ULNRC-06259, Results of Analysis Capsule W, Reactor Vessel Radiation Surveillance Program

Results of Analysis Capsule W, Reactor Vessel Radiation Surveillance Program

ML15288A518
Person / Time
Site:	Callaway
Issue date:	10/15/2015
From:	Maglio S Ameren Missouri, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-06259 WCAP-18001-NP, Rev. 0
Download: ML15288A518 (315)

Text

2/JIJlereII Cal laway Plant MISSOURI October 15, 2015 ULNRC-06259 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-000 1 10 CFR 50.61(b) 10 CFR 50 Appendix H Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

RENEWED FACILITY OPERATING LICENSE NPF-30 RESULTS OF ANALYSIS OF CAPSULE W FROM AMEREN MISSOURI (UNION ELECTRIC) CALLAWAY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Per 10 Cf R 50, Appendix H, this letter transmits the report, WCAP- 18001-NP, Revision 0, which provides the results of the testing of surveillance Capsule W from the Callaway Unit 1 reactor vessel.

Capsule W was removed during Refuel 20 (November 2014) at 25.75 Effective full-Power Years (EFPY). All surveillance capsules have now been removed from the Callaway Unit 1 reactor vessel, and ex-vessel dosimetry has been installed for future measurement of vessel fluence. Post irradiation mechanical testing of the Capsule W Charpy V-notch and tensile specimens was performed, and a fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDf/3-VT database.

A summary of results includes the calculated peak clad/base metal vessel fluence after 25.75 EFPY of plant operation was 1.49 x i nlcm2 (E> 1.0 MeV), and the fluence received by Capsule W was 5.98 x i nlcm2 (E> 1.0 MeV). The report also includes a brief summary of the Charpy V-notch testing results. Appendix E to the report presents an upper-shelf energy evaluation, using the screening criteria of 10 CFR 50 Appendix G, that documents a conclusion that with consideration of surveillance data, all beltline and extended beltline materials exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY).

P0 Box 620  : Fulton, MO 65251 : AmerenMissouri.com STARS Alliance

ULNRC- 06259 October 15, 2015 Page 2 With the receipt of WCAP- 18001-NP, Revision 0, the Callaway Pressure and Temperature Limits Report (PTLR) document will require revision. The PTLR revision is anticipated to be ready for transmittal to the NRC, in accordance with Technical Specification 5.6.6.c, during the first quarter of 2016.

This letter does not contain new commitments.

If there are any questions, please contact Jim Nurrenbem at 314-225-1908.

Sincerely, Scott A. Maglio Manager, Regulatory Affairs JPKI Attachment WCAP-18001-NP, Revision 0, Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program

ULNRC- 06259 October 15, 2015 Page 3 cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission

$201 NRC Road Steedman, MO 65077 Mr. L. John Klos Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 08H4 Washington, DC 20555-0001

ULNRC- 06259 October 15, 2015 Page 4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 6100 Western Place, Suite 1050 fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Other Situations ULNRC Distribution:

f. M. Diya D. W. Neterer L. H. Graessle T. E. Herrmann B. L. Cox M. A. McLachlan S. A. Maglio M. M. Hall T. B. Elwood M. G. Hoehn, III J. D. Nurrenbem Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. John ONeill (Pillsbury Winthrop Shaw Pittman LLP)

Missouri Public Service Commission

Attachment to ULNRC-06259 WCAP-1$OO 1-NP, Revision 0, Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program 310 Pages

Westinghouse Non-Proprietary Class 3 WCAP-18001-NP September 2015 Revision 0 Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program

Westinghouse Non-Proprietary Class 3 WCAP-18001-NP Revision 0 Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program Elaine M. Ruminski*

Materials Center of Excellence Arzu Alpan*

Nuclear Operations & Radiation Analysis September 2015 Reviewers: Elliot J. Long*

Materials Center of Excellence Eugene T. Hayes*

Nuclear Operations & Radiation Analysis Approved: David B. Love*, Acting Manager Materials Center of Excellence Laurent P. Houssay*, Manager Nuclear Operations & Radiation Analysis

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2015 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES .................................................................................................................................... vii EXECUTIVE

SUMMARY

........................................................................................................................... x 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 W ..................................................................... 5-1 5.1 OVERVIEW .................................................................................................................... 5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS........................................................... 5-2 5.3 TENSILE TEST RESULTS ............................................................................................. 5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ....................................................... 6-1

6.1 INTRODUCTION

........................................................................................................... 6-1 6.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 6-2 6.3 NEUTRON DOSIMETRY .............................................................................................. 6-4 6.4 CALCULATIONAL UNCERTAINTIES ........................................................................ 6-5 7 SURVEILLANCE CAPSULE REMOVAL

SUMMARY

............................................................ 7-1 8 REFERENCES ............................................................................................................................. 8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ............................................................................................. A-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS .................................... B-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD ........................................................ C-1 APPENDIX D CALLAWAY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION ....

........................................................................................................................................ D-1 APPENDIX E CALLAWAY UNIT 1 UPPER-SHELF ENERGY EVALUATION................................ E-1 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 4-1 Chemical Composition (wt. %) of the Callaway Unit 1 Reactor Vessel Surveillance Materials (Unirradiated)................................................................................................... 4-3 Table 4-2 Heat Treatment History of the Callaway Unit 1 Reactor Vessel Surveillance Materials . 4-4 Table 5-1 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation) ........................ 5-5 Table 5-2 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation) ........................... 5-6 Table 5-3 Charpy V-Notch Data for the Callaway Unit 1 Surveillance Program Weld Metal (Heat #

90077) Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)................................ 5-7 Table 5-4 Charpy V-Notch Data for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) ............................................ 5-8 Table 5-5 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

(Longitudinal Orientation) ............................................................................................... 5-9 Table 5-6 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

(Transverse Orientation) ................................................................................................ 5-10 Table 5-7 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Surveillance Program Weld Metal (Heat #90077) Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) .....

....................................................................................................................................... 5-11 Table 5-8 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)................ 5-12 Table 5-9 Effect of Irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Callaway Unit 1 Reactor Vessel Surveillance Capsule W Materials ........................................................................................................................ 5-13 Table 5-10 Comparison of the Callaway Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper-Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions ..................................................................................................................... 5-14 Table 5-11 Tensile Properties of the Callaway Unit 1 Capsule W Reactor Vessel Surveillance Materials Irradiated to 5.98 x 1019 n/cm2 (E > 1.0 MeV) .............................................. 5-15 Table 6-1 Calculated Fast Neutron Fluence Rate (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 ..................................................................... 6-7 Table 6-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections .................................... 6-8 Table 6-3 Calculated Iron Atom Displacement Rate at the Surveillance Capsule Center and at Core Midplane for Cycles 1 Through 20 .................................................................................. 6-9 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 v Table 6-4 Calculated Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections ........................................................... 6-10 Table 6-5 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence Rates (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface ........................................................... 6-11 Table 6-6 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface ..................................................................... 6-12 Table 6-7 Calculated Azimuthal Variation of Maximum Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface ..................................................................... 6-13 Table 6-8 Calculated Azimuthal Variation of Maximum Iron Atom Displacements at the Reactor Vessel Clad/Base Metal Interface .................................................................................. 6-14 Table 6-9 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Callaway Unit 1 ............................................................................................................................. 6-15 Table 6-10 Calculated Surveillance Capsule Lead Factors .............................................................. 6-15 Table 6-11 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface.............................................................................................. 6-16 Table 6-12 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface ......................................................................................................................... 6-19 Table 7-1 Surveillance Capsule Withdrawal Summary.................................................................... 7-1 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors .................................... A-9 Table A-2 Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor .......................................................................................................................... A-10 Table A-3 Surveillance Capsules U, Y, V, X, and W Fluence Rates for Cj Calculation, Core Midplane Elevation ....................................................................................................................... A-15 Table A-4 Surveillance Capsules U, Y, V, X, and W Cj Factors, Core Midplane Elevation .......... A-16 Table A-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U .............. A-17 Table A-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y............... A-18 Table A-7 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V............... A-19 Table A-8 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X .............. A-20 Table A-9 Measured Sensor Activities and Reaction Rates for Surveillance Capsule W .............. A-21 Table A-10 Least-Squares Evaluation of Dosimetry in Surveillance Capsule U (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycle 1 Irradiation ................................................. A-22 Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule Y (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 4 Irradiation .............................. A-23 Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 8 Irradiation .............................. A-24 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 vi Table A-13 Least-Squares Evaluation of Dosimetry in Surveillance Capsule X (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 10 Irradiation ............................ A-25 Table A-14 Least-Squares Evaluation of Dosimetry in Surveillance Capsule W (31.5° Azimuth, Core Midplane - Single Capsule Holder) Cycles 1 Through 20 Irradiation ......................... A-26 Table A-15 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions ..................................................................................................... A-27 Table A-16 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... A-27 Table C-1 Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH ................................................ C-2 Table C-2 Upper-Shelf L.E. Value (mils) Fixed in CVGRAPH ...................................................... C-2 Table D-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Callaway Unit 1 Surveillance Data ................................................................................................. D-4 Table D-2 Callaway Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line ................. D-5 Table E-1 Callaway Unit 1 Pressure Vessel 1/4T Fast Neutron Fluence Calculation ..................... E-1 Table E-2 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 35 EFPY ....................... E-3 Table E-3 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 54 EFPY ....................... E-4 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 vii LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Callaway Unit 1 Reactor Vessel .............. 4-5 Figure 4-2 Capsule W Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeter .................................................................................................................. 4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ................................................ 5-16 Figure 5-1(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued ............................ 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ................................................ 5-18 Figure 5-2(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued ............................ 5-19 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ................................................ 5-20 Figure 5-3(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued ............................ 5-21 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) .................................................... 5-22 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued ............................... 5-23 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) .................................................... 5-24 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued ............................... 5-25 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) .................................................... 5-26 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued ............................... 5-27 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ........................................................ 5-28 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued ................................... 5-29 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ............................................. 5-30 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued ........................ 5-31 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 viii Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ........................................................ 5-32 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued ................................... 5-33 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material ......................................................................................... 5-34 Figure 5-10(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued ..................................................................... 5-35 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material .............................................................................. 5-36 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued .......................................................... 5-37 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material ......................................................................................... 5-38 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued ..................................................................... 5-39 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ............................................................ 5-40 Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) ............................................................... 5-41 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ........................................................ 5-42 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material .................................................................................................. 5-43 Figure 5-17 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV) ... 5-44 Figure 5-18 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV) ....... 5-45 Figure 5-19 Tensile Properties for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV) ........... 5-46 Figure 5-20 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) .............................................................................. 5-47 Figure 5-21 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) ................................................................................. 5-48 Figure 5-22 Fractured Tensile Specimens from the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ............................................................................. 5-49 Figure 5-23 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Longitudinal Orientation) ........................................................................... 5-50 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 ix Figure 5-24 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Transverse Orientation) .............................................................................. 5-51 Figure 5-25 Engineering Stress-Strain Curves for Callaway Unit 1 Program Weld Metal (Heat #

90077) Tensile Specimens ............................................................................................. 5-52 Figure 6-1 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules and 12.5° Neutron Pad Configuration ....................................... 6-22 Figure 6-2 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Single Capsule Holder and 20.0° Neutron Pad Configuration .................................................. 6-23 Figure 6-3 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Dual Capsule Holder and 22.5° Neutron Pad Configuration .................................................. 6-24 Figure 6-4 Callaway Unit 1 r,,z Reactor Geometry Section View at 31.5° Azimuthal Angle with Surveillance Capsule...................................................................................................... 6-25 Figure E-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence ..................................................................................... E-2 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 x EXECUTIVE

SUMMARY

The purpose of this report is to document the testing results of surveillance Capsule W from Callaway Unit 1. Capsule W was removed at 25.75 Effective Full Power Years (EFPY) and post-irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database. Capsule W received a fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) after irradiation to 25.75 EFPY. The peak clad/base metal interface vessel fluence after 25.75 EFPY of plant operation was 1.49 x 1019 n/cm2 (E > 1.0 MeV).

This evaluation led to the following conclusions: 1) The measured percent decreases in upper-shelf energy for the surveillance plate (longitudinal and transverse orientations) and weld materials contained in Callaway Unit 1 Capsule W are less than the Regulatory Guide 1.99, Revision 2 [Ref. 1] predictions.

2) The Callaway Unit 1 surveillance plate is judged to be non-credible and the surveillance weld (Heat #

90077) is judged to be credible. This credibility evaluation can be found in Appendix D. 3) With consideration of surveillance data, all beltline and extended beltline materials exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2]. The upper-shelf energy evaluation is presented in Appendix E.

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

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 1-1 1

SUMMARY

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

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

Appendix C presents the CVGRAPH, Version 6.0, Charpy V-notch plots for Capsule W and previous capsules, along with the program input data.

Capsule W received an average fast neutron fluence (E > 1.0 MeV) of 5.98 x 1019 n/cm2 after 25.75 effective full power years (EFPY) of plant operation.

