradiation Location Time Time [EFPY]

U 31.5° End of Cycle 1 1.07 Y 290 End of Cycle 5 4.79 V 290 End of Cycle 9 9.78 X 31.50 EndofCycle 12 13.83 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

Appendix A

11 A-2 The passive neutron sensors included in the evaluations of Surveillance Capsules U, Y. V, and X are summarized as follows:

• The cobalt-aluminum measurements for this plant include both bare wire and c.idmium-coxered sensors Since all the dosimetry monitors were accommodated within the dosimeter block centered at the radial, azimuthal, and axial center of the material test specimen array, gradient corrections were not required for these reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table A-I.

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

• The measured specific activity of each monitor,

• the physical characteristics of each monitor,

• the operating history of the reactor,

• the energy response of each monitor, and

• the neutron energy spectrum at the monitor location.

Appendix A

A-3 The radiometric counting of the neutron sensors from Capsules U and Y was carried out at the Westinghouse Analytical Services Laboratory at the Waltz Mill Site (Reference A-2). The radiometric counting of the sensors from Capsule V was completed at the Pace Analytical Laboratory, also located at the Waltz Mill Site. Capsule X's radiometric sensor counting was completed by Pace Analytical Services, located at the Waltz Mill Site. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples.

In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

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

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

A R =A No F Y ' C, [I -e-I] [e-Ad]

P,e where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,,f (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 O(E > 1.0 MeV) during irradiation period j to the time weighted average O(E > 1.0 MeV) over the entire irradiation period.

=

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

t = Length of irradiation periodj (sec).

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

Appendix A

lo A-4 and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio

[P3]/[Pwf] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio C,, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, C, is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional C, term should be employed.

The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.

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

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

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 23 5U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 23'U and 2 37Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Wolf Creek fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V Capsule X U Impunty/Pu Build-in 0.87 0.84 0.80 0.76 238 U(y,f) 0.97 0.97 0.97 0.97 Net 238 U Correction 0.84 0.81 0.78 0.73 237 lNp(Yf) 0.99 0.99 0.99 0.99 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules U, Y, V, and X are given in Table A4. In Table A-4, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with the applied corrections for 238 U impurities, plutonium build-in, and gamma ray induced fission effects.

Appendix A

A-5 A.1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as O(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, R +SR a=E(G )( 0+/-c+/-6, relates a set of measured reaction rates, R,, to a single neutron spectrum, O., through the multigroup dosimeter reaction cross-section, oyg, each with an uncertainty 6. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the Wolf Creek surveillance capsule dosimetry, the FERRET code (Reference A-3) was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (O(E > 1.0 MeV) and dpa) along with associated uncertainties for the four in-vessel capsules withdrawn to date.

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

I - The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2 - The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3 - The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Wolf Creek application, the calculated neutron spectrum was obtained from the results of plant specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library (Reference A-4). The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations byASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances.

The assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance."

Appendix A

A-6 The following provides a summary of the uncertainties associated with the least squares evaluation of the Wolf Creek surveillance capsule sensor sets.

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

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

Reaction Uncertainty 63 Cu(nQ)60Co 5%

54Fe(np) 54 Mn 5%

58Ni(np) 58 Co 5%

238U(nf) 137Cs 10%

2 3 7Np(n,f)' 37 Cs 10%

5 9 Co(n,y)6OCo 5%

These uncertainties are given at the I c level.

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

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

Appendix A

A-7 These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

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

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

Mgg*= R 2 +R *Rg *Pgg-where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg' specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

P. = ((J-6]S,. + 0e-"

where 2

(g _ g,)

2y72 Appendix A

A-8 The first term m the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term)

The value of 8 is 1 0 when g = g', and is 0.0 otherwise The set of parameters defining the input covariance matrix for the Wolf Creek calculated spectra was as follows Flux Normalization Uncertainty (R.) 15%

Flux Group Uncertainties (R., Rg.)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0 68 eV) 52%

Short Range Correlation (0)

(E >00055 MeV) 09 (0.68 eV < E < 0.0055 MeV) 05 (E<0.68eV) 05 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0 68 eV<E<0.0055 MeV) 3 (E<0.68eV) 2 Appendix A

A-9 A.1.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Wolf Creek surveillance capsules withdrawn to date are provided in Tables A-5 and A-6. In Table A-5, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates.

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

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

Further comparisons of the measurement results with calculations are given in Tables A-7 and A-8.

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

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.93-1.30 for the 20 samples included in the data set.

The overall average M/C ratio for the entire set of Wolf Creek data is 1.08 with an associated standard deviation of 8.2%.

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

Table A-9 has been included to address current and projected (through 54 EFPY) neutron fluences (E >1.0 MeV) experienced by each of the circumferential and vertical welds modeled for this project.

Appendix A

A-10 Table A- I Nuclear Parameters Used In the Evaluation of Neutron Sensors Notes: The 90% response range is defined such that, in the neutron spectrum characteristic of the Wolf Creek surveillance capsules, approximately 90%

of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5%

of the total response due to neutrons with energies above the upper limit.

Appendix A

A- Il Table A-2 Monthly Thermal Generation During the First Twelve Fuel Cycles Of The Wolf Creek Reactor (Reactor Power of 3411 MWt June 12, 1985 -May 17, 1993; 3450 MWt May 17, 1993 -November 2, 1994; and 3565 MWt thereafter)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1985 6 356676 1988 8 2533606 1991 10 0 1985 7 1025780 1988 9 2450165 1991 II 0 1985 8 1643803 1988 10 492163 1991 12 0 1985 9 2053023 1988 11 0 1992 1 1268945 1985 10 2086772 1988 12 0 1992 2 1524407 1985 11 2366472 1989 1 2095086 1992 3 321390 1985 12 2368666 1989 2 2113705 1992 4 2446580 1986 1 2480479 1989 3 2535552 1992 5 2534299 1986 2 2005668 1989 4 2454150 1992 6 2453249 1986 3 2513225 1989 5 2498149 1992 7 2535304 1986 4 933250 1989 6 2448863 1992 8 2531360 1986 5 2341310 1989 7 2493515 1992 9 2453478 1986 6 1670026 1989 8 2534633 1992 10 2534881 1986 7 2210358 1989 9 2453774 1992 11 2296524 1986 8 2439547 1989 10 2516573 1992 12 2535128 1986 9 2406802 1989 11 2450503 1993 1 2534643 1986 10 1219774 1989 12 2536033 1993 2 2288546 1986 11 0 1990 1 2534772 1993 3 282997 1986 12 650000 1990 2 2017613 1993 4 0 1987 1 1533313 1990 3 599723 1993 5 1124412 1987 2 2192444 1990 4 0 1993 6 2453687 1987 3 2471746 1990 5 1003923 1993 7 2535510 1987 4 2247475 1990 6 2442569 1993 8 2535563 1987 5 2436662 1990 7 2515109 1993 9 2453641 1987 6 2250313 1990 8 2534494 1993 10 2532824 1987 7 2066874 1990 9 2453417 1993

- 11 2435990 1987 8 2527262 1990 10 2533710 1993 12 2557464 1987 9 1954923 1990 11 2421081 1994 1 2051879 1987 10 0 1990 12 2531359 1994 2 2312015 1987 11 0 1991 1 2363291 1994 3 2561261 1987 12 0 1991 2 1840498 1994 4 2531248 1988 1 1216547 1991 3 1969185 1994 5 2597022 1988 2 956585 1991 4 1506284 1994 6 2520895 1988 3 2526972 1991 -5 1692964 1994 7 2573456 1988 4 2452604 1991 6 2434282 1994 8 2577876 1988 5 2533966 1991 7 2534580 1994 9 1098290 1988 6 2451743 1991 8 2466385 1994 10 0 1988 7 2531412 1991 9 1221097 1994 11 2306676 Appendix A

