ML031060076: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot change)
(StriderTol Bot change)
 
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:-                     WELF CREEK
{{#Wiki_filter:-
                                    '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
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:==
==Subject:==
Line 24: Line 25:
Surveillance capsule X was withdrawn from the Wolf Creek Generating Station reactor vessel on April 12, 2002, at the end of Refuel 12. The capsule was withdrawn to determine the vessel integrity after being subjected to neutron radiation exposure equivalent to the peak end-of-life (extended) fluence on the inside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.
Surveillance capsule X was withdrawn from the Wolf Creek Generating Station reactor vessel on April 12, 2002, at the end of Refuel 12. The capsule was withdrawn to determine the vessel integrity after being subjected to neutron radiation exposure equivalent to the peak end-of-life (extended) fluence on the inside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Appendix H to 10 CFR 50 requires that a report be submitted to the Nuclear Regulatory Commission for each capsule withdrawn.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.
Appendix H to 10 CFR 50 requires that a report be submitted to the Nuclear Regulatory Commission for each capsule withdrawn.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.
There are no commitments contained in this correspondence.
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
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-1 6028 Revision 0 March 2003 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         )
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
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
©2003 Westinghouse Electric Company LLC All Rights Reserved


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


==SUMMARY==
==SUMMARY==
..................                                                                                  ix 1    
ix 1  


==SUMMARY==
==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
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 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
.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)
iv LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the Wolf Creek Reactor Vessel Surveillance Materials (Unirradiated)..............................................................
(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)
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 1 0'9 n/cm2 (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 1 0'9 n/cm2 (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/cm2 (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/cm2 (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)
(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
(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/cm2 (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/cm2 (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.)
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
Table 5-11 Tensile Properties of the Wolf Creek Capsule XReactor Vessel Surveillance Materials Irradiated to 3.49 x 10'9 n/cm2 (E> 1.0MeV).....................................  
..................... 5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center.
6-12 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface.
6-16 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall.
6-20 Table 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
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.)
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) ...................................................
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  
5-30 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.................
                                                                                                                  ...... 5-31 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)..................
5-32 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).
5-33 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
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:
viii PREFACE This report has been technically reviewed and verified by:
Reviewer:
Reviewer:
Sections I through 5, 7, 8, Appendices B, C and D         A.R. Rawluszki Section 6 and Appendix A                                  D. M. Chapman
Sections I through 5, 7, 8, Appendices B, C and D Section 6 and Appendix A A.R. Rawluszki D. M. Chapman


ix EXECUTIVE  
ix EXECUTIVE  


==SUMMARY==
==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/cm2.
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.
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.
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    
1-1 1  


==SUMMARY==
==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:
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
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.
* 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 36.110F and an irradiated 50 ft-lb transition temperature of 67.750F. This results in a 30 ft-lb transition temperature increase of 61.061F and a 50 ft-lb transition temperature increase of 67.640F 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 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 55.970F 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 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 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.
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.
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
Summary of Results


1-2
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 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 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.
* 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:
* 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 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 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.
        -      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:
* 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
Calculated:
* 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:
Vessel inner radius* = 3.51 x 1019 n/cm2 Vessel 1/4 thickness = 2.09 x 10'9n/cm2 Vessel 3/4 thickness = 7.42 x 1018 n/cm2
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)
      *Clad/base metal interface. (From Table 6-2)
Summary of Results
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.
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.
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
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.
removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor vessel and discusses the analysis of the data.
Introduction
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.
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).
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.
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===
===Background===
 
4-1 4
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.
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).
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 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.
Line 123: Line 162:
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.
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:
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)
2.5% Ag, 97.5% Pb Melting Point: 579TF (3041C) 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590'F (3100C)
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule X is shown in Figure 4-2.
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule X is shown in Figure 4-2.
Description of Program
Description of Program


4-3 Table 4-1     Chemical Composition (wt%) of the Wolf Creek Reactor Vessel Surveillance Materials (Unirradiated)(Y)
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)
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:
R2508-3 C
(a)   Data obtained from WCAP-10015 and duplicated herein for completeness.
0.20 0.11 Mn 1.45 1.46 P
(b)  Weld wire Type B4, Heat Number 90146, Flux Type Linde 124, and Flux Lot Number 1061.
0008 0 005 S
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.
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)
(b)
Data obtained from WCAP-10015 and duplicated herein for completeness.
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
Description of Program


4-4 Table 4-2   Heat Treatment History of the Wolf Creek Reactor Vessel Surveillance Maierial cm   raur         -                  ea                 ! Poit~
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:
Lower Shell Plate Austemutized @
(a)     This table was taken from WCAP-10015t 4 ].
4 hrs.
(b)      The stress relief heat treatment received by the surveillance test plate and weldment have been simulated.
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)
(b)
This table was taken from WCAP-10015t4].
The stress relief heat treatment received by the surveillance test plate and weldment have been simulated.
Description of Program
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
4-5 Table 4-3 Chemical Composition (  
                                                ,........... :      ,  , ,pecimen
) of four Charpy Specimens from Wolf Creek Capsule X Concentration in Weight Percent Weld Metal Specimens  
                                                                            .. 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
-Base Metal S
,pecimen 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
4-6 Tble44 4
                                                            ~wAlly
C' Results fom L  
            ,,,3i',><,,   ,$<S tand       ard7st (wt                               *          '2'  "''
~wAlly Sti NIST Certified Reference7
nt       .        '"NIST           =             NIT-Ce'dMeasured                 Certified-       Measureds3 Al                 0.021         0.020           0.083             0.078 Co                 0.032         0.032           0.300             0.314 Cr                 0.694         0 705           0.300             0.310 Cu                 0 042         0.042           0.500             0.500 Fe               95.600         95.100         95.300           97.300 Mn                 0.660         0.662           1.040             1.070 Mo                 0.190         0.153         0.068             0.053 Ni                 2.000         2 080           0.590             0 608 P                 0.014         0.015           0.041             0.035 Si                 0 222         0.179           0.390             0 212 Sn                 0.010         0.011           0.016             0.018 Ti                 0 020         0.021           0.097             0.039 V                 0 011         0.013           0.040             0.040 Zr                 0.009         0.003           0.190             0.120 C                 0.383         0.380           0.160             0.160 S                 0.014         0 013           0.036             0.035 Description of Program
,,,3i',><,,  
,$<S tand ard7st (wt  
'2 '
nt  
'"NIST  
=
NIT-Ce'dMeasured Certified-Measureds3 Al 0.021 0.020 0.083 0.078 Co 0.032 0.032 0.300 0.314 Cr 0.694 0 705 0.300 0.310 Cu 0 042 0.042 0.500 0.500 Fe 95.600 95.100 95.300 97.300 Mn 0.660 0.662 1.040 1.070 Mo 0.190 0.153 0.068 0.053 Ni 2.000 2 080 0.590 0 608 P
0.014 0.015 0.041 0.035 Si 0 222 0.179 0.390 0 212 Sn 0.010 0.011 0.016 0.018 Ti 0 020 0.021 0.097 0.039 V
0 011 0.013 0.040 0.040 Zr 0.009 0.003 0.190 0.120 C
0.383 0.380 0.160 0.160 S
0.014 0 013 0.036 0.035 Description of Program


4-7 Table 4-S   Chemical Residts from Low Alloy Steel NIST Certified Reference lStandards (wt%)       .. ........
4-7 Table 4-S Chemical Residts from Low Alloy Steel NIST Certified Reference lStandards (wt%)
Element.                 NIST 363 ",                       NIST-364 Certified '     Measured             ied Al           0.240           0.306           0.008             0.013 Co             0.048           0.048           0.150             0.153 Cr           1.310           1.320           0.060             0.066 Cu             0.100           0.100           0.240             0.248 Fe           94.400           96.800           96.700           96.900 Mn             1.500           1.540           0.250             0.257 Mo             0.028           0.026           0.490             0.426 Ni             0.300           3.150           0.140             0.140 P             0.029           0.029           0.010             0.010 Si           0.740           0.485           0.060             0 058 Sn             0.104           0.101           0.008             0.008 Ti             0.050           0.054           0.240             0.237 V             0.310           0.313           0.100             0.107 Zr             0.049           0.067           0.068             0.034 C             0.620           0.610           0.870             0.850 S             0.007           0.006           0.025             0.023 Description of Program
Element.
NIST 363 ",
NIST-364 Certified '
Measured ied Al 0.240 0.306 0.008 0.013 Co 0.048 0.048 0.150 0.153 Cr 1.310 1.320 0.060 0.066 Cu 0.100 0.100 0.240 0.248 Fe 94.400 96.800 96.700 96.900 Mn 1.500 1.540 0.250 0.257 Mo 0.028 0.026 0.490 0.426 Ni 0.300 3.150 0.140 0.140 P
0.029 0.029 0.010 0.010 Si 0.740 0.485 0.060 0 058 Sn 0.104 0.101 0.008 0.008 Ti 0.050 0.054 0.240 0.237 V
0.310 0.313 0.100 0.107 Zr 0.049 0.067 0.068 0.034 C
0.620 0.610 0.870 0.850 S
0.007 0.006 0.025 0.023 Description of Program


4-8 REACTOR VESSEL
4-8 REACTOR VESSEL
                                                                    /-CORE BARREL
/-CORE BARREL
                                                                        -NEUTRON     PAD CAPSULE U 270*
-NEUTRON PAD CAPSULE U 270*
90*
90*
10 PLAN VIEW Figure 4-1     Arrangement of Surveillance Capsules in the Wolf Creek Reactor Vessel Description of Program
10 PLAN VIEW Figure 4-1 Arrangement of Surveillance Capsules in the Wolf Creek Reactor Vessel Description of Program


5-1 5       TESTING OF SPECIMENS FROM CAPSULE X 5.1     OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Center. Testing was performed in accordance with 10CFR50, Appendices G and H121, ASTM Specification E185-82491, and Westinghouse Procedure RMF 84 02E10], Revision 2 as modified by Westinghouse RMF Procedures 81021"1, Revision 1, and 8103('21, Revision 1.
5-1 5
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-10015 t 4 ]. No discrepancies were found Examination of the two low-melting point 5797F (304'C) and 590'F (310'C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 5791F (3041C).
TESTING OF SPECIMENS FROM CAPSULE X 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Center. Testing was performed in accordance with 10CFR50, Appendices G and H121, ASTM Specification E185-82491, and Westinghouse Procedure RMF 84 02E10], Revision 2 as modified by Westinghouse RMF Procedures 81021"1, Revision 1, and 8103('21, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-10015t 4]. No discrepancies were found Examination of the two low-melting point 5797F (304'C) and 590'F (3 10'C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 5791F (3041C).
The Charpy impact tests were performed per ASTM Specification E23-98t 131 and RMF Procedure 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 930-I instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix B), the load of general yielding (PGA), the time to general yielding (toy), the maximum load (PM), and the time to maximum load (tM) can be determined Under some test conditions, a sharp drop in load indicative of fast fracture was observed The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).
The Charpy impact tests were performed per ASTM Specification E23-98t 131 and RMF Procedure 8103 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 930-I instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix B), the load of general yielding (PGA), the time to general yielding (toy), the maximum load (PM), and the time to maximum load (tM) can be determined Under some test conditions, a sharp drop in load indicative of fast fracture was observed The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).
The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load (EM).
The energy at maximum load (EM) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load (EM).
The yield stress (cry) was calculated from the three-point bend formula having the following expression ayr= (PGor*L)/fB *(W- a)2 *C]                                               (1) where:     L     =     distance between the specimen supports in the impact machine B     =     the width of the specimen measured parallel to the notch W     =     height of the specimen, measured perpendicularly to the notch a     =     notch depth The constant C is dependent on the notch flank angle (f), notch root radius (p) and the type of loading (i.e.,
The yield stress (cry) was calculated from the three-point bend formula having the following expression ayr= (PGor*L)/fB *(W-a)2 *C]
(1) where:
L  
=
distance between the specimen supports in the impact machine B  
=
the width of the specimen measured parallel to the notch W  
=
height of the specimen, measured perpendicularly to the notch a  
=
notch depth The constant C is dependent on the notch flank angle (f), notch root radius (p) and the type of loading (i.e.,
pure bending or three-point bending). In three-point bending, for a Charpy specimen in which + = 450 and p = 0.010 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),
pure bending or three-point bending). In three-point bending, for a Charpy specimen in which + = 450 and p = 0.010 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),
Testing of Specimens from Capsule X
Testing of Specimens from Capsule X


5-2 ar=(PGy*L)/[B*(W-a)' $121]=(3.305 *PG,*W)/fB *(W-a)2]
5-2 ar=(PGy*L)/[B *(W-a)' $121]=(3.305 *PG,*W)/fB *(W-a)2]
(2)
(2)
For the Charpy specimen, B = 0 394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:
For the Charpy specimen, B = 0 394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:
cry = 33.3
cry = 33.3
* POy                                               (3) where cry is in units of psi and PGY is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.
* POy (3) where cry is in units of psi and PGY is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.
The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:
The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:
A=B*(W-a)=0.1241 sq.in.                                               (4)
A=B*(W-a)=0.1241 sq.in.
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance withASTM Specification E23-98 andA370-97aI'4 . The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
(4)
Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-99['5 1and E21-92 (1998)['61, and Procedure RMF 8102. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance withASTM Specification E23-98 andA370-97aI'4. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
Extension measurements were made with a linear variable displacement transducer extensometer.
Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-99['5 1 and E21-92 (1998)['61, and Procedure RMF 8102. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
The extensometer knife-edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 perASTM E83-93 117'.
Extension measurements were made with a linear variable displacement transducer extensometer. The extensometer knife-edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 perASTM E83-93 117'.
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 5500F. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +20F.
thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 5500 F.
During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +20 F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
Testing of Specimens from Capsule X
Testing of Specimens from Capsule X


5-3 5.2       CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule X, which received a fluence of 3.49 x 1019 n/cm 2 (E> 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.
5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule X, which received a fluence of 3.49 x 1019 n/cm2(E> 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:
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 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 67.750F. This results in a 30 ft-lb transition temperature increase of 61.06'F and a 50 ft-lb transition temperature increase of 67.640F 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 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 55.970F and an irradiated 50 ft-lb transition temperature of 97.04'F. This results in a 30 ft-lb transition temperature increase of 53.960F 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.
Irradiation of the weld metal (heat number 90146) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 10.660F and an irradiated 50 ft-lb transition temperature of 54.730F.
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.
This results in a 30 ft-lb transition temperature increase of 68.360F and a 50 ft-lb transition temperature increase of 75.380F Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-74.340F and an irradiated 50 ft-lb transition temperature of -52.920F. This results in a 30 ft-lb transition temperature increase of 69.660F and a 50 ft-lb transition temperature increase of 61.080F.
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 (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 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.
Line 195: Line 298:
Testing of Specimens from Capsule X
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:
It 5-4 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.
              -    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.
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 21'1 predictions led to the following conclusions:
            -    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 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 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.
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.
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 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.
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.
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 t2k The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule X materials is shown in Figures 5-13 through 5-16 and shows an increasingly ductile or tougher appearance with increasing test temperature.
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 data from WCAP-10015141, WCAP-1 1553"5' and WCAP-13365, Rev. 1[61 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.
Testing of Specimens from Capsule X
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.
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 (longitudinal orientation) indicated that irradiation to 3.49 x 1 09 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 dataE43. 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 lower Shell Plate R2508-3 (Transverse orientation) indicated that irradiation to 3.49 x 1 O'9 n/cm2 (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 results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 3.49 x 1019 n/cm2 (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.
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.
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
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)
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/cm2 (E> 1.0 MeV)
Sample             Temperature               Impact Energy       Lateral Expansion         Shear Number           OF           0C ft-lbs       Joules   mils           nun           %
(Longitudinal Orientation)
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
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)
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)
Sample           Temperature         J       Impact Energy     Lateral Expansion     Shear Number         OF     J   0 C           ft-lbs       joules   mils                   %
(Transverse Orientation)
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
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)
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           &deg;C         ft-lbs       Joules       mils         mm     %
Sample Temperature Impact Energy Lateral Expansion Shear Number OF  
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
&deg;C ft-lbs Joules mils mm AW47
-75  
-59 4
5 0
0.00 10 AW52  
-35  
-37 1 3 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 7 1 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)
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     %
Sample Temperature Impact Energy Lateral Expansion Shear Number OF 0C Ft-lbs Joules mils mm AH53
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
-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)     __,',__
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).
Churp           Normalized Energie Enry(lt/n)Yield'                                               'im n Test           .Load                                                                             FastY Timeto       MaxMaL "-,Max.,       Fracto< ?(Arrest Sample    Temp.  ,;          Charpy        Max.                                                                                  Yield,,:, Flow,,
(Longitudinal Orientation)
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)
Churp Normalized Energie Enry(lt/n)Yield'  
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
'im n FastY Test  
.Load Timeto MaxMaL "-,Max.,
Fracto<  
?(Arrest Yield,,:,
Flow,,
Sample Temp.
Charpy Max.
Prop,,
PPY 3Yield tGo' 3
P M, IM LoadPr Lo d P Stress, ;
Stress 0o.
(F)
E/kA, b'
Qb' (msec)
(Ib)
(mssec)',
(Ib s(Ib)
"',.ksi) k, 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
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)     >___>__
-.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)
chiarpy       No tlized EftergiesYedTm                                 t   as.
(Transverse Orientation) chiarpy No tlized EftergiesYedTm t
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)
as.
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
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)
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
CharpY..
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-
Normalized Energies Energy:ftbi) ed Time to I Fast l'Test YLiad lTim to M
    -- , -- -1.        -- ---            -
M I
A Yeld Flow p
Te Charpy Max."
Prp PGYrop.
Yield t y Load PM t
jLoadPF 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
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,,
[Table 5-8 Instrumented Charpy, Impact Test Results for the Wolf Creek Heat-Affected-Zone (HAZy Metal 1'd;Sh~d f*~
hP,.           Normlized .nr..e.
I,6.fl is, iii'-,.i n Itxi,,
            ')        sa: emerge~o>>ss3?>Tet                                            ;  ?'T~ t o        fv~ ls  Mal.toC, Tii"        Vinci  i Fa': tl!';t,, Arrest Yield v      ; . Flow,
hP,.
* Charpy         MaL       PrPop,             GYI I   Yield tGV  Lad P.         M     I AS ad a PF          LoaI Stress ay     Stres (3F)     (   llNo.
Normlized.nr..e.
lb) E3WA             EA     , jA               Qb)     ,      P   *(b),       .            i)           *ksi3(k~i)
emerge~o>>ss3?>  
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
')
sa:
Tii"
: toC, Fa':  
?
t ls i tl!';t,,
v Tet
'T~ o fv~
Mal.
Vinci Arrest Yield
: Flow, Charpy MaL
: PrPop, GYI I GV I AS Yield t Lad P.
a M
F ad P LoaI Stress ay Stres llNo.
(3F)
( lb) E3WA EA,
jA Qb)
P  
*(b),
i)
(k~i)  
*ksi3 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
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
.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().
,SurvFeillance Mate'r's().
                                                                                                            '      . "    ,    '"..y   ,,,,',
y  
Averag   30Q (t-lbAverage ft-lb)             Avcrage 3 mI l       er 1&deg;               AveragecS50 ft.Ibj ft-lb(,                         Energ Absorption~a Matenal               Transition Temperature (,)         Expansion Temperature (0       5) Transition Temperature                             at Full Shear (ft-Jb)
, Averag Q ft-lb)
      '5l:5555,5,',
Avcrage 3 mI l er 1&deg; AveragecS ft-lb(,
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.)
50 ft.Ibj 30 (t-lbAverage Energ Absorption~a Matenal Transition Temperature (,)
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.)
Expansion Temperature (0
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)
: 5) Transition Temperature at Full Shear (ft-Jb)
HAZMetal                       -144.01     -74.34       69.66     -89.86       -27.62     62.24       -114           -52.92         61.08       161               135       -26
Unrdae Iid
: 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)
'e ATi:
: 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).
Unirradiated Irradiated, AT, I nin-adiated Irradiated, AT Umaiae rradiated E
'5l:5555,5,',
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
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:
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......
E   0l.MeV)                 ____
30 ft-lb Transition TUpper Shelf Energy e,,,._jTemperature Shift' Decrease Material  
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:
'Capsule FluencyPredicted 1
(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.
Measured  
(b)   Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 (See Appendix C)
-' Predicted Measured:
(c) Values are based on the definition of upper shelf energy given in ASTM E185-82 (d) The fluence values presented here are the calculated values, not the best estimate values.
E 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
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
5-16 Table 5-11 Tensile Properties of the Wolf Creek Capsule X Reactor Vessel Surveillance Material Trrafllnted hr i
                                                                                                                                  ,,,-,l2 {. s 1 n .
49 x AQ l 9,,,-,l2 {. s 1 n Material Sample Test 0.2%
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)           (%)             (%)           (%)
Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp.
__              (ksi)
Yield Strength Load Stress (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.)
Strength Elongation Elongation in Area (OF)
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.)
Strength (ksi)
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
(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)
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                                 -
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
1.0 4-,F           ___
0 219 0
I 0__ 10 -3000           20-0 0-                 0 0
148 0  
                  ~177
-2495 0
                  -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#
11 0
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)
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  
-3000 20-0 0
10 0-0 I
0__
~177
-300  
-200  
-100 0
100 Temperature in Curve Leged I 0-~
2 0-----
a 0
~
4 Data Setgs) Plotted 200 300 400 500 600 Degrees F i
5
-~
Curve Plant Capsule Material I
WCI UNIRR PLATE SA533BI 2
C!C U
PLATE SA533BI 3
WfCI Y
PLATE SA533BI 4
W'I V
PLATE SA533BI 5
iC!
X PLATE SA533BI OriL Heat#
LT C4935-2 LT C4935-2 LT C4935-2 LT C4935-2 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
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)
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
Testing of Specimens from Capsule X