Irradiation of the reactor vessel Lower Shell Plate R2708-1 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 55.9F and an irradiated 50 ft-lb transition temperature of 92.3F. This results in a 30 ft-lb transition temperature increase of 58.6F and a 50 ft-lb transition temperature increase of 69.1F for the longitudinally oriented specimens.

Irradiation of the reactor vessel Lower Shell Plate R2708-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 77.1F and an irradiated 50 ft-lb transition temperature of 139.3F. This results in a 30 ft-lb transition temperature increase of 95.2F and a 50 ft-lb transition temperature increase of 111.2F for the transversely oriented specimens.

Irradiation of the Surveillance Program Weld Metal (Heat # 90077) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 12.8F and an irradiated 50 ft-lb transition temperature of 42.0F. This results in a 30 ft-lb transition temperature increase of 65.8F and a 50 ft-lb transition temperature increase of 60.7F.

Irradiation of the Heat Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -8.6F and an irradiated 50 ft-lb transition temperature of 30.6F.

This results in a 30 ft-lb transition temperature increase of 91.6F and a 50 ft-lb transition temperature increase of 84.3F.

The average upper-shelf energy (USE) of Lower Shell Plate R2708-1 (longitudinal orientation) resulted in an average energy decrease of 9 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 117 ft-lb for the longitudinally oriented specimens.

The average upper-shelf energy of Lower Shell Plate R2708-1 (transverse orientation) resulted in an average energy decrease of 21 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 83 ft-lb for the transversely oriented specimens.

The average upper-shelf energy of the Surveillance Program Weld Metal (Heat # 90077) Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 103 ft-lb for the weld metal specimens.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 1-2 The average upper-shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 5 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 101 ft-lb for the HAZ Material.

Comparisons of the measured 30 ft-lb shift in transition temperature values and upper-shelf energy decreases to those predicted by Regulatory Guide 1.99, Revision 2 [Ref. 1] for the Callaway Unit 1 reactor vessel surveillance materials are presented in Table 5-10.

Based on the credibility evaluation presented in Appendix D, the Callaway Unit 1 surveillance plate is non-credible and the surveillance weld material (Heat # 90077) is credible.

Based on the upper-shelf energy evaluation in Appendix E, all beltline and extended beltline materials contained in the Callaway Unit 1 reactor vessel exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2].

The maximum calculated 54 EFPY (end-of-license renewal) neutron fluence (E > 1.0 MeV) for the Callaway Unit 1 reactor vessel beltline using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the Guide) is as follows:

Calculated (54 EFPY): Vessel clad/base metal interface fluence* = 2.91 x 1019 n/cm2 Vessel 1/4 thickness fluence = 1.73 x 1019 n/cm2

• This fluence value is documented in Table 6-6.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 2-1 2 INTRODUCTION This report presents the results of the examination of Capsule W, the fifth capsule removed and tested in the continuing surveillance program, which monitors the effects of neutron irradiation on the Ameren Missouri Callaway Unit 1 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Callaway Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Company. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are contained in WCAP-9842 [Ref.

3], Union Electric Company Callaway Unit No. 1 Reactor Vessel Radiation Surveillance Program. The surveillance program was originally planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73 [Ref. 4], Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Capsule W was removed from the reactor after 25.75 EFPY of exposure and shipped to the Westinghouse Materials Center of Excellence 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 and post-irradiation data obtained from surveillance Capsule W removed from the Callaway Unit 1 reactor vessel and discusses the analysis of the data.

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Westinghouse Non-Proprietary Class 3 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 Callaway Unit 1 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 [Ref. 5]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop-weight nil-ductility transition temperature (NDTT per ASTM E208 [Ref. 6]) or the temperature 60F 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 Section XI of the ASME Code

[Ref. 5]. The KIc curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIc curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Callaway Unit 1 reactor vessel radiation surveillance program, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (RTNDT) 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 (initial RTNDT + M + RTNDT) 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.

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Westinghouse Non-Proprietary Class 3 4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Callaway Unit 1 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel, as shown in Figure 4-1, between the neutron shielding pads and the vessel wall, at various azimuthal locations. The vertical center of the capsules is opposite the vertical center of the core. The capsules contain specimens made from the following:

Lower Shell Plate R2708-1 (longitudinal orientation)

Lower Shell Plate R2708-1 (transverse orientation)

Weld metal fabricated with weld wire Heat Number 90077, Linde Type 124 flux, Lot Number 1061, which is identical to that used in the actual fabrication of the intermediate to lower shell circumferential weld and is the equivalent heat number used in the actual fabrication of the intermediate shell and lower shell longitudinal weld seams; however, these vessel welds used Linde Type 0091 flux in their fabrication Weld heat affected zone (HAZ) material of Lower Shell Plate R2708-1 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 relieving treatment on the test material. Test specimens were also removed from weld metal of a stress relieved weldment joining Lower Shell Plate R2708-1 and adjacent Intermediate Shell Plate R2707-1.

All heat-affected zone specimens were obtained from the weld heat affected zone of Lower Shell Plate R2708-1.

Charpy V-notch impact specimens from Lower Shell Plate R2708-1 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major rolling direction).

The core-region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular (normal) 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 R2708-1 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented perpendicular to the welding direction.

Compact Tension (CT) specimens from Lower Shell Plate R2708-1 were machined in both the longitudinal and transverse orientations. CT specimens from the weld metal were machined with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399 [Ref. 7].

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Westinghouse Non-Proprietary Class 3 4-2 All six capsules contain 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 (237Np) and Uranium (238U) were placed in the capsules to measure the integrated flux at specific neutron energy levels.

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

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

The chemical composition and the heat treatment of the unirradiated surveillance materials in Capsule W are presented in Tables 4-1 and 4-2, respectively. The data in Tables 4-1 and 4-2 was obtained from the original surveillance program report, WCAP-9842 [Ref. 3].

Capsule W was removed after 25.75 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile and compact tension specimens, dosimeters, and thermal monitors.

Figures 4-1 and 4-2 detail the arrangement of the surveillance capsules in the reactor and the placement of specimens within Capsule W, respectively.

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Westinghouse Non-Proprietary Class 3 4-3 Table 4-1 Chemical Composition (wt. %) of the Callaway Unit 1 Reactor Vessel Surveillance Materials (Unirradiated)

Lower Shell Plate R2708-1(a) Weld Metal(a)

Element WCAP-9842 WCAP-11374 WCAP-9842 WCAP-11374 Analysis Analysis Analysis Analysis

[Ref. 3] [Ref. 14] [Ref. 3] [Ref. 14]

C 0.220 0.230 0.150 0.110 Mn 1.470 1.320 1.370 1.280 P 0.006 <0.005 0.005 0.005 S 0.014 0.016 0.008 0.016 Si 0.250 0.220 0.440 0.470 Ni 0.590 0.550 0.070 0.060 Mo 0.570 0.490 0.540 0.500 Cr 0.050 0.060 0.040 0.040 Cu 0.070 0.060 0.060 0.030 Al 0.025 -- 0.003 --

Co 0.013 0.006 0.011 0.004 Pb -- -- <0.001 --

W <0.010 -- <0.010 --

Ti <0.010 <0.001 <0.010 <0.001 Zr <0.001 -- <0.001 --

V 0.003 0.010 0.004 <0.010 Sn 0.002 -- 0.003 --

As 0.001 -- <0.001 --

Cb <0.010 -- <0.010 --

N 0.008 -- 0.007 --

Sb -- -- 0.0015 --

B <0.001 -- 0.001 --

Note:

(a) Data obtained from WCAP-15400, Revision 0 [Ref. 17].

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Westinghouse Non-Proprietary Class 3 4-4 Table 4-2 Heat Treatment History of the Callaway Unit 1 Reactor Vessel Surveillance Materials Material(a) Temperature(a) Time(a) Coolant(a)

Austenitized @ 1600 +/- 25 4 hrs. Water-Quenched (871°C)

Lower Shell Plate R2708-1 Tempered @ 1225 +/- 25 (663°C) 4 hrs. Air-Cooled Stress Relieved @ 1150 +/- 50 13 hrs. Furnace-Cooled (621°C)

Surveillance Program Weld Stress Relieved @ 1150 +/- 50 7 hrs. 45 min. Furnace-Cooled Metal (Heat # 90077) (621°C)

Note:

(a) Data obtained from WCAP-9842 [Ref. 3].

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Westinghouse Non-Proprietary Class 3 4-5 Figure 4-1 Arrangement of Surveillance Capsules in the Callaway Unit 1 Reactor Vessel WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 4-6 Figure 4-2 Capsule W Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeter WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-1 5 TESTING OF SPECIMENS FROM CAPSULE W 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed at the Westinghouse Materials Center of Excellence Hot Cell Facility. Testing was performed in accordance with 10 CFR 50, Appendix H [Ref. 2] and ASTM Specification E185-82 [Ref.

8].

Capsule W was opened upon receipt 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-9842 [Ref. 3]. All of the items were in their proper locations.

Examination of the thermal monitors indicated that 1 of the 3 temperature monitors had melted, as described below:

Top Thermal Monitor, the 579°F (304°C) was melted.

Middle Thermal Monitor, the 590°F (310°C) was not melted.

Bottom Thermal Monitor, the 579°F (304°C) was not melted.

Based on this examination, the maximum temperature to which the specimens were exposed was less than 590°F (310°C), but greater than 579°F (304°C).

The Charpy impact tests were performed per ASTM Specification E185-82 [Ref. 8] and E23-12c [Ref. 9]

on a Tinius-Olsen Model 74, 358J machine. The Charpy machine striker was instrumented with an Instron Impulse system. Instrumented testing and calibration were performed to ASTM E2298-13a [Ref.

10].

The instrumented striker load signal data acquisition rate was 819 kHz with data acquired for 10 ms.

From the load-time curve, the load of general yielding (Fgy), the maximum load (Fm) and the time to maximum load were 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 brittle fracture load (Fbf). The termination load after the fast load drop is identified as the arrest load (Fa). Fgy, Fm, Fbf, and Fa were determined per the guidance in ASTM Standard E2298-13a [Ref. 10].

The energy at maximum load (Wm) was determined by integrating the load-time record to the maximum load point. 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 (WP) is the difference between the total energy (Wt) and the energy at maximum load (Wm). Wt is compared to the dial energy (KV). Wt derived from the instrumented striker were all within 15% of the calibrated dial energy values as required in ASTM E2298-13a [Ref. 10].

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Westinghouse Non-Proprietary Class 3 5-2 Percent shear was determined from post-fracture photographs using the ratio-of-areas method in compliance with ASTM E23-12c [Ref. 9] and A370-13 [Ref. 11]. The lateral expansion was measured using a dial gage rig similar to that shown in the same specifications.

Tensile tests were performed on a 250 KN Instron screw driven tensile machine (Model 5985) per ASTM E185-82 [Ref. 8]. Testing met ASTM Specifications E8/E8M-13a [Ref. 12] or E21-09 [Ref. 13]. Load was applied through a clevis and pin connection. The strain rate obtained met the requirements of ASTM E8/E8M-13a [Ref. 12] and ASTM E21-09 [Ref. 13].

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 11-inch hot zone. Tensile specimens were soaked at temperature (+/-5ºF) for a minimum of 20 minutes before testing. All tests were conducted in air.

The tensile specimens were, nominally, 4.230 inches long with a 1.00 inch gage section and a reduced section of 1.50 inches long by 0.250 inch in diameter, as documented in Figure 2-2 of WCAP-9842 [Ref.

3]. The yield load, ultimate load, fracture load, uniform elongation and elongation at fracture were determined directly from the load-extension curve. The yield strength (0.2% offset method), ultimate tensile strength and fracture strength were calculated using the original cross-sectional area. Yield point elongation (YPE) was calculated as the difference in strain between the upper yield strength and the onset of uniform strain hardening using the methodology described in E8/E8M-13a [Ref. 12]. The final diameter and final gage length were determined from post-fracture photographs. This final diameter measurement was used to calculate the fracture stress (true stress at fracture) and the percent reduction in area. The final and original gage lengths were used to calculate total elongation after fracture.

5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule W, which received a fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) in 25.75 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with the unirradiated and previously withdrawn capsule results as shown in Figures 5-1 through 5-12. The unirradiated and previously withdrawn capsule results were taken from WCAP-9842 [Ref. 3], WCAP-11374 [Ref. 14], WCAP-12946 [Ref. 15], WCAP-14895 [Ref. 16] and WCAP-15400 [Ref. 17]. The previous capsules, along with the original program unirradiated material input data, were updated using CVGRAPH, Version 6.0. This accounts for the differences in measured values of 30 ft-lb and 50 ft-lb transition temperature between the results documented in this report and those shown in prior Callaway Unit 1 capsule reports.

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

Irradiation of the reactor vessel Lower Shell Plate R2708-1 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 55.9F and an irradiated 50 ft-lb transition temperature of 92.3F. This results in a 30 ft-lb transition temperature increase of 58.6F and a 50 ft-lb transition temperature increase of 69.1F for the longitudinally oriented specimens.