A-12 Table A-2 cont'd Monthly Thermal Generation during the First Twelve Fuel Cycles Of The Wolf Creek Reactor (Reactor Power of 3411 MWt June 12, 1985 - May 17, 1993; 3450 MWt May 17, 1993 -November 2, 1994; and 3565 MWt thereafter)

Thra I I Thermal I l Thermal Thermal Generation Generatio Generation n

Year Month I (MWt-hr) Year Mn19 onth (MWt-hr)

. . I I I Tllt Year Month l (MWt-IIr) 1994 12 2636320 I1998 _

i 2632 2394162 2001 2562211 1995 4 1995 2 1 2639803 2383600 1998 1998 3 2649959 2001 2631325 4 2001 6 2565829 1995 3 2020996 1998 2564698 5 2001 7 2651544 1995 4 2543603 1998 2649737 6 2561103 2001 8 2651448 1995 5 2634001 1998 7 2642611 2001 9 2565781 1995 6 2524794 1998 _8 I 2649749 2001 10 1995 7 2632369 1998 1 1 2654261 9 2564818 2001 1995 8 2634129 1998 10 2001 2001 11 I 2651470 1995 9 I 2468895 1998 I I10 25i 51237 2649145, 2001

• 1-

_ I

_ 11 V) 2650720 1995 10 2637097 1998 12 2002 43012 2388288 1995 I1 2549850 I 2002 2 1999 26116340 1836759 1995 12 2640215 3 58213 1996 1 2481700 1999 3 2 t2613 200 1996 2 0 1999 4 104209 1996 3 0 1999 5 174 40059 1996 4 1784195 1999 6 25( 64943 1996 5 2622027 1999 7 264 49843 I.

1996 1 6 2349911 1999 8 251 4800 I VV0 ____

7---.-.-- 26428 I11 1qqq

__ _ __I_ a _

oUt L 1 I 1996 8 2603985 1999 10 2652619 1996 9 2564802 1999 11 2543480 1996 10 2621302 1999 12 2624035 iu 1l _ Z563J1 2000 l 2642844 1996 12 l 2649864 2000 2 2468736

-1 '__. I - I --- I I I 77 I I I Ilhaxq Ifuln nd:rnOlca inn- I i i iI I

-' L0JUL I IYY Z 2393586

__ I___

2000n

_vv A .

IOc~noI I^D j~

.-. I n 1997 1997 33OTA2i1 2650251 2EOo 2000 rn 5 2638828 4 1

1997 4 12564428 1 2000 6 2564614 IUU-7 I Iz I 'ii. o~n ,- I - -- [- -. tU-I II

. I7I7I -J zz I aIzv Z()()(l)

/1I I -7 7e AC4q1)

ILI~¶1C 1997 6 25-63505 2000 8 2650187 1CV-1997 7 2648185 2000 2064968 =

1997 8 2649138 2000 10 0 1997 9 2563192 2000 1898243 _

1997 10 18863 2000 12 2650536 1997 11 0 2001 1 2649695 i af7 _

I I I I - ._

i 7I I IZ z4w(fil/ I Vfull ozarcrn 1 inno i -

0I.;Ia I 2649875 1 2001 'A I OS144z I,, _I2 Zt97 I 2001 I)__ _ _ _ I ~.)IUUJO I II Appendix A

A-13 Table A-3 Calculated C, Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel _(E > 1.0 MeV) [n/cm 2 -s _ _C Cycle Capsule Capsule Capsule Capsule X Capsule Capsule Y Capsule Capsule U Y ,V U V X

] 9.33E+10 8.73E+ 10 8.73E+ 10 9.33E+I 0 1.00 1.11 1.20 1.16 2 9.08E+ 10 9.08E+ I0 I.OOE+ 11 1.15 1.25 1.25 3 7.67E+10 7.67E+10 8.33E+10 0.97 1.06 1.04 4 7.30E+ 10 7.30E+ 10 8.04E+I 0 0.93 1.01 1.00 5 7.12E+I 0 7.12E+ 10 7.60E+10 0.90 0.98 0.95 6 6.66E+ 10 7.05E+10 0.92 0.88 7 6.43E+10 6.98E+10 0.89 0.87 8 7.44E+10 7.91E+10 1.02 0.98 9 6.22E+10 7.10E+10 0.86 0.88 10 8.64E+ 10 1.08 11 8.25E+ 10 1.03 12 8.29E+1 0 1.03 Average 9.33E+10 7.88E+10 7.27E+ 10 8.04E+ 10 1.00 1.00 1.00 1.00 Appendix A

A-14 Table A-4 Measured Sensor Activities and Reaction Rates Surveillance Capsule U

. Radiall Radially

.justed Ad Ajse Measured Saturated Saturated Reaction Activity Acti ivity Rate Reaction Location (dpsfg) dpsg) (dps[g) (rps/atom) 63Cu (n,a) 6OCo Top 4.44E+04 3.54E+05 3.54E+05 5 41E-17 Center 4.40E+04 3 51E+05 3.5 1E+05 5.36E-17 Bottom 4.75E+04 3.79E+05 3.79E+05 5.78E-17 Average 5.52E-17 54Fe (n,p) 54Mn Top 1.5 1E+06 3.50E+06 3.50E+06 5.55E-15 Center 1.50E+06 3.48E+06 3.48E+06 5.52E-15 Bottom 1.80E+06 4.18E+06 4.18E+06 6.62E-15 Average 5.90E-15 58Ni (n,p) 5SCo Top 1.64E+07 5.43E+07 5.43E+07 7.77E-15 Center 1.61E+07 5.33E+07 5.33E+07 7.62E-15 Bottom 1.76E+07 5.82E+07 5.82E+07 8.33E-15 Average 7.91E-15 (n,f) 37 2U Cs (Cd) Middle 1.43E+05 5.90E+06 I 5.90E+06 3.87E-14 238U (n,f) 13 7Cs 235 (Cd) 9 Including U, 23 pu, andy,fission corrections: 3.26E-14 237Np (nf) 137 CS (Cd) Middle 1.24E+06 5.12E+07 5.12E+07 3 26E-13 237Np (nf) 137 CS (Cd) Including Yfis sion correction. 3.23E-13 59 Co (nY) 60Co Top 1.04E+07 8.30E+07 8 30E+07 5.42E-12 Middle 1.OOE+07 7.98E+07 7.98E+07 5 21E-12 Bottom l.OlE+07 8.06E+07 8.06E+07 5.26E-12 Average 5.30E-12 5 9Co (ny) 6WCo (Cd) Top 5.27E+06 4.21E+07 4.2 1E+07 2.75E-12 Middle 5.14E+06 4.1OE+07 4.10E+07 2.68E-12 Bottom 4 89E+06 3 90E+07 3.90E+07 2.55E-12 I Average 2.66E-12 Notes: 1) Measured specific activities are indexed to a counting date of February 2, 1987.

2) The average 238U (n,f) reaction rate of 3.26E-14 includes a correction factor of 0.87 to account for plutonium build-in and an additional factor of 0.97 to account for photo-fission effects in the sensor

3) The average 237 Np (n,f) reaction rate of 3.23E-13 includes a correction factor of 0.99 to account for photo-fission effects in the sensor Appendix A

A-15 I I ,

Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule Y Radially Radially Adjusted Adjusted