5-19 LOWER SHELL PLATE R-2508-3 (LONGITUDINAL)
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
CYGRAPH 41 Hyperbolhc Tangent Curve Printed at 095013 on 01-10-20)03 Results Curve Fluence T o 50z Shear d-T o 50%. Shear I
              -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
0 3843 0
_. 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)
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 Temperature in Degrees F Cune legend 120-3 0 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 2 500 600 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
Testing of Specimens from Capsule X


5.20 LOWER SHELL PLATE R2508-3 (TRANSVERSE)
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 .^ ^
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
I U
4..W U
94 2
0 219 0
96 3
0 219 0
94 4
0 219 0
88 5
0 219 0
95 U
2 2
25B 0
3739
-6 5654 1
55X97 0
34.32 23.79 59.55 35.39 8149 54.53 90.59 53.96 97.04 0
2523 4716 5627 62.71 C1 7
a)
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
-30  
                                                                                            *vtI~
-200  
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)
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10 20---
3 0 4 ^-
5 Data Set(s) Plotted Matenal Curve Plant Capulle Ori HtIPf r---n -
Tlva  
*vtI~
I OC!
UNIRR PLATE SA533BI 2
TCI U
PLATE SA533BI 3
TO Y
PLATE SA533BI 4
WT V
PLATE SA533B1 5
TI X
PLATE SA533BI TL C4935-2 TL C4935-2 TL C4935-2 TL C4935-2 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
Testing of Specimens from Capsule X


5-21 LOWER SHELL PLATE R2508-3 (TRANSVERSE)
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       &deg;       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_
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  
          ~1007-CO     50
&deg; 605 2
              -300     -200       -100           0         100         200       300         400   500 600 Temperature in Degrees F Curve legend I   O-           2 0--               30                    4   -              5         -
0 72B6 3
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)
0 7548 4
0 61.41 5
0 64.44 0
481 743
-663
-a6 2544 357 67.84 9a79 1024 0
1025 42.39 68.34 7695 1507_
~1007-CO 50
-300  
-200  
-100 0
100 200 300 400 Temperature in Degrees F Curve legend I
O-2 0--
3 0 4
5 500 600 Data Set(s) Plotted Material Curve Plant Capsule Ori Heatf I
IC UNIRR 2
lCI U
3 W!a Y
4 WC!
V 5
WC!
x PLATE SA533BI PLATE SA533BI PLATE SA533BI PLATE SA533BI PLATE SA533BI TL C4935-2 TL C4935-2 TL C4935-2 TL C4935-2 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
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)
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
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--
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
              -300       -200       -100           0       100           200         300         400     500       600 Temperature in Degrees F Curve legend Io      -          2 0--             30                     4     -              5 V.-
0 219 0
Data Set(s) Plotted Curve       Plant     Capsule           Material              Ori   Heat#
100 0  
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
-57.69 2
0 219 0
92  
-8  
-3047 3
0 219 0
94  
-6  
-1259 4
0 2.19 0
89  
-Ll  
-1136 5
0 219 0
93  
-7 10.66 0
-20.64 2721 644 45.09 2082 46.33 3179 68.36 5473 0
27.09 41.47 5Z44 7538 a) z 0--
-300  
-200  
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve legend I o 2 0--
30 4
5 V.-
Data Set(s) Plotted Material Curve Plant Capsule Ori Heat#
I WlC UNIER 2
WC]
U 3
WC1 Y
4 WC1 V
5 lCI x
WD van Wm Wma Wm WIRE HEAT NO.90146 WIRE HEAT NO.90146 WIRE HEAT NO.90146 WMRE HEAT NO.90146 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
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 T o LE35 d-T o LE35 0
              -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
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----
3 0 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)
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 2
0 3
0 4
0 5
0
-7394 20.94 2076 2109 0
94B9 55.03 9471 9504 U) a)
C) 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#
-300  
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
-200  
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1
2 C---
3 0 4 ^
5 Data Set(s) Plotted Curve Plant Capsule Material Or.
Heat#
I UC UNIRR 2
TOC U
3 WCI Y
4 WTOI V
5 KCI X
WED WELD WmL WmL WIRE HEAT NO.90146 WIRE HEAT N090146 TIRE HEAT NO.90146 WIRE HEAT NO.90146 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
5-26 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)
5-27 HEAT AFFECTED ZONE CVGRAPH 41 Hyperbolic Tangent Curve Printed at 10(3855 on 01-10-2003 RBults d-USE T o LE35 d-T o LE35 Curve nuence USE I
0 8466 0  
-896 2
0 6726 26  
-54.79 3
0 97.96 1329  
-6051 4
0 6114  
-16.51  
-436 5
0 72.49  
-l116  
-2762 0
35.07 29.35 4626 6224 U)
P--
P--
Ct
Ct
    ;4
;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
-300  
-200  
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve legend I
o-20--
3 0 4 -
5 Data Set(s) Plotted Material Curve Plant Capsule 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 .%-<
5-28 HEAT AFFECTED ZONE CVGRAPH 41 Hyperbolic Tangent Curve Pnnted at 1a4139 on 01-10-2003 Results Curve Fluence T o 50z. Shear
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.
.I-T A 5rdx.%-<
I 0
2 0
3 0
4 0
5 0
T 0 50z. Shr
-77.81
-20.47
-47.81
-30.46
_m2s 0
57.34 30 47.34 65.62 C.
U)
U)
C.)
C.)
I
I
              -300     -200       -100             0         100           200     300         400       500     600 Temperature in Degrees F Curve legend I0              2 0--                   30                      4^                 5     ~
-300  
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
-200  
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve legend I 0 2 0--
3 0 4 ^
5  
~
Data Set(s) Plotted Ilaterial Curve Plant Capsule Ori.
Ueat Material On llptI I
WC1 UNIRR 2
lC!
U 3
lCI Y
4 lCI V
5 liCI X
HEAT AFFD ZONE HEAT AFFD ZONE HEAT AFF'D ZONE HEAT AFFD ZONE HEAT AFFD ZONE WERE HEAT NO.90146 lIRE HEAT NO.S0146 WME HEAT NO.90146 WIRE HEAT NO090146 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)
5-29 AL48,-50F AL57, 00F AL53, 250F AL52, 400F AL56, 500F AL55, 750F 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
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)
5-30 AT46,-750F AT50,-250 F AT60,15-F AT56, 500 F AT54, 750F 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
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
5-31 AW47, -75 0F AW52, -350F AW53, 0F AW51, 250F AW58, 500F AW46, 750F AW59, 1000F AW57, 1250F AW55, 125TF AW48, 150TF AW60, 160TF AW56, 2000F AW50,2250F AW49, 2500F AW54, 2500F 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
5-32 AH53, -1750F AH51, -1000F AH50,-750 F AH49,-500F AH58,-500F AH52, -25 0F 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
5-33 (O C) 0 50 100 150 200 250 300 120 110 100 Cn 90 l 80 La 70 C-,
              -A           A         =                           A_     600  Cs l 80 La
60 50 40 I
    - 70        -_                          AA                            500 C-,
I I
60            0                     0                     0 400 0                         -
I I
50          0.2% YIELD STRENGTH 300 40 LEGEND:
I l_
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
ULTIMATE TENSILE STRENGTH
    >-   50 1-
-A
    "_. 40 3 30 20 10 0
-A A  
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)
A_
A A 0
0 0
0 0.2% YIELD STRENGTH 800 700 600 Cs 500 400 300 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
Figure 5-17 0
100 200 300 400 500 600 TEMPERATURE (OF)
Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Longitudinal Orientation)
Testing of Specimens from Capsule X
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                         _
5-34 (O C) 0 50 100 150 200 250 300 120 110 100 C-1 La I-~
La I-~  70                                                                          500 60                  0                 0                          0
90 80 70 I
                        %Y         S2                                               400 50        0 2% YIELD STRENGTH 300 40 LEGEND:
I I
A 0 UNIRRADIATED 19     2           0
I I
                  *.      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
I I_
_                                          A A
I&
50 W-4
ULTIMATE TENSILE STRENGTH 2
    -J 0-4 40 C-,
A-2 0
30   _      o                     TOTAL ELONGATION           2 0
0 0
10 UNIFORM ELONGATION 0              I           I         I           I       I 0           100       200       300         400     500         600 TEMPERATURE (OF)
% Y S2 0 2% YIELD STRENGTH 800 700 600 500 400 300 60 50 40 LEGEND:
Figure 5-18   Tensile Properties for Wolf Creek Reactor Vessel Lower Shell Plate R2508-3 (Transverse Orientation)
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 I-1 W-4
-J 0-4 C-,
60 50 40 30 2
REDUCTION IN AREA A
A A
o TOTAL ELONGATION 2
0 UNIFORM ELONGATION I
I I
I I
10 0
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
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-,
5-35 (0C) 0 50 100 150 200 250 300 120 110 100 90 80 70 It Lin C-,
70                                                                    500 60  -        0.2% YIELD STRENGTH                         2/
I I
400 50 300 40 LEGEND:
I I
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-
I I
  -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)
I ULTIMATE TENSILE STRENGTH A  
't 0.2% YIELD STRENGTH 2/
800 700 600
-d 0l-60 50 40 500 400 300 LEGEND:
A 0 UNIRRADIATED 19 2
0 As IRRADIATED TO A FLUENCE OF 3.49 X 10 nlcm (E>1.OMeV) AT 550 F I-1 t-40-J-
-j I--
80 70 60 50 40 30 REDUCTION IN AREA TOTAL ELONGATION 2
UNIFORM ELONGATION
*I I
I I
I 20 10 0
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
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)
Il 5-36 Specimen ALI0 Tested at 750F Specimen AL 1I Tested at 3 000 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
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)
5-37 Specimen ATI 0 Tested at 750F Specimen AT 1 Tested at 3000F 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
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
5-38 Specimen AWIO Tested at 750F Specimen AW 1I Tested at 300'F Specimen AW12 Tested at 5500F 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
5-39 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.X-100 90 80 70 y
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
60 to I
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)
40 30 20 10 0
AL-10 75 F 0
0.05 0 1 0.15 STRAIN, INAN 0.2 025 03 100 90 80 70 y
60 co50 Uj 40 30 20 10 0
STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.X' AL-11 300 F 0
005 01 0.15 STRAIN, INAN 0.2 025 0.3 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
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
5-40 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.-X" 100' 90 80-70-Se 60-LO 50-w o
0          005         01         0 15             02 025 03 STRAIN, INAN Figure 5-23        Continued Testing of Specimens from Capsule X
40 30 20 10 0
0 Figure 5-23 AL-12 550 F 005 01 0 15 STRAIN, INAN 02 025 03 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,
5-41 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE *X-100 90 80 70 e
ED 50 40 AT-10 30 75 F 20 10 0
60 U,
0        005         01             015           02     025           03 STRAIN, INAN STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE X" 100 -
ED 50 40 30 20 10 0
AT-10 75 F 0
005 01 015 02 025 STRAIN, INAN 03 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE X" 100 -
90 -
90 -
80 -
80 -
70 -
70 -
i   60   -
i 60 -
(6 co   50 -
(6 co 50 -
I   40 -
I 40 -
AT-1 1 30 -                         300 F 20 -
30 -
20 -
10 -
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)
0-AT-1 1 300 F 0
0.05 0.1 0.15 STRAIN, IN/IN 02 0.25 03 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
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
.1 5-42 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE"X" 100 90 80 70 u4 60 CO50 w
CO50 40 AT-12 30 550 F 20 10 0
40 30 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
AT-12 550 F 0
005 0.1 0.15 0.2 0.25 STRAIN, INAN 0.3 Figure 5-24 Continued Testing of Specimens from Capsule X


543 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE *X' 100 90 80 70
543 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE *X' 100 90 80 70 to~4 60 to 50 w
    ~4to  60 to 50 w
,,, 40 30 20 10 0
    ,,, 40 AW-10 30 75 F 20 10 0
AW-10 75 F 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
005 0.1 0.15 0.2 025 03 STRAIN, ININ 100 90 80 70 60 t) 50 0
  $      60 t)    50 400 30                          AW-11 300 F 20 10 0
40 30 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
0 STRESS-STRAIN CURVE BEAVER VALLEY UNIT 2 W CAPSULE AW-11 300 F 005 01 0.15 STRAIN, ININ 02 025 03 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
5-44 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE X' 100 90 80 70 9
0           0 05        01            0.15           0.2 0.25 03 STRAIN, ININ Figure 5     Continued Testing of Specimens from Capsule X
60 to 50 40 30 20 10 0
AW-12 550 F 0
0 05 0 1 0.15 0.2 0.25 STRAIN, ININ 03 Figure 5 Continued Testing of Specimens from Capsule X


6-1 6       RADIATION ANALYSIS AND NEUTRON DOSIMETRY
6-1 6
RADIATION ANALYSIS AND NEUTRON DOSIMETRY


==6.1 INTRODUCTION==
==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.
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.
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.
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."
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]
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
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.
.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&deg;, 121.50, 238.50, and 301.5&deg; (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.
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&deg;, 121.50, 238.50, and 301.5&deg; (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.
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.
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 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:
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, 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 &#xa2;(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.
* 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 &#xa2;(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 --
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 --
Line 459: Line 1,060:
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 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.
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.
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 library1241. 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
Radiation Analysis and Neutron Dosimetry


Line 473: Line 1,074:
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.
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.
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
Reaction'Rates (rps/atoin)
                        ' 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
MC Reaction Measued Calculated Ratio 63Cu(n,a)6 Co 4.65E-17 4.34E-17 1.07
                                                        % 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
' 4Fe(n,p)54Mn 4.72E-15 4.79E-15 0.99 58Ni(n,p)"Co 6.49E-15 6.71E-15 0.97 23SU(np)l 7Cs (Cd) 3.01E-14 2.56E-14 1.18 27Np(n,f)'3 7Cs (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.
+/- 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.
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
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).
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 -
2 - Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.
Comparison of calculations with benchmark measurements from the Pool Cntical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).
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.
2 -
4 - Comparisons of the plant specific calculations with all available dosimetry results from the Wolf Creek surveillance program.
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 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 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.
Line 490: Line 1,098:
Radiation Analysis and Neutron Dosimetry
Radiation Analysis and Neutron Dosimetry


6-7 Capsule       Vessel IR PCA Comparisons                                                 3%             3%
6-7 Capsule Vessel IR PCA Comparisons 3%
H. B. Robinson Comparisons                                       3%             3%
3%
Analytical Sensitivity Studies                                 10%             11%
H. B. Robinson Comparisons 3%
Additional Uncertainty for Factors not Explicitly Evaluated     5%             5%
3%
Net Calculational Uncertainty                                   12%             13%
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.
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.
Therefore, the resultant uncertainty was random and no systematic bias was applied to the analytical results.
Line 500: Line 1,113:
Radiation Analysis and Neutron Dosimetry
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
--- - - 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 D
  -c R Axis (cm)
-c DORT Geometry R Axis (cm)
Radiation Analysis and Neutron Dosimetry
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
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
.Z I
.      C C,
C C,
R Axis (cm)
R Axis (cm)
Radiation Analysis and Neutron Dosimetry
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
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
        ,  O R Axis (cm)
R Axis (cm)
Radiation Analysis and Neutron Dosimetry
Radiation Analysis and Neutron Dosimetry


Line 518: Line 1,131:


6-12 Table 6-I Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)
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]
Cycle 1
Time                  [n/cmA2-sec]
2 3
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.
4 5
6 7
8 9
10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time
[EFPY]
1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00 Flux
[E>1.0 MeV]
[n/cmA2-sec]
29 Deg 31.5 Deg 8.73E+10 9.33E+10 9.08E+10 1.OOE+11 7.67E+10 8.33E+10 7.30E+10 8.04E+10 7.12E+10 7.60E+10 6.66E+10 7.05E+10 6.44E+10 6.98E+10 7.44E+10 7.91E+10 6.22E+10 7.1 OE+10 8.19E+10 8.64E+10 7.23E+10 8.25E+10 7.29E+10 8.29E+10 6.97E+10 7.79E+10 7.29E+10 8.29E+10 7.29E+10 8 29E+10 7.29E+10 8.29E+10 7.29E+10 8.29E+10 7.29E+10 8.29E+10 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
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)
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]
Cumulative Fluence Operating
Time                   [n/cmA2-sec]
[E>1.0 MeV]
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
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]
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
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
[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]
6-15 Table 6-1 cont'd Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENTS Cycle 1
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.
2 3
4 5
6 7
8 9
10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time
[EFPY]
1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00 Displacements
[dpa]
29 Deg 31.5 Deg 5.79E-03 6.1 8E-03 9.61 E-03 1.04E-02 1.28E-02 1.39E-02 1.79E-02 1.95E-02 2.32E-02 2.52E-02 2.74E-02 2.96E-02 3.25E-02 3.52E-02 3.80E-02 4.1 OE-02 4.32E-02 4.69E-02 4.99E-02 5.40E-02 5.59E-02 6.09E-02 6.20E-02 6.78E-02 6.96E-02 7.63E-02 8.96E-02 9.91 E-02 1.08E-01 1.19E-01 1.43E-01 1.60E-01 1.79E-01 2.01 E-01 2.15E-01 2.42E-01 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
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]
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
Time                       [n/cmA2-sec]
[E>1.0 MeV]
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
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]
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 Cycle 1
Time                         [n/cmA 2]
2 3
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
4 5
6 7
8 9
10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time
[EFPY]
1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00 0 Deg 4.26E+1 7 7.11E+17 9.50E+1 7 1.39E+18 1.84E+18 2.14E+18 2.45E+1 8 2.80E+18 3.14E+18 3.55E+1 8 3.95E+1 8 4.33E+18 4.86E+1 8 6.12E+18 7.26E+1 8 9.53E+1 8 1.18E+19 1.41 E+19 1.58E+19 Maximum Pressure Vessel Fluence
[E>1.0 MeV]
[n/cmA 2]
15 Deg 30 Deg 6.32E+17 7.36E+17 1.04E+18 1.21 E+18 1.39E+18 1.62E+18 1.99E+18 2.27E+18 2.63E+18 2.96E+18 3.14E+18 3.50E+18 3.66E+18 4.14E+18 4.28E+18 4.85E+18 4.74E+18 5.53E+18 5.40E+18 6.36E+18 5 96E+18 7.13E+18 6.54E+18 7.91 E+18 7.31 E+18 8.88E+18 9.23E+18 1.14E+19 1.09E+19 1.37E+19 1.44E+19 1.83E+19 1.78E+1 9 2.28E+1 9 2.12E+19 2.74E+19 2.38E+19 3.08E+19 45 Deg 7.44E+17 1.29E+18 1.69E+18 2.37E+1 8 3.03E+18 3.55E+18 4.18E+18 4.79E+18 5.54E+1 8 6.33E+1 8 7.19E+18 8.1 OE+18 9.20E+1 8 1.22E+19 1.49E+19 2.03E+1 9 2.57E+19 3.11E+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]
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 Cycle 1
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
2 3
4 5
6 7
8 9
10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time
[EFPY]
1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32 00 40 00 48.00 54.00
[dpa/sec]
Maximum Iron Atom Displacements 0 Deg 1.95E-11 2.16E-1 1 1.75E-11 1.90E-1 1 1.79E-11 1.48E-11 1.27E-11 1.46E-1 1 1.21 E-11 1.55E-11 1.43E-1 1 1.40E-1 1 1.45E-1 1 1.40E-11 1.40E-1 1 1.40E-11 1.40E-11 1.40E-1 1 1.40E-1 1 15 Deg 2.87E-1 1 3.06E-1 1 2.51 E-11 2.58E-1 1 2.56E-1 1 2.44E-1 1 2.08E-11 2.54E-1 1 1.68E-1 1 2.40E-1 1 2.02E-1 1 2.1 OE-11 2.1 OE-1 1 2.1OE-1 1 2.1 OE-11 2.1OE-1 1 2.1 OE-11 2.10E-1 1 2.10E-1 1 30 Deg 3.35E-1 1 3.56E-1 1 2.95E-1 1 2.82E-1 1 2.73E-1 1 2.61 E-11 2.58E-1 1 2.93E-1 1 2.43E-1 1 3.06E-1 1 2.77E-1 1 2.79E-1 1 2.67E-1 1 2.79E-1 1 2.79E-1 1 2.79E-1 1 2.79E-1 1 2 79E-11 2.79E-1 1 45 Deg 3.47E-1 1 4.22E-1 1 2.93E-1 1 3.01 E-1 1 2.70E-1 1 2.57E-1 1 2.57E-1 1 2.61 E-11 2.73E-1 1 3.OOE-1 1 3.13E-1 1 3.37E-11 3.07E-1 1 3.37E-1 1 3.37E-11 3.37E-1 1 3.37E-1 1 3.37E-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]
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 Cycle 1
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
2 3
4 5
6 7
8 9
10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time
[EFPY]
1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00
[dpa]
Maximum Iron Atom Displacements 0 Deg 6.62E-04 1.11 E-03 1.48E-03 2.16E-03 2.85E-03 3.33E-03 3.82E-03 4.36E-03 4.89E-03 5.53E-03 6.15E-03 6.75E-03 7.57E-03 9.55E-03 1.13E-02 1.49E-02 1.84E-02 2.20E-02 2.46E-02 15 Deg 9.74E-04 1.60E-03 2.14E-03 3.07E-03 4.05E-03 4.84E-03 5.64E-03 6.59E-03 7.31 E-03 8.32E-03 9.19E-03 1.01 E-02 1.1 3E-02 1.42E-02 1.69E-02 2.22E-02 2.75E-02 3.28E-02 3.68E-02 30 Deg 1.14E-03 1.87E-03 2.49E-03 3.51 E-03 4.56E-03 5.41 E-03 6.40E-03 7.50E-03 8.54E-03 9.82E-03 1.1 OE-02 1.22E-02 1.37E-02 1.77E-02 2.12E-02 2.82E-02 3.53E-02 4.23E-02 4.76E-02 45 Deg 1.18E-03 2.04E-03 2.67E-03 3.75E-03 4.79E-03 5.62E-03 6.61 E-03 7.59E-03 8.76E-03 1.OOE-02 1.14E-02 1.28E-02 1.45E-02 1.93E-02 2.36E-02 3.21 E-02 4.06E-02 4.91 E-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)
'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 RAD1IUS AZIMUTHALANGLE (cm) 00 150 300 45 l
Within The Reactor Vessel Wall RADIUS                         AZIMUTHALANGLE (cm)           .0&deg;             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
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&deg; 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
6-21 Table 6-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Wolf Creek Capsule A
                                      ,EPY]
Irradiation Time Fluence D(E.O MeY)
1        .s           .Xfnn IK                  A [
Iron Displacements
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
,EPY]  
                  -And Location                       StatusLead                     Factor -
.s 1
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&deg;)                     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.
.Xfnn f1 I K A [ 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&deg;)
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.
(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
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.
7-1 7
Table 7     Recommended Suellance Capsule With                 Schedule CapsCap       Cpsul Location         edFacor(         Wit       wal EFPYV       Fluence(Wcr(a  2
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(
U               58.50                 4.25                   1.07             3.16 x loll (c)
Wit wal EFPYV Fluence(Wcr 2 )
Y                 2410                 3.93                   4.79               1.19 x 1019 (c)
(a U
V               60.10                 4.02                   9.78               2.22 x 10'9 (c)
58.50 4.25 1.07 3.16 x loll (c)
X               238.50                 4.30                   13.83             3.49 x 1019 (c)
Y 2410 3.93 4.79 1.19 x 1019 (c)
W               121.50                 4.11                 Standby                   (d)
V 60.10 4.02 9.78 2.22 x 10'9 (c)
Z               301.50                 4.11                 Standby                   (d)
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.
Notes (a) Updated in Capsule X dosimetry analysis (b) Effective Full Power Years (EFPY) from plant startup.
(c) Plant specific evaluation.
(c) Plant specific evaluation.
Line 576: Line 1,291:
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
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
8-1 8
REFERENCES
: 1. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, 1988.
: 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.
: 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.
: 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.
: 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.
: 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.
: 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.
: 7. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.
: 8. Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G. FractureToughness Criteria for ProtectionAgainst Failure
: 8.
: 9. ASTM El 85-82, StandardPracticefor Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.
Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G. Fracture Toughness Criteria for Protection Against Failure
: 9. ASTM El 85-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.
: 10. Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 2.
: 10. Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 2.
: 11. Procedure RMF 8102, Tensile Testing, Revision 1.
: 11. Procedure RMF 8102, Tensile Testing, Revision 1.
: 12. Procedure RMF 8103, Charpy Impact 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.
: 13. ASTM E23-98, Standard Test Methodfor Notched Bar Impact Testing of Metallic Materials, ASTM, 1998.
: 14. ASTM A370-97a, StandardTest Methods and Definitionsfor Mechanical Testing of Steel Products, ASTM, 1997.
: 14. ASTM A370-97a, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 1997.
References
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.
Il 8-2 16 ASTM E21-92 (1998), Standard TestMethodsforElevated Temperature Tension Tests ofMetallzc Materials, ASTM, 1998 17 ASTM E83-93, Standard Practice for Verification and Classification of Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
: 18. ASTM E 185-79, StandardPracticefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels
: 18. ASTM E 185-79, Standard Practice for 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.
: 19. WCAP-143 70, Use of the Hyperbolic Tangent Function for 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.
: 20. Regulatory Guide RG-1.190, Calculatonal and Dosimetry Methods for Determining Pressure 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.
21 WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCSHeatup and Cooldown Limit Curves, January 1996.
22 WCAP- 15557, Revision 0, Qualificationof the Westinghouse Pressure Vessel Neutron Fluence EvaluationMethodology, August 2000.
22 WCAP-15557, Revision 0, Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology, August 2000.
: 23. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two-and Three-Dimensional Discrete OrdinatesNeutron/PhotonTransport Code System, August 1996.
: 23. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two-and Three-Dimensional Discrete Ordinates Neutron/Photon Transport 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.
: 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
References
Line 607: Line 1,324:
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-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:
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]
Capsule ID Azimuthal Withdrawal Irradiation Location Time Time [EFPY]
U                   31.5&deg;         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.
U 31.5&deg; 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
Appendix A


Line 614: Line 1,334:
* 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 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 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 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.
* 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
Appendix A


Line 625: Line 1,341:
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.
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:
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]
A R =A No F Y  
' C, [I -e-I] [e-Ad]
P,e where:
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).
R  
A     =     Measured specific activity (dps/gm).
=
No   =     Number of target element atoms per gram of sensor.
Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,,f (rps/nucleus).
F   =     Weight fraction of the target isotope in the sensor material.
A  
Y   =     Number of product atoms produced per reaction.
=
Pj   =     Average core power level during irradiation period j (MW).
Measured specific activity (dps/gm).
Pref=       Maximum or reference power level of the reactor (MW).
No  
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.
=
        =
Number of target element atoms per gram of sensor.
X      Decay constant of the product isotope (1/sec).
F  
t   =     Length of irradiation periodj (sec).
=
td   =     Decay time following irradiation period j (sec).
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
Appendix A


lo A-4 and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
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
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.
[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.
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.
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.
Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.
Corrections were also made to the 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:
Corrections were also made to the 23'U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the 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.
Correction Capsule U Capsule Y Capsule V Capsule X U Impunty/Pu Build-in 0.87 0.84 0.80 0.76 238U(y,f) 0.97 0.97 0.97 0.97 Net 2 3 8U Correction 0.84 0.81 0.78 0.73 237lNp(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.
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 238U impurities, plutonium build-in, and gamma ray induced fission effects.
Appendix A
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.
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.
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:
The application of the least squares methodology requires the following input:
Line 667: Line 1,404:
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.
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:
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%
Reaction Uncertainty 6 3 Cu(nQ)60Co 5%
54Fe(np) 54 Mn                       5%
54Fe(np) 54 Mn 5%
58Ni(np) 58 Co                      5%
58 Ni(np)58Co 5%
238U(nf) 137Cs 10%
238U(nf) 137Cs 10%
2 3 7Np(n,f)' 37 Cs                    10%
237Np(n,f)' 37Cs 10%
5 9 Co(n,y)6OCo 5%
59Co(n,y)6OCo 5%
These uncertainties are given at the I c level.
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.
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.
Line 681: Line 1,418:
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.
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:
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:
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-"
P.
where 2
[
(g _ g,)
= [J-6]S,.  
2y72 Appendix A
+ 0e-"
where (g _ g,) 2 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)
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%
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.)
Flux Group Uncertainties (R., Rg.)
(E > 0.0055 MeV)                                   15%
(E > 0.0055 MeV) 15%
(0.68 eV < E < 0.0055 MeV)                           29%
(0.68 eV < E < 0.0055 MeV) 29%
(E < 0 68 eV)                                       52%
(E < 0 68 eV) 52%
Short Range Correlation (0)
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 >00055 MeV) 09 (0.68 eV < E < 0.0055 MeV) 0 5 (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
(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.
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.
Line 708: Line 1,447:
Appendix A
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%
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.
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
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)
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
Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr)
                                        -                                        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
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)
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
Thra I
Year                Month I (MWt-hr)                                     Year                 Mn19 onth      (MWt-hr)
I Thermal Generation (MWt-hr)
          .    .        I                                                                       I I  Tllt                                      Year            Month    l (MWt-IIr) 1994                     12                   2636320                     I1998 _
Year I
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
l Year Month I Mn19 I
                                                                                                                                                                *1-
Tllt I
_        I
I onth 1994 12 2636320 I1998 2632 1995 1995 1995 1995 1995 1995 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.
1 2
1996           1         6                   2349911                   1999                     8             251 4800 I VV0   ____
3 4
7---.-.--          26428 I11 1qqq
5 6
__  _      __I_     a      _
2639803 2383600 2020996 2543603 2634001 2524794 1998 1998 1998 1998 1998 1998 3
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
4 5
                                                      -1     '__. I -     I      ---          I                                                 I I 77 I                     I           I     Ilhaxq                    Ifuln                                   nd:rnOlca inn- I         i                       i                           iI                               I
6 7
                                                                                                          -'                   L0JUL             I IYY                      Z                   2393586
_8 Thermal Generatio n
__              I___
(MWt-hr) 2394162 2649959 2564698 2649737 2561103 2642611 2649749 2564818 Year 2001 2001 2001 2001 2001 2001 2001 2001 Month 4
2000n
6 7
_vv A         .
8 9
IOc~noI I^D     j~
1 10 1
                                                                                                                                    .-. I n 1997             1997   33OTA2i1          2650251                       2EOo 2000         rn 5             2638828               4 1
Thermal Generation l (MWt-IIr) 2562211 2631325 2565829 2651544 2651448 2565781 2654261 1995 1995 1995 I
1997                        4            12564428                   1       2000                     6             2564614 IUU-7         I         Iz           I     'ii. o~n ,-         I         - --      [-                           -. tU-I           II
7 8
    . I7I7I                  -J                   zz I aIzv                   Z()()(l)
9 2632369 2634129 2468895 1998 1998 9
                                                                                /1I I                    -7           7e AC4q1)
10 2001 2001 11 2651470
ILI~&#xb6;1C 1997                       6                   25-63505                   2000                     8             2650187 1CV-1997                       7                   2648185                   2000                                     2064968                               =
*1-I I
1997                       8                   2649138                     2000                   10                     0 1997                       9                   2563192                   2000                                     1898243                               _
1995 1995 1995 10 I1 12 2637097 2549850 2640215 1998 1998 1999 I
1997                     10                       18863                   2000                     12             2650536 1997                       11                           0                 2001                       1             2649695 i af7                                                                                                                                           _
I10 11 12 I
I          I I         I                               -          ._
25i 261 1996 1
i 7I             I         IZ                     z4w(fil/           I       Vfull                                   ozarcrn               1 inno             i                                                               -
2481700 1999 3
0I.;Ia                 I                 2649875             1       2001                       'A         I OS144z I,,           _I2 Zt97      I       2001           I)__       _    _  _  I   ~.)IUUJO               I         II Appendix A
2 1996 2
0 1999 4
10 1996 3
0 1999 5
174 1996 4
1784195 1999 6
25(
1996 5
2622027 1999 7
264 1996 1
6 2349911 1999 8
251
: 2649145, 51237 43012 16340 58213 t2613 4209 40059 64943 49843 4800 I.
2001 2002 2002 200 I VV0 I
7 26428 I11 1 qqq a
oUt
__I_
L 1 I 1996 8
2603985 1999 10 2652619 1996 9
2564802 1999 11 2543480 1996 10 2621302 1999 12 2624035 iu 1 l Z563J1 2000 l
2642844 1996 l 12 2649864 2000 2
2468736 V) 2 3
2650720 2388288 1836759 I
I
-1 '__. -
I I
I 77 I I
Ilhaxq I Ifuln nd:rnOlca inn-I i
i iI  
-' I L0JUL I
I YY Z
2393586 2000n A
IOc~noI n
I___ _vv I^D j~ I 1997 1
3OTA2i 2 EOo rn 1997 3
2650251 2000 5
2638828 1997 4
12564428 1 2000 6
2564614 4
1 I
I I
UU-7 I
Iz I 'ii. o~n,-
I
[-
tU-I I7I7I
-J zz I aIzv Z()()(l)
-7 7e AC4q 1)
ILI~&#xb6;1C  
/1I I 1CV-1997 6
25-63505 2000 8
2650187 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 Zt9 7 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
A-13 Table A-3 Calculated C, Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel
  ]     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
_(E  
> 1.0 MeV) [n/cm2 -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
A-14 Table A-4 Measured Sensor Activities and Reaction Rates Surveillance Capsule U Radiall Radially Ad
                                                                                  .          Radiall         Radially
.justed Ajse Measured Saturated Saturated Reaction Activity Acti ivity Rate Reaction Location (dpsfg) dpsg)
                                  .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.
(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 2U (n,f) 3 7Cs (Cd)
Middle 1.43E+05 5.90E+06 I 5.90E+06 3.87E-14 238U (n,f) 13 7Cs (Cd)
Including 235U, 239 pu, and y,fission corrections:
3.26E-14 237Np (nf) 137CS (Cd)
Middle 1.24E+06 5.12E+07 5.12E+07 3 26E-13 237Np (nf) 137CS (Cd)
Including Yfis sion correction.
3.23E-13 59Co (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 59Co (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
: 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
: 3) The average 237Np (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 ,
A-15 I
Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule Y Radially     Radially Adjusted     Adjusted
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
.Mea re. d Saturated Saturated Reaction Activity.
.                                    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.
Activity -
.Activity..
Rate Reaction aLocation (dP  
)sa-......
(dps[R)..  
-(dpsfg)
(rps/atom) 63Cu (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) 58Co 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) 137CS (Cd)
Middle 5.43E+05 5.33E+06 I 5.33E+06 3.50E-14 238u (nf) 137CS (Cd)
Including 235U, 239pu, and yfission corrections:
2.84E-14 237Np (nf) 13 7Cs (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 59Co (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.
: 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.
: 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
Appendix A