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Westinghouse Non-Proprietary Class 3 5-3 Irradiation of the reactor vessel Lower Shell Plate R2708-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 77.1F and an irradiated 50 ft-lb transition temperature of 139.3F. This results in a 30 ft-lb transition temperature increase of 95.2F and a 50 ft-lb transition temperature increase of 111.2F for the transversely oriented specimens.

Irradiation of the Surveillance Program Weld Metal (Heat # 90077) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 12.8F and an irradiated 50 ft-lb transition temperature of 42.0F. This results in a 30 ft-lb transition temperature increase of 65.8F and a 50 ft-lb transition temperature increase of 60.7F.

Irradiation of the Heat Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -8.6F and an irradiated 50 ft-lb transition temperature of 30.6F.

This results in a 30 ft-lb transition temperature increase of 91.6F and a 50 ft-lb transition temperature increase of 84.3F.

The irradiated upper-shelf energy of Lower Shell Plate R2708-1 (longitudinal orientation) resulted in an average energy decrease of 9 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 117 ft-lb for the longitudinally oriented specimens.

The average upper-shelf energy of Lower Shell Plate R2708-1 (transverse orientation) resulted in an average energy decrease of 21 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 83 ft-lb for the transversely oriented specimens.

The average upper-shelf energy of the Surveillance Program Weld Metal (Heat #90077) Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 103 ft-lb for the weld metal specimens.

The average upper-shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 5 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 101 ft-lb for the HAZ Material.

Comparisons of the measured 30 ft-lb shift in transition temperature values and upper-shelf energy decreases to those predicted by Regulatory Guide 1.99, Revision 2 [Ref. 1] for the Callaway Unit 1 reactor vessel surveillance materials are presented in Table 5-10.

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Westinghouse Non-Proprietary Class 3 5-4 The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-13 through 5-16. The fractures show an increasingly ductile or tougher appearance with increasing test temperature. Load-time records for the individual instrumented Charpy specimens are contained in Appendix B.

With consideration of the surveillance data, all beltline materials exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2]. This evaluation can be found in Appendix E.

5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule W irradiated to 5.98 x 1019 n/cm2 (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-17 through 5-19.

The results of the tensile tests performed on the Lower Shell Plate R2708-1 (longitudinal orientation) indicated that irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) caused increases in the 0.2 percent offset yield strength and the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-17 and Table 5-11.

The results of the tensile tests performed on the Lower Shell Plate R2708-1 (transverse orientation) indicated that irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) caused increases in the 0.2 percent offset yield strength and the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-18 and Table 5-11.

The results of the tensile tests performed on the Surveillance Program Weld Metal (Heat # 90077) indicated that irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) caused increases in the 0.2 percent offset yield strength and the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-19 and Table 5-11.

The fractured tensile specimens for the Lower Shell Plate R2708-1 (longitudinal orientation) material are shown in Figure 5-20; the fractured tensile specimens for the Lower Shell Plate R2708-1 (transverse orientation) material are shown in Figure 5-21; the fractured tensile specimens for the Surveillance Program Weld Metal (Heat # 90077) are shown in Figure 5-22. The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-25.

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Westinghouse Non-Proprietary Class 3 5-5 Table 5-1 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)

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

CL32 0 -18 7 9 6 0.2 5 CL38 20 -7 14 19 9 0.2 10 CL31 35 2 10 14 7 0.2 15 CL45 40 4 42 57 30 0.8 25 CL39 50 10 38 52 27 0.7 25 CL40 50 10 33 45 23 0.6 20 CL34 72 22 41 56 31 0.8 30 CL37 90 32 38 52 29 0.7 40 CL44 100 38 52 71 38 1.0 40 CL36 120 49 62 84 49 1.2 60 CL41 150 66 77 104 58 1.5 70 CL33 170 77 101 137 70 1.8 85 CL42 200 93 102 138 67 1.7 95 CL43 230 110 122 165 76 1.9 100 CL35 250 121 126 171 76 1.9 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-6 Table 5-2 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation)

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

CT31 0 -18 17 23 13 0.3 5 CT42 20 -7 11 15 6 0.2 5 CT36 40 4 20 27 15 0.4 10 CT40 50 10 21 28 14 0.4 10 CT43 60 16 32 43 22 0.6 20 CT44 60 16 28 38 18 0.5 15 CT38 72 22 32 43 22 0.6 25 CT32 100 38 40 54 29 0.7 25 CT34 120 49 38 52 30 0.8 25 CT35 150 66 46 62 34 0.9 45 CT39 170 77 50 68 40 1.0 55 CT37 200 93 67 91 46 1.2 75 CT45 230 110 74 100 55 1.4 100 CT41 250 121 92 125 68 1.7 100 CT33 275 135 82 111 62 1.6 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-7 Table 5-3 Charpy V-Notch Data for the Callaway Unit 1 Surveillance Program Weld Metal (Heat # 90077) Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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

CW45 -25 -32 5 7 2 0.1 5 CW36 0 -18 25 34 18 0.5 25 CW40 10 -12 15 20 14 0.4 30 CW32 20 -7 36 49 26 0.7 35 CW34 25 -4 55 75 43 1.1 55 CW42 35 2 26 35 22 0.6 45 CW33 40 4 60 81 46 1.2 60 CW44 50 10 69 94 52 1.3 65 CW43 60 16 51 69 43 1.1 60 CW38 72 22 77 104 56 1.4 75 CW37 120 49 89 121 76 1.9 90 CW31 150 66 87 118 76 1.9 90 CW35 170 77 99 134 73 1.9 95 CW39 200 93 96 130 77 2.0 100 CW41 250 121 115 156 81 2.1 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-8 Table 5-4 Charpy V-Notch Data for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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

CH41 -50 -46 14 19 9 0.2 10 CH44 -40 -40 17 23 11 0.3 15 CH42 -25 -32 33 45 24 0.6 30 CH37 0 -18 30 41 33 0.8 25 CH43 10 -12 50 68 34 0.9 50 CH34 20 -7 45 61 30 0.8 55 CH40 35 2 56 76 39 1.0 60 CH33 50 10 53 72 40 1.0 70 CH39 60 16 41 56 31 0.8 75 CH31 70 21 49 66 38 1.0 75 CH35 72 22 94 127 61 1.5 70 CH36 100 38 100 136 72 1.8 90 CH38 150 66 107 145 78 2.0 100 CH32 170 77 93 126 68 1.7 100 CH45 200 93 104 141 65 1.7 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)

Total Energy to General Total Dial Fracture Test Instrumented Difference, Max Maximum Time to Yield Arrest Sample Energy, Load, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fa Number KV Fbf

(°F) Wt (%) Wm (lb) (msec) Fgy (lb)

(ft-lb) (lb)

(ft-lb) (ft-lb) (lb)

CL32 0 7 6.04 14 3.62 4100 0.09 3500 3700 0 CL38 20 14 11.66* 17* 9.45* 3900* 0.20* 3400* 3800* 0*

CL31 35 10 8.09* 19* 3.41* 4000* 0.09* 3500* 3400* 0*

CL45 40 42 39.5 6 35.91 4400 0.6 3200 4200 0 CL39 50 38 34.97 8 27.30 4200 0.48 3100 4000 300 CL40 50 33 29.89 9 27.45 4200 0.48 3100 4200 0 CL34 72 41 37.13 9 35.29 4300 0.60 3000 4200 300 CL37 90 38 31.64* 17* 23.28* 4100* 0.43* 3000* 3900* 600*

CL44 100 52 45.70 12 34.43 4200 0.61 3000 4000 1000 CL36 120 62 56.08 10 34.12 4200 0.60 3000 4000 1200 CL41 150 77 72.07 6 33.25 4100 0.60 2900 3800 2300 CL33 170 101 96.47 4 33.16 4100 0.61 2900 3000 1900 CL42 200 102 99.18 3 33.02 4100 0.61 2800 3200 2900 CL43 230 122 118.63 3 32.75 4000 0.60 2700 0 0 CL35 250 126 121.98 3 39.34 4100 0.72 2700 0 0

• The difference between instrumented Charpy and dial values was greater than 15%. The values were not adjusted per Reference 10.

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Westinghouse Non-Proprietary Class 3 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation)

Total Energy to General Total Dial Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Sample Energy, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa Number KV

(°F) Wt (%) Wm (lb) (msec) Fgy (lb) (lb)

(ft-lb)

(ft-lb) (ft-lb) (lb)

CT31 0 17 16.72 2 3.66 4200 0.09 3400 4100 0 CT42 20 11 10.40 5 3.47 4100 0.09 3500 3900 0 CT36 40 20 18.71 6 17.13 4100 0.31 3300 4100 0 CT40 50 21 19.90 5 16.73 4100 0.31 3200 4000 0 CT43 60 32 29.73 7 24.57 4200 0.43 3300 4100 0 CT44 60 28 25.02 11 19.64 4200 0.35 3300 4100 0 CT38 72 32 28.79 10 25.20 4200 0.43 3200 4200 300 CT32 100 40 35.79 11 23.33 4100 0.43 3100 3800 800 CT34 120 38 33.78 11 26.50 4000 0.48 3000 4000 1000 CT35 150 46 42.48 8 26.00 4000 0.48 2900 3800 1500 CT39 170 50 46.78 6 22.50 3900 0.43 2900 3900 2300 CT37 200 67 64.51 4 22.55 4000 0.43 3100 3600 2300 CT45 230 74 71.86 3 2.23 4100 0.09 2600 0 0 CT41 250 92 89.35 3 28.67 4300 0.56 2800 0 0 CT33 275 82 80.20 2 31.79 3900 0.60 2500 0 0 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Surveillance Program Weld Metal (Heat #90077)

Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

Total Energy to General Total Dial Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Sample Energy, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa Number KV

(°F) Wt (%) Wm (lb) (msec) Fgy (lb) (lb)

(ft-lb)

(ft-lb) (ft-lb) (lb)

CW45 -25 5 3.40* 32* 2.25* 3900* 0.08* 3500* 3600* 0*

CW36 0 25 23.11 8 3.56 4100 0.09 3600 3100 600 CW40 10 15 10.97* 27* 3.66* 4100* 0.09* 3300* 3700* 400*

CW32 20 36 32.30 10 3.26 4000 0.09 3000 3800 400 CW34 25 55 52.28 5 33.92 4100 0.60 3100 3800 1800 CW42 35 26 24.25 7 12.88 3800 0.26 3200 3600 1500 CW33 40 60 56.77 5 33.68 4100 0.60 3100 3900 2600 CW44 50 69 66.04 4 33.55 4100 0.60 2900 3700 2000 CW43 60 51 49.13 4 33.23 4000 0.60 2900 3800 1700 CW38 72 77 73.97 4 35.29 4200 0.60 3100 4000 2500 CW37 120 89 87.14 2 32.60 3900 0.61 2800 2000 1500 CW31 150 87 84.34 3 31.49 3800 0.60 2800 2700 2200 CW35 170 99 95.12 4 31.18 3800 0.60 2700 2100 1800 CW39 200 96 93.08 3 31.24 3800 0.60 2700 0 0 CW41 250 115 110.20 4 42.51 3700 0.83 2300 0 0

• The difference between instrumented Charpy and dial values was greater than 25%. The values were not discarded as required by Reference 10.

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Westinghouse Non-Proprietary Class 3 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

Total Energy to General Total Dial Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Sample Energy, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa Number KV

(°F) Wt (%) Wm (lb) (msec) Fgy (lb) (lb)

(ft-lb)

(ft-lb) (ft-lb) (lb)

CH41 -50 14 12.48 11 3.70 4500 0.09 3600 3900 0 CH44 -40 17 15.84 7 3.77 4500 0.09 3500 4100 0 CH42 -25 33 30.46 8 3.96 4500 0.09 3500 4300 0 CH37 0 30 27.84 7 3.52 4100 0.09 3200 3900 900 CH43 10 50 45.98 8 35.62 4300 0.60 3100 4200 1600 CH34 20 45 41.11 9 3.38 4200 0.09 3200 3800 1500 CH40 35 56 52.88 6 35.79 4100 0.61 3400 3900 1100 CH33 50 53 50.56 5 34.93 4100 0.61 3200 4100 2200 CH39 60 41 39.69 3 3.52 4000 0.09 3200 3600 2500 CH31 70 49 47.13 4 14.67 3900 0.29 3000 3200 2200 CH35 72 94 87.72 7 36.59 4400 0.60 3400 3700 1600 CH36 100 100 97.91 2 34.32 4200 0.61 3000 2000 1700 CH38 150 107 102.46 4 32.29 4000 0.61 2700 0 0 CH32 170 93 90.96 2 31.60 3800 0.60 2600 0 0 CH45 200 104 101.06 3 32.27 4000 0.60 2900 0 0 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-13 Table 5-9 Effect of Irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Callaway Unit 1 Reactor Vessel Surveillance Capsule W Materials Average 30 ft-lb Transition Average 35 mil Lateral Expansion Average 50 ft-lb Transition Average Energy Absorption at Material Temperature(a) (°F) Temperature(a) (°F) Temperature(a) (°F) Full Shear(b) (ft-lb)

Unirradiated Irradiated T Unirradiated Irradiated T Unirradiated Irradiated T Unirradiated Irradiated E Lower Shell Plate R2708-1 -2.7 55.9 58.6 26.7 88.0 61.3 23.2 92.3 69.1 126 117 -9 (Longitudinal)

Lower Shell Plate R2708-1 -18.1 77.1 95.2 28.9 132.5(c) 103.6 28.1 139.3 111.2 104 83 -21 (Transverse)

Surveillance Weld Material -53.0 12.8 65.8 -21.7 34.7 56.4 -18.7 42.0 60.7 107 103 -4 (Heat #90077)

Heat Affected

-100.2 -8.6 91.6 -42.1 27.1 69.2 -53.7 30.6 84.3 106 101 -5 Zone Material Notes:

(a) Average value is determined by CVGRAPH (see Appendix A).