.Mea re.d Saturated Saturated Reaction Activity. Activity - .Activity.. Rate Reaction . aLocation (dP )sa-...... (dps[R) .. -(dpsfg) (rps/atom) 63 Cu (na) 6OCo Top 1.37E+05 3.38E+05 3.38E+05 5.16E-17 Center 1.20E+05 2.96E+05 2.96E+05 4.52E-17 Bottom 1.21E+05 2.99E+05 2.99E+05 4.56E-17 Average 4.75E-17 54Fe (n,p) 4Mn Top 1.66E+06 3.05E+06 3.05E+06 4.84E-15 Center 1.49E+06 2.74E+06 2.74E+06 4.34E-15 Bottom 1.48E+06 2.72E+06 2.72E+06 4.31E-15 Average 4.50E-15 58Ni (n,p) 5 8 Co Top 8.04E+06 4.53E+07 4.53E+07 6.48E-15 Center 7.38E+06 4.16E+07 4.16E+07 5.95E-15 Bottom 7.33E+06 4.13E+07 4.13E+07 5.91E-15 Average 6.12E15 238U (nf) 137 CS (Cd) Middle 5.43E+05 5.33E+06 I 5.33E+06 3.50E-14 238u 37 5 239 Including 23U, pu, and yfission corrections: 2.84E-14 (nf) 1 CS (Cd) 237Np (nf) 13 7 Cs (Cd) Middle 4.40E+06 4.32E+07 4.32E+07 2.76E-13 237Np (n f) 137CS (Cd) Including y,fis sion correction 2.73E-13 59Co (n,y) 6Co Top 2.59E+07 6.39E+07 6.39E+07 4.17E-12 Bottom 2.57E+07 6.34E+07 6.34E+07 4.14E-12 Average 4.16E-12 59 Co (n,y) 60Co (Cd) Top 1.30E+07 3.21E+07 3.21E+07 2.09E-12 Middle 1.36E+07 3.36E+07 3.36E+07 2.19E-12

. Bottom 1.39E+07 3.43E+07 3.43E+07 2.24E-12 Average 2.17E-12 Notes: 1) Measured specific activities are indexed to a counting date of February 19, 1992.

2) The average 238U (nf) reaction rate of 2.84E-14 includes a correction factor of 0.84 to account for plutonium build-in and an additional factor of 0.97 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.73E-13 includes a correction factor of 0.99 to account for photo-fission effects in the sensor.

Appendix A

A- 16 Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule V

-,.Radially Radially.

'Adjusted Adjusted Measured Saturated

. Reaction ActiAity ' Activity Rate Reaction Location (dpsR) (dpsJg) (dp/lg) (rpslatom) 63CU (n,a) 6OCo Top 1.64E+05 2.79E+05 2.79E+05 4.25E-17 Center 1 6lE+05 2.74E+05 2.74E+05 4.17E-17 Bottom 1.85E+05 3.14E+05 3.144E+05 4.79E- 17 Average 4.41E-17 54Fe (n,p) 54 Mn Top 1.35E+06 2 67E+06 2.67E+06 4.24E-15 Center 1 37E+06 2.71E+06 2.71E+06 4.30E-15 Bottom 1.5 E+06 2.9906 6 2.99E+06 Average 4.74E- 15 4.43E-15 58Ni (n,p) 5"Co Top 4.01E+06 4.38E+07 4.38E+07 6.27E-15 Center 4.00E+06 4.37E+07 4.37E+07 6.25E-15 Bottom 4.37E+06 4.77E+07 4.77E+07 Average 6 83E-15 6.45E-15 23 8 U (n,f) '"Cs (Cd) Middle 1.14E+06 5.91E+06 5.91E+06 3.88E-14 23U (n,f) Cs (Cd) Including 235U, 239 Pu, and yfission corrections: 3.01E-14 Np (n,f) '3CS (Cd) Middle 8.16E+06 4.23E+07 4 23E+07 2.70E-13 237Np (n,f) '-"Cs (Cd)

Including y,fission correction: 2.67E-13 59 Co (n,y) 6Co Top 2.78E+07 4.72E+07 4.72E+07 3 08E-12 Middle 3.13E+07 5.32E+07 5.32E+07 3.47E-12 Bottom 2 63E+07 4.47E+07 4.47E+07 2.92E- 12 Average 3.16E-12 59 Co (n,y) 6'Co (Cd) Top 1.66E+07 2.82E+07 2.82E+07 1.84E-12 Middle 1.62E+07 2.75E+07 2.75E+07 1.80E-12 Bottom 1.57E+07 2.67E+07 2.67E+07 1.74E-12 Average 1.79E-12 Notes: 1) Measured specific activities are indexed to a counting date of May 18, 1998.

2) The average 238U (n,f) reaction rate of 3.01E-14 includes a correction factor of 0.80 to account for plutonium build-in and an additional factor of 0.97 to account for photo-fission effects in the sensor

3) The average 237Np (nf) reaction rate of 2.67E-13 includes a correction factor of 0.99 to account for photo-fission effects in the sensor.

Appendix A

A-17 I

Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule X

.. ...I Radially Radially

.,-Adjusted Adjusted

'Measueid Saturated Saturated , Rection Actv-y A tiity Activity,. Rate Reaction Location (dpsg), (dps/g) ., (rpslatom) 63 Cu (n,a) 'Co Top 2.39E+05 3.27E+05 3.27E+05 4.99E-17 Center 2.17E+05 2.97E+05 2.97E+05 4.53E-17 Bottom 2.13E+05 2.92E+05 2.92E+05 4 45E-17 Average 4.66E-17 54Fe (n,p) 54Mn Top 2.13E+06 3.19E+06 3.19E+06 5.06E-15 Center 1.92E+06 2.88E+06 2.88E+06 4.56E-15 Bottom 1.9 1E+06 2.86E+06 2.86E+06 4.54E-15 Average 4.72E-15 58Ni (n,p) 58Co Top 8.37E+06 4.75E+07 4.75E+07 6.8 1E-15 Center 7.8 1E+06 4.44E+07 - 4.44E+07 6.35E-15 Bottom 7.75E+06 4.40E+07 4.40E+07 6.30E-15 Average 6.49E-15 238U (nf) 13 7 Cs (Cd) Middle 1.65E+06 I 6.25E+06 6.25E+06 4.11E-14 238U (nf) 137Cs (Cd) Including 235U, 239Pu, and y,fission corrections' 3.02E-14 237Np (nf) 137Cs (Cd) Middle 1.07E+07 4.06E+07 4.06E+07 2.59E-13 237Np (nf) 137Cs (Cd) Including -Yfission correction- 2.56E-13 59 60Co Top 4.42E+07 6.05E+07 Co (n,Y) 6.05E+07 3.95E-12 Bottom 4.44E+07 6.08E+07 6.08E+07 3.96E-12 Average 3.96E-12 59 Co (ny) 6OCo (Cd) Top 2.46E+07 3.37E+07 3.37E+07 2.20E-12 Middle 2.27E+07 3.1 lE+07 3.11E+07 2.03E-12 Bottom 2.40E+07 3.28E+07 3.28E+07 2.14E-12 Average 2.12E-12 Notes: 1) Measured specific activities are indexed to a counting date of September 20, 2002.

2) The average 238U(n,f) reaction rate of 3.02E-14 includes a correction factor of 0.76 to account for plutonium build-in and an additional factor of 0.97 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.56E-13 includes a correction factor of 0.99 to account for photo-fission effects in the sensor.

Appendix A

1.

A-18 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At the Surveillance Capsule Center Capsule U Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate MI/C M/BE 6 3 Cu(n,u)6OCo 5.52E- 17 4 81E-17 5.39E- 17 1.15 1.02 54 Fe(np) 4 Mn 5.89E- 15 5.45E- 15 5.89E- 15 1.08 1.00 5 8Ni(n,p)"Co 7.91E-15 7 65E-15 8.16E-15 1.03 0.97 238U(n,f) 37 Cs (Cd) 3.26E-14 2.96E-14 3.16E-14 1.10 1 03 2 37 Np(n'f) 37 1 Cs (Cd) 3.23E- 13 2.92E- 13 3.17E- 13 1.11 1.02 59 Co(nY)6 0Co 5.29E- 12 4.22E- 12 5.20E- 12 1.25 1.02 5 9 Co(n,7)'Co (Cd) 2.66E- 12 2.92E- 12 2.70E-12 0.91 0.98 Capsule Y Reac tion Rate[rpsa tom]