A- 16 Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule V
A-16 Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule V
                                                                                            -,.Radially           Radially.
-,.Radially Radially.
                                                                                              'Adjusted         Adjusted Measured                   Saturated
'Adjusted Adjusted Measured Saturated Reaction ActiAity Activity Rate Reaction Location (dpsR)
                                                                                        .                        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)
(dpsJg)
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.
(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-1 7 Average 4.41E-17 54Fe (n,p) 54Mn 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 4.74E-15 Average 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 6 83E-15 Average 6.45E-15 238U (n,f) '"Cs (Cd)
Middle 1.14E+06 5.91E+06 5.91E+06 3.88E-14 23U (n,f)
Cs (Cd)
Including 235U, 239Pu, 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 59Co (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 59Co (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
: 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.
: 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.
Line 774: Line 1,725:
A-17 I
A-17 I
Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule X
Table A-4 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule X
                                                        .. ...I                           Radially       Radially
..... I Radially Radially
                                                                                          .,-Adjusted     Adjusted
.,-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.
'Measueid Saturated Saturated,
: 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.
Rection Actv-y A
tiity Activity,.
Rate Reaction Location (dpsg),
(dps/g).,
(rpslatom) 63Cu (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 7Cs (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 Co (n,Y) 60Co Top 4.42E+07 6.05E+07 6.05E+07 3.95E-12 Bottom 4.44E+07 6.08E+07 6.08E+07 3.96E-12 Average 3.96E-12 59Co (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.
: 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
Appendix A
Line 783: Line 1,745:
1.
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]
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 MI/C M/BE 63Cu(n,u)6OCo 5.52E-17 4 81E-17 5.39E-17 1.15 1.02 54Fe(np) 4 Mn 5.89E-15 5.45E-15 5.89E-15 1.08 1.00 58Ni(n,p)"Co 7.91E-15 7 65E-15 8.16E-15 1.03 0.97 238U(n,f) 37Cs (Cd) 3.26E-14 2.96E-14 3.16E-14 1.10 1 03 237Np(n'f) 137Cs (Cd) 3.23E-13 2.92E-13 3.17E-13 1.11 1.02 5 9Co(nY)60Co 5.29E-12 4.22E-12 5.20E-12 1.25 1.02 5 9Co(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 MI/BE 6 3 Cu(nax)6 0Co 4.74E-17 4.24E-17 4.53E-17 1.12 1.05 54Fe(n,p)54 Mn 4.50E-15 4.68E-15 4.65E-15 0.96 0.97 "Ni(n,p) 58Co 6.1 IE-l5 6.56E-15 6.45E-15 0.93 0.95 238 U(nf)'37Cs (Cd) 2.84E-14 2.51E-14 2.52E-14 1.13 1.12 237Np(nf)137Cs (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) 60Co (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
Best Reaction Measured Calculated Estimate M/C M/BE 63Cu(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
- Fe(n,p)ftMn 4.43E-15 4.35E-15 4.64E-15 1.02 0.95 58Ni(n,p)"8Co 6.45E-15 6.09E-15 6.56E-15 1.06 0.98 238U(n,f) 37Cs (Cd) 3.01E-14 2.32E-14 2.57E-14 1.30 1.18 237Np(nf)'37Cs (Cd) 2.67E-13 2.26E-13 2.60E-13 1.18 1.03 59Co(n,y)6oCo 3.15E-12 3.18E-12 3.12E-12 0.99 1.01 59Co(ny)ICo (Cd) 1.79E-12 2.21E-12 1.82E-12 0.81 0.98 Appendix A


A-19 Capsule X
A-19 Capsule X Reac ion Rate [rps/ tom]
              .______          __      Reac ion Rate [rps/ tom]
Best Reaction Measured Calculated Estimate M/C MIBE 63Cu(n,a)60Co 4.6513-17 4.34E-17 4.53E-17 1.07 1.03 54Fe(n,p)5 4Mn 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 238U(n,f) 37Cs (Cd) 2 3 7 Np(n 1f)3 7Cs (Cd) 3.01E-14 2.56E-14 2.62E-14 1.18 1.15 5 9 Co(n,y)6oCo 2.56E-13 2.50E-13 2.57E-13 1.02 1.00 5 9Co(n,-y) 6 0Co (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
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
.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 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.
Capsule ID Calculated E'stimateh BE/C U
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.
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 Atom DisplacementRate [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
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
A-21 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule U Capsule Y Capsule V Capsule X 63Cu(n,a)60Co 1.15 1.12 1.11 1.07 54Fe(n,p) Mn 1.08 0.96 1.02 0.99 58Ni(np)58Co 1.03 0.93 1.06 0.97 238U(np)137Cs (Cd) 1.10 1.13 1.30 1.18 237Np(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%.
% Standard Deviation 4.0 9.2 9.7  
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)
==8.0 Notes==
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.
: 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%.
: 2. Intermediate Circumferential weld location at -38.35 cm below core centerline and at an azimuth of 45&deg; to document the maximum neutron fluence.
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/cm2-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&deg; to document the maximum neutron fluence.
Appendix A
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-23 A.2 Appendix A References A-I.
A-2. REAC-SAP-172, "Analysis of Neutron Dosimetry from Wolf Creek - Capsules U, Y, and V,"
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.
Perock, J. D. April, 1998.
A-3. A. Schmittroth, FERRETDataAnalysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
A-3.
A-4. RSIC Data Library Collection DLC- 178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.
A. Schmittroth, FERRETData Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
A-4.
RSIC Data Library Collection DLC-1 78, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.
Appendix A
Appendix A