(b) Irradiated value is a calculated average of Capsule W test data at greater than or equal to 95% shear.

(c) The upper shelf lateral expansion for Capsule W was fixed, based on the average of three (3) data points at 100% shear.

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Westinghouse Non-Proprietary Class 3 5-14 Table 5-10 Comparison of the Callaway Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper-Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Capsule 30 ft-lb Transition Upper-Shelf Energy Fluence Temperature Shift Decrease Material Capsule (x1019 n/cm2, E Predicted(a) Measured(b) Predicted(a) Measured(b)

> 1.0 MeV) (°F) (°F) (%) (%)

U 0.313 30.0 0(c) 14.5 2 Lower Shell Plate Y 1.18 46.0 25.1 20.0 6 R2708-1 (Longitudinal) V 2.32 54.0 16.3 23.0 0 X 3.08 57.1 25.9 25.0 5 W 5.98 63.2 58.6 29.0 7 U 0.313 30.0 26.1 14.5 11 Y 1.18 46.0 46.7 20.0 13 Lower Shell Plate V 2.32 54.0 45.2 23.0 3 R2708-1 (Transverse)

X 3.08 57.1 30.8 25.0 5 W 5.98 63.2 95.2 29.0 20 U 0.313 21.7 66.2 14.5 6 Y 1.18 33.3 35.0 20.0 9 Surveillance Weld V 2.32 39.0 46.2 23.0 3 Material (Heat # 90077)

X 3.08 41.2 49.7 25.0 3 W 5.98 45.7 65.8 29.0 4 U 0.313 --- 66.3 --- 0(d)

Y 1.18 --- 56.5 --- 14 Heat Affected Zone V 2.32 --- 56.8 --- 0 Material X 3.08 --- 42.3 --- 0(d)

W 5.98 --- 91.6 --- 5 Notes:

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

(b) Calculated by CVGRAPH, Version 6.0 using measured Charpy data (See Appendix C).

(c) This RTNDT was calculated to be negative (-5.9°F). Physically, this should not occur; therefore, a conservative value of zero degrees F is shown in this table.

(d) USE values were calculated to have increased. Physically, this should not occur; therefore, conservative values of zero percent are shown in this table.

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Westinghouse Non-Proprietary Class 3 5-15 Table 5-11 Tensile Properties of the Callaway Unit 1 Capsule W Reactor Vessel Surveillance Materials Irradiated to 5.98 x 1019 n/cm2 (E > 1.0 MeV) 0.2% Fracture Test Ultimate Fracture Fracture Uniform Total Reduction Sample Yield True Material Temp. Strength Load Strength Elongation Elongation in Area Number Strength Stress

(°F) (ksi) (kip) (ksi) (%) (%) (%)

(ksi) (ksi)

CL7 72 81.2 101.6 3.4 68.8 182.5 9.1 22.3 62.1 Lower Shell Plate R2708-1 CL8 175 77.8 96.3 3.2 64.3 186.0 9.5 22.6 65.0 (Longitudinal)

CL9 550 73.0 94.8 3.5 70.4 171.9 8.2 18.3 58.5 CT7 72 80.8 101.8 3.5 71.7 165.7 8.8 21.1 57.0 Lower Shell Plate R2708-1 CT8 175 77.1 96.4 2.8 57.9 134.2 9.1 20.7 57.5 (Transverse)

CT9 550 72.3 94.2 3.5 70.7 145.5 7.1 15.6 51.0 CW7 72 79.1 94.8 3.0 60.9 179.2 9.1 25.9 65.9 Surveillance Weld Material CW8 175 76.6 90.8 2.8 57.8 181.9 10.0 25.5 68.6 (Heat # 90077)

CW9 550 74.1 91.8 3.2 64.6 178.7 6.8 19.2 63.5 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-16 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-17 Figure 5-1(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-18 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-19 Figure 5-2(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-20 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-21 Figure 5-3(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-22 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-23 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-24 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-25 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-26 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-27 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-28 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-29 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-30 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-31 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-32 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-33 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-34 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-35 Figure 5-10(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-36 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-37 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-38 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-39 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-40 CL32, 0°F CL38, 20°F CL31, 35°F CL45, 40°F CL39, 50°F CL40, 50°F CL34, 72°F CL37, 90°F CL44, 100°F CL36, 120°F CL41, 150°F CL33, 170°F CL42, 200°F CL43, 230°F CL35, 250°F Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-41 CT31, 0°F CT42, 20°F CT36, 40°F CT40, 50°F CT43, 60°F CT44, 60°F CT38, 72°F CT32, 100°F CT34, 120°F CT35, 150°F CT39, 170°F CT37, 200°F CT45, 230°F CT41, 250°F CT33, 275°F Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-42 CW45, -25°F CW36, 0°F CW40, 10°F CW32, 20°F CW34, 25°F CW42, 35°F CW33, 40°F CW44, 50°F CW43, 60°F CW38, 72°F CW37, 120°F CW31, 150°F CW35, 170°F CW39, 200°F CW41, 250°F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-43 CH41, -50°F CH44, -40°F CH42, -25°F CH37, 0°F CH43, 10°F CH34, 20°F CH40, 35°F CH33, 50°F CH39, 60°F CH31, 70°F CH35, 72°F CH36, 100°F CH38, 150°F CH32, 170°F CH45, 200°F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-44 Figure 5-17 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 5-45 Figure 5-18 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 5-46 Figure 5-19 Tensile Properties for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 5-47 CL7 - Tested at 72°F CL8 - Tested at 175°F CL9 - Tested at 550°F Figure 5-20 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-48 CT7 - Tested at 72°F CT8 - Tested at 175°F CT9 - Tested at 550°F Figure 5-21 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-49 CW7 - Tested at 72°F CW8 - Tested at 175°F CW9 - Tested at 550°F Figure 5-22 Fractured Tensile Specimens from the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-50 Tensile Specimen CL7, Tested at 72°F Tensile Specimen CL8, Tested at 175°F Tensile Specimen CL9, Tested at 550°F Figure 5-23 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-51 Tensile Specimen CT7, Tested at 72°F Tensile Specimen CT8, Tested at 175°F Tensile Specimen CT9, Tested at 550°F Figure 5-24 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-52 Tensile Specimen CW7, Tested at 72°F Tensile Specimen CW8, Tested at 175°F Tensile Specimen CW9, Tested at 550°F Figure 5-25 Engineering Stress-Strain Curves for Callaway Unit 1 Program Weld Metal (Heat #

90077) Tensile Specimens WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates (Sn) transport analysis performed for the Callaway Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant- and fuel-cycle-specific basis. An evaluation of the most recent dosimetry sensor set from Capsule W, withdrawn at the end of the 20th plant operating cycle, is provided. In addition, the sensor sets from the previously withdrawn and analyzed capsules (U, Y, V, and X) were re-analyzed and are presented. 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 form the basis for projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY) at 3565 MWt.

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.

However, in recent years, 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 E853-13, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, [Ref. 18] 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 E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom [Ref. 19]. 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 [Ref. 1].

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on nuclear cross-section data derived from ENDF/B-VI and used the latest available calculational tools.

Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. 20]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040, Revision 4, [Ref. 21]. As an improvement, instead of the fluence rate synthesis technique, three-dimensional transport calculations were performed.

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Westinghouse Non-Proprietary Class 3 6-2 6.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Callaway Unit 1 reactor vessel is shown in Figure 4-

1. Six irradiation capsules attached to the neutron pad are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.5°, 61.0°,

121.5°, 238.5°, 241.0°, and 301.5°, as shown in Figure 4-1. These full-core positions correspond to the octant symmetric locations shown in Figures 6-2 and 6-3: 29.0° from the core cardinal axes (for the 61.0° and 241.0° capsules) and 31.5° from the core cardinal axes (for the 58.5°, 121.5°, 238.5°, and 301.5° capsules). The stainless steel specimen containers are 1.6-inch by 1.25-inch and are approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the approximate 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 significant effect on both the spatial distribution of neutron fluence rate and the neutron spectrum in the vicinity of the capsules. However, the capsules are far enough apart that they do not interfere with one another. 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 Callaway Unit 1 reactor vessel and surveillance capsules, a series of fuel-cycle-specific forward transport calculations were carried out using a three-dimensional geometrical reactor model. For the Callaway Unit 1 transport calculations, the r,,z models depicted (given as r, plan view) in Figures 6-1 (12.5° neutron pad configuration), 6-2 (20.0° neutron pad configuration), and 6-3 (22.5° neutron pad configuration) were utilized since, with the exception of the neutron pads, capsules, and associated support structures, the reactor is octant symmetric.

The r,z section view depicted in Figure 6-4 shows the model having an axial span from an elevation approximately 6 feet below the bottom of the active fuel to 5 feet above the top of the active fuel. These r,,z models include the core, the reactor internals, the neutron pads including explicit representations of octants not containing surveillance capsules and octants with surveillance capsules at 29° and/or 31.5°,

the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing these analytical models, nominal design dimensions were employed for the various structural components with one exception; the minimum pressure vessel thickness was used. Water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,,z reactor models for the 20.0° and 22.5° neutron pad configurations consisted of 183 radial, 110 azimuthal, and 206 axial intervals. The geometric mesh description of the r,,z reactor model for the 12.5° neutron pad configuration consisted of 183 radial, 110 azimuthal, and 204 axial intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,,z calculations was set at a value of 0.001.

The core power distributions used in the plant-specific transport analysis for each of the first 20 fuel cycles at Callaway Unit 1 included cycle-dependent fuel assembly initial enrichments, burnups, and axial WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-3 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 fluence rate, 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 RAPTOR-M3G discrete ordinates code [Ref. 22] and the BUGLE-96 cross-section library [Ref. 23]. The BUGLE-96 library provides a coupled 47-neutron, 20-gamma-group cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion, and angular discretization was modeled with an S16 order of angular quadrature.

Energy- and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-12. In Tables 6-1 and 6-2, the calculated exposure rates and integral exposures expressed in terms of fast neutron fluence rate (E > 1.0 MeV) and fast neutron fluence (E > 1.0 MeV), respectively, are given at the radial and azimuthal center of the surveillance capsule positions, i.e., for the 29.0° and 31.5° dual capsule holder locations and 31.5° single capsule holder location. In Tables 6-3 and 6-4, the calculated exposure rates and integral exposures expressed in terms of iron atom displacement rate (dpa/s) and iron atom displacements (dpa), respectively, are given at the radial and azimuthal center of the surveillance capsule positions, i.e., for the 29.0° and 31.5° dual capsule holder locations and 31.5° single capsule holder location. In Tables 6-2 and 6-4, the calculated integral exposures expressed for future projections, in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements, respectively, are given at the radial and azimuthal center of the surveillance capsule positions. These results, representative of the average axial exposure of the material specimens, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projections into the future.

Similar information, in terms of calculated fast neutron fluence rate (E > 1.0 MeV), fast neutron fluence (E > 1.0 MeV), dpa/s, and dpa, are provided in Tables 6-5 through 6-8, for the reactor vessel inner radius at four azimuthal locations, as well as the maximum exposure observed within the octant. The vessel data given in Tables 6-5 through 6-8 were taken at the clad/base metal interface and represent maximum calculated exposure levels on the vessel. From the data provided in Table 6-6, it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the 20th fuel cycle (i.e., after 25.75 EFPY of plant operation) was 1.49E+19 n/cm2.

These data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end of the 20th fuel cycle, as well as future projections to 32, 35, 40, 48, 54, and 60 EFPY. The calculations account for the uprate from 3411 MWt to 3565 MWt that occurred prior to Cycle 3. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 21 is representative of future plant operation. The future projections are based on the current reactor power level of 3565 MWt.

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Westinghouse Non-Proprietary Class 3 6-4 The calculated fast neutron exposures for the six surveillance capsules withdrawn from the Callaway Unit 1 reactor are provided in Table 6-9. These neutron exposure levels are based on the plant- and fuel-cycle-specific neutron transport calculations performed for the Callaway Unit 1 reactor. From the data provided in Table 6-9, Capsule W received a fast neutron fluence (E > 1.0 MeV) of 5.98E+19 n/cm2 after exposure through the end of the 20th fuel cycle (i.e., after 25.75 EFPY).