Best Reaction Measured Calculated Estimate M/C MI/BE 63 Cu(nax)6 0 Co 4.74E- 17 4.24E- 17 4.53E-17 1.12 1.05 54Fe(n,p) 54Mn 4.50E-15 4.68E-15 4.65E-15 0.96 0.97 "Ni(n,p) 5 8 Co 6.1 IE-l5 6.56E-15 6.45E-15 0.93 0.95 238 U(nf)'3 7 Cs (Cd) 2.84E-14 2.51E-14 2.52E-14 1.13 1.12 23 7 Np(nf)137 Cs (Cd) 2.73E-13 2.45E-13 2.62E-13 1.11 1.04 59Co(n,y)6 Co 4.15E- 12 3.48E-12 4.08E- 12 1.20 1.02 59Co(n,y)60 Co (Cd) 2.17E-12 2.42E-12 2.21E-12 0.90 0.98 Capsule V Reacton Rate r s/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 6 3Cu(n,a)6 Co 4.40E- 17 3.98E- 17 4.30E- 17 1.11 1.02

- Fe(n,p)ftMn 4.43E- 15 4.35E-15 4.64E-15 1.02 0.95 5 8Ni(n,p)"8 Co 6.45E-15 6.09E- 15 6.56E- 15 1.06 0.98 238U(n,f) 37 Cs (Cd) 3.01E-14 2.32E-14 2.57E-14 1.30 1.18 23 7Np(nf)'3 7 Cs (Cd) 2.67E-13 2.26E-13 2.60E-13 1.18 1.03 59 Co(n,y)6oCo 3.15E- 12 3.18E-12 3.12E- 12 0.99 1.01 5 9 Co(ny)ICo (Cd) 1.79E-12 2.21E-12 1.82E-12 0.81 0.98 Appendix A

A-19 Capsule X

.______ __ Reac ion Rate [rps/ tom]

Best Reaction Measured Calculated Estimate M/C MIBE 63Cu(n,a) 60 Co 4.6513-17 4.34E- 17 4.53E- 17 1.07 1.03 54 5 4 Fe(n,p) Mn 4.72E- 15 4.79E- 15 4.84E- 15 0.99 0.98 "Ni(n ,p)5 8 Co 6.49E- 15 6.71 E- 15 6.73E- 15 0.97 0.96 238 U(n,f) 37Cs (Cd) 2 37 Np(n 1f)3 7Cs (Cd) 3.01E-14 2.56E- 14 2.62E-14 1.18 1.15 59 Co(n,y)6oCo 2.56E-13 2.50E- 13 2.57E- 13 1.02 1.00 59 Co(n,-y) 6 0 Co (Cd) 3.99E-12 3.56E- 12 3.93E- 12 1.12 1.02 2.12E-12 2.47E1-12 2.16E-12 0.86 0.98 Appendix A

.1 A-20 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center 4v_> 1.0 MeY) Incm2-sl

-. Uncertainty.

'est Capsule ID Calculated E'stimateh BE/C U 9.40E+10 L.OOE+11 6% 1 06 Y 7 93E+10 8.02E+10 6% 1 01 V 7.31E+10 8.20E+10 6% 1 12 X 8.09E+10 8.31E+10 6% 1.03 Notes: 1) Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Iron AtomDisplacementRate [dpasl-Best Uncertainty Capsule 11- Calculated Estimate . H) BE/C U 1.82E-10 1.94E-10 8% 1.07 Y 1 53E-10 1.58E-10 8% 1 03 V 1 40E-10 1.58E-10 8% 1 13 X 1.55E-10 1 61E-10 8% 1.04 Notes 1) Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation penod.

Appendix A

A-21 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule U Capsule Y Capsule V Capsule X 6 3 Cu(n,a) 6 0Co 1.15 1.12 1.11 1.07 54Fe(n,p) Mn 1.08 0.96 1.02 0.99 5 8Ni(np) 5 8Co 1.03 0.93 1.06 0.97 2 38 U(np)13 7 Cs (Cd) 1.10 1.13 1.30 1.18 23 7Np(nf) 3 7Cs (Cd) 1.11 1.11 1.18 1.02 Average 1.09 1.05 1.13 1.05

% Standard Deviation 4.0 9.2 9.7 8.0 Notes: 1) The overall average M/C ratio for the set of 20 sensor measurements is 1.08 with an associated standard deviation of 8.2%.

Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID O(E > 1.0 MeV) dpa/s U 1.07 1.07 Y 1.02 1.03 V 1.14 1.12 X 1.04 1.03 Average 1.07 1.06

% Standard Deviation 4.9 4.1 Appendix A

__ 11 A-22 Table A-9 Current and Projected Neutron Fluences (E > 1.0 MeV) Experienced by the Intermediate and Upper Circumferential Welds Fluences In/cm 2 -sec)

Cumulative Circumferential Vertical Operations Time (EFPY) Intermediate Upper 00 300 13.83 7.97E+ 18 2.53E+ 17 4.33E+ 18 7.91 E+18 15.53 9.05E+18 2.96E+17 4.86E+18 8.88E+ 18 20.00 1.20E+ 19 4.OOE+ 17 6.12E+ 18 1.14E+19 24.00 1.47E+19 4.93E+ 17 7.26E+ 18 1.37E+19 32.00 2.01E+19 6.80E+17 9.53E+18 1.83E+19 40.00 2.54E+ 1 9 8.66E+ 17 1.18E+19 2.28E+ 19 48.00 3.07E+ 19 1.05E+ 18 1.41E+19 2.74E+19 54 00 3.47E+ 19 I.19E+18 1.58E+ 19 3.08E+19 I Upper Circumferential weld location at 235.97 cm above core centerline and at an azimuth of 450 to document the maximum neutron fluence.

2. Intermediate Circumferential weld location at -38.35 cm below core centerline and at an azimuth of 45° to document the maximum neutron fluence.

Appendix A

A-23 A.2 Appendix A References A-I. Regulatory Guide RG-l.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

A-2. REAC-SAP-172, "Analysis of Neutron Dosimetry from Wolf Creek - Capsules U, Y, and V,"

Perock, J. D. April, 1998.

A-3. A. Schmittroth, FERRETDataAnalysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

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

Appendix A

B-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS

• Specimen prefix "AL" denotes Lower Plate, Longitudinal Orientation

• Specimen prefix "AT" denotes Lower 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 D0-4000 00-n 3000 00 0

-J 2000.00-1000 00-000l ~~~~~~~it -.t--l 000 1 00 200 300 4 00 500 6 00 Time-1 (ms)

AL48, -500 F 5000 00 4000 00 as (5' 3000 00 0

-J m

2000.00 1000 Do 0 00 4 -

0.00 1 00 200 300 400 500 600 Time-I (ms)

AL57, 0F 5000 00-4000 00

.0

,, 3000 00-0

-J 2000 00 1000 00 200 300 600 Time-1 (ms)

AL53, 250 F Appendix B

~fiLl B-2 1

5000 00i 4000 00 1f l

.0

-jo 000 1 00 200 300 400 500 6 0o Time-1 (ms)

AL52, 40 0 F 1I 5000ooj 4000 00 I 40 m',3000 00 I 0

2000 001}

100000 0 00 0 00 100 200 Ann --

4- uu b6U 0 Tine-1 (ms)

AL56, 500 F oa3 0

-j 000 100 200 300 400 500 6 00 Time-1 (ms)

AL55, 75TF Appendix B

B-3 5000 00-4000 00 Z 300000 0

-j 2000 00 1000 00 0 00 000 100 2.00 300 400 500 600 Time-1 (ms)

AL59, 1100 F 5000 00 4000 00 300000 2000 00 1000 00 000 000 1 00 2 00 3 00 4 00 5 00 6 00 Time-1 (ms)

AL5O, 1350 F 5000 00 4000 00

= 3000 00 2000 00 X -

1000 00 I I I I IS 000 000 1 00 200 300 400 500 600 Time-i (ms)

AL60, 150TF Appendix B

'I B-4 n 3000 00 2000 00 1000 00.

0 00o 01DO 1 00 200 300 400 500 600 Time-I (ms)

.__ AL49, 175 0 F 5000 00 4000 00 Q .