B-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS
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
* 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
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)
-J 2000.00-1000 00-000l
AL48, -500 F 5000 00 4000 00 as (5' 3000 00 0
~~~~~~~it  
            -J m
-.t--l 000 1 00 200 300 Time-1 (ms)
2000.00 1000 Do 0 00 4       -
AL48, -500F 4 00 500 6 00 5000 00 4000 00 as (5' 3000 00 0
0.00 1 00     200         300           400  500 600 Time-I (ms)
-J 2000.00 1000 Do 0 00 m
AL57, 0F 5000 00-4000 00
4  
            .0
-
              ,, 3000 00-0
0.00 1 00 200 300 Time-I (ms)
            -J 2000 00 1000 00 200         300                   600 Time-1 (ms)
AL57, 0F 400 500 600 5000 00-4000 00
AL53, 250 F Appendix B
.0
,, 3000 00-0
-J 2000 00 1000 00 200 300 Time-1 (ms)
AL53, 250F 600 Appendix B


                                                                                        ~fiLl B-2 1
~fiLl B-2 1
5000 00i 4000 00 1f l
5000 00i 4000 00 l 1f
                .0
.0
                -jo 000   1 00   200         300       400   500     6 0o Time-1 (ms)
-jo 000 1 00 200 300 400 500 6
AL52, 40 0 F 1I 5000ooj 4000 00 I 40 m',3000 00 I 0
Time-1 (ms)
AL52, 400F 1I 5000ooj 40 4000 00 I m',3000 00 I 0
2000 001}
2000 001}
100000 0 00 0 00    100   200                   Ann         --
100000 0 000 00 100 200 Ann 0o 0
4-    uu    b6U  0 Tine-1 (ms)
Tine-1 (ms)
AL56, 500 F oa3 0
AL56, 500F 4 -
          -j 000     100   200         300       400   500   6 00 Time-1 (ms)
uu b6U oa3 0-j 000 100 200 300 400 500 Time-1 (ms)
AL55, 75TF Appendix B
AL55, 75TF 6 00 Appendix B


B-3 5000 00-4000 00 Z 300000 0
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)
-j 2000 00 1000 00 0 00 000 1 00 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)
AL59, 1100F 5000 00 4000 00 300000 2000 00 1000 00 000 0 00 1 00 2 00 3 00 4 00 5 00 6 00 Time-1 (ms)
AL5O, 1350  F 5000 00 4000 00
AL5O, 1350F 5000 00 4000 00
          = 3000 00 2000 00                   X       -
= 3000 00 2000 00 X
1000 00 I   I   I I   IS 000 000   1 00   200         300     400 500 600 Time-i (ms)
1000 00 I I
I I
IS 000 000 1 00 200 300 400 500 600 Time-i (ms)
AL60, 150TF Appendix B
AL60, 150TF Appendix B


                                                                                      'I B-4 n   3000 00 2000 00 1000 00.
'I B-4 n 3000 00 2000 00 1000 00.
0 00o 01DO      1 00   200        300      400  500    600 Time-I (ms)
0 00 o 01 5000 00 4000 00 Q
                            .__            AL49, 175 0 F 5000 00 4000 00 Q           .
3000 00 0,
* 3000 00 0,
2000 00 1000000 0000 001 5000
2000 00 1000000 0000 0013      1 00    2 00        3 00      4 00  5 00  6 00 Time-I (ms)
.o 4000 00
AL51, 1900 F 5000      .o 4000 00
.0.
          .0.
co 3000 00 0
co 3000 00 0
          -Jl 200000 100000 000 )
-Jl 200000 100000 000 )
000      1 00   200         300       400   500   6 00 Time-I (ms)
DO 1 00 200 300 400 500 Time-I (ms)
AL46, 2250   F Appendix B
AL49, 1750F 600 3
1 00 2 00 3 00 4 00 5 00 Time-I (ms)
AL51, 1900F 6 00 000 1 00 200 300 400 500 Time-I (ms)
AL46, 2250 F 6 00 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)
B-5 5000 00 400000 n
AL54,   2500 F 50000Do 4000 00
3000 00 20X00 0 0 0 00 1.00 2.00 3 00 4 00 5 00 6 00 Time-i (ms)
          '7 300000 2000DO0l 1000 00 l   ,    ,    ,    ,    ,          =
AL54, 2500 F 50000 Do 4000 00
000 000     1.00       200           300     400   500   60o Time-I (ms)
'7 300000 2000DO0l 1 000 00 l  
AL58, 2750 F 5000 00 0   c 0000   100       2.0           3 00     400   500   600 Time-1 (ms)
=
000 000 1.00 200 300 400 500 60o Time-I (ms)
AL58, 2750F 5000 00 0
c 0000 100 2.0 3 00 400 500 600 Time-1 (ms)
AL47, 300TF Appendix B
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 _  _
'I B-6 5000 00 4000 00 Z-u 3000-00 03 0
000       1 00         200         300           400                  500          6 00 Time-i (ms)
2000 00 1000 00 non ll 1111 r
AT46, -750 F 5000 00-4000 00
v Ad at 111111-ant  
          .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)
l t
AT50, -250 F 3
Am -
A r  
-sB eL at It 000 1 00 200 300 Time-i (ms)
AT46, -750F 400 500 6 00 5000 00-4000 00
.0 3 3000 00-03
-J 2000 00-1000 00-0 00' 000 1 00 200 300 Time-1 (ms)
AT50, -250F 4 00 5 00 6 00 3
0
0
            -j 000       1Lo0       200           300         400                 500           600 Time-i (ms)
-j 000 1Lo0 200 300 400 500 Time-i (ms)
AT60, 15-F Appendix B
AT60, 15-F 600 Appendix B


B-7 5000 Co 4000 00 n 3000 00
B-7 5000 Co 4000 00 n
          -J 2000 00 1000 00 0 00 0 00 1 00 200        300    400  500  600 Time-1 (ms)
3000 00
AT56, 50TF 5000 00-4000 00 x 3000 00-
-J 2000 00 1000 00 0 00 0
          -J 2000 00-1000 00 0 00 0
5000 00-4000 00 x
Time-1 (ms)
3000 00-
AT54, 750 F 5000 00 4000 00
-J 2000 00-1000 00 0 00 0
          -J 3000 00 2000 00 1 000 00 000 Il.O 2 00       3 00   4 00 5 00 600 Time-1 (ms)
5000 00 4000 00
AT53, 100 0 F Appendix B
-J 3000 00 2000 00 1 000 00 00 1 00 200 300 400 500 Time-1 (ms)
AT56, 50TF 600 Time-1 (ms)
AT54, 750F 000 Il.O 2 00 3 00 4 00 5 00 Time-1 (ms)
AT53, 1000F 600 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)
'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)
AT59, 125SF 5000 00 4000 00 30000 0 2000 00 100000 0 00 000 1 00 200 300 400 500 600 Time-1 (ms)
AT58, 150 0 F 5000 00 4000 00
AT58, 1500F 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)
          , 3000 00 2000 00 1000 00Il 0000 000 100   2 00       3 00     4 00 5 00 6 00 Time-I (ms)
AT48, 1750F Appendix B
AT48, 1750  F Appendix B


B-9 5000 00 4000 00 .
B-9
            .0 3000.00 0
.0 0
            -. 1 2000 00 1000 00 0 00             I   ,    ,    ,    ,          ,    ,
-. 1 5000 00 4000 00.
ha .    .    .
3000.00 2000 00 1000 00 0 00 00i'0 I
00i'0          1 00       2.00           3 00             4 00      50o          600 Time-1 (ms)
ha 1 00 2.00 3 00 Time-1 (ms)
AT51, 1750 F 5000 00 4000 00 m 3000 00
AT51, 1750F 4 00 50o 600 5000 00 4000 00 m 3000 00
            -0 2000 00 1000 00 II'''
-0 2000 00 1000 00 II'''
lilt     ,-
lilt 000 1 00 200 300 Time-1 (ms)
000           1 00       200             300             4 00      500        6 00 Time-1 (ms)
AT57, 2000F 4 00 500 6 00 5000 00 4000 00
AT57, 2000 F 5000 00 4000 00
,, 3000 00
            ,, 3000 00
-.1 2000 00
          -.1 2000 00
&#xb6;000 00' l
                &#xb6;000 00' l           UUt.         . I     I   I       I     I     I       -1 1 .
UUt.
1 .
I I
000           1 00       2.00           3 00             400        500        600 Time-1 (ms)
I I
AT52, 225 0F Appendix B
I I  
-1 1
1 000 1 00 2.00 3 00 Time-1 (ms)
AT52, 225 0F 400 500 600 Appendix B


it B-10 5000 00 4000 00 n     3000 00 2000 00 1000 00 0 00 Time-1 (ms)
it B-10 5000 00 4000 00 n
AT49, 250 0 F 5000 00 4000 00
3000 00 2000 00 1000 00 0 00 5000 00 4000 00
            .0.
.0.
as   3000 00 0
as 3000 00 0
2000 00 1000 00 0 00 00 0 1 00 200         300       400 500 600 Time-I (ms)
2000 00 1000 00 0 00 00 5000 00 4000 00 30
AT47, 2750  F 5000 00 4000 00 30          .
-~3000 00.
              -~3000  00.
0o Time-1 (ms)
0o              ,
AT49, 2500F 0
000   1 00 200         300       4 00 500 6 00 Time-I (ms)
1 00 200 300 400 500 Time-I (ms)
AT55, 300 0 F Appendix B
AT47, 2750F 600 000 1 00 200 300 4 00 500 Time-I (ms)
AT55, 3000F 6 00 Appendix B


B-Il 5000 OOf 4000 00 7 3000 001 0
B-Il 5000 OO f 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)
-j 2000 00 1000 00 l 000 0 00 5000 00l 4000 001 a,
                                  ..            AW47, -750 F 5000 00l 4000 001 a,
-J 2000 00 1000 00 00 c
            -J 2000 00 1000 00 00c    -o.            I                         I-0.00       1 0o   200       300       4 00      5 00  600 Time-1 (ms)
-o.
AW52, -35 0 F 5000 00.
1 00 200 300 Time-1 (ms)
4000 00 A0 3000 00-0
AW47, -750 F 4 00 5 00 600 I
          -J 2000 00]
I-0.00 1 0o 200 300 Time-1 (ms)
000         1 00   200       300       400       500   6 00 Time-I (ms)
AW52, -350 F 4 00 5 00 600 5000 00.
AW53, 0F Appendix B
4000 00 A0 3000 00-0-J 2000 00]
000 1 00 200 300 400 500 Time-I (ms)
AW53, 0F 6 00 Appendix B


1.
1.
B-12 4000 2
B-12 4000 2
            'o 3000 03
'o 3000 03
            -J 2000 000 I 00 200       300     400 500 6 00 Time-1 (ms)
-J 2000 000 I 00 200 300 400 500 Time-1 (ms)
AW51, 25 0 F 5000 00 4000 00 n
AW51, 250F 6 00 5000 00 4000 00 n
            'a   3000 00-0
'a 0
            -J 2000.00-1000 001 000 1 00 200         300     400 500 6 00 Time-I (ms)
-J 03
AW58, 500 F 03
-J 3000 00-2000.00-1000 001 000 1 00 200 300 400 500 Time-I (ms)
          -J 2.00       3 00             600 Time-i (ms)
AW58, 500F 6 00 2.00 3 00 Time-i (ms)
AW46, 75TF Appendix B
AW46, 75TF 600 Appendix B


B-13 0
B-13 0
            -J 0 00 1 00 2 00       3 00       4 00   5 00  6 00 Time-1 (ms) 5000 4000
-J 0 00 1 00 2 00 3 00 4 00 5 00 Time-1 (ms) 6 00 5000 4000
            .s 3000 00 0
.s 0
            -j 2000 00 1000 00 000   1 00 200         300       400   500   600 A,
-j A,
Time-i (ms)
0-J 3000 00 2000 00 1000 00 000 1 00 200 300 400 500 Time-i (ms)
AW57, 125TF 0
AW57, 125TF 600 000 1 00 200 3.00 400 500 Time-I (ms)
          -J 000  1 00 200       3.00     400   500   6 00 Time-I (ms)
AW55, 125TF 6 00 Appendix B
AW55, 125TF Appendix B


II B- 14 5000 00 4000 00 c 300000 0
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)
-J 2000 00 100000 0 00 000 1 00 200 300 400 500 600 Time-i (Ms)
AW48, 150 0 F 5000 00 400000
AW48, 1500F 5000 00 400000
            -3000     00 2000 00 100000 000 000 1 00 200       300     400   500 600 Time-I (ms)
-3000 00 2000 00 100000 000 000 1 00 200 300 400 500 600 Time-I (ms)
AW60, 1600 F 500000o 400000
AW60, 1600F 500000o 400000 300000 2000 00 100000 000 000 1 00 200 300 400 500 600 rTme-i (Ms)
          -    300000 2000 00 100000 000 000 1 00   200       300       400 500 600 rTme-i (Ms)
AW56, 2000F Appendix B
AW56, 200 0 F Appendix B


B-15
B-15
              .0 0
.0 0
              -J 000   1.00   200       300       400   500 6 00 Time-i (ms)
-J 000 1.00 200 300 400 500 Time-i (ms)
AW5O, 2250  F 5000 00 4000 00 80
AW5O, 2250F 6 00 5000 00 4000 00 80 3000 00-0
            ,,    3000 00-0
-j 2000 001 Time-1 (ms)
            -j 2000 001 Time-1 (ms)
AW49, 250TF 4000 00 Z-a 3000 00 0
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)
-J 2000 00 1 000 00-0000 0.00 1 00 2.00 3 00 4 00 5 00 Time-1 (ms)
AW54, 2500   F Appendix B
AW54, 2500 F 600 Appendix B


II B-16 4000 00 n     3000 00 s
II B-16 4000 00 n
              -J 2000 00 1000 00-0 oo4 0 (01    00  200        300      400  500  6 00 Time-I (ms)
3000 00 s
AH53, -1750    F 5000 DE1 4000 00
-J 2000 00 1000 00-0 oo4 0 (
            -      3000 00
5000 DE1 4000 00 3000 00
          -J*.
-J*.
2000 00 100000 0 001-00 0   1 00   200         300       400   500   600 Time-1 (ms)
2000 00 100000 0 001-00 5000 00 4000 00 r 300000]-
AH51, -1000   F 5000 00 4000 00 r      300000]-
X, I
X,              I o                II 2000 0011 1000 00 000     1 00   200         300       400   500   6 00 Time-1 (ms)
o II 2000 0011 1000 00 01 00 200 300 400 500 Time-I (ms)
AH50, F Appendix B
AH53, -1750F 6 00 0
1 00 200 300 400 500 Time-1 (ms)
AH51, -1000 F 600 000 1 00 200 300 400 500 Time-1 (ms)
AH50, F 6 00 Appendix B


B-17 5000 00 4000 00 7   3000 00 0
B-17 5000 00 4000 00 7
2000 00O 1000 00 n nn
3000 00 0
                        ^ AA u uu -                i     -  - -- - -    -
2000 00O 1000 00
0 00   1 00 200           300         400  5 00  6 00 Time-I (ms)
^ AA n nn u uu i
AH49, -500 F 5000 00 4000 00 2
0 00 1 00 200 300 Time-I (ms)
              .0 r     3000 00 0
AH49, -500 F 400 5 00 6 00 5000 00 4000 00 2
            -J 2000 00 1000.00-0 00 0X Time-i (ms)
.0 r
AH58, -50F 5000 00 4000 00
3000 00 0
(     3000 00 0
-J 2000 00 1000.00-0 00 0X 5000 00 4000 00
          -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
3000 00 0
-J 2000 00 1000.00 Time-i (ms)
AH58, -50F uuu-i 000 1 00 200 300 Time-1 (is)
AH52, -250F 4 00 5 00 6 00 Appendix B


B-18
B-18
                .0m 0
.0m 0
              -j 000   1 00 200         300     400 500 600 Time-I (ms)
-j 000 1 00 200 300 400 500 Time-I (ms)
AH47, 0F 5000 00 4000 00 c, 3000 00 0
AH47, 0F 600 5000 00 4000 00 c, 3000 00 0-2 2000 00 1000 00.
            -2 2000 00 1000 00.
0 00 01 5000 00 4000 00
0 00 01 Time-1 (ms)
.\\
AH55, 0F 5000 00 4000 00
          .\
m' 3000 00 0
m' 3000 00 0
          -j 000   1 00   200       300       400 500 600 Time-I (ms)
-j Time-1 (ms)
AH59, 250  F Appendix B
AH55, 0F 000 1 00 200 300 400 500 Time-I (ms)
AH59, 250F 600 Appendix B


B-19 n
B-19 n
0
0
          -J Time-I (ms)
-J Time-I (ms)
AH48, 350 F 5000 00 4000 00-n 3000 00 0
AH48, 350F 5000 00 4000 00-n 3000 00 0
            -J 2000 00-1000 co 000-                         I               I             I     I
-J 2000 00-1000 co 000-5000 00 4000 00 X5 3000 00 0-J 2000 00 1000 00
                        )o    100   2 00           3 00        400      500        6 00 Time-i (ms)
)o I
AH54, 500F 5000 00 4000 00 X5 3000 00 0
I I
            -J I
I 100 2 00 3 00 Time-i (ms)
2000 00 1000 00 nnn. l                                           ,  .      ,
AH54, 500F 400 500 6 00 I
nil       ,    , ,  ,    ,        ,
nnn.
000     1 00   200           300         400      500        600 Time-1 (rns)
l nil 000 1 00 200 300 Time-1 (rns)
AH60, 1000 F Appendix B
AH60, 1000F 400 500 600 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)
'I B-20 5000 00 4000 00 D, 3000 00-0 000 1 00 200 300 400 500 Time-1 (ms)
AH46, 150 0 F 000   1 00   200       300       400   500   6 00 Time-1 (ms)
AH46, 1500F 6 00 000 1 00 200 300 400 500 Time-1 (ms) 6 00 Q
Q 0
0
          -j 000   1 00   200       300       400   5 00 600 Time-i (ms)
-j 000 1 00 200 300 400 5 00 Time-i (ms)
AH56, 2000  F Appendix B
AH56, 2000F 600 Appendix B


c-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Appendix C
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:
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."
"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.
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.
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.
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] ,             ____-'_
Table C-1 Upper Shelf Energy Values Fixed in C"VGRAPH Ift-lb],
Uaps                       e-._k
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.)
                                                        ,          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.)
Lower Shell Plate             94               96               94                 88           95 R2508-3 (Trans.)
Weld Metal 100 92 94 89 93 (heat # 90146)
Weld Metal                 100             92               94                 89           93 (heat # 90146)
HAZ Material 161 140 200 167 135 Appendix C
HAZ Material               161             140             200                 167         135 Appendix C


II CAPSULE X (LONGITUDINAL ORIENTATION)
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:
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09:13:31 on 12-12-2002 Page 1 Coefficients of Curve 1 I
300  --                                            ----                                                        .       -
A = 72.09 B = 69.9 C = 85.66 T0 = 95.8 Upper Shelf Energy: 142 Fixed Material:
U]     250-200 10 15 0-V                                                                        0                 0 i too0 100~l 5fF
Equation is CVN = A + B I tanh((T - TO)/C I Temp. at 30 ft-lbs 36.1 Temp. at 50 ft-lbs:
            -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
PLATE SA533B1 Heat Number C4935-2 I
      -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
Capsule: X Total Fluence:
                                              *** Data continued on next page "**
67.7 Lower Orientation: LT Shelf Energy 2.19 Fixed U]
10 V
300 250-200 15 0-0 0
i too0 100~l 5fF
-300
-200  
-100 0
100 200 300 Temperature in Degrees Data Set(s) Plotted Plant WC1 Cap: X Material: PLATE SA533BI OrL LT Heat Charpy V-Notch Data Input CVN Energy Computed CVN Energy 400 F
500 600 II: C4935-2 Temperature
-50 0
25 40 50 75 110 135 150 2
13 21 37 53 43 74 108 133 6.69 15.69 24.66 32.07 37.92 55.45 8358 102.02 11L23 Differential
-4.69
-2.69
-3 66 4.92 15.07
-12.45
-958 5.97 2L76
*** Data continued on next page "**
C-2
C-2


CAPSULE X (LONGITUDINAL ORIENTATION)
CAPSULE X (LONGITUDINAL ORIENTATION)
Page 2 Material: PLATE SA533BI           Heat Number. C4935-2       Orientation: LT Capsule: X     Total Fluence:
Page 2 Material: PLATE SA533BI Heat Number. C4935-2 Orientation: LT Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input CVN Energy Computed CVN Energy Differential 100 12298  
-22.98 122 12804  
-6.04 150 135.47 1452 146 13828 7.71 135 139.9  
-4.9 137 140.2  
-3.82 SUM of RESIDUALS = -86 Temperature 175 190 225 250 275 300 C-3


II CAPSULE X (LONGITUDINAL ORIENTATION)
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:
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09.23:49 on 12-12-2002 Page 1 Coefficients of Curve 1 I
200        ----
A = 37.66 B = 36.66 C = 72.74 T0 = 7825 Upper Shelf L.E. 74.32 Material: PLATE Equation is LK = A + B * [ tanh((T - TO)/C) I Temperature at LK 35:
Wi 150 0
72.9 Lo SA533B1 Heat Number C4935-2 Capsule: X Total Fluence:
wer Shelf LE-I Fixed Orientation: LT Wi 200 150 0
50 0
50 0
0 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
-.jUU
  -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       "**
-200  
-100 0
100 200 300 Temperature in Degrees Data Set(s) Plotted Plantk WC1 Cap: X Material PLATE SA533B1 OriL LT H6 Charpy V-Notch Data Input Lateral Expansion Computed LEK 400 F
500 600 it # C4935-2 Temperature
-50 0
25 40 50 75 110 135 150 0
5 12 24 33 29 47 67 74 3.09 8.64 14.77 19.98 24.09 36.02 52.71 61.59 65.37 Differential
-3.09
-3.64
-2.77 4.01 8.9
-7.02
-5.71 5.4 8.62 me Data continued on next page "**
C-4
C-4


CAPSULE X (LONGITUDINAL ORIENTATION)
CAPSULE X (LONGITUDINAL ORIENTATION)
Page 2 Material: PLATE SA533B1           Heat Number C4935-2       Orientation: LT Capsule: X     Total Fluence Charpy V-Notch Data (Continued)
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
Input Lateral Expansion Computed LE Differential 64 69.52  
-552 67 71.07  
-4.07 71 73.04  
-2.04 75 73.67 132 75 73.99 1
75 7415 B4 SUM of RESIDUALS =-3.79 Temperature 175 190 225 250 275 300 C-5


CAPSULE X (LONGITUDINAL ORIENTATION)
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
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09:2558 on 12-12-2002 Page 1 Coefficients of Curve I I
-4    4 2--3C
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 Capsule: X Total Fluence Orientation: LT
              )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
-4 4
    -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
V])
                                        ""  Data continued on next page ****
6aC 2--3C Temperature
)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 Input Percent Shear Computed Percent Shear Differential
-50 0
25 40 50 75 110 135 150 2
5 10 15 20 30 50 65 90 1.65 5.4 10.63 14.96 18.59 30.47 5.18 67.68 75.59
.34
-. 4
-.63
.03 L4
-.47
-218
-2.68 14.4 Data continued on next page ****
C-6
C-6


CAPSULE X (LONGITUDINAL ORIENTATION)
CAPSULE X (LONGITUDINAL ORIENTATION)
Page 2 Material: PLATE SA533B1             Heat Number. C4935-2     Orientation: LT Capsule: X     Total Fluence:
Page 2 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: LT Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input Percent Shear Computed Percent Shear Differential 75 85.59  
-10.59 85 89.78  
-4.78 100 95.62 4.37 100 97.67 2.32 100 98.77 122 100 99.35  
.64 SUM of RESIDUALS = 255 Temperature 175 190 225 250 275 300 C-7


II CAPSULE X (TRANSVERSE ORIENTATION)
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:
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
Y)     25 10 A4m     2C0~
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 Material: PLATE SA533B1 Heat Number C4935-2 Capsule: X Total Fluence:
P-e
97 Lower Shelf Energy: 2.19 Fixed Orientation: TL Y) 25 10 A4m 2C P-e 15 z0 Cz; 10 5
;,    15 z0 Cz;     10 5
Tempera
          -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
      -75                                  5                                        433
-25 15 50 75 100 125 150 175 0~
      -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
-300  
-200  
-100 0
100 200 300 400 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 Lture Input CVN Energy Computed CVN Energy 5
433 11 8.36 15 15.85 30 27.49 41 38.82 52 51.51 55 63.78 67 74.06 79 81.69 500 600 Differential