Updated lead factors for the Callaway Unit 1 surveillance capsules are provided in Table 6-10. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric radial and azimuthal center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-10, the lead factors for capsules that have been withdrawn from the reactor (Capsules U, Y, V, X, Z, and W) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules.

Table 6-11 presents the maximum fast neutron fluences (E > 1.0 MeV) and Table 6-12 presents the maximum dpa for pressure vessel materials.

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 serve to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least-squares evaluation comparisons is documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule W, which was withdrawn from Callaway Unit 1 at the end of the 20th fuel cycle, is summarized below.

Reaction Rate (rps/atom)

Reaction M/C Measured (M) Calculated (C)

Cu-63(n,)Co-60 4.52E-17 3.94E-17 1.15 Fe-54(n,p)Mn-54 4.65E-15 4.32E-15 1.08 Ni-58(n,p)Co-58 6.75E-15 6.06E-15 1.11 U-238(Cd)(n,f)Cs-137 2.82E-14 2.32E-14 1.21 Np-237(Cd)(n,f)Cs-137 2.41E-13 2.30E-13 1.05 Average 1.12

% standard deviation 5.6 The measured-to-calculated (M/C) reaction rate ratios for the Capsule W threshold reactions range from 1.05 to 1.21, and the average M/C ratio is 1.12 5.6% (1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Callaway Unit 1.

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Westinghouse Non-Proprietary Class 3 6-5 6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Callaway Unit 1 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

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

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

4. Comparisons of the plant-specific calculations with all available dosimetry results from the Callaway Unit 1 surveillance program.

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

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

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

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

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Westinghouse Non-Proprietary Class 3 6-6 Description Capsule and Vessel IR PCA Comparisons 3%

H. B. Robinson Comparisons 3%

Analytical Sensitivity Studies 11%

Additional Uncertainty for Factors not Explicitly 5%

Net Calculational Uncertainty 13%

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

Therefore, the resultant uncertainty was treated as random, and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons described in Appendix A support these uncertainty assessments for Callaway Unit 1.

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Westinghouse Non-Proprietary Class 3 6-7 Table 6-1 Calculated Fast Neutron Fluence Rate (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 Fluence Rate (n/cm2-s)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 8.79E+10 9.42E+10 9.34E+10 2 1.15 2.21 7.76E+10 8.09E+10 8.01E+10 3 1.23 3.43 8.26E+10 9.51E+10 9.45E+10 4 1.26 4.70 7.16E+10 7.92E+10 7.86E+10 5 1.27 5.97 7.10E+10 7.71E+10 7.65E+10 6 1.32 7.29 7.13E+10 7.57E+10 7.51E+10 7 1.30 8.59 6.66E+10 7.11E+10 7.05E+10 8 1.35 9.94 6.58E+10 7.12E+10 7.06E+10 9 1.19 11.13 6.37E+10 6.80E+10 6.74E+10 10 1.36 12.49 6.79E+10 7.21E+10 7.15E+10 11 1.38 13.87 6.42E+10 6.99E+10 6.93E+10 12 1.35 15.22 6.57E+10 7.15E+10 7.09E+10 13 1.26 16.48 7.80E+10 8.73E+10 8.66E+10 14 1.22 17.70 6.76E+10 7.35E+10 7.29E+10 15 1.29 18.99 6.81E+10 7.49E+10 7.43E+10 16 1.42 20.41 6.44E+10 6.80E+10 6.73E+10 17 1.35 21.76 6.65E+10 7.18E+10 7.12E+10 18 1.33 23.09 6.10E+10 6.61E+10 6.56E+10 19 1.36 24.45 6.09E+10 6.44E+10 6.38E+10 20 1.30 25.75 5.72E+10 6.03E+10 5.98E+10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-8 Table 6-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections Fluence (n/cm2)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 2.92E+18 3.13E+18 3.10E+18 2 1.15 2.21 5.75E+18 6.07E+18 6.02E+18 3 1.23 3.43 8.95E+18 9.76E+18 9.68E+18 4 1.26 4.70 1.18E+19 1.29E+19 1.28E+19 5 1.27 5.97 1.46E+19 1.60E+19 1.59E+19 6 1.32 7.29 1.76E+19 1.92E+19 1.90E+19 7 1.30 8.59 2.03E+19 2.21E+19 2.19E+19 8 1.35 9.94 2.32E+19 2.51E+19 2.49E+19 9 1.19 11.13 2.56E+19 2.77E+19 2.75E+19 10 1.36 12.49 2.85E+19 3.08E+19 3.05E+19 11 1.38 13.87 3.13E+19 3.38E+19 3.35E+19 12 1.35 15.22 3.41E+19 3.69E+19 3.66E+19 13 1.26 16.48 3.72E+19 4.03E+19 4.00E+19 14 1.22 17.70 3.98E+19 4.32E+19 4.28E+19 15 1.29 18.99 4.25E+19 4.62E+19 4.58E+19 16 1.42 20.41 4.54E+19 4.92E+19 4.88E+19 17 1.35 21.76 4.82E+19 5.23E+19 5.19E+19 18 1.33 23.09 5.08E+19 5.51E+19 5.46E+19 19 1.36 24.45 5.34E+19 5.79E+19 5.74E+19 20 1.30 25.75 5.58E+19 6.03E+19 5.98E+19 Future 32.00 6.78E+19 7.33E+19 7.27E+19 Future 35.00 7.36E+19 7.95E+19 7.89E+19 Future 40.00 8.32E+19 8.99E+19 8.92E+19 Future 48.00 9.87E+19 1.07E+20 1.06E+20 Future 54.00 1.10E+20 1.19E+20 1.18E+20 Future 60.00 1.22E+20 1.31E+20 1.30E+20 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-9 Table 6-3 Calculated Iron Atom Displacement Rate at the Surveillance Capsule Center and at Core Midplane for Cycles 1 Through 20 Iron Atom Displacement Rate (dpa/s)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 1.75E-10 1.88E-10 1.86E-10 2 1.15 2.21 1.54E-10 1.60E-10 1.59E-10 3 1.23 3.43 1.65E-10 1.90E-10 1.89E-10 4 1.26 4.70 1.42E-10 1.57E-10 1.56E-10 5 1.27 5.97 1.41E-10 1.53E-10 1.51E-10 6 1.32 7.29 1.41E-10 1.50E-10 1.48E-10 7 1.30 8.59 1.32E-10 1.41E-10 1.39E-10 8 1.35 9.94 1.30E-10 1.41E-10 1.40E-10 9 1.19 11.13 1.26E-10 1.34E-10 1.33E-10 10 1.36 12.49 1.34E-10 1.43E-10 1.41E-10 11 1.38 13.87 1.27E-10 1.38E-10 1.37E-10 12 1.35 15.22 1.30E-10 1.42E-10 1.40E-10 13 1.26 16.48 1.54E-10 1.73E-10 1.71E-10 14 1.22 17.70 1.34E-10 1.45E-10 1.44E-10 15 1.29 18.99 1.35E-10 1.48E-10 1.47E-10 16 1.42 20.41 1.27E-10 1.34E-10 1.33E-10 17 1.35 21.76 1.31E-10 1.42E-10 1.40E-10 18 1.33 23.09 1.21E-10 1.31E-10 1.29E-10 19 1.36 24.45 1.20E-10 1.27E-10 1.26E-10 20 1.30 25.75 1.13E-10 1.19E-10 1.18E-10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-10 Table 6-4 Calculated Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections Iron Atom Displacements (dpa)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 5.82E-03 6.24E-03 6.18E-03 2 1.15 2.21 1.14E-02 1.21E-02 1.19E-02 3 1.23 3.43 1.78E-02 1.94E-02 1.93E-02 4 1.26 4.70 2.35E-02 2.57E-02 2.55E-02 5 1.27 5.97 2.91E-02 3.18E-02 3.15E-02 6 1.32 7.29 3.50E-02 3.80E-02 3.77E-02 7 1.30 8.59 4.04E-02 4.38E-02 4.34E-02 8 1.35 9.94 4.59E-02 4.98E-02 4.94E-02 9 1.19 11.13 5.07E-02 5.49E-02 5.44E-02 10 1.36 12.49 5.64E-02 6.10E-02 6.04E-02 11 1.38 13.87 6.20E-02 6.70E-02 6.64E-02 12 1.35 15.22 6.75E-02 7.30E-02 7.24E-02 13 1.26 16.48 7.37E-02 7.99E-02 7.92E-02 14 1.22 17.70 7.88E-02 8.55E-02 8.47E-02 15 1.29 18.99 8.43E-02 9.15E-02 9.07E-02 16 1.42 20.41 9.00E-02 9.75E-02 9.67E-02 17 1.35 21.76 9.56E-02 1.04E-01 1.03E-01 18 1.33 23.09 1.01E-01 1.09E-01 1.08E-01 19 1.36 24.45 1.06E-01 1.15E-01 1.14E-01 20 1.30 25.75 1.10E-01 1.19E-01 1.18E-01 Future 32.00 1.34E-01 1.45E-01 1.44E-01 Future 35.00 1.46E-01 1.57E-01 1.56E-01 Future 40.00 1.65E-01 1.78E-01 1.76E-01 Future 48.00 1.95E-01 2.11E-01 2.09E-01 Future 54.00 2.18E-01 2.35E-01 2.33E-01 Future 60.00 2.41E-01 2.60E-01 2.57E-01 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-11 Table 6-5 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence Rates (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface Cycle Total Fluence Rate (n/cm2-s)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 1.25E+10 1.93E+10 2.17E+10 2.32E+10 2.32E+10 2 1.15 2.21 1.22E+10 1.88E+10 1.91E+10 1.86E+10 2.17E+10 3 1.23 3.43 1.23E+10 1.87E+10 2.07E+10 2.81E+10 2.81E+10 4 1.26 4.70 1.06E+10 1.51E+10 1.78E+10 1.99E+10 1.99E+10 5 1.27 5.97 9.89E+09 1.49E+10 1.76E+10 1.93E+10 1.93E+10 6 1.32 7.29 9.05E+09 1.53E+10 1.76E+10 1.80E+10 1.86E+10 7 1.30 8.59 9.26E+09 1.47E+10 1.70E+10 1.77E+10 1.78E+10 8 1.35 9.94 9.44E+09 1.44E+10 1.67E+10 1.82E+10 1.82E+10 9 1.19 11.13 9.93E+09 1.44E+10 1.58E+10 1.56E+10 1.67E+10 10 1.36 12.49 9.54E+09 1.48E+10 1.68E+10 1.74E+10 1.78E+10 11 1.38 13.87 9.22E+09 1.40E+10 1.67E+10 1.86E+10 1.86E+10 12 1.35 15.22 9.24E+09 1.45E+10 1.72E+10 1.90E+10 1.90E+10 13 1.26 16.48 9.13E+09 1.48E+10 2.00E+10 2.24E+10 2.24E+10 14 1.22 17.70 8.47E+09 1.43E+10 1.72E+10 1.91E+10 1.91E+10 15 1.29 18.99 9.39E+09 1.46E+10 1.73E+10 1.97E+10 1.97E+10 16 1.42 20.41 8.61E+09 1.41E+10 1.61E+10 1.62E+10 1.71E+10 17 1.35 21.76 9.52E+09 1.43E+10 1.65E+10 1.74E+10 1.74E+10 18 1.33 23.09 8.79E+09 1.37E+10 1.55E+10 1.66E+10 1.66E+10 19 1.36 24.45 8.70E+09 1.37E+10 1.54E+10 1.49E+10 1.63E+10 20 1.30 25.75 8.86E+09 1.29E+10 1.44E+10 1.35E+10 1.52E+10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-12 Table 6-6 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface Cycle Total Fluence (n/cm2)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 4.15E+17 6.40E+17 7.20E+17 7.71E+17 7.71E+17 2 1.15 2.21 8.47E+17 1.30E+18 1.40E+18 1.42E+18 1.53E+18 3 1.23 3.43 1.32E+18 2.02E+18 2.19E+18 2.51E+18 2.51E+18 4 1.26 4.70 1.74E+18 2.62E+18 2.90E+18 3.30E+18 3.30E+18 5 1.27 5.97 2.14E+18 3.22E+18 3.61E+18 4.07E+18 4.07E+18 6 1.32 7.29 2.51E+18 3.86E+18 4.33E+18 4.82E+18 4.82E+18 7 1.30 8.59 2.89E+18 4.45E+18 5.02E+18 5.53E+18 5.53E+18 8 1.35 9.94 3.29E+18 5.06E+18 5.72E+18 6.30E+18 6.30E+18 9 1.19 11.13 3.66E+18 5.60E+18 6.32E+18 6.89E+18 6.89E+18 10 1.36 12.49 4.07E+18 6.23E+18 7.03E+18 7.63E+18 7.63E+18 11 1.38 13.87 4.46E+18 6.83E+18 7.74E+18 8.41E+18 8.41E+18 12 1.35 15.22 4.84E+18 7.42E+18 8.44E+18 9.19E+18 9.19E+18 13 1.26 16.48 5.19E+18 8.00E+18 9.22E+18 1.01E+19 1.01E+19 14 1.22 17.70 5.51E+18 8.54E+18 9.87E+18 1.08E+19 1.08E+19 15 1.29 18.99 5.89E+18 9.12E+18 1.06E+19 1.16E+19 1.16E+19 16 1.42 20.41 6.27E+18 9.75E+18 1.13E+19 1.23E+19 1.23E+19 17 1.35 21.76 6.68E+18 1.04E+19 1.20E+19 1.30E+19 1.30E+19 18 1.33 23.09 7.04E+18 1.09E+19 1.26E+19 1.37E+19 1.37E+19 19 1.36 24.45 7.41E+18 1.15E+19 1.33E+19 1.44E+19 1.44E+19 20 1.30 25.75 7.77E+18 1.20E+19 1.39E+19 1.49E+19 1.49E+19 Future 32.00 9.45E+18 1.47E+19 1.68E+19 1.80E+19 1.80E+19 Future 35.00 1.03E+19 1.59E+19 1.83E+19 1.95E+19 1.95E+19 Future 40.00 1.16E+19 1.80E+19 2.07E+19 2.20E+19 2.20E+19 Future 48.00 1.37E+19 2.14E+19 2.45E+19 2.60E+19 2.60E+19 Future 54.00 1.54E+19 2.39E+19 2.73E+19 2.91E+19 2.91E+19 Future 60.00 1.70E+19 2.64E+19 3.02E+19 3.21E+19 3.21E+19 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-13 Table 6-7 Calculated Azimuthal Variation of Maximum Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface Cycle Total Iron Atom Displacement Rate (dpa/s)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 1.93E-11 2.95E-11 3.33E-11 3.66E-11 3.66E-11 2 1.15 2.21 1.88E-11 2.87E-11 2.94E-11 2.93E-11 3.30E-11 3 1.23 3.43 1.90E-11 2.86E-11 3.19E-11 4.43E-11 4.43E-11 4 1.26 4.70 1.64E-11 2.32E-11 2.74E-11 3.13E-11 3.13E-11 5 1.27 5.97 1.53E-11 2.28E-11 2.70E-11 3.05E-11 3.05E-11 6 1.32 7.29 1.40E-11 2.34E-11 2.70E-11 2.84E-11 2.84E-11 7 1.30 8.59 1.43E-11 2.26E-11 2.61E-11 2.79E-11 2.79E-11 8 1.35 9.94 1.46E-11 2.21E-11 2.56E-11 2.87E-11 2.87E-11 9 1.19 11.13 1.54E-11 2.21E-11 2.44E-11 2.47E-11 2.55E-11 10 1.36 12.49 1.48E-11 2.27E-11 2.58E-11 2.74E-11 2.74E-11 11 1.38 13.87 1.43E-11 2.15E-11 2.57E-11 2.92E-11 2.92E-11 12 1.35 15.22 1.43E-11 2.23E-11 2.65E-11 3.00E-11 3.00E-11 13 1.26 16.48 1.41E-11 2.27E-11 3.08E-11 3.53E-11 3.53E-11 14 1.22 17.70 1.31E-11 2.19E-11 2.64E-11 3.01E-11 3.01E-11 15 1.29 18.99 1.45E-11 2.23E-11 2.67E-11 3.11E-11 3.11E-11 16 1.42 20.41 1.33E-11 2.16E-11 2.48E-11 2.56E-11 2.62E-11 17 1.35 21.76 1.47E-11 2.20E-11 2.55E-11 2.75E-11 2.75E-11 18 1.33 23.09 1.36E-11 2.10E-11 2.39E-11 2.62E-11 2.62E-11 19 1.36 24.45 1.35E-11 2.10E-11 2.36E-11 2.35E-11 2.49E-11 20 1.30 25.75 1.37E-11 1.97E-11 2.21E-11 2.13E-11 2.32E-11 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-14 Table 6-8 Calculated Azimuthal Variation of Maximum Iron Atom Displacements at the Reactor Vessel Clad/Base Metal Interface Cycle Total Iron Atom Displacements (dpa)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 6.41E-04 9.79E-04 1.11E-03 1.22E-03 1.22E-03 2 1.15 2.21 1.31E-03 1.99E-03 2.15E-03 2.25E-03 2.34E-03 3 1.23 3.43 2.04E-03 3.10E-03 3.38E-03 3.96E-03 3.96E-03 4 1.26 4.70 2.69E-03 4.02E-03 4.47E-03 5.20E-03 5.20E-03 5 1.27 5.97 3.31E-03 4.93E-03 5.55E-03 6.43E-03 6.43E-03 6 1.32 7.29 3.89E-03 5.91E-03 6.67E-03 7.60E-03 7.60E-03 7 1.30 8.59 4.47E-03 6.81E-03 7.72E-03 8.72E-03 8.72E-03 8 1.35 9.94 5.09E-03 7.75E-03 8.81E-03 9.94E-03 9.94E-03 9 1.19 11.13 5.67E-03 8.58E-03 9.73E-03 1.09E-02 1.09E-02 10 1.36 12.49 6.30E-03 9.55E-03 1.08E-02 1.20E-02 1.20E-02 11 1.38 13.87 6.90E-03 1.05E-02 1.19E-02 1.33E-02 1.33E-02 12 1.35 15.22 7.49E-03 1.14E-02 1.30E-02 1.45E-02 1.45E-02 13 1.26 16.48 8.04E-03 1.23E-02 1.42E-02 1.59E-02 1.59E-02 14 1.22 17.70 8.54E-03 1.31E-02 1.52E-02 1.70E-02 1.70E-02 15 1.29 18.99 9.13E-03 1.40E-02 1.63E-02 1.82E-02 1.82E-02 16 1.42 20.41 9.72E-03 1.49E-02 1.74E-02 1.94E-02 1.94E-02 17 1.35 21.76 1.03E-02 1.59E-02 1.84E-02 2.05E-02 2.05E-02 18 1.33 23.09 1.09E-02 1.67E-02 1.94E-02 2.16E-02 2.16E-02 19 1.36 24.45 1.15E-02 1.76E-02 2.04E-02 2.26E-02 2.26E-02 20 1.30 25.75 1.20E-02 1.84E-02 2.13E-02 2.35E-02 2.35E-02 Future 32.00 1.46E-02 2.25E-02 2.59E-02 2.84E-02 2.84E-02 Future 35.00 1.59E-02 2.44E-02 2.81E-02 3.08E-02 3.08E-02 Future 40.00 1.80E-02 2.76E-02 3.18E-02 3.47E-02 3.47E-02 Future 48.00 2.13E-02 3.28E-02 3.77E-02 4.11E-02 4.11E-02 Future 54.00 2.38E-02 3.66E-02 4.21E-02 4.58E-02 4.58E-02 Future 60.00 2.63E-02 4.05E-02 4.65E-02 5.05E-02 5.05E-02 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-15 Table 6-9 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Callaway Unit 1 Irradiation Fluence Iron Atom Irradiation Capsule Time (E > 1.0 MeV) Displacements Cycle(s)