• 3000 00 0,

2000 00 1000000 0000 0013 1 00 2 00 3 00 4 00 5 00 6 00 Time-I (ms)

AL51, 1900 F 5000 .o 4000 00

.0.

co 3000 00 0

-Jl 200000 100000 000 )

000 1 00 200 300 400 500 6 00 Time-I (ms)

AL46, 2250 F Appendix B

B-5 5000 00 400000 n 3000 00 20X0000 0 00 1.00 2.00 3 00 4 00 5 00 6 00 Time-i (ms)

AL54, 2500 F 50000Do 4000 00

'7 300000 2000DO0l 1000 00 l , , , , , =

000 000 1.00 200 300 400 500 60o Time-I (ms)

AL58, 2750 F 5000 00 0 c 0000 100 2.0 3 00 400 500 600 Time-1 (ms)

AL47, 300TF Appendix B

'I B-6 5000 00 4000 00 at Z- Ad u 3000-00 v 03 0 r 2000 00 1000 00 non1111 111111-ll ' ant + _ l _ t Am - A r - -sB - . - eL -- at - - It _ _

000 1 00 200 300 400 500 6 00 Time-i (ms)

AT46, -750 F 5000 00-4000 00

.0 3 3000 00-03

-J 2000 00-1000 00-0 00' 000 1 00 200 300 4 00 5 00 6 00 Time-1 (ms)

AT50, -250 F 3

0

-j 000 1Lo0 200 300 400 500 600 Time-i (ms)

AT60, 15-F Appendix B

B-7 5000 Co 4000 00 n 3000 00

-J 2000 00 1000 00 0 00 0 00 1 00 200 300 400 500 600 Time-1 (ms)

AT56, 50TF 5000 00-4000 00 x 3000 00-

-J 2000 00-1000 00 0 00 0

Time-1 (ms)

AT54, 750 F 5000 00 4000 00

-J 3000 00 2000 00 1 000 00 000 Il.O 2 00 3 00 4 00 5 00 600 Time-1 (ms)

AT53, 100 0 F Appendix B

'I B-8 5000 00 4000 00 3000 00 2000 00 1000 00 000 000 1 00 200 300 400 500 600 Time-i (ms)

AT59, 125SF 5000 00 4000 00 300000 2000 00 100000 0 00 000 1 00 200 300 400 500 600 Time-1 (ms)

AT58, 150 0 F 5000 00 4000 00

, 3000 00 2000 00 1000 00Il 0000 000 100 2 00 3 00 4 00 5 00 6 00 Time-I (ms)

AT48, 1750 F Appendix B

B-9 5000 00 4000 00 .

.0 3000.00 0

-. 1 2000 00 1000 00 0 00 I , , , , , ,

ha . . .

00i'0 1 00 2.00 3 00 4 00 50o 600 Time-1 (ms)

AT51, 1750 F 5000 00 4000 00 m 3000 00

-0 2000 00 1000 00 II

lilt ,-

000 1 00 200 300 4 00 500 6 00 Time-1 (ms)

AT57, 2000 F 5000 00 4000 00

,, 3000 00

-.1 2000 00

¶000 00' l UUt. . I I I I I I -1 1 .

1 .

000 1 00 2.00 3 00 400 500 600 Time-1 (ms)

AT52, 225 0F Appendix B

it B-10 5000 00 4000 00 n 3000 00 2000 00 1000 00 0 00 Time-1 (ms)

AT49, 250 0 F 5000 00 4000 00

.0.

as 3000 00 0

2000 00 1000 00 0 00 00 0 1 00 200 300 400 500 600 Time-I (ms)

AT47, 2750 F 5000 00 4000 00 30 .

-~3000 00.

0o ,

000 1 00 200 300 4 00 500 6 00 Time-I (ms)

AT55, 300 0 F Appendix B

B-Il 5000 OOf 4000 00 7 3000 001 0

-j 2000 00 1000 00 l 000 0 00 1 00 200 300 4 00 5 00 600 Time-1 (ms)

.. AW47, -750 F 5000 00l 4000 001 a,

-J 2000 00 1000 00 00c -o. I I-0.00 1 0o 200 300 4 00 5 00 600 Time-1 (ms)

AW52, -35 0 F 5000 00.

4000 00 A0 3000 00-0

-J 2000 00]

000 1 00 200 300 400 500 6 00 Time-I (ms)

AW53, 0F Appendix B

1.

B-12 4000 2

'o 3000 03

-J 2000 000 I 00 200 300 400 500 6 00 Time-1 (ms)

AW51, 25 0 F 5000 00 4000 00 n

'a 3000 00-0

-J 2000.00-1000 001 000 1 00 200 300 400 500 6 00 Time-I (ms)

AW58, 500 F 03

-J 2.00 3 00 600 Time-i (ms)

AW46, 75TF Appendix B

B-13 0

-J 0 00 1 00 2 00 3 00 4 00 5 00 6 00 Time-1 (ms) 5000 4000

.s 3000 00 0

-j 2000 00 1000 00 000 1 00 200 300 400 500 600 A,

Time-i (ms)

AW57, 125TF 0

-J 000 1 00 200 3.00 400 500 6 00 Time-I (ms)

AW55, 125TF Appendix B

II B- 14 5000 00 4000 00 c 300000 0

-J 2000 00 100000 0 00 000 1 00 200 300 400 500 600 Time-i (Ms)

AW48, 150 0 F 5000 00 400000

-3000 00 2000 00 100000 000 000 1 00 200 300 400 500 600 Time-I (ms)

AW60, 1600 F 500000o 400000

- 300000 2000 00 100000 000 000 1 00 200 300 400 500 600 rTme-i (Ms)

AW56, 200 0 F Appendix B

B-15

.0 0

-J 000 1.00 200 300 400 500 6 00 Time-i (ms)

AW5O, 2250 F 5000 00 4000 00 80

,, 3000 00-0

-j 2000 001 Time-1 (ms)

AW49, 250TF 4000 00 Z-a 3000 00 0

-J 2000 00 1000 00-0000 0.00 1 00 2.00 3 00 4 00 5 00 600 Time-1 (ms)

AW54, 2500 F Appendix B

II B-16 4000 00 n 3000 00 s

-J 2000 00 1000 00-0 oo4 0 (01 00 200 300 400 500 6 00 Time-I (ms)

AH53, -1750 F 5000 DE1 4000 00

- 3000 00

-J*.

2000 00 100000 0 001-00 0 1 00 200 300 400 500 600 Time-1 (ms)

AH51, -1000 F 5000 00 4000 00 r 300000]-

X, I o II 2000 0011 1000 00 000 1 00 200 300 400 500 6 00 Time-1 (ms)

AH50, F Appendix B

B-17 5000 00 4000 00 7 3000 00 0

2000 00O 1000 00 n nn

^ AA u uu - i - - -- - - -

0 00 1 00 200 300 400 5 00 6 00 Time-I (ms)

AH49, -500 F 5000 00 4000 00 2

.0 r 3000 00 0

-J 2000 00 1000.00-0 00 0X Time-i (ms)

AH58, -50F 5000 00 4000 00

( 3000 00 0

-J 2000 00 1000.00 uuu-i 000 1 00 200 300 4 00 5 00 6 00 Time-1 (is)

AH52, -25 0 F Appendix B

B-18

.0m 0

-j 000 1 00 200 300 400 500 600 Time-I (ms)

AH47, 0F 5000 00 4000 00 c, 3000 00 0

-2 2000 00 1000 00.