.66 2.63
-.85 25 217
.48
-. 78
-7.06
-2.69 Data continued on next page he c-8


CAPSULE X (TRANSVERSE ORIENTATION)
CAPSULE X (TRANSVERSE ORIENTATION)
Page 2 Material: PLATE SA533B1           Heat Number. C4935-2     Orientation: TL Capsule: X     Total Fluence:
Page 2 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: TL Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input CVN Energy Computed CYN Energy Differential 98 8169 16.3 88 86.85 L14 91 9013  
.86 93 9214  
.85 101 93.33 7.66 96 94.03 1.96 SUM of RESIDUALS = 17.86 Temperature 175 200 225 250 275 300 C-9


II CAPSULE X (TRANSVERSE ORIENTATION)
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:
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09:33:46 on 12-12-2002 Page 1 Coefficients of Curve 1 I
200-7 M
A = 32.72 B = 3L72 C = 942 TO = 95.62 Upper Shelf LE: 64.44 Material: PLATE Equation is: LE. = A + B I [ tanh((T - TO)/C) I Temperature at LE 35:
"-4
102.4 Lo SA533B1 Heat Number. C4935-2 Capsule: X Total Fluence:
. F-     150 100 F-f CZ 0       0         -
ower Shelf LE I Fixed Orientation: TL M
P~                                                                                               I 0
"-4
                                                                    . X If   __    I             I I-300 Ir I      -  I
. F-200-7 150 100 0
                        -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
0 P~
        -75                                   0                                         2.65                               -2.65
I F-f CZ
        -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
. X 0
                                              *** Data continued on next page "**
I Ir If I
I I
I
-300  
-200  
-100 0
100 200 300 400 500 600 Temperature in I Data Set(s) Plotted Plant WC1 Cap: X Material: PLATE SA533B1
)egrees F Ori: TL Heat # C4935-2 Charpy V-Notch Data Temperature Input Lateral Expansion
-75
-25 15 50 75 100 125 150 175 0
4 8
20 29 36 38 50 53 Computed LE 2.65 5.54 10.7 1845 25.88 3419 42.3 49.23 54.51 Differential
-2.65
-1.54
-2.7 L54 311 1.8
-4.3
.76
-151
*** Data continued on next page "**
C-1O
C-1O


CAPSULE X (TRANSVERSE ORIENTATION)
CAPSULE X (TRANSVERSE ORIENTATION)
Page 2 Material: PLATE SA533B1           Heat Number. C4935-2       Orientation: TL Capsule: X     Total Fluence:
Page 2 Material: PLATE SA533B1 Heat Number. C4935-2 Orientation: TL Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input lateral Expansion Computed LE.
Differential 57 5451 2.48 51 582  
-72 69 60.61 8.38 60 6213  
-2.13 66 63.06 2.93 61 6362  
-2.62 SUJM of RESIDUALS = -366 Temperature 175 200 225 250 275 300 C-11


CAPSULE X (TRANSVERSE ORIENTATION)
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:
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 rE SA533BI Heat Number. C4935-2 0:
;.-4     ou a) 0 cD       60 4a)
Capsule: X Total Fluence:
  -      40 2F
Material: PLA1 rientation: TL
          -30 DO
;.-4 ou a)0 cD 60 4a) 40 2F
                    .      a     ,
-30 Temperature
                    -2U0
,\\
              ,\
a DO
                                    -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
-2U0  
      -75                                   2                                         1.83
-100 0
      -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
100 200 300 Temperature in Degrees Data Set(s) Plotted Plant-WC1 Cap: X Materia1: PLATE SA533B1 Oi: TL Heat Charpy V-Notch Data Input Percent Shear Computed Percent Shear 400 F
                                            '** Data continued on next page     Ir' C-12
500 600
# C4935-2 Differential
-75
-25 15 50 75 100 125 150 175 2
5 15 25 40 45 55 65 60 1.83 5.35 1208 23 3421 47.52 6118 7329 8269 16
-.35 2.91 L99 5.78
-252
-6.18 4-29
-Z69 Data continued on next page Ir' C-12


CAPSULE X (TRANSVERSE ORIENTATION)
CAPSULE X (TRANSVERSE ORIENTATION)
Page 2 Material: PLATE SA533B1             Heat Number C4935-2       Orientatj ion: TL Capsule: X     Total Fluence:
Page 2 Material:
Temperature 175 200 225 250 275 300 PLATE SA533B1 Heat Number C4935-2 Orientatj Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input Percent Shear Computed Percent Shear 90 82.69 95 8926 100 93.54 100 9618 100 97.77 100 98.7 Sul ion: TL Differential 7.3 5.73 6.45 3.81 2Z2 129 hi of RESIDUALS = 17.61 C-13


                                                                                                                                  'I CAPSULE X (WELD)
'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) ]
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 0916:18 on 01-10-2003 Page 1 Coefficients of Curve 1 l
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:
A = 47.59 B = 45.4 C = 9538 TO = 49.68 Equation is CVN = A + B * [ tanh((T - TO)/C) ]
Capsule: X       Total Fluence:
Upper Shelf Energy: 93 Fixed Temp. at 30 ft-lbs 10.6 Temp. at 50 ft-lbs 54.7 Material: WELD Heat Number WIRE HEAT NO.90146 Capsule: X Total Fluence:
30   09--                                                     -a C')    250-20f C-) 10                                                                           n (F         --                                                      -I
Lower Shelf Energy: 219 Fixed Orientation:
            -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
C')
      -75                                 4                                         839                             -4.39
C-)
      -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 *'*
30 09--  
-a 250-20f 10 n
(F  
-I
-300
-200  
-100 0
100 200 300 Temperature in Degrees Data Set(s) Plotted Plant. WC1 Cap.: X Material WELD Or.
Heat lh WERE HI Charpy V-Notch Data Input CVN Energy Computed CVN Energy 400 F
500 600 EAT NO.90146 Temperature
-75
-35 0
25 50 75 100 125 125 4
13 23 37 53 64 68 78 76 839 1535 25.88 361 47.74 59.37 6954 77.47 77.47 Differential
-4.39
-235
-2188
.89 5.25 4.62
-1.54
.52
-147 Data continued on next page *'*
C-14
C-14


CAPSULE X (WELD)
CAPSULE X (WELD)
Page 2 Material: WE)LD             Heat Number. WIRE HEAT NO.90146     0rientation:
Page 2 Material: WE Temperature 150 160 200 225 250 250
Capsule: X     Total Fluence:
)LD Heat Number. WIRE HEAT NO.90146 0
Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input CVN Energy Computed CVN Energy 67 8312 89 84.82 84 8927 102 90.75 96 91.65 94 91.65 Si rientation:
Differential
-16.12 4.17
-5.27 1124 4.34 2.34 JM of RESIDUAIS = -.65 C-15


II CAPSULE X (WELD)
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
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09.1850 on 01-10-2003 Page 1 Coefficients of Curve 1 I
* 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:
A = 31.67 B = 30.67 C = 75.77 T0 = 45 Upper Shelf LE 6234 Material: WELD Equation is LE = A + B
Capsule X       Total Fluence 200  --                              -                                      ---                            .      -
* I tanh((T - TO)/C) I Temperature at LE 35:
M) 150 pS--
532 Lower Shelf LE. 1 Fixed Heat Number WIRE HEAT NO.90146 Orientation:
10 0
Capsule X Total Fluence M) pS--
50                                                                   _______            -        -    .
200 150 10 0
e9/
50 e9/
I     -      I           1-
I I
          -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
1-
    -75                                   0                                          3.47
-300
    -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
-200
                                            **    Data continued on next page     '"
-100 0
C-16
100 200 300 400 F
500 600 Temperature in Degrees Data Set(s) Plotted PlantL WC1 Cap: X Material WELD Ori.
Heat #: WIRE HEAT NO.90146 Charpy V-Notch Data Temperature Input Lateral Expansion
-75
-35 0
25 50 75 100 125 0
5 15 25 35 45 51 53 55 Computed LK 3.47 7.62 15.33 23.76 33.69 4322 50.7 55.72 55.72 Differential
-3.47
-2.62
-.33 123 1L3 177
.29
-2.72
-.72 Data continued on next page C-16


CAPSULE X (WELD)
CAPSULE X (WELD)
Page 2 Material: WELD                     Heat Number. WIRE HEAT NO.90146   Orientation:
Page 2 Temperature 150 160 200 225 250 250 Material: WELD Heat Number. WIRE HEAT NO.90146 Capsule: X Total Fluence:
Capsule: X     Total Fluence:
Charpy V-Notch Data (Continued)
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
Input Lateral Expansion Computed LK 54 58.73 62 59.53 54 6L34 64 6182 64 62.07 69 62.07 Orientation:
I '    1 C-17
Differential
-4.73 2.46
-7.34 217 1.92 6.92 SULM of RESIDUALS = -386 I
I 1
C-17


II CAPSULE X (WELD)
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:
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 09X.43 on 01-10-2003 Page 1 Coefficients of Curve 1 I
Capsule: X       Total Fluence:
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 Capsule: X Total Fluence:
Orientation:
pCi C-)
pCi C-)
AH
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
-300  
    -75                             10                                       10X84                           -.84
-200  
    -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       -
-100 0
C-18
100 200 300 400 500 Temperature in Degrees F 600 Data Set(s) Plotted Plantk WC1 Cap: X Material: WELD Ori Heat #: WIRE HEAT NO.90146 Charpy V-Notch Data Input Percent Shear Computed Percent Shear Temperature
-75
-35 0
25 50 75 100 125 125 10 20 45 50 60 80 85 95 95 10X84 2262 38.64 5213 65.33 76.52 84.93 90.69 9069 Differential
-.84
-2.62 6.35
-213
-5.33 3.47
.06 43 43 Data continued on next page C-18


CAPSULE X (WELD)
CAPSULE X (WELD)
Page 2 Material: I ELD                 Heat Number. WIRE HEAT NO.90146     Orientation:
Page 2 Material: I Temperature 150 160 200 225 250 250 ELD Heat Number. WIRE HEAT NO.90146 Or Capsule: X Total Fluence:
Capsule: X     Total Fluence:
Charpy V-Notch Data (Continued)
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
Input Percent Shear Computed Percent Shear 85 94.4 95 95.45 95 98.05 100 98.86 100 99.34 100 99.34 Differential
-9.4
-.45
-3.05 113
.65
.65 S -2.91 ientation:
SUM of RESIDUAL' C-19


Ii CAPSULE X (HAZ)
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
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10:32:26 on 12-12-2002 Page 1 Coefficients of Curve I I
* 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:
A = 68.59 B = 66.4 C = 56.86 TO = -36.56 l
Capsule: X       Total Fluence:
Equation is CVN = A + B
300-
* I tanh((T - TO)/C) I Upper Shelf Energy: 135 Fixed Temp. at 30 ft-lbs  
: 9)     250 10 p---q I
-74.3 Temp. at 50 ft-lbs  
zq       200
-52.9 Material: HEAT AFFD ZONE Heat Number: WIRE HEAT NO.90146 Capsule: X Total Fluence:
  -4     15                                           1 Q)                                                          0             0     ___
Lower Shelf Orientation:
0                   I-
Energy: 2.19 Fixed 9) 10 p---q I
;i 4
zq
: 10.                                           I C-)
-4 Q)
50
;i4 C-)
                                .70 0-   -            --                        -            -I
300-250 200 15 1
            --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
0 0
      -175                                   6
0 I-
      -100                                                                               32                                  2.78 15                                          15.0 7
: 10.
      -75                                   34                                                                              -.07
I 50
      -50                                                                               29.49                                  4.5 29                                          531 9
.70 0-  
      -50                                   88                                                                            -2419
-I
      -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 ***
--OW  
-)u  
-100 0
100 200 300 Temperature in Degrees Data Set(s) Plotted Plant. WC1 Cap: X Material: HEAT AFFD ZONE OrL Heat # M Charpy V-Notch Data ire Input CVN Energy Computed CVN Energy 400 F
500 600 Temperat.
-175
-100
-75
-50
-50
-25 0
0 25 6
15 34 29 88 50 92 152 98 32 15.0 29.4 531 531 81.9 106.
106.2 1213 7
9 9
9 1
WIRE HEAT NO.90146 Differential 2.78
-.07 4.5
-2419 34B
-3L91
-1424 45.75
-23.33 e*** Data continued on next page ***
C-20
C-20


CAPSULE X (HAZ)
CAPSULE X (HAZ)
Page 2 Material: HEAT AFPD ZONE           Heat Number. WIRE HEAT NO.90146     Orientation:
Page 2 Material: HEAT Temperature 35 50 100 150 200 200 AFPD ZONE Heat Number. WIRE HEAT NO.90146 Capsule X Total Fluence:
Capsule X     Total Fluence:
Charpy V-Notch Data (Continued)
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
Input CVN Energy Computed CVN Energy 132 125.08 140 128.96 146 133.91 140 134.81 124 134.96 29 134.96 St Orientation:
Differential 6.91 11.03 1a0B 518
-10.96
-5.96 JM of RESIDUAIS = 12.36 C-21


II CAPSULE X (HAZ)
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:
P4 a)
Capsule: X       Total Fluence:
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:
2007 15 P4 100 a) 5Y 0
Capsule: X Total Fluence:
5~~~01 0
2007 15 100 5Y 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
0 5~~~01
      -175                                 0                                            L411
-;UU
      -100                                 5                                                                               -141 6.0,                               -1.07
-W0u Plant WC1 Cap.:
      -75                                12                                                08
-100 0
      -50                                12                                                                                  .11 222 :1                             -1021
100 200 Temperature in Degi Data Set(s) Plotted X
      -50                                42                                          222 1
Material HEAT AFFD ZONE Ori.:
      -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
Charpy V-Notch Data ut Lateral Expansion Compute 300 rees 400 F
                                            **  Data continued on next page
500 600 Temperature Inp
* C-22
-175
-100
-75
-50
-50
-25 0
0 25 0
5 12 12 42 21 44 70 52 L41 6.0, 222 222 361 510' 510:
61.4' Heat # WIRE HEAT NO90146 d LE Differential 1
-141
-1.07 08  
.11
:1  
-1021 1
19.78 6
-15.6 3  
-7.03 3
18.96 5  
-9.45 Data continued on next page C-22


CAPSULE X (HAZ)
CAPSULE X (HAZ)
Page 2 Material: HEAT AFFD ZONE             Heat Number. WIRE HEAT NO.90146   Orientation:
Page 2 Temperature 35 50 100 150 200 200 Material: HEAT AFFD ZONE Heat Number. WIRE HEAT NO.90146 Capsule: X Total Fluence:
Capsule: X     Total Fluence:
Charpy V-Notch Data (Continued)
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
Input Lateral Expansion Computed LE.
70 6428 70 67.34 70 715 71 72.31 72 72.46 74 72.46 Orientation:
Differential 5.71 2.65
-1.5
-1.31
-.46 153 SUM of RESIDUALS =.67 C-23


I, CAPSULE X (HAZ)
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:
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  
Capsule: X       Total Fluence:
-12.1 Material: HEAT AFFD ZONE Heat Number WYIRE HEAT NO.90146 Capsule: X Total Fluence:
DU I
Orientation:
30 C) 04 0xr
C) 04 Tempera
          -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
-175
    -175I                                0                                         2
-10
    -100 I                                2                                                                             -2 a4                                -1.4
-75
    -75                                  5                                        8.37
-50
    -50                                  15                                                                            -337 19.15                                -4.15
-50
    -50                                  30                                      1915
-25 r
    -25                                                                                                                1084 20                                      38.03                                18.03
4 0
    -25                                50                                      61.39                                11.39 50                                100                                      61.39 65                                                                            38.6 80.47                                15.47
50
                                          '** Data continued on next page
-25 DU I
30 0xr
-300  
-200  
-100 0
100 200 300 400 500 Temperature in Degrees F Data Set(s) Plotted Plant WC1 Cap.: X Material: HEAT AFFD ZONE Ori:
Heat P WYIRE HEAT NO.90146 Charpy V-Notch Data ture Input Percent Shear Computed Percent Shear Diff I
I 0
2 5
15 30 20 50 100 65 2
a4 8.37 19.15 1915 38.03 61.39 61.39 80.47 600 erential
-2
-1.4
-337
-4.15 1084 18.03 11.39 38.6 15.47
'** Data continued on next page
* C-24
* C-24


CAPSULE X (HAZ)
CAPSULE X (HAZ)
Page 2 Material: HEAT AFFD ZONE             Heat Number. WIRE HEAT NO.90146     Orientation:
Page 2 Material: HEAT AFFD ZONE Heat Number. WIRE HEAT NO.90146 Orientation:
Capsule: X     Total Fluence:
Capsule: X Total Fluence:
Charpy V-Notch Data (Continued)
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
Input Percent Shear Computed Percent Shear Differential 75 85.77  
-10.77 100 91.43 8.56 100 9862 137 100 99.79  
.2 100 99.96  
.03 100 99.96  
.03 SUM of RESIDUAIS = -518 Temperature 35 50 100 150 200 200 C-25


D-O APPENDIX D WOLF CREEK SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D
D-O APPENDIX D WOLF CREEK SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D
Line 1,267: Line 2,493:
Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.
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 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 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:
The Wolf Creek reactor vessel consists of the following beltline region materials:
* Intermediate   Shell Plates R2005-1, 2, 3
Intermediate Shell Plates R2005-1, 2, 3
* Lower Shell   Plates R2508-1, 2, 3
* Lower Shell Plates R2508-1, 2, 3 Intermediate & Lower Shell Longitudinal Weld Seams (Heat # 90146),
* Intermediate   & Lower Shell Longitudinal Weld Seams (Heat # 90146),
Intermediate to Lower Shell Circumferential Weld Seam (Heat # 90146).
* Intermediate to Lower Shell Circumferential Weld Seam (Heat # 90146).
Appendix D
Appendix D


Line 1,279: Line 2,504:
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.
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.
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.
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.
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.
Criterion 3:
When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 280F for welds and 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.
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.
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.
Line 1,288: Line 2,515:
Appendix D
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
D-3 TABLE D-1 Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule Data Material  
* 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
'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) = 39.10F 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)
* 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]
* X( FF2) = (213.362) * (4.828) = 44.10F Notes-(a) f = fluence Calculated fluence from Section 6 of this report [x 1019 n/cm2, E > 1.0 MeV]
(b) FF = fluence factor = f(O28 .0 'logf .
(b)
(c)     ARTNDT values are the measured 30 ft-lb shift values taken from Figures 5-1, 5-4 and 5.7, herein [IF]
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.
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
Appendix D
Line 1,299: Line 2,547:
II D-4 Table D-2:
II D-4 Table D-2:
Wolf Creek Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials.
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)
CF.
(Slopeb.W                 ARTTNT       ART T(I           N2FK 7 2F(Weld)
Measured Predicted ScMtret 0
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.
(alse Material  
-,-Capsule FF Metals)
(Slopeb.W ARTTNT ART T(I N2FK 7 2 F(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 +/- 1 a of 28 0F scatter band for the surveillance weld data, therefore the weld data is also credible per the third criterion.
Appendix D
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.
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 +/- 250F.
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.