(EFPY) (n/cm2) (dpa)

U 1 1.05 3.13E+18 6.24E-03 Y 1-4 4.70 1.18E+19 2.35E-02 V 1-8 9.94 2.32E+19 4.59E-02 X 1-10 12.49 3.08E+19 6.10E-02 Z(a) 1-13 16.48 4.00E+19 7.92E-02 W 1-20 25.75 5.98E+19 1.18E-01 Note:

(a) Capsule Z was placed in storage.

Table 6-10 Calculated Surveillance Capsule Lead Factors Capsule Location Status Lead Factor 58.5º (Capsule U) Withdrawn EOC 1 4.06 241° (Capsule Y) Withdrawn EOC 4 3.58 61° (Capsule V) Withdrawn EOC 8 3.67 238.5° (Capsule X) Withdrawn EOC 10 4.03 301.5° (Capsule Z)(a) Withdrawn EOC 13 3.98 121.5° (Capsule W) Withdrawn EOC 20 4.01 Note:

(a) Capsule Z was placed in storage.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-16 Table 6-11 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Outlet Inlet Total Nozzle to Nozzle to Upper Shell Cycle Cycle Upper Upper Upper Upper Shell Longitudinal Cycle Length Time Shell Shell Shell Longitudinal Welds at (EFPY)

(EFPY) Welds - Welds - Plates Weld at 90° 210° and Lowest Lowest 330° Extent Extent 1 1.05 1.05 1.11E+15 1.26E+15 8.60E+15 4.84E+15 7.13E+15 2 1.15 2.21 2.38E+15 2.71E+15 1.94E+16 1.16E+16 1.65E+16 3 1.23 3.43 3.77E+15 4.29E+15 3.10E+16 1.75E+16 2.52E+16 4 1.26 4.70 4.96E+15 5.64E+15 4.08E+16 2.29E+16 3.32E+16 5 1.27 5.97 6.06E+15 6.90E+15 4.91E+16 2.74E+16 3.99E+16 6 1.32 7.29 7.25E+15 8.25E+15 5.87E+16 3.24E+16 4.80E+16 7 1.30 8.59 8.30E+15 9.43E+15 6.50E+16 3.60E+16 5.34E+16 8 1.35 9.94 9.45E+15 1.07E+16 7.50E+16 4.14E+16 6.15E+16 9 1.19 11.13 1.05E+16 1.19E+16 8.24E+16 4.61E+16 6.80E+16 10 1.36 12.49 1.17E+16 1.33E+16 9.36E+16 5.22E+16 7.73E+16 11 1.38 13.87 1.29E+16 1.46E+16 1.03E+17 5.72E+16 8.48E+16 12 1.35 15.22 1.41E+16 1.60E+16 1.13E+17 6.27E+16 9.34E+16 13 1.26 16.48 1.54E+16 1.75E+16 1.24E+17 6.77E+16 1.02E+17 14 1.22 17.70 1.65E+16 1.87E+16 1.34E+17 7.25E+16 1.10E+17 15 1.29 18.99 1.77E+16 2.01E+16 1.44E+17 7.81E+16 1.19E+17 16 1.42 20.41 1.89E+16 2.15E+16 1.55E+17 8.40E+16 1.28E+17 17 1.35 21.76 2.02E+16 2.30E+16 1.66E+17 9.01E+16 1.37E+17 18 1.33 23.09 2.14E+16 2.43E+16 1.76E+17 9.59E+16 1.45E+17 19 1.36 24.45 2.25E+16 2.57E+16 1.86E+17 1.02E+17 1.54E+17 20 1.30 25.75 2.36E+16 2.69E+16 1.95E+17 1.07E+17 1.62E+17 Future 32.00 2.91E+16 3.31E+16 2.43E+17 1.34E+17 2.02E+17 Future 35.00 3.17E+16 3.61E+16 2.66E+17 1.47E+17 2.22E+17 Future 40.00 3.61E+16 4.11E+16 3.05E+17 1.69E+17 2.54E+17 Future 48.00 4.31E+16 4.91E+16 3.66E+17 2.03E+17 3.06E+17 Future 54.00 4.84E+16 5.51E+16 4.12E+17 2.29E+17 3.45E+17 Future 60.00 5.36E+16 6.12E+16 4.59E+17 2.55E+17 3.83E+17 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-17 Table 6-11 (cont.) Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Total Upper Int. Shell Int. Shell Cycle Int. Shell Cycle Shell to Long. to Lower Cycle Length Int. Shell Long.

Time Int. Shell Welds at Shell (EFPY) Plates Weld at (EFPY) Circ. 210° and Circ.

90° Weld 330° Weld 1 1.05 1.05 8.60E+15 7.70E+17 4.15E+17 7.20E+17 7.70E+17 2 1.15 2.21 1.94E+16 1.53E+18 8.47E+17 1.40E+18 1.53E+18 3 1.23 3.43 3.10E+16 2.51E+18 1.32E+18 2.19E+18 2.51E+18 4 1.26 4.70 4.08E+16 3.30E+18 1.74E+18 2.90E+18 3.30E+18 5 1.27 5.97 4.91E+16 4.08E+18 2.14E+18 3.61E+18 4.08E+18 6 1.32 7.29 5.87E+16 4.82E+18 2.51E+18 4.33E+18 4.82E+18 7 1.30 8.59 6.50E+16 5.53E+18 2.88E+18 5.02E+18 5.53E+18 8 1.35 9.94 7.50E+16 6.30E+18 3.29E+18 5.72E+18 6.30E+18 9 1.19 11.13 8.24E+16 6.89E+18 3.66E+18 6.32E+18 6.89E+18 10 1.36 12.49 9.36E+16 7.63E+18 4.07E+18 7.03E+18 7.63E+18 11 1.38 13.87 1.03E+17 8.41E+18 4.45E+18 7.73E+18 8.41E+18 12 1.35 15.22 1.13E+17 9.18E+18 4.83E+18 8.44E+18 9.18E+18 13 1.26 16.48 1.24E+17 1.00E+19 5.19E+18 9.21E+18 1.00E+19 14 1.22 17.70 1.34E+17 1.08E+19 5.51E+18 9.86E+18 1.08E+19 15 1.29 18.99 1.44E+17 1.16E+19 5.89E+18 1.06E+19 1.16E+19 16 1.42 20.41 1.55E+17 1.23E+19 6.27E+18 1.13E+19 1.23E+19 17 1.35 21.76 1.66E+17 1.30E+19 6.67E+18 1.20E+19 1.30E+19 18 1.33 23.09 1.76E+17 1.37E+19 7.03E+18 1.26E+19 1.37E+19 19 1.36 24.45 1.86E+17 1.43E+19 7.40E+18 1.33E+19 1.43E+19 20 1.30 25.75 1.95E+17 1.49E+19 7.76E+18 1.38E+19 1.49E+19 Future 32.00 2.43E+17 1.80E+19 9.44E+18 1.68E+19 1.80E+19 Future 35.00 2.66E+17 1.95E+19 1.02E+19 1.83E+19 1.95E+19 Future 40.00 3.05E+17 2.20E+19 1.16E+19 2.06E+19 2.20E+19 Future 48.00 3.66E+17 2.60E+19 1.37E+19 2.44E+19 2.60E+19 Future 54.00 4.12E+17 2.90E+19 1.53E+19 2.73E+19 2.90E+19 Future 60.00 4.59E+17 3.20E+19 1.70E+19 3.02E+19 3.20E+19 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-18 Table 6-11 (cont.) Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Lower Total Lower Cycle Lower Shell to Cycle Shell Cycle Length Lower Shell Lower Time Long.