0 00 01 Time-1 (ms)

AH55, 0F 5000 00 4000 00

.\

m' 3000 00 0

-j 000 1 00 200 300 400 500 600 Time-I (ms)

AH59, 250 F Appendix B

B-19 n

0

-J Time-I (ms)

AH48, 350 F 5000 00 4000 00-n 3000 00 0

-J 2000 00-1000 co 000- I I I I

)o 100 2 00 3 00 400 500 6 00 Time-i (ms)

AH54, 500F 5000 00 4000 00 X5 3000 00 0

-J I

2000 00 1000 00 nnn. l , . ,

nil , , , , , ,

000 1 00 200 300 400 500 600 Time-1 (rns)

AH60, 1000 F Appendix B

'I B-20 5000 00 4000 00 D, 3000 00-0 000 1 00 200 300 400 500 6 00 Time-1 (ms)

AH46, 150 0 F 000 1 00 200 300 400 500 6 00 Time-1 (ms)

Q 0

-j 000 1 00 200 300 400 5 00 600 Time-i (ms)

AH56, 2000 F Appendix B

c-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Appendix C

C-l Contained in Table C-I are the upper shelf energy values used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 4.1. The definition for Upper Shelf Energy (USE) is given in ASTM E185-82, Section 4.18, and reads as follows:

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

If there are specimens tested in set of three at each temperature Westinghouse reports the set having the highest average energy as the USE (usually unirradiated material). If the specimens were not tested in sets of three at each temperature Westinghouse reports the average of all 100% shear Charpy data as the USE.

Hence, the USE values reported in Table C-1 and used to generate the Charpy V-notch curves were determined utilizing this methodology.

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

Table C-1 Upper Shelf Energy Values Fixed in C"VGRAPH Ift-lb] , ____-'_

Uaps e-._k

, u A; Material Unirradiated, apsule U Capsule Y Capsule V C sule X Lower Shell Plate 148 145 131 129 142 R2508-3 (Long.)

Lower Shell Plate 94 96 94 88 95 R2508-3 (Trans.)

Weld Metal 100 92 94 89 93 (heat # 90146)

HAZ Material 161 140 200 167 135 Appendix C

II CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09:13:31 on 12-12-2002 Page 1 Coefficients of Curve 1 I A = 72.09 B = 69.9 C = 85.66 T0 = 95.8 Equation is CVN = A + B I tanh((T - TO)/C I Upper Shelf Energy: 142 Fixed Temp. at 30 ft-lbs 36.1 Temp. at 50 ft-lbs: 67.7 Lower Shelf Energy 2.19 Fixed Material: PLATE SA533B1 Heat Number C4935-2 I Orientation: LT Capsule: X Total Fluence:

300 -- ---- . -

U] 250-200 10 15 0-V 0 0 i too0 100~l 5fF

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plant WC1 Cap: X Material: PLATE SA533BI OrL LT Heat II: C4935-2 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential

-50 2 6.69 -4.69 0 13 15.69 -2.69 25 21 24.66 -3 66 40 37 32.07 4.92 50 53 37.92 15.07 75 43 55.45 -12.45 110 74 8358 -958 135 108 102.02 5.97 150 133 11L23 2L76

• • • Data continued on next page "**

C-2

CAPSULE X (LONGITUDINAL ORIENTATION)

Page 2 Material: PLATE SA533BI Heat Number. C4935-2 Orientation: LT Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input CVN Energy Computed CVN Energy Differential 175 100 12298 -22.98 190 122 12804 -6.04 225 150 135.47 1452 250 146 13828 7.71 275 135 139.9 -4.9 300 137 140.2 -3.82 SUM of RESIDUALS = -86 C-3

II CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09.23:49 on 12-12-2002 Page 1 Coefficients of Curve 1 I A = 37.66 B = 36.66 C = 72.74 T0 = 7825 Equation is LK = A + B * [ tanh((T - TO)/C) I Upper Shelf L.E. 74.32 Temperature at LK 35: 72.9 Lower Shelf LE- I Fixed Material: PLATE SA533B1 Heat Number C4935-2 Orientation: LT Capsule: X Total Fluence:

200 ----

Wi 150 0

50 0

0 0 -- ---- --

-.jUU -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plantk WC1 Cap: X Material PLATE SA533B1 OriL LT H6it # C4935-2 Charpy V-Notch Data Temperature Input Lateral Expansion Computed LEK Differential

-50 0 3.09 0 5 -3.09 8.64 -3.64 25 12 14.77 40 24 -2.77 19.98 4.01 50 33 24.09 75 8.9 29 36.02 -7.02 110 47 52.71 135 -5.71 67 61.59 5.4 150 74 65.37 8.62 me Data continued on next page "**

C-4

CAPSULE X (LONGITUDINAL ORIENTATION)

Page 2 Material: PLATE SA533B1 Heat Number C4935-2 Orientation: LT Capsule: X Total Fluence Charpy V-Notch Data (Continued)

Temperature Input Lateral Expansion Computed LE Differential 175 64 69.52 -552 190 67 71.07 -4.07 225 71 73.04 -2.04 250 75 73.67 132 275 75 73.99 1 300 75 7415 B4 SUM of RESIDUALS =-3.79 C-5

CAPSULE X (LONGITUDINAL ORIENTATION)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09:2558 on 12-12-2002 Page 1 Coefficients of Curve I I A = 50 B = 50 C = 76.71 T0 = 106.64 Equation is Shear/ = A + B * [ tanh((T - T0)/C) I Temperature at 50x Shear 106.6 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: LT Capsule: X Total Fluence V]) 6aC

-4 4 2--3C

)O -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plant WC1 Cap: X MateriaL: PLATE SA533B1 Ori: LT Heat # C4935-2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-50 2 1.65 .34 0 5 5.4 -. 4 25 10 10.63 -.63 40 15 14.96 .03 50 20 18.59 L4 75 30 30.47 -.47 110 50 5.18 -218 135 65 67.68 -2.68 150 90 75.59 14.4

"" Data continued on next page ****

C-6

CAPSULE X (LONGITUDINAL ORIENTATION)

Page 2 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: LT Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input Percent Shear Computed Percent Shear Differential 175 75 85.59 -10.59 190 85 89.78 -4.78 225 100 95.62 4.37 250 100 97.67 2.32 275 100 98.77 122 300 100 99.35 .64 SUM of RESIDUALS = 255 C-7

II CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09.13:45 on 01-10-2003 Page 1 Coefficients of Curve I A= 4859 B = 46.4 C = 9028 T0 = 94.31 l Equation is CVN = A + B * [ tanh((T - TO)/C) I Upper Shelf Energy: 95 Fixed Temp. at 30 ft-lbs 55.9 Temp. at 50 ft-lbs 97 Lower Shelf Energy: 2.19 Fixed Material: PLATE SA533B1 Heat Number C4935-2 Orientation: TL Capsule: X Total Fluence:

Y) 25 10 A4m 2C0~

P-e

, 15 z0 Cz; 10 5

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plantl WCI Cap.: X Material PLATE SA533B1 OrL TL Heat A.C4935-2 Charpy V-Notch Data Tempera Lture Input CVN Energy Computed CVN Energy Differential

-75 5 433

-25 11 .66 8.36 2.63 15 15 15.85 50 30 -.85 27.49 25 75 41 38.82 100 52 217 51.51 .48 125 55 63.78 150 67 -. 78 74.06 -7.06 175 79 81.69 -2.69 Data continued on next page he c-8

CAPSULE X (TRANSVERSE ORIENTATION)

Page 2 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: TL Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input CVN Energy Computed CYN Energy Differential 175 98 8169 16.3 200 88 86.85 L14 225 91 9013 .86 250 93 9214 .85 275 101 93.33 7.66 300 96 94.03 1.96 SUM of RESIDUALS = 17.86 C-9

II CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09:33:46 on 12-12-2002 Page 1 Coefficients of Curve 1 I A = 32.72 B = 3L72 C = 942 TO = 95.62 Equation is: LE. = A + B I [ tanh((T - TO)/C) I Upper Shelf LE: 64.44 Temperature at LE 35: 102.4 Loower Shelf LE I Fixed Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: TL Capsule: X Total Fluence:

200-7 M

"-4

. F- 150 100 F-f CZ 0 0 -

P~ I 0

. X If __ I I I-300 Ir I - I

-200 -100 0 100 200 300 400 500 600 Temperature in I)egrees F Data Set(s) Plotted Plant WC1 Cap: X Material: PLATE SA533B1 Ori: TL Heat # C4935-2 Charpy V-Notch Data Temperature Input Lateral Expansion Computed LE Differential