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 250F. 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.
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.
The Wolf Creek surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to the Wolf Creek surveillance program.

Latest revision as of 10:53, 16 January 2025

Submittal of Test Results for Withdrawal of Surveillance Capsule X from Reactor Vessel
ML031060076
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
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)
v • d • e


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 Revision 0 March 2003 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 1 0'9 n/cm2 (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 1 0'9 n/cm2 (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/cm2 (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/cm2 (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/cm2 (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/cm2 (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/cm2 (E> 1.0MeV).....................................

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

6-12 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface.

6-16 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall.

6-20 Table 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 Section 6 and Appendix A A.R. Rawluszki 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/cm2.

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 36.110F and an irradiated 50 ft-lb transition temperature of 67.750F. This results in a 30 ft-lb transition temperature increase of 61.061F and a 50 ft-lb transition temperature increase of 67.640F 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 55.970F 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/cm2 Vessel 3/4 thickness = 7.42 x 1018 n/cm2

  • 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 (3100C)

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)

(b)

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

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)

(b)

This table was taken from WCAP-10015t4].

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

-Base Metal S

,pecimen 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

~wAlly Sti NIST Certified Reference7

,,,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 67.750F. This results in a 30 ft-lb transition temperature increase of 61.06'F and a 50 ft-lb transition temperature increase of 67.640F 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 55.970F and an irradiated 50 ft-lb transition temperature of 97.04'F. This results in a 30 ft-lb transition temperature increase of 53.960F 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 10.660F and an irradiated 50 ft-lb transition temperature of 54.730F.

This results in a 30 ft-lb transition temperature increase of 68.360F and a 50 ft-lb transition temperature increase of 75.380F Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-74.340F and an irradiated 50 ft-lb transition temperature of -52.920F. This results in a 30 ft-lb transition temperature increase of 69.660F and a 50 ft-lb transition temperature increase of 61.080F.

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 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.

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 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 the extended life of the vessel (54 EFPY) as required by IOCFR50, Appendix G t2k The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule X materials is shown in Figures 5-13 through 5-16 and shows an increasingly ductile or tougher appearance with increasing test temperature.

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 data from WCAP-10015141, WCAP-1 1553"5' and WCAP-13365, Rev. 1[61 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.

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 1 09 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 dataE43. 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 1 O'9 n/cm2 (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/cm2 (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/cm2 (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 1 3 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 7 1 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 FastY Test

.Load Timeto MaxMaL "-,Max.,

Fracto<

?(Arrest Yield,,:,

Flow,,

Sample Temp.

Charpy Max.

Prop,,

PPY 3Yield tGo' 3

P M, IM LoadPr Lo d P Stress, ;

Stress 0o.

(F)

E/kA, b'

Qb' (msec)

(Ib)

(mssec)',

(Ib s(Ib)

"',.ksi) k, 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 l'Test YLiad lTim to M

M I

A Yeld Flow p

Te Charpy Max."

Prp PGYrop.

Yield t y Load PM t

jLoadPF 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.

emerge~o>>ss3?>

')

sa:

Tii"

toC, Fa':

?

t ls i tl!';t,,

v Tet

'T~ o fv~

Mal.

Vinci Arrest Yield

Flow, Charpy MaL
PrPop, GYI I GV I AS Yield t Lad P.

a M

F ad P LoaI Stress ay Stres llNo.

(3F)

( lb) E3WA EA,

jA Qb)

P

  • (b),

i)

(k~i)

  • ksi3 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 Q ft-lb)

Avcrage 3 mI l er 1° AveragecS ft-lb(,

50 ft.Ibj 30 (t-lbAverage Energ Absorption~a Matenal Transition Temperature (,)

Expansion Temperature (0

5) Transition Temperature at Full Shear (ft-Jb)

Unrdae Iid

'e ATi:

Unirradiated Irradiated, AT, I nin-adiated Irradiated, AT Umaiae rradiated E

'5l:5555,5,',

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 Trrafllnted hr i

49 x AQ 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

-3000 20-0 0

10 0-0 I

0__

~177

-300

-200

-100 0

100 Temperature in Curve Leged I 0-~

2 0-----

a 0

~

4 Data Setgs) Plotted 200 300 400 500 600 Degrees F i

5

-~

Curve Plant Capsule Material I

WCI UNIRR PLATE SA533BI 2

C!C U

PLATE SA533BI 3

WfCI Y

PLATE SA533BI 4

W'I V

PLATE SA533BI 5

iC!

X PLATE SA533BI OriL Heat#

LT C4935-2 LT C4935-2 LT C4935-2 LT C4935-2 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 T o 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 Temperature in Degrees F Cune legend 120-3 0 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 2 500 600 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 2

0 219 0

96 3

0 219 0

94 4

0 219 0

88 5

0 219 0

95 U

2 2

25B 0

3739

-6 5654 1

55X97 0

34.32 23.79 59.55 35.39 8149 54.53 90.59 53.96 97.04 0

2523 4716 5627 62.71 C1 7

a)

-30

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10 20---

3 0 4 ^-

5 Data Set(s) Plotted Matenal Curve Plant Capulle Ori HtIPf r---n -

Tlva

  • vtI~

I OC!

UNIRR PLATE SA533BI 2

TCI U

PLATE SA533BI 3

TO Y

PLATE SA533BI 4

WT V

PLATE SA533B1 5

TI X

PLATE SA533BI TL C4935-2 TL C4935-2 TL C4935-2 TL C4935-2 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 2

0 72B6 3

0 7548 4

0 61.41 5

0 64.44 0

481 743

-663

-a6 2544 357 67.84 9a79 1024 0

1025 42.39 68.34 7695 1507_

~1007-CO 50

-300

-200

-100 0

100 200 300 400 Temperature in Degrees F Curve legend I

O-2 0--

3 0 4

5 500 600 Data Set(s) Plotted Material Curve Plant Capsule Ori Heatf I

IC UNIRR 2

lCI U

3 W!a Y

4 WC!

V 5

WC!

x PLATE SA533BI PLATE SA533BI PLATE SA533BI PLATE SA533BI PLATE SA533BI TL C4935-2 TL C4935-2 TL C4935-2 TL C4935-2 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 2

0 219 0

92

-8

-3047 3

0 219 0

94

-6

-1259 4

0 2.19 0

89

-Ll

-1136 5

0 219 0

93

-7 10.66 0

-20.64 2721 644 45.09 2082 46.33 3179 68.36 5473 0

27.09 41.47 5Z44 7538 a) z 0--

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve legend I o 2 0--

30 4

5 V.-

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

I WlC UNIER 2

WC]

U 3

WC1 Y

4 WC1 V

5 lCI x

WD van Wm Wma Wm WIRE HEAT NO.90146 WIRE HEAT NO.90146 WIRE HEAT NO.90146 WMRE HEAT NO.90146 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 T o 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----

3 0 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 2

0 3

0 4

0 5

0

-7394 20.94 2076 2109 0

94B9 55.03 9471 9504 U) a)

C) a)

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1

2 C---

3 0 4 ^

5 Data Set(s) Plotted Curve Plant Capsule Material Or.

Heat#

I UC UNIRR 2

TOC U

3 WCI Y

4 WTOI V

5 KCI X

WED WELD WmL WmL WIRE HEAT NO.90146 WIRE HEAT N090146 TIRE HEAT NO.90146 WIRE HEAT NO.90146 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 d-USE T o LE35 d-T o LE35 Curve nuence USE I

0 8466 0

-896 2

0 6726 26

-54.79 3

0 97.96 1329

-6051 4

0 6114

-16.51

-436 5

0 72.49

-l116

-2762 0

35.07 29.35 4626 6224 U)

P--

Ct

4

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve legend I

o-20--

3 0 4 -

5 Data Set(s) Plotted Material Curve Plant Capsule 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.%-<

I 0

2 0

3 0

4 0

5 0

T 0 50z. Shr

-77.81

-20.47

-47.81

-30.46

_m2s 0

57.34 30 47.34 65.62 C.

U)

C.)

I

-300

-200

-100 0

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

3 0 4 ^

5

~

Data Set(s) Plotted Ilaterial Curve Plant Capsule Ori.

Ueat Material On llptI I

WC1 UNIRR 2

lC!

U 3

lCI Y

4 lCI V

5 liCI X

HEAT AFFD ZONE HEAT AFFD ZONE HEAT AFF'D ZONE HEAT AFFD ZONE HEAT AFFD ZONE WERE HEAT NO.90146 lIRE HEAT NO.S0146 WME HEAT NO.90146 WIRE HEAT NO090146 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, 00F AL53, 250F AL52, 400F AL56, 500F AL55, 750F 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,-750F AT50,-250 F AT60,15-F AT56, 500 F AT54, 750F 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 0F AW52, -350F AW53, 0F AW51, 250F AW58, 500F AW46, 750F AW59, 1000F AW57, 1250F AW55, 125TF AW48, 150TF AW60, 160TF AW56, 2000F AW50,2250F AW49, 2500F AW54, 2500F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Wolf Creek Reactor Vessel Weld Metal Testing of Specimens from Capsule X

5-32 AH53, -1750F AH51, -1000F AH50,-750 F AH49,-500F AH58,-500F AH52, -25 0F 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 50 100 150 200 250 300 120 110 100 Cn 90 l 80 La 70 C-,

60 50 40 I

I I

I I

I l_

ULTIMATE TENSILE STRENGTH

-A

-A A

=

A_

A A 0

0 0

0 0.2% YIELD STRENGTH 800 700 600 Cs 500 400 300 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

Figure 5-17 0

100 200 300 400 500 600 TEMPERATURE (OF)

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 50 100 150 200 250 300 120 110 100 C-1 La I-~

90 80 70 I

I I

I I

I I_

I&

ULTIMATE TENSILE STRENGTH 2

A-2 0

0 0

% Y S2 0 2% YIELD STRENGTH 800 700 600 500 400 300 60 50 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 I-1 W-4

-J 0-4 C-,

60 50 40 30 2

REDUCTION IN AREA A

A A

o TOTAL ELONGATION 2

0 UNIFORM ELONGATION I

I I

I I

10 0

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 50 100 150 200 250 300 120 110 100 90 80 70 It Lin C-,

I I

I I

I I

I ULTIMATE TENSILE STRENGTH A

't 0.2% YIELD STRENGTH 2/

800 700 600

-d 0l-60 50 40 500 400 300 LEGEND:

A 0 UNIRRADIATED 19 2

0 As IRRADIATED TO A FLUENCE OF 3.49 X 10 nlcm (E>1.OMeV) AT 550 F I-1 t-40-J-

-j I--

80 70 60 50 40 30 REDUCTION IN AREA TOTAL ELONGATION 2

UNIFORM ELONGATION

  • I I

I I

I 20 10 0

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 750F Specimen AL 1I Tested at 3 000 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 750F Specimen AT 1 Tested at 3000F 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 750F Specimen AW 1I Tested at 300'F Specimen AW12 Tested at 5500F 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 20 10 0

AL-10 75 F 0

0.05 0 1 0.15 STRAIN, INAN 0.2 025 03 100 90 80 70 y

60 co50 Uj 40 30 20 10 0

STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAP.X' AL-11 300 F 0

005 01 0.15 STRAIN, INAN 0.2 025 0.3 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 20 10 0

0 Figure 5-23 AL-12 550 F 005 01 0 15 STRAIN, INAN 02 025 03 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 30 20 10 0

AT-10 75 F 0

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

90 -

80 -

70 -

i 60 -

(6 co 50 -

I 40 -

30 -

20 -

10 -

0-AT-1 1 300 F 0

0.05 0.1 0.15 STRAIN, IN/IN 02 0.25 03 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 CO50 w

40 30 20 10 0

AT-12 550 F 0

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

543 STRESS-STRAIN CURVE WOLF CREEK UNIT 1 CAPSULE *X' 100 90 80 70 to~4 60 to 50 w

,,, 40 30 20 10 0

AW-10 75 F 0

005 0.1 0.15 0.2 025 03 STRAIN, ININ 100 90 80 70 60 t) 50 0

40 30 20 10 0

0 STRESS-STRAIN CURVE BEAVER VALLEY UNIT 2 W CAPSULE AW-11 300 F 005 01 0.15 STRAIN, ININ 02 025 03 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 30 20 10 0

AW-12 550 F 0

0 05 0 1 0.15 0.2 0.25 STRAIN, ININ 03 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 library1241. 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 23SU(np)l 7Cs (Cd) 3.01E-14 2.56E-14 1.18 27Np(n,f)'3 7Cs (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 D

-c DORT Geometry 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)

Cycle 1

2 3

4 5

6 7

8 9

10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time

[EFPY]

1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00 Flux

[E>1.0 MeV]

[n/cmA2-sec]

29 Deg 31.5 Deg 8.73E+10 9.33E+10 9.08E+10 1.OOE+11 7.67E+10 8.33E+10 7.30E+10 8.04E+10 7.12E+10 7.60E+10 6.66E+10 7.05E+10 6.44E+10 6.98E+10 7.44E+10 7.91E+10 6.22E+10 7.1 OE+10 8.19E+10 8.64E+10 7.23E+10 8.25E+10 7.29E+10 8.29E+10 6.97E+10 7.79E+10 7.29E+10 8.29E+10 7.29E+10 8 29E+10 7.29E+10 8.29E+10 7.29E+10 8.29E+10 7.29E+10 8.29E+10 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 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time

[EFPY]

1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00 Displacements

[dpa]

29 Deg 31.5 Deg 5.79E-03 6.1 8E-03 9.61 E-03 1.04E-02 1.28E-02 1.39E-02 1.79E-02 1.95E-02 2.32E-02 2.52E-02 2.74E-02 2.96E-02 3.25E-02 3.52E-02 3.80E-02 4.1 OE-02 4.32E-02 4.69E-02 4.99E-02 5.40E-02 5.59E-02 6.09E-02 6.20E-02 6.78E-02 6.96E-02 7.63E-02 8.96E-02 9.91 E-02 1.08E-01 1.19E-01 1.43E-01 1.60E-01 1.79E-01 2.01 E-01 2.15E-01 2.42E-01 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 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time

[EFPY]

1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00 0 Deg 4.26E+1 7 7.11E+17 9.50E+1 7 1.39E+18 1.84E+18 2.14E+18 2.45E+1 8 2.80E+18 3.14E+18 3.55E+1 8 3.95E+1 8 4.33E+18 4.86E+1 8 6.12E+18 7.26E+1 8 9.53E+1 8 1.18E+19 1.41 E+19 1.58E+19 Maximum Pressure Vessel Fluence

[E>1.0 MeV]

[n/cmA 2]

15 Deg 30 Deg 6.32E+17 7.36E+17 1.04E+18 1.21 E+18 1.39E+18 1.62E+18 1.99E+18 2.27E+18 2.63E+18 2.96E+18 3.14E+18 3.50E+18 3.66E+18 4.14E+18 4.28E+18 4.85E+18 4.74E+18 5.53E+18 5.40E+18 6.36E+18 5 96E+18 7.13E+18 6.54E+18 7.91 E+18 7.31 E+18 8.88E+18 9.23E+18 1.14E+19 1.09E+19 1.37E+19 1.44E+19 1.83E+19 1.78E+1 9 2.28E+1 9 2.12E+19 2.74E+19 2.38E+19 3.08E+19 45 Deg 7.44E+17 1.29E+18 1.69E+18 2.37E+1 8 3.03E+18 3.55E+18 4.18E+18 4.79E+18 5.54E+1 8 6.33E+1 8 7.19E+18 8.1 OE+18 9.20E+1 8 1.22E+19 1.49E+19 2.03E+1 9 2.57E+19 3.11E+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 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time

[EFPY]

1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32 00 40 00 48.00 54.00

[dpa/sec]

Maximum Iron Atom Displacements 0 Deg 1.95E-11 2.16E-1 1 1.75E-11 1.90E-1 1 1.79E-11 1.48E-11 1.27E-11 1.46E-1 1 1.21 E-11 1.55E-11 1.43E-1 1 1.40E-1 1 1.45E-1 1 1.40E-11 1.40E-1 1 1.40E-11 1.40E-11 1.40E-1 1 1.40E-1 1 15 Deg 2.87E-1 1 3.06E-1 1 2.51 E-11 2.58E-1 1 2.56E-1 1 2.44E-1 1 2.08E-11 2.54E-1 1 1.68E-1 1 2.40E-1 1 2.02E-1 1 2.1 OE-11 2.1 OE-1 1 2.1OE-1 1 2.1 OE-11 2.1OE-1 1 2.1 OE-11 2.10E-1 1 2.10E-1 1 30 Deg 3.35E-1 1 3.56E-1 1 2.95E-1 1 2.82E-1 1 2.73E-1 1 2.61 E-11 2.58E-1 1 2.93E-1 1 2.43E-1 1 3.06E-1 1 2.77E-1 1 2.79E-1 1 2.67E-1 1 2.79E-1 1 2.79E-1 1 2.79E-1 1 2.79E-1 1 2 79E-11 2.79E-1 1 45 Deg 3.47E-1 1 4.22E-1 1 2.93E-1 1 3.01 E-1 1 2.70E-1 1 2.57E-1 1 2.57E-1 1 2.61 E-11 2.73E-1 1 3.OOE-1 1 3.13E-1 1 3.37E-11 3.07E-1 1 3.37E-1 1 3.37E-11 3.37E-1 1 3.37E-1 1 3.37E-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 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 Projection Projection Projection Projection Projection Projection Projection Cumulative Operating Time

[EFPY]

1.07 1.75 2.43 3.57 4.79 5.82 7.12 8.33 9.78 11.10 12.47 13.83 15.53 20.00 24.00 32.00 40.00 48.00 54.00

[dpa]

Maximum Iron Atom Displacements 0 Deg 6.62E-04 1.11 E-03 1.48E-03 2.16E-03 2.85E-03 3.33E-03 3.82E-03 4.36E-03 4.89E-03 5.53E-03 6.15E-03 6.75E-03 7.57E-03 9.55E-03 1.13E-02 1.49E-02 1.84E-02 2.20E-02 2.46E-02 15 Deg 9.74E-04 1.60E-03 2.14E-03 3.07E-03 4.05E-03 4.84E-03 5.64E-03 6.59E-03 7.31 E-03 8.32E-03 9.19E-03 1.01 E-02 1.1 3E-02 1.42E-02 1.69E-02 2.22E-02 2.75E-02 3.28E-02 3.68E-02 30 Deg 1.14E-03 1.87E-03 2.49E-03 3.51 E-03 4.56E-03 5.41 E-03 6.40E-03 7.50E-03 8.54E-03 9.82E-03 1.1 OE-02 1.22E-02 1.37E-02 1.77E-02 2.12E-02 2.82E-02 3.53E-02 4.23E-02 4.76E-02 45 Deg 1.18E-03 2.04E-03 2.67E-03 3.75E-03 4.79E-03 5.62E-03 6.61 E-03 7.59E-03 8.76E-03 1.OOE-02 1.14E-02 1.28E-02 1.45E-02 1.93E-02 2.36E-02 3.21 E-02 4.06E-02 4.91 E-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]

.s 1

.Xfnn f1 I K A [ 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 2 )

(a 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, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.
8.

Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix G. Fracture Toughness Criteria for Protection Against Failure

9. ASTM El 85-82, Standard Practice for 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 Bar Impact Testing of Metallic Materials, ASTM, 1998.
14. ASTM A370-97a, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 1997.

References

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

18. ASTM E 185-79, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels
19. WCAP-143 70, Use of the Hyperbolic Tangent Function for Fitting Transition Temperature Toughness Data, T. R. Mager, et al, May 1995.
20. Regulatory Guide RG-1.190, Calculatonal and Dosimetry Methods for Determining Pressure 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 Overpressure Mitigating System Setpoints and RCSHeatup and Cooldown Limit Curves, January 1996.

22 WCAP-15557, Revision 0, Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology, August 2000.

23. RSICC Computer Code Collection CCC-650, DOORS 3.1, One, Two-and Three-Dimensional Discrete Ordinates Neutron/Photon Transport 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 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 23'U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the 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 238U(y,f) 0.97 0.97 0.97 0.97 Net 2 3 8U Correction 0.84 0.81 0.78 0.73 237lNp(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 238U 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 6 3 Cu(nQ)60Co 5%

54Fe(np) 54 Mn 5%

58 Ni(np)58Co 5%

238U(nf) 137Cs 10%

237Np(n,f)' 37Cs 10%

59Co(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 (g _ g,) 2 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) 0 5 (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 Generation (MWt-hr)

Year I

l Year Month I Mn19 I

Tllt I

I onth 1994 12 2636320 I1998 2632 1995 1995 1995 1995 1995 1995 i

1 2

3 4

5 6

2639803 2383600 2020996 2543603 2634001 2524794 1998 1998 1998 1998 1998 1998 3

4 5

6 7

_8 Thermal Generatio n

(MWt-hr) 2394162 2649959 2564698 2649737 2561103 2642611 2649749 2564818 Year 2001 2001 2001 2001 2001 2001 2001 2001 Month 4

6 7

8 9

1 10 1

Thermal Generation l (MWt-IIr) 2562211 2631325 2565829 2651544 2651448 2565781 2654261 1995 1995 1995 I

7 8

9 2632369 2634129 2468895 1998 1998 9

10 2001 2001 11 2651470

  • 1-I I

1995 1995 1995 10 I1 12 2637097 2549850 2640215 1998 1998 1999 I

I10 11 12 I

25i 261 1996 1

2481700 1999 3

2 1996 2

0 1999 4

10 1996 3

0 1999 5

174 1996 4

1784195 1999 6

25(

1996 5

2622027 1999 7

264 1996 1

6 2349911 1999 8

251

2649145, 51237 43012 16340 58213 t2613 4209 40059 64943 49843 4800 I.

2001 2002 2002 200 I VV0 I

7 26428 I11 1 qqq a

oUt

__I_

L 1 I 1996 8

2603985 1999 10 2652619 1996 9

2564802 1999 11 2543480 1996 10 2621302 1999 12 2624035 iu 1 l Z563J1 2000 l

2642844 1996 l 12 2649864 2000 2

2468736 V) 2 3

2650720 2388288 1836759 I

I

-1 '__. -

I I

I 77 I I

Ilhaxq I Ifuln nd:rnOlca inn-I i

i iI

-' I L0JUL I

I YY Z

2393586 2000n A

IOc~noI n

I___ _vv I^D j~ I 1997 1

3OTA2i 2 EOo rn 1997 3

2650251 2000 5

2638828 1997 4

12564428 1 2000 6

2564614 4

1 I

I I

UU-7 I

Iz I 'ii. o~n,-

I

[-

tU-I I7I7I

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

-7 7e AC4q 1)

ILI~¶1C

/1I I 1CV-1997 6

25-63505 2000 8