(EFPY) Shell Long. Vessel (EFPY) Weld at Plates Weld at Head 210° and 90° Circ.

330° Weld 1 1.05 1.05 7.71E+17 4.15E+17 7.20E+17 9.07E+14 2 1.15 2.21 1.53E+18 8.44E+17 1.39E+18 1.86E+15 3 1.23 3.43 2.51E+18 1.32E+18 2.19E+18 3.08E+15 4 1.26 4.70 3.30E+18 1.74E+18 2.90E+18 4.06E+15 5 1.27 5.97 4.08E+18 2.14E+18 3.61E+18 4.99E+15 6 1.32 7.29 4.82E+18 2.51E+18 4.33E+18 5.92E+15 7 1.30 8.59 5.53E+18 2.89E+18 5.02E+18 6.89E+15 8 1.35 9.94 6.30E+18 3.29E+18 5.72E+18 7.91E+15 9 1.19 11.13 6.89E+18 3.66E+18 6.32E+18 8.73E+15 10 1.36 12.49 7.63E+18 4.07E+18 7.03E+18 9.79E+15 11 1.38 13.87 8.41E+18 4.46E+18 7.74E+18 1.08E+16 12 1.35 15.22 9.19E+18 4.84E+18 8.44E+18 1.19E+16 13 1.26 16.48 1.01E+19 5.19E+18 9.22E+18 1.30E+16 14 1.22 17.70 1.08E+19 5.51E+18 9.87E+18 1.39E+16 15 1.29 18.99 1.16E+19 5.89E+18 1.06E+19 1.49E+16 16 1.42 20.41 1.23E+19 6.27E+18 1.13E+19 1.59E+16 17 1.35 21.76 1.30E+19 6.68E+18 1.20E+19 1.69E+16 18 1.33 23.09 1.37E+19 7.04E+18 1.26E+19 1.78E+16 19 1.36 24.45 1.44E+19 7.41E+18 1.33E+19 1.88E+16 20 1.30 25.75 1.49E+19 7.77E+18 1.39E+19 1.96E+16 Future 32.00 1.80E+19 9.45E+18 1.68E+19 2.38E+16 Future 35.00 1.95E+19 1.03E+19 1.83E+19 2.58E+16 Future 40.00 2.20E+19 1.16E+19 2.07E+19 2.92E+16 Future 48.00 2.60E+19 1.37E+19 2.45E+19 3.46E+16 Future 54.00 2.90E+19 1.54E+19 2.73E+19 3.87E+16 Future 60.00 3.21E+19 1.70E+19 3.02E+19 4.27E+16 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-19 Table 6-12 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements (dpa)

Outlet Inlet Cycle Total Nozzle to Nozzle to Upper Shell Cycle Length Time Upper Upper Upper Shell Longitudinal Upper (EFPY) (EFPY) Shell Shell Longitudinal Welds at Shell Plates Welds - Welds - Weld at 90° 210° and Lowest Lowest 330° Extent Extent 1 1.05 1.05 6.96E-06 7.89E-06 1.52E-05 1.09E-05 1.28E-05 2 1.15 2.21 1.48E-05 1.68E-05 3.39E-05 2.37E-05 2.95E-05 3 1.23 3.43 2.36E-05 2.68E-05 5.45E-05 3.72E-05 4.52E-05 4 1.26 4.70 3.10E-05 3.52E-05 7.17E-05 4.88E-05 5.94E-05 5 1.27 5.97 3.79E-05 4.30E-05 8.64E-05 5.95E-05 7.17E-05 6 1.32 7.29 4.53E-05 5.14E-05 1.03E-04 7.09E-05 8.61E-05 7 1.30 8.59 5.18E-05 5.88E-05 1.15E-04 8.10E-05 9.59E-05 8 1.35 9.94 5.89E-05 6.69E-05 1.32E-04 9.22E-05 1.10E-04 9 1.19 11.13 6.51E-05 7.39E-05 1.45E-04 1.02E-04 1.22E-04 10 1.36 12.49 7.27E-05 8.25E-05 1.65E-04 1.14E-04 1.38E-04 11 1.38 13.87 8.01E-05 9.10E-05 1.81E-04 1.26E-04 1.52E-04 12 1.35 15.22 8.78E-05 9.97E-05 1.99E-04 1.37E-04 1.67E-04 13 1.26 16.48 9.57E-05 1.09E-04 2.18E-04 1.49E-04 1.83E-04 14 1.22 17.70 1.03E-04 1.16E-04 2.35E-04 1.60E-04 1.97E-04 15 1.29 18.99 1.10E-04 1.25E-04 2.54E-04 1.72E-04 2.13E-04 16 1.42 20.41 1.18E-04 1.34E-04 2.73E-04 1.84E-04 2.29E-04 17 1.35 21.76 1.26E-04 1.43E-04 2.92E-04 1.96E-04 2.45E-04 18 1.33 23.09 1.33E-04 1.51E-04 3.10E-04 2.08E-04 2.61E-04 19 1.36 24.45 1.40E-04 1.59E-04 3.27E-04 2.19E-04 2.76E-04 20 1.30 25.75 1.47E-04 1.67E-04 3.43E-04 2.30E-04 2.90E-04 Future 32.00 1.81E-04 2.05E-04 4.27E-04 2.83E-04 3.62E-04 Future 35.00 1.97E-04 2.24E-04 4.67E-04 3.09E-04 3.96E-04 Future 40.00 2.24E-04 2.55E-04 5.35E-04 3.52E-04 4.54E-04 Future 48.00 2.67E-04 3.04E-04 6.42E-04 4.21E-04 5.46E-04 Future 54.00 3.00E-04 3.41E-04 7.23E-04 4.72E-04 6.15E-04 Future 60.00 3.32E-04 3.78E-04 8.04E-04 5.24E-04 6.84E-04 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-20 Table 6-12 (cont.) Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements (dpa)

Cycle Total Int. Shell Cycle Length Time Upper Shell to Int. Shell Int. Shell to Int. Shell Long. Welds (EFPY) (EFPY) Int. Shell Circ. Long. Weld Lower Shell Plates at 210° and Weld at 90° Circ. Weld 330° 1 1.05 1.05 1.52E-05 1.22E-03 6.41E-04 1.11E-03 1.22E-03 2 1.15 2.21 3.39E-05 2.34E-03 1.31E-03 2.15E-03 2.33E-03 3 1.23 3.43 5.45E-05 3.96E-03 2.04E-03 3.38E-03 3.96E-03 4 1.26 4.70 7.17E-05 5.20E-03 2.69E-03 4.47E-03 5.20E-03 5 1.27 5.97 8.64E-05 6.43E-03 3.31E-03 5.55E-03 6.43E-03 6 1.32 7.29 1.03E-04 7.60E-03 3.89E-03 6.67E-03 7.60E-03 7 1.30 8.59 1.15E-04 8.72E-03 4.46E-03 7.72E-03 8.72E-03 8 1.35 9.94 1.32E-04 9.94E-03 5.09E-03 8.81E-03 9.94E-03 9 1.19 11.13 1.45E-04 1.09E-02 5.66E-03 9.73E-03 1.09E-02 10 1.36 12.49 1.65E-04 1.20E-02 6.29E-03 1.08E-02 1.20E-02 11 1.38 13.87 1.81E-04 1.33E-02 6.90E-03 1.19E-02 1.33E-02 12 1.35 15.22 1.99E-04 1.45E-02 7.48E-03 1.30E-02 1.45E-02 13 1.26 16.48 2.18E-04 1.59E-02 8.03E-03 1.42E-02 1.59E-02 14 1.22 17.70 2.35E-04 1.70E-02 8.53E-03 1.52E-02 1.70E-02 15 1.29 18.99 2.54E-04 1.82E-02 9.12E-03 1.63E-02 1.82E-02 16 1.42 20.41 2.73E-04 1.94E-02 9.71E-03 1.74E-02 1.94E-02 17 1.35 21.76 2.92E-04 2.05E-02 1.03E-02 1.84E-02 2.05E-02 18 1.33 23.09 3.10E-04 2.16E-02 1.09E-02 1.94E-02 2.16E-02 19 1.36 24.45 3.27E-04 2.26E-02 1.15E-02 2.04E-02 2.26E-02 20 1.30 25.75 3.43E-04 2.35E-02 1.20E-02 2.13E-02 2.35E-02 Future 32.00 4.27E-04 2.84E-02 1.46E-02 2.59E-02 2.84E-02 Future 35.00 4.67E-04 3.08E-02 1.59E-02 2.81E-02 3.08E-02 Future 40.00 5.35E-04 3.47E-02 1.80E-02 3.18E-02 3.47E-02 Future 48.00 6.42E-04 4.10E-02 2.13E-02 3.76E-02 4.10E-02 Future 54.00 7.23E-04 4.57E-02 2.38E-02 4.20E-02 4.57E-02 Future 60.00 8.04E-04 5.05E-02 2.63E-02 4.64E-02 5.05E-02 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-21 Table 6-12 (cont.) Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements (dpa)

Cycle Total Lower Shell Lower Shell Cycle Length Time Lower Shell Lower Shell Long. Weld to Lower (EFPY) (EFPY) Long. Weld Plates at 210° and Vessel Head at 90° 330° Circ. Weld 1 1.05 1.05 1.22E-03 6.41E-04 1.11E-03 5.76E-06 2 1.15 2.21 2.33E-03 1.30E-03 2.14E-03 1.17E-05 3 1.23 3.43 3.96E-03 2.04E-03 3.38E-03 1.96E-05 4 1.26 4.70 5.20E-03 2.69E-03 4.47E-03 2.57E-05 5 1.27 5.97 6.43E-03 3.31E-03 5.55E-03 3.16E-05 6 1.32 7.29 7.60E-03 3.89E-03 6.67E-03 3.74E-05 7 1.30 8.59 8.72E-03 4.47E-03 7.72E-03 4.35E-05 8 1.35 9.94 9.94E-03 5.09E-03 8.81E-03 4.99E-05 9 1.19 11.13 1.09E-02 5.67E-03 9.73E-03 5.50E-05 10 1.36 12.49 1.20E-02 6.30E-03 1.08E-02 6.17E-05 11 1.38 13.87 1.33E-02 6.90E-03 1.19E-02 6.81E-05 12 1.35 15.22 1.45E-02 7.49E-03 1.30E-02 7.49E-05 13 1.26 16.48 1.59E-02 8.04E-03 1.42E-02 8.20E-05 14 1.22 17.70 1.70E-02 8.54E-03 1.52E-02 8.78E-05 15 1.29 18.99 1.82E-02 9.13E-03 1.63E-02 9.41E-05 16 1.42 20.41 1.94E-02 9.72E-03 1.74E-02 1.00E-04 17 1.35 21.76 2.05E-02 1.03E-02 1.84E-02 1.06E-04 18 1.33 23.09 2.16E-02 1.09E-02 1.94E-02 1.12E-04 19 1.36 24.45 2.26E-02 1.15E-02 2.04E-02 1.18E-04 20 1.30 25.75 2.35E-02 1.20E-02 2.13E-02 1.23E-04 Future 32.00 2.84E-02 1.46E-02 2.59E-02 1.50E-04 Future 35.00 3.08E-02 1.59E-02 2.81E-02 1.62E-04 Future 40.00 3.47E-02 1.80E-02 3.18E-02 1.84E-04 Future 48.00 4.10E-02 2.13E-02 3.77E-02 2.17E-04 Future 54.00 4.58E-02 2.38E-02 4.21E-02 2.43E-04 Future 60.00 5.05E-02 2.63E-02 4.65E-02 2.68E-04 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-22 Figure 6-1 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules and 12.5° Neutron Pad Configuration WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-23 Figure 6-2 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Single Capsule Holder and 20.0° Neutron Pad Configuration WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-24 Figure 6-3 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Dual Capsule Holder and 22.5° Neutron Pad Configuration WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-25 Figure 6-4 Callaway Unit 1 r,,z Reactor Geometry Section View at 31.5° Azimuthal Angle with Surveillance Capsule WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 7-1 7 SURVEILLANCE CAPSULE REMOVAL

SUMMARY

The following surveillance capsule removal summary (Table 7-1) meets the requirements of ASTM E185-82 [Ref. 8]. Note that all capsules have been removed from the Callaway Unit 1 reactor vessel.