-75 0 2.65 -2.65

-25 4 5.54 -1.54 15 8 10.7 -2.7 50 20 1845 L54 75 29 25.88 311 100 36 3419 1.8 125 38 42.3 -4.3 150 50 49.23 .76 175 53 54.51 -151

• • • Data continued on next page "**

C-1O

CAPSULE X (TRANSVERSE ORIENTATION)

Page 2 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: TL Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input lateral Expansion Computed LE. Differential 175 57 5451 2.48 200 51 582 -72 225 69 60.61 8.38 250 60 6213 -2.13 275 66 63.06 2.93 300 61 6362 -2.62 SUJM of RESIDUALS = -366 C-11

CAPSULE X (TRANSVERSE ORIENTATION)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09:36:03 on 12-12-2002 Page 1 Coefficients of Curve I A = 50 B = 50 C = 90.18 TO = 104.46 Equation is: Shear/ = A+ B * [ tanh((T - TO)/C) I Temperature at 50n/ Shear 104.4 Material: PLA1rE SA533BI Heat Number. C4935-2 0:rientation: TL Capsule: X Total Fluence:

.-4 ou a) 0 cD 60 4a)

- 40 2F

-30 DO

. a ,

-2U0

,\

-100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plant- WC1 Cap: X Materia1: PLATE SA533B1 Oi: TL Heat # C4935-2 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-75 2 1.83

-25 5 16 5.35 -.35 15 15 1208 50 2.91 25 23 L99 75 40 3421 100 5.78 45 47.52 -252 125 55 6118 150 -6.18 65 7329 4-29 175 60 8269 -Z69

'** Data continued on next page Ir' C-12

CAPSULE X (TRANSVERSE ORIENTATION)

Page 2 Material: PLATE SA533B1 Heat Number C4935-2 Orientatj ion: TL Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input Percent Shear Computed Percent Shear Differential 175 90 82.69 7.3 200 95 8926 5.73 225 100 93.54 6.45 250 100 9618 3.81 275 100 97.77 2Z2 300 100 98.7 129 Sul hi of RESIDUALS = 17.61 C-13

'I CAPSULE X (WELD)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 0916:18 on 01-10-2003 Page 1 Coefficients of Curve 1 l A = 47.59 B = 45.4 C =9538 TO = 49.68 Equation is CVN = A + B * [ tanh((T - TO)/C) ]

Upper Shelf Energy: 93 Fixed Temp. at 30 ft-lbs 10.6 Temp. at 50 ft-lbs 54.7 Lower Shelf Energy: 219 Fixed Material: WELD Heat Number WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

30 09-- -a C') 250-20f C-) 10 n (F -- -I

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plant. WC1 Cap.: X Material WELD Or. Heat lhWERE HI EAT NO.90146 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential

-75 4 839 -4.39

-35 13 1535 -235 0 23 25.88 -2188 25 37 361 .89 50 53 47.74 5.25 75 64 59.37 4.62 100 68 6954 -1.54 125 78 77.47 .52 125 76 77.47 -147 Data continued on next page *'*

C-14

CAPSULE X (WELD)

Page 2 Material: WE)LD Heat Number. WIRE HEAT NO.90146 0rientation:

Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input CVN Energy Computed CVN Energy Differential 150 67 8312 -16.12 160 89 84.82 4.17 200 84 8927 -5.27 225 102 90.75 1124 250 96 91.65 4.34 250 94 91.65 2.34 SiJM of RESIDUAIS = -.65 C-15

II CAPSULE X (WELD)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09.1850 on 01-10-2003 Page 1 Coefficients of Curve 1 I A = 31.67 B = 30.67 C = 75.77 T0 = 45 Equation is LE = A + B

• I tanh((T - TO)/C) I Upper Shelf LE 6234 Temperature at LE 35: 532 Lower Shelf LE. 1 Fixed Material: WELD Heat Number WIRE HEAT NO.90146 Orientation:

Capsule X Total Fluence 200 -- - --- . -

M) 150 pS--

10 0

50 _______ - - .

e9/

I - I 1-

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted PlantL WC1 Cap: X Material WELD Ori. Heat #: WIRE HEAT NO.90146 Charpy V-Notch Data Temperature Input Lateral Expansion Computed LK Differential

-75 0 3.47

-35 5 -3.47 7.62 -2.62 0 15 15.33 25 25 -.33 23.76 123 50 35 33.69 75 45 4322 1L3 100 51 177 50.7 125 53 55.72 .29 55 -2.72 55.72 -.72

• • Data continued on next page '"

C-16

CAPSULE X (WELD)

Page 2 Material: WELD Heat Number. WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input Lateral Expansion Computed LK Differential 150 54 58.73 -4.73 160 62 59.53 2.46 200 54 6L34 -7.34 225 64 6182 217 250 64 62.07 1.92 250 69 62.07 6.92 SULM of RESIDUALS = -386 I

I ' 1 C-17

II CAPSULE X (WELD)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09X.43 on 01-10-2003 Page 1 Coefficients of Curve 1 I A = 50 B = 50 C = 9L24 T0 = 21.09 Equation is Shear/ = A + B * [ tanh((T - TO)/C) I Temperature at 50z Shear 21 Materia] WELD Heat Number WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

pCi C-)

AH

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plantk WC1 Cap: X Material: WELD Ori Heat #: WIRE HEAT NO.90146 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential

-75 10 10X84 -.84

-35 20 2262 -2.62 0 45 38.64 6.35 25 50 5213 -213 50 60 65.33 -5.33 75 80 76.52 3.47 100 85 84.93 .06 125 95 90.69 43 125 95 9069 43 Data continued on next page -

C-18

CAPSULE X (WELD)

Page 2 Material: I ELD Heat Number. WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input Percent Shear Computed Percent Shear Differential 150 85 94.4 -9.4 160 95 95.45 -.45 200 95 98.05 -3.05 225 100 98.86 113 250 100 99.34 .65 250 100 99.34 .65 SUM of RESIDUAL' S -2.91 C-19

Ii CAPSULE X (HAZ)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10:32:26 on 12-12-2002 Page 1 Coefficients of Curve I I A = 68.59 B = 66.4 C = 56.86 TO = -36.56 l Equation is CVN = A + B

• I tanh((T - TO)/C) I Upper Shelf Energy: 135 Fixed Temp. at 30 ft-lbs -74.3 Temp. at 50 ft-lbs -52.9 Lower Shelf Energy: 2.19 Fixed Material: HEAT AFFD ZONE Heat Number: WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

300-

9) 250 10 p---q I

zq 200

-4 15 1 Q) 0 0 ___

0 I-

i 4

10. I C-)

50

.70 0- - -- - -I

--OW -)u -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set(s) Plotted Plant. WC1 Cap: X Material: HEAT AFFD ZONE OrL Heat # MWIRE HEAT NO.90146 Charpy V-Notch Data Temperat. ire Input CVN Energy Computed CVN Energy Differential

-175 6

-100 32 2.78 15 15.0 7

-75 34 -.07

-50 29.49 4.5 29 531 9

-50 88 -2419

-25 531 9 34B 50 0 92 81.9 1 -3L91 106. -1424 0 152 106.2 25 98 45.75 1213 -23.33 e*** Data continued on next page ***

C-20

CAPSULE X (HAZ)

Page 2 Material: HEAT AFPD ZONE Heat Number. WIRE HEAT NO.90146 Orientation:

Capsule X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input CVN Energy Computed CVN Energy Differential 35 132 125.08 6.91 50 140 128.96 11.03 100 146 133.91 1a0B 150 140 134.81 518 200 124 134.96 -10.96 200 29 134.96 -5.96 StJM of RESIDUAIS = 12.36 C-21