2650187 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 Zt9 7 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/cm2 -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 Ad

.justed 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 2U (n,f) 3 7Cs (Cd)

Middle 1.43E+05 5.90E+06 I 5.90E+06 3.87E-14 238U (n,f) 13 7Cs (Cd)

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

3.26E-14 237Np (nf) 137CS (Cd)

Middle 1.24E+06 5.12E+07 5.12E+07 3 26E-13 237Np (nf) 137CS (Cd)

Including Yfis sion correction.

3.23E-13 59Co (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 59Co (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 237Np (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) 63Cu (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) 58Co 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) 137CS (Cd)

Middle 5.43E+05 5.33E+06 I 5.33E+06 3.50E-14 238u (nf) 137CS (Cd)

Including 235U, 239pu, and yfission corrections:

2.84E-14 237Np (nf) 13 7Cs (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 59Co (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-1 7 Average 4.41E-17 54Fe (n,p) 54Mn 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 4.74E-15 Average 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 6 83E-15 Average 6.45E-15 238U (n,f) '"Cs (Cd)

Middle 1.14E+06 5.91E+06 5.91E+06 3.88E-14 23U (n,f)

Cs (Cd)

Including 235U, 239Pu, 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 59Co (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 59Co (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) 63Cu (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 7Cs (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 Co (n,Y) 60Co Top 4.42E+07 6.05E+07 6.05E+07 3.95E-12 Bottom 4.44E+07 6.08E+07 6.08E+07 3.96E-12 Average 3.96E-12 59Co (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 63Cu(n,u)6OCo 5.52E-17 4 81E-17 5.39E-17 1.15 1.02 54Fe(np) 4 Mn 5.89E-15 5.45E-15 5.89E-15 1.08 1.00 58Ni(n,p)"Co 7.91E-15 7 65E-15 8.16E-15 1.03 0.97 238U(n,f) 37Cs (Cd) 3.26E-14 2.96E-14 3.16E-14 1.10 1 03 237Np(n'f) 137Cs (Cd) 3.23E-13 2.92E-13 3.17E-13 1.11 1.02 5 9Co(nY)60Co 5.29E-12 4.22E-12 5.20E-12 1.25 1.02 5 9Co(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 6 3 Cu(nax)6 0Co 4.74E-17 4.24E-17 4.53E-17 1.12 1.05 54Fe(n,p)54 Mn 4.50E-15 4.68E-15 4.65E-15 0.96 0.97 "Ni(n,p) 58Co 6.1 IE-l5 6.56E-15 6.45E-15 0.93 0.95 238 U(nf)'37Cs (Cd) 2.84E-14 2.51E-14 2.52E-14 1.13 1.12 237Np(nf)137Cs (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) 60Co (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 63Cu(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 58Ni(n,p)"8Co 6.45E-15 6.09E-15 6.56E-15 1.06 0.98 238U(n,f) 37Cs (Cd) 3.01E-14 2.32E-14 2.57E-14 1.30 1.18 237Np(nf)'37Cs (Cd) 2.67E-13 2.26E-13 2.60E-13 1.18 1.03 59Co(n,y)6oCo 3.15E-12 3.18E-12 3.12E-12 0.99 1.01 59Co(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)60Co 4.6513-17 4.34E-17 4.53E-17 1.07 1.03 54Fe(n,p)5 4Mn 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 238U(n,f) 37Cs (Cd) 2 3 7 Np(n 1f)3 7Cs (Cd) 3.01E-14 2.56E-14 2.62E-14 1.18 1.15 5 9 Co(n,y)6oCo 2.56E-13 2.50E-13 2.57E-13 1.02 1.00 5 9Co(n,-y) 6 0Co (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

'est Uncertainty.

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 Atom DisplacementRate [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 63Cu(n,a)60Co 1.15 1.12 1.11 1.07 54Fe(n,p) Mn 1.08 0.96 1.02 0.99 58Ni(np)58Co 1.03 0.93 1.06 0.97 238U(np)137Cs (Cd) 1.10 1.13 1.30 1.18 237Np(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/cm2-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, FERRETData Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-4.

RSIC Data Library Collection DLC-1 78, "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 Time-1 (ms)

AL48, -500F 4 00 500 6 00 5000 00 4000 00 as (5' 3000 00 0

-J 2000.00 1000 Do 0 00 m

4

-

0.00 1 00 200 300 Time-I (ms)

AL57, 0F 400 500 600 5000 00-4000 00

.0

,, 3000 00-0

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

AL53, 250F 600 Appendix B

~fiLl B-2 1

5000 00i 4000 00 l 1f

.0

-jo 000 1 00 200 300 400 500 6

Time-1 (ms)

AL52, 400F 1I 5000ooj 40 4000 00 I m',3000 00 I 0

2000 001}

100000 0 000 00 100 200 Ann 0o 0

Tine-1 (ms)

AL56, 500F 4 -

uu b6U oa3 0-j 000 100 200 300 400 500 Time-1 (ms)

AL55, 75TF 6 00 Appendix B

B-3 5000 00-4000 00 Z 300000 0

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

AL59, 1100F 5000 00 4000 00 300000 2000 00 1000 00 000 0 00 1 00 2 00 3 00 4 00 5 00 6 00 Time-1 (ms)

AL5O, 1350F 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 00 o 01 5000 00 4000 00 Q

3000 00 0,

2000 00 1000000 0000 001 5000

.o 4000 00

.0.

co 3000 00 0

-Jl 200000 100000 000 )

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

AL49, 1750F 600 3

1 00 2 00 3 00 4 00 5 00 Time-I (ms)

AL51, 1900F 6 00 000 1 00 200 300 400 500 Time-I (ms)

AL46, 2250 F 6 00 Appendix B

B-5 5000 00 400000 n

3000 00 20X00 0 0 0 00 1.00 2.00 3 00 4 00 5 00 6 00 Time-i (ms)

AL54, 2500 F 50000 Do 4000 00

'7 300000 2000DO0l 1 000 00 l

=

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

AL58, 2750F 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 Z-u 3000-00 03 0

2000 00 1000 00 non ll 1111 r

v Ad at 111111-ant

+

l t

Am -

A r

-sB eL at It 000 1 00 200 300 Time-i (ms)

AT46, -750F 400 500 6 00 5000 00-4000 00

.0 3 3000 00-03

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

AT50, -250F 4 00 5 00 6 00 3

0

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

AT60, 15-F 600 Appendix B

B-7 5000 Co 4000 00 n

3000 00

-J 2000 00 1000 00 0 00 0

5000 00-4000 00 x

3000 00-

-J 2000 00-1000 00 0 00 0

5000 00 4000 00

-J 3000 00 2000 00 1 000 00 00 1 00 200 300 400 500 Time-1 (ms)

AT56, 50TF 600 Time-1 (ms)

AT54, 750F 000 Il.O 2 00 3 00 4 00 5 00 Time-1 (ms)

AT53, 1000F 600 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 30000 0 2000 00 100000 0 00 000 1 00 200 300 400 500 600 Time-1 (ms)

AT58, 1500F 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, 1750F Appendix B

B-9

.0 0

-. 1 5000 00 4000 00.

3000.00 2000 00 1000 00 0 00 00i'0 I

ha 1 00 2.00 3 00 Time-1 (ms)

AT51, 1750F 4 00 50o 600 5000 00 4000 00 m 3000 00

-0 2000 00 1000 00 II

lilt 000 1 00 200 300 Time-1 (ms)

AT57, 2000F 4 00 500 6 00 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 Time-1 (ms)

AT52, 225 0F 400 500 600 Appendix B

it B-10 5000 00 4000 00 n

3000 00 2000 00 1000 00 0 00 5000 00 4000 00

.0.

as 3000 00 0

2000 00 1000 00 0 00 00 5000 00 4000 00 30

-~3000 00.

0o Time-1 (ms)

AT49, 2500F 0

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

AT47, 2750F 600 000 1 00 200 300 4 00 500 Time-I (ms)

AT55, 3000F 6 00 Appendix B

B-Il 5000 OO f 4000 00 7 3000 001 0

-j 2000 00 1000 00 l 000 0 00 5000 00l 4000 001 a,

-J 2000 00 1000 00 00 c

-o.

1 00 200 300 Time-1 (ms)

AW47, -750 F 4 00 5 00 600 I

I-0.00 1 0o 200 300 Time-1 (ms)

AW52, -350 F 4 00 5 00 600 5000 00.

4000 00 A0 3000 00-0-J 2000 00]

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

AW53, 0F 6 00 Appendix B

1.

B-12 4000 2

'o 3000 03

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

AW51, 250F 6 00 5000 00 4000 00 n

'a 0

-J 03

-J 3000 00-2000.00-1000 001 000 1 00 200 300 400 500 Time-I (ms)

AW58, 500F 6 00 2.00 3 00 Time-i (ms)

AW46, 75TF 600 Appendix B

B-13 0

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

.s 0

-j A,

0-J 3000 00 2000 00 1000 00 000 1 00 200 300 400 500 Time-i (ms)

AW57, 125TF 600 000 1 00 200 3.00 400 500 Time-I (ms)

AW55, 125TF 6 00 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, 1500F 5000 00 400000

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

AW60, 1600F 500000o 400000 300000 2000 00 100000 000 000 1 00 200 300 400 500 600 rTme-i (Ms)

AW56, 2000F Appendix B

B-15

.0 0

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

AW5O, 2250F 6 00 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 1 000 00-0000 0.00 1 00 2.00 3 00 4 00 5 00 Time-1 (ms)

AW54, 2500 F 600 Appendix B

II B-16 4000 00 n

3000 00 s

-J 2000 00 1000 00-0 oo4 0 (

5000 DE1 4000 00 3000 00

-J*.

2000 00 100000 0 001-00 5000 00 4000 00 r 300000]-

X, I

o II 2000 0011 1000 00 01 00 200 300 400 500 Time-I (ms)

AH53, -1750F 6 00 0

1 00 200 300 400 500 Time-1 (ms)

AH51, -1000 F 600 000 1 00 200 300 400 500 Time-1 (ms)

AH50, F 6 00 Appendix B

B-17 5000 00 4000 00 7

3000 00 0

2000 00O 1000 00

^ AA n nn u uu i

0 00 1 00 200 300 Time-I (ms)

AH49, -500 F 400 5 00 6 00 5000 00 4000 00 2

.0 r

3000 00 0

-J 2000 00 1000.00-0 00 0X 5000 00 4000 00

(

3000 00 0

-J 2000 00 1000.00 Time-i (ms)

AH58, -50F uuu-i 000 1 00 200 300 Time-1 (is)

AH52, -250F 4 00 5 00 6 00 Appendix B

B-18

.0m 0

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

AH47, 0F 600 5000 00 4000 00 c, 3000 00 0-2 2000 00 1000 00.

0 00 01 5000 00 4000 00

.\\

m' 3000 00 0

-j Time-1 (ms)

AH55, 0F 000 1 00 200 300 400 500 Time-I (ms)

AH59, 250F 600 Appendix B

B-19 n

0

-J Time-I (ms)

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

-J 2000 00-1000 co 000-5000 00 4000 00 X5 3000 00 0-J 2000 00 1000 00

)o I

I I

I 100 2 00 3 00 Time-i (ms)

AH54, 500F 400 500 6 00 I

nnn.

l nil 000 1 00 200 300 Time-1 (rns)

AH60, 1000F 400 500 600 Appendix B

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

AH46, 1500F 6 00 000 1 00 200 300 400 500 Time-1 (ms) 6 00 Q

0

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

AH56, 2000F 600 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 Upper Shelf Energy: 142 Fixed Material:

Equation is CVN = A + B I tanh((T - TO)/C I Temp. at 30 ft-lbs 36.1 Temp. at 50 ft-lbs:

PLATE SA533B1 Heat Number C4935-2 I

Capsule: X Total Fluence:

67.7 Lower Orientation: LT Shelf Energy 2.19 Fixed U]

10 V

300 250-200 15 0-0 0

i too0 100~l 5fF

-300

-200

-100 0

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

500 600 II: C4935-2 Temperature

-50 0

25 40 50 75 110 135 150 2

13 21 37 53 43 74 108 133 6.69 15.69 24.66 32.07 37.92 55.45 8358 102.02 11L23 Differential

-4.69

-2.69

-3 66 4.92 15.07

-12.45

-958 5.97 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)

Input CVN Energy Computed CVN Energy Differential 100 12298

-22.98 122 12804

-6.04 150 135.47 1452 146 13828 7.71 135 139.9

-4.9 137 140.2

-3.82 SUM of RESIDUALS = -86 Temperature 175 190 225 250 275 300 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 Upper Shelf L.E. 74.32 Material: PLATE Equation is LK = A + B * [ tanh((T - TO)/C) I Temperature at LK 35:

72.9 Lo SA533B1 Heat Number C4935-2 Capsule: X Total Fluence:

wer Shelf LE-I Fixed Orientation: LT Wi 200 150 0

50 0

0 0 --

-.jUU

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plantk WC1 Cap: X Material PLATE SA533B1 OriL LT H6 Charpy V-Notch Data Input Lateral Expansion Computed LEK 400 F

500 600 it # C4935-2 Temperature

-50 0

25 40 50 75 110 135 150 0

5 12 24 33 29 47 67 74 3.09 8.64 14.77 19.98 24.09 36.02 52.71 61.59 65.37 Differential

-3.09

-3.64

-2.77 4.01 8.9

-7.02

-5.71 5.4 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)

Input Lateral Expansion Computed LE Differential 64 69.52

-552 67 71.07

-4.07 71 73.04

-2.04 75 73.67 132 75 73.99 1

75 7415 B4 SUM of RESIDUALS =-3.79 Temperature 175 190 225 250 275 300 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 Capsule: X Total Fluence Orientation: LT

-4 4

V])

6aC 2--3C Temperature

)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 Input Percent Shear Computed Percent Shear Differential

-50 0

25 40 50 75 110 135 150 2

5 10 15 20 30 50 65 90 1.65 5.4 10.63 14.96 18.59 30.47 5.18 67.68 75.59

.34

-. 4

-.63

.03 L4

-.47

-218

-2.68 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)

Input Percent Shear Computed Percent Shear Differential 75 85.59

-10.59 85 89.78

-4.78 100 95.62 4.37 100 97.67 2.32 100 98.77 122 100 99.35

.64 SUM of RESIDUALS = 255 Temperature 175 190 225 250 275 300 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 Material: PLATE SA533B1 Heat Number C4935-2 Capsule: X Total Fluence:

97 Lower Shelf Energy: 2.19 Fixed Orientation: TL Y) 25 10 A4m 2C P-e 15 z0 Cz; 10 5

Tempera

-75

-25 15 50 75 100 125 150 175 0~

-300

-200

-100 0

100 200 300 400 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 Lture Input CVN Energy Computed CVN Energy 5

433 11 8.36 15 15.85 30 27.49 41 38.82 52 51.51 55 63.78 67 74.06 79 81.69 500 600 Differential

.66 2.63

-.85 25 217

.48

-. 78

-7.06

-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)

Input CVN Energy Computed CYN Energy Differential 98 8169 16.3 88 86.85 L14 91 9013

.86 93 9214

.85 101 93.33 7.66 96 94.03 1.96 SUM of RESIDUALS = 17.86 Temperature 175 200 225 250 275 300 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 Upper Shelf LE: 64.44 Material: PLATE Equation is: LE. = A + B I [ tanh((T - TO)/C) I Temperature at LE 35:

102.4 Lo SA533B1 Heat Number. C4935-2 Capsule: X Total Fluence:

ower Shelf LE I Fixed Orientation: TL M

"-4

. F-200-7 150 100 0

0 P~

I F-f CZ

. X 0

I Ir If I

I I

I

-300

-200

-100 0

100 200 300 400 500 600 Temperature in I Data Set(s) Plotted Plant WC1 Cap: X Material: PLATE SA533B1

)egrees F Ori: TL Heat # C4935-2 Charpy V-Notch Data Temperature Input Lateral Expansion

-75

-25 15 50 75 100 125 150 175 0

4 8

20 29 36 38 50 53 Computed LE 2.65 5.54 10.7 1845 25.88 3419 42.3 49.23 54.51 Differential

-2.65

-1.54

-2.7 L54 311 1.8

-4.3

.76

-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)

Input lateral Expansion Computed LE.

Differential 57 5451 2.48 51 582

-72 69 60.61 8.38 60 6213

-2.13 66 63.06 2.93 61 6362

-2.62 SUJM of RESIDUALS = -366 Temperature 175 200 225 250 275 300 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 rE SA533BI Heat Number. C4935-2 0:

Capsule: X Total Fluence:

Material: PLA1 rientation: TL

.-4 ou a)0 cD 60 4a) 40 2F

-30 Temperature

,\\

a DO

-2U0

-100 0

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

500 600

  1. C4935-2 Differential

-75

-25 15 50 75 100 125 150 175 2

5 15 25 40 45 55 65 60 1.83 5.35 1208 23 3421 47.52 6118 7329 8269 16

-.35 2.91 L99 5.78

-252

-6.18 4-29

-Z69 Data continued on next page Ir' C-12

CAPSULE X (TRANSVERSE ORIENTATION)

Page 2 Material:

Temperature 175 200 225 250 275 300 PLATE SA533B1 Heat Number C4935-2 Orientatj Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Input Percent Shear Computed Percent Shear 90 82.69 95 8926 100 93.54 100 9618 100 97.77 100 98.7 Sul ion: TL Differential 7.3 5.73 6.45 3.81 2Z2 129 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 Material: WELD Heat Number WIRE HEAT NO.90146 Capsule: X Total Fluence:

Lower Shelf Energy: 219 Fixed Orientation:

C')

C-)

30 09--

-a 250-20f 10 n

(F

-I

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant. WC1 Cap.: X Material WELD Or.

Heat lh WERE HI Charpy V-Notch Data Input CVN Energy Computed CVN Energy 400 F

500 600 EAT NO.90146 Temperature

-75

-35 0

25 50 75 100 125 125 4

13 23 37 53 64 68 78 76 839 1535 25.88 361 47.74 59.37 6954 77.47 77.47 Differential

-4.39

-235

-2188

.89 5.25 4.62

-1.54

.52

-147 Data continued on next page *'*

C-14

CAPSULE X (WELD)

Page 2 Material: WE Temperature 150 160 200 225 250 250

)LD Heat Number. WIRE HEAT NO.90146 0

Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Input CVN Energy Computed CVN Energy 67 8312 89 84.82 84 8927 102 90.75 96 91.65 94 91.65 Si rientation:

Differential

-16.12 4.17

-5.27 1124 4.34 2.34 JM 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 Upper Shelf LE 6234 Material: WELD Equation is LE = A + B

  • I tanh((T - TO)/C) I Temperature at LE 35:

532 Lower Shelf LE. 1 Fixed Heat Number WIRE HEAT NO.90146 Orientation:

Capsule X Total Fluence M) pS--

200 150 10 0

50 e9/

I I

1-

-300

-200

-100 0

100 200 300 400 F

500 600 Temperature in Degrees Data Set(s) Plotted PlantL WC1 Cap: X Material WELD Ori.

Heat #: WIRE HEAT NO.90146 Charpy V-Notch Data Temperature Input Lateral Expansion

-75

-35 0

25 50 75 100 125 0

5 15 25 35 45 51 53 55 Computed LK 3.47 7.62 15.33 23.76 33.69 4322 50.7 55.72 55.72 Differential

-3.47

-2.62

-.33 123 1L3 177

.29

-2.72

-.72 Data continued on next page C-16

CAPSULE X (WELD)

Page 2 Temperature 150 160 200 225 250 250 Material: WELD Heat Number. WIRE HEAT NO.90146 Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Input Lateral Expansion Computed LK 54 58.73 62 59.53 54 6L34 64 6182 64 62.07 69 62.07 Orientation:

Differential

-4.73 2.46

-7.34 217 1.92 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 Capsule: X Total Fluence:

Orientation:

pCi C-)

AH

-300

-200

-100 0

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

-75

-35 0

25 50 75 100 125 125 10 20 45 50 60 80 85 95 95 10X84 2262 38.64 5213 65.33 76.52 84.93 90.69 9069 Differential

-.84

-2.62 6.35

-213

-5.33 3.47

.06 43 43 Data continued on next page C-18

CAPSULE X (WELD)

Page 2 Material: I Temperature 150 160 200 225 250 250 ELD Heat Number. WIRE HEAT NO.90146 Or Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Input Percent Shear Computed Percent Shear 85 94.4 95 95.45 95 98.05 100 98.86 100 99.34 100 99.34 Differential

-9.4

-.45

-3.05 113

.65

.65 S -2.91 ientation:

SUM of RESIDUAL' 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 Material: HEAT AFFD ZONE Heat Number: WIRE HEAT NO.90146 Capsule: X Total Fluence:

Lower Shelf Orientation:

Energy: 2.19 Fixed 9) 10 p---q I

zq

-4 Q)

i4 C-)

300-250 200 15 1

0 0

0 I-

10.

I 50

.70 0-

-I

--OW

-)u

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant. WC1 Cap: X Material: HEAT AFFD ZONE OrL Heat # M Charpy V-Notch Data ire Input CVN Energy Computed CVN Energy 400 F

500 600 Temperat.

-175

-100

-75

-50

-50

-25 0

0 25 6

15 34 29 88 50 92 152 98 32 15.0 29.4 531 531 81.9 106.

106.2 1213 7

9 9

9 1

WIRE HEAT NO.90146 Differential 2.78

-.07 4.5

-2419 34B

-3L91

-1424 45.75

-23.33 e*** Data continued on next page ***

C-20

CAPSULE X (HAZ)

Page 2 Material: HEAT Temperature 35 50 100 150 200 200 AFPD ZONE Heat Number. WIRE HEAT NO.90146 Capsule X Total Fluence:

Charpy V-Notch Data (Continued)

Input CVN Energy Computed CVN Energy 132 125.08 140 128.96 146 133.91 140 134.81 124 134.96 29 134.96 St Orientation:

Differential 6.91 11.03 1a0B 518

-10.96

-5.96 JM of RESIDUAIS = 12.36 C-21

II CAPSULE X (HAZ)

P4 a)

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 15 100 5Y 0

0 5~~~01

-;UU

-W0u Plant WC1 Cap.:

-100 0

100 200 Temperature in Degi Data Set(s) Plotted X

Material HEAT AFFD ZONE Ori.:

Charpy V-Notch Data ut Lateral Expansion Compute 300 rees 400 F

500 600 Temperature Inp

-175

-100

-75

-50

-50

-25 0

0 25 0

5 12 12 42 21 44 70 52 L41 6.0, 222 222 361 510' 510:

61.4' Heat # WIRE HEAT NO90146 d LE Differential 1

-141

-1.07 08

.11

1

-1021 1

19.78 6

-15.6 3

-7.03 3

18.96 5

-9.45 Data continued on next page C-22

CAPSULE X (HAZ)

Page 2 Temperature 35 50 100 150 200 200 Material: HEAT AFFD ZONE Heat Number. WIRE HEAT NO.90146 Capsule: X Total Fluence:

Charpy V-Notch Data (Continued)

Input Lateral Expansion Computed LE.

70 6428 70 67.34 70 715 71 72.31 72 72.46 74 72.46 Orientation:

Differential 5.71 2.65

-1.5

-1.31

-.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 Capsule: X Total Fluence:

Orientation:

C) 04 Tempera

-175

-10

-75

-50

-50

-25 r

4 0

50

-25 DU I

30 0xr

-300

-200

-100 0

100 200 300 400 500 Temperature in Degrees F Data Set(s) Plotted Plant WC1 Cap.: X Material: HEAT AFFD ZONE Ori:

Heat P WYIRE HEAT NO.90146 Charpy V-Notch Data ture Input Percent Shear Computed Percent Shear Diff I

I 0

2 5

15 30 20 50 100 65 2

a4 8.37 19.15 1915 38.03 61.39 61.39 80.47 600 erential

-2

-1.4

-337

-4.15 1084 18.03 11.39 38.6 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)

Input Percent Shear Computed Percent Shear Differential 75 85.77

-10.77 100 91.43 8.56 100 9862 137 100 99.79

.2 100 99.96

.03 100 99.96

.03 SUM of RESIDUAIS = -518 Temperature 35 50 100 150 200 200 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 280F 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) = 39.10F 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) = 44.10F Notes-(a) f = fluence Calculated fluence from Section 6 of this report [x 1019 n/cm2, E > 1.0 MeV]

(b)

FF = fluence factor = f(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 2 F(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 +/- 1 a 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 +/- 250F.

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 250F. 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

Retrieved from "https://kanterella.com/w/index.php?title=ML031060076&oldid=5422929"