II CAPSULE X (HAZ)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10-3328 on 12-12-2002 Page 1 Coefficients of Curve I A = 36.74 B = 35.74 C = 58.51 TO = -24.75 Equation is: LE = A + B I tanh((T - TO)/C) I Upper Shelf LE. 7W49 Temperature at LE 35: -27.6 Lower Shelf LE.: I Fixed MateriaL: HEAT AFFD ZONE Heat Number WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

2007 15 P4 100 a) 5Y 0

5~~~01 0

-;UU -W0u -100 0 100 200 300 400 500 600 Temperature in Degi rees F Data Set(s) Plotted Plant WC1 Cap.: X Material HEAT AFFD ZONE Ori.: Heat # WIRE HEAT NO90146 Charpy V-Notch Data Temperature Input Lateral Expansion Compute d LE Differential

-175 0 L411

-100 5 -141 6.0, -1.07

-75 12 08

-50 12 .11 222 :1 -1021

-50 42 222 1

-25 19.78 21 3616 0 44 -15.6 510' 3 -7.03 0 70 510: 3 18.96 25 52 61.4' 5 -9.45

• • Data continued on next page

• C-22

CAPSULE X (HAZ)

Page 2 Material: HEAT AFFD ZONE Heat Number. WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input Lateral Expansion Computed LE. Differential 35 70 6428 5.71 50 70 67.34 2.65 100 70 715 -1.5 150 71 72.31 -1.31 200 72 72.46 -.46 200 74 72.46 153 SUM of RESIDUALS = .67 C-23

I, CAPSULE X (HAZ)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10.34:31 on 12-12-2002 Page 1 Coefficients of Curve I A = 50 B = 50 C = 52.52 TO = -1Z18 Equation is: Shear/ = A + B I [ tanh((T - T0)/C) I Temperature at 50. Shear -12.1 Material: HEAT AFFD ZONE Heat Number WYIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

DU I

30 C) 04 0xr

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F r Data Set(s) Plotted Plant WC1 Cap.: X Material: HEAT AFFD ZONE 4 Ori: Heat P WYIRE HEAT NO.90146 Charpy V-Notch Data Tempera ture Input Percent Shear Computed Percent Shear Diff erential

-175I 0 2

-100 I 2 -2 a4 -1.4

-75 5 8.37

-50 15 -337 19.15 -4.15

-50 30 1915

-25 1084 20 38.03 18.03

-25 50 61.39 11.39 50 100 61.39 65 38.6 80.47 15.47

'** Data continued on next page

• C-24

CAPSULE X (HAZ)

Page 2 Material: HEAT AFFD ZONE Heat Number. WIRE HEAT NO.90146 Orientation:

Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Temperature Input Percent Shear Computed Percent Shear Differential 35 75 85.77 -10.77 50 100 91.43 8.56 100 100 9862 137 150 100 99.79 .2 200 100 99.96 .03 200 100 99.96 .03 SUM of RESIDUAIS = -518 C-25

D-O APPENDIX D WOLF CREEK SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D

D-1 INTRODUCTION:

Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

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

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

EVALUATION:

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

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements", as follows:

"the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."

The Wolf Creek reactor vessel consists of the following beltline region materials:

• Intermediate Shell Plates R2005-1, 2, 3

• Lower Shell Plates R2508-1, 2, 3

• Intermediate & Lower Shell Longitudinal Weld Seams (Heat # 90146),

• Intermediate to Lower Shell Circumferential Weld Seam (Heat # 90146).

Appendix D

11 D-2 Per WCAP-10015, the Wolf Creek surveillance program was based on ASTM E185-79. When the surveillance program material was selected it was believed that copper and phosphorus were elements most important to embrittlement of the reactor vessel steels. Lower shell plate R2508-3 had the highest initial RTNDT and the lowest USE of all plate materials in the beltline region. In addition, lower shell plate R2508-3 had approximately the same copper and phosphorus content as the other beltline plate materials.

Therefore, based on the highest initial RTNDT and the lowest USE, lower shell plate was chosen for the surveillance program.

The weld material in the Wolf Creek surveillance program was made of the same wire as all the reactor vessel beltline welds, thus it was chosen as the surveillance weld material.

Hence, Criterion 1 is met for the Wolf Creek reactor vessel.

Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy of the Wolf Creek surveillance materials unambiguously. Hence, the Wolf Creek surveillance program meets this criterion.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 280 F for welds and 17TF for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 28TF for welds and less than 17TF for the plate.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2 In addition, the recommended NRC methods for determining credibility will be followed.

The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998.

At this meeting the NRC presented five cases. Of the five cases Case 1 ("Surveillance data available from plant but no other source') most closely represents the situation listed above for Wolf Creek surveillance weld metal. Note, for the plate materials, the straight forward method of Regulatory Guide 1.99, Revision 2 will be followed.

Appendix D

D-3 TABLE D-1 Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule Data Material 'Capsule Capsule j<) FF~b) ARF!() FF*ARTN,. . FFa Lower Shell U 0.316 0.684 36.46 24.94 0 468 Plate R2508-3 Y 1.19 1.05 16.03 16.83 1.10 (Longitudinal) V 2.22 1.22 52.03 63.48 1.49 X 3.49 1.33 61.06 81.21 1.77 Lower Shell U 0 316 0 684 23 79 16.27 0 468 Plate R2508-3 Y 1.19 1.05 35.39 37.16 1.10 (Transverse) V 2.22 1.22 54.53 66.53 1.49 X 3.49 1.33 53.96 71.77 1.77 SUM: 378.19 9.656 CFR2508 3 = X(FF

• ARTNDT) ( FF2 ) = (378 19) * (9.656) = 39.10 F Surveillance Weld U 0.316 0 684 27.21 18 612 0.468 Material Y 1.19 1.05 45.09 47.34 1.10 V 2.22 1.22 46.3 56.49 1.49 X 3.49 1.33 68.36 90.92 1.77 SUM: 213.362 4.828 CF Starr Weld =(FF

• ARTT)

• X( FF2 ) = (213.362) * (4.828) = 44.10 F 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(O28 .0 'logf .

(c) ARTNDT values are the measured 30 ft-lb shift values taken from Figures 5-1, 5-4 and 5.7, herein [IF]

The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2.

Appendix D

II D-4 Table D-2:

Wolf Creek Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials.

CF. Measured Predicted ScMtret 0 (alse Material -,-Capsule FF Metals)

(Slopeb.W ARTTNT ART T(I N2FK 7 2F(Weld)

Lower Shell Plate U 39.1 0.684 36.46 26.74 9.72 Yes R2508-3 Y 39.1 1.05 16.03 41.06 -25.03 No (Longitudinal) V 39.1 1.22 52.03 47.70 4.33 Yes X 39.1 1.33 61.06 52.00 9 06 Yes Lower Shell Plate U 39.1 0.684 23.79 26.74 -2.95 Yes R2508-3 Y 39.1 1.05 35.39 41.06 -5.67 Yes (Transverse) V 39.1 1.22 54.53 47.70 6.83 Yes X 39.1 1.33 53.96 52.00 1.96 Yes U 44.1 0.684 27.21 30.16 -2.95 Yes Vessel Beltline Y 44.1 1.05 45.09 46.31 -1.22 Yes Welds (Heat # 90146) V 44.1 1.22 46.3 53.80 -7.5 Yes X 44.1 1.33 68.36 58.65 9.71 Yes Table D-2 indicates that only one data point falls outside the +/- la of 17 0F scatter band for the lower shell plate R2508-3 surveillance data. One out of 8 data point is still consider credible. No weld data points fall outside the +/- 1a of 28 0F scatter band for the surveillance weld data, therefore the weld data is also credible per the third criterion.

Appendix D

D-5 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 250 F.

The capsule specimens are located in the reactor between the neutron pad and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wvall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 250 F. Hence, this criterion is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The Wolf Creek surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to the Wolf Creek surveillance program.

CONCLUSION:

Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B and 10 CFR 50.61, the Wolf Creek surveillance plate and weld data is credible.

Appendix D