Table 7-1 Surveillance Capsule Withdrawal Summary Capsule Fluence Capsule Capsule Lead Withdrawal Capsule ID Status(a) (n/cm2, E > 1.0 Location Factor(a) EFPY(b, c)

MeV)(c)

Withdrawn U 58.5º 4.07 1.05 0.313 (EOC 1)

Withdrawn Y 241º 3.58 4.70 1.18 (EOC 4)

Withdrawn V 61º 3.68 9.94 2.32 (EOC 8)

Withdrawn X(d) 238.5º 4.03 12.49 3.08 (EOC 10)

Withdrawn W(e) 121.5º 4.02 25.75 5.98 (EOC 20)

Withdrawn Z(f) 301.5º 3.98 16.48 4.00 (EOC 13)

Notes:

(a) Updated in Capsule W dosimetry analysis; see Table 6-10.

(b) EFPY from plant startup.

(c) Updated in Capsule W dosimetry analysis; see Table 6-9.

(d) Capsule X satisfies the 60-year EOLR requirements for 54 EFPY of operation.

(e) Capsule W satisfies the 80, 100, 120 year potential future EOL requirements for 72, 90 and 108 EFPY terms of operation, respectively. The Capsule W fluence value is equal to the projected peak vessel fluence at approximately 115.4 EFPY.

(f) Capsule Z has been placed in the Callaway Unit 1 spent fuel pool per MRP-326 [Ref. 24]. The fluence at the time of withdrawal and the lead factor were updated as part of the Capsule W dosimetry analysis. No specific recommendations are provided herein for this capsule. It should remain in the spent fuel pool for potential testing or reinsertion for further irradiation and/or additional metallurgical testing, if required.

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Westinghouse Non-Proprietary Class 3 8-1 8 REFERENCES

1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

2. 10 CFR 50, Appendix G, Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.

3. Westinghouse Report WCAP-9842, Revision 0, Union Electric Company Callaway Unit No. 1 Reactor Vessel Radiation Surveillance Program, May 1981.

4. ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, ASTM, 1973.

5. Appendix G of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

6. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM.

7. ASTM E399, Test Method for Plane-Strain Fracture Toughness of Metallic Materials, ASTM.

8. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.

9. ASTM E23-12c, Standard Test Methods for Notched Bar Impact Testing of Metallic Materials, ASTM, 2012.

10. ASTM E2298-13a, Standard Test Method for Instrumented Impact Testing of Metallic Materials, ASTM, 2013.

11. ASTM A370-13, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 2013.

12. ASTM E8/E8M-13a, Standard Test Methods for Tension Testing of Metallic Materials, ASTM, 2013.

13. ASTM E21-09, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, ASTM, 2009.

14. Westinghouse Report WCAP-11374, Revision 1, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

15. Westinghouse Report WCAP-12946, Revision 0, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1991.

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Westinghouse Non-Proprietary Class 3 8-2

16. Westinghouse Report WCAP-14895, Revision 0, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, July 1997.

17. Westinghouse Report WCAP-15400, Revision 0, Analysis of Capsule X from the Ameren-UE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 2000.

18. ASTM E853-13, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, ASTM, 2014

19. ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706 (ID), ASTM, 1994.

20. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

21. WCAP-14040, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

22. WCAP-16083-NP, Revision 1, Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry, April 2013.

23. RSICC Data Library Collection DLC-185, BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, July 1999.

24. Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)

Guidelines (MRP-326). EPRI, Palo Alto, CA: 2011. 1022871.

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS A.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn and analyzed to date at Callaway Unit 1 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. A-1]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least-squares adjusted values to within 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report.

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

Capsule Azimuthal Withdrawal Time Irradiation Time Location (EFPY) 58.5º (Capsule U) End of Cycle 1 1.05 241° (Capsule Y) End of Cycle 4 4.70 61° (Capsule V) End of Cycle 8 9.94 238.5° (Capsule X) End of Cycle 10 12.49 121.5° (Capsule W) End of Cycle 20 25.75 The passive neutron sensors included in the evaluations of surveillance Capsules U, Y, V, X, and W are summarized as follows:

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Westinghouse Non-Proprietary Class 3 A-2 Reaction Of Capsule Capsule Capsule Capsule Capsule Sensor Material Interest U Y V X W 63 Copper Cu(n,)60Co X X X X X 54 Iron Fe(n,p)54Mn X X X X X 58 58 Nickel Ni(n,p) Co X X X X X 238 Uranium-238(Cd) U(n,f)FP X X X X X 237 Neptunium-237(Cd) Np(n,f)FP X X X X X Cobalt-Aluminum(1) 59 Co(n,)60Co X X X X X Note:

1. The cobalt-aluminum and uranium monitors for this plant include both bare and cadmium-covered sensors.

Pertinent physical and nuclear characteristics of the passive neutron sensors analyzed are listed in Table A-1 for Capsules U, Y, V, X, and W.

The use of passive monitors such as those listed above do not yield a direct measure of the energy-dependent neutron fluence rate 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 fluence rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron fluence rate level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

the measured specific activity of each monitor, the physical characteristics of each monitor, the operating history of the reactor, the energy response of each monitor, and the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules U, Y, V, and X are documented in References A-2, A-3, A-4, and A-5, respectively. The radiometric counting of the sensors from Capsule W was carried out by Pace Analytical Services, Inc. The radiometric counting followed established ASTM procedures.

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

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

A R

Pj - t j - t d, j N0 F Y C j [1 - e ] [e ]

P ref where:

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

A = Measured specific activity (dps/g).

N0 = Number of target element atoms per gram of sensor.

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

Y = Number of product atoms produced per reaction.

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

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

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

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

tj = Length of irradiation period j (sec).

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

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

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in fluence rate level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, the additional Cj term should be employed. The impact of changing fluence rate 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 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-4 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 fluence rates and the computed values for Cj are listed in Tables A-3 and A-4, respectively, for Capsules U, Y, V, X, and W. These fluence rates represent the capsule- and cycle-dependent results at the radial and azimuthal center of the respective capsules at core 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 cadmium-covered measurements to account for the presence of 235U impurities in the sensors, as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the capsule irradiations. The correction factors corresponding to the Callaway Unit 1 fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V Capsule X Capsule W 235 U Impurity/Pu Build-in 0.8720 0.8385 0.7989 0.7737 0.6840 238 U(,f) 0.9652 0.9662 0.9665 0.9659 0.9689 238 Net U Correction 0.8417 0.8102 0.7721 0.7473 0.6627 238 Np(,f) Correction 0.9900 0.9902 0.9903 0.9901 0.9910 The correction factors for Capsules U, Y, V, X, and W, were applied in a multiplicative fashion to the decay-corrected cadmium-covered uranium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules U, Y, V, X, and W, are given in Tables A-5 through A-9. In Tables A-5 through A-9, the measured specific activities, decay-corrected saturated specific activities, and computed reaction rates for each sensor are listed. The cadmium-covered fission sensor reaction rates are listed both with and without the applied corrections for 235U impurities, plutonium build-in, and gamma-ray-induced fission effects.

A.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best-estimates for key exposure parameters such as fluence rate (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 i R i ( ig ig )( g g )

g WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-5 relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross-sections, ig, each with an uncertainty . 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 Callaway Unit 1 surveillance capsule dosimetry, the FERRET code [Ref. A-6] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (fluence rate (E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.

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

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

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

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

For the Callaway Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A.1.1.

The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [Ref. A-7].

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance [Ref. A-8].

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Callaway Unit 1 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 ensured 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:

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Westinghouse Non-Proprietary Class 3 A-6 Reaction Uncertainty 63 Cu(n,)60Co 5%

54 Fe(n,p)54Mn 5%

58 Ni(n,p)58Co 5%

238 U(n,f)FP 10%

237 Np(n,f)FP 10%

59 Co(n,)60Co 5%

These uncertainties are given at the 1 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 recent cross-section evaluations, and they have been tested for 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 Callaway Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

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

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

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

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

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

59 Co(n,)60Co 0.79-3.59%

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

Calculated Neutron Spectrum The neutron spectra inputs 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 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-7 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:

M gg' R 2n R g

• R g'

• 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:

Pgg = [1 - ] gg + e-H where (g g' ) 2 H

2 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term). The value of is 1.0 when g = g, and is 0.0 otherwise.

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

Fluence Rate Normalization Uncertainty (Rn) 15%

Fluence Rate Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Fluence Rate Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-8 A.1.3 Comparisons of Measurements and Calculations Results of the least-squares evaluations of the dosimetry from the Callaway Unit 1 surveillance capsules withdrawn to date are provided in Tables A-10, A-11, A-12, A-13, and A-14 for Capsules U, Y, V, X, and W, respectively. In these tables, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in these tabulations 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. Additionally, comparisons of the calculated and best-estimate values of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the BE/C ratios observed for each of the capsules. Note that for Capsule X, the cadmium-covered uranium monitor was discarded because it was outside the expected values.

The data comparisons provided in Tables A-10 through A-14 show that the adjustments to the calculated spectra are relatively small and 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, the calculational uncertainty is specified as 13% at the 1 level.

Further comparisons of the measurement results with calculations are given in Tables A-15 and A-16.

These comparisons are given on two levels. In Table A-15, 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-16, calculations of fast neutron exposure rates in terms of fast neutron fluence rate (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 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, for the individual threshold foils considered in the least-squares analysis, the average M/C comparisons for fast neutron reactions range from 1.00 to 1.17 in the data set. The overall average M/C ratio for the entire set of Callaway Unit 1 data is 1.07 with an associated standard deviation of 7.9%.

In the comparisons of best-estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.94 to 1.10 for neutron fluence rate (E > 1.0 MeV) and from 0.95 to 1.09 for iron atom displacement rate. The overall average BE/C ratios for neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are 1.02 with a standard deviation of 5.9% and 1.02 with a standard deviation of 5.2%, respectively.

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

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Westinghouse Non-Proprietary Class 3 A-9 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors 90%

Reaction Atomic Target Product Fission

Response

of Weight Atom Half-life Yield Range(a)

Interest (g/g-atom) Fraction (days) (%)

(MeV) 63 Cu (n,) 60Co 63.546 0.6917 1925.5 n/a 5 - 12 54 Fe (n,p) 54Mn 55.845 0.05845 312.11 n/a 2-9 58 Ni (n,p) 58Co 58.693 0.68077 70.82 n/a 2-8 238 U (n,f) 137Cs 238.051 0.9996 10983.07 6.02 1-7 237 Np (n,f) 137Cs 237.048 1.0 10983.07 6.17 0.3 - 4 59 Co (n,) 60Co 58.933 0.0015 1925.5 n/a non-threshold Note:

(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Callaway Unit 1 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 [Ref. A-9].

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Westinghouse Non-Proprietary Class 3 A-10 Table A-2 Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 1 Cycle 2 Cycle 3 Cycle 4 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Oct-84 73090 Apr-86 558152 Nov-87 776234 May-89 459251 Nov-84 878649 May-86 1628231 Dec-87 2498378 Jun-89 2230340 Dec-84 1001525 Jun-86 1891846 Jan-88 2400761 Jul-89 2646374 Jan-85 1825824 Jul-86 2279207 Feb-88 1944633 Aug-89 2567249 Feb-85 1997026 Aug-86 2236572 Mar-88 2534237 Sep-89 2346488 Mar-85 2093839 Sep-86 2395748 Apr-88 1401495 Oct-89 2607932 Apr-85 1476721 Oct-86 2478466 May-88 2444952 Nov-89 2529924 May-85 2351261 Nov-86 2000621 Jun-88 2547644 Dec-89 2644439 Jun-85 1990891 Dec-86 2476793 Jul-88 2624336 Jan-90 2595499 Jul-85 2204437 Jan-87 2310037 Aug-88 2590033 Feb-90 2367162 Aug-85 2226770 Feb-87 2278365 Sep-88 2295718 Mar-90 2604203 Sep-85 2399444 Mar-87 2310093 Oct-88 2526306 Apr-90 2557687 Oct-85 2206396 Apr-87 99068 Nov-88 2355533 May-90 2394320 Nov-85 1940798 May-87 1527462 Dec-88 1913183 Jun-90 2193115 Dec-85 2212562 Jun-87 2442877 Jan-89 2640944 Jul-90 2587861 Jan-86 2445738 Jul-87 2426722 Feb-89 2391524 Aug-90 2614245 Feb-86 2161005 Aug-87 2466901 Mar-89 2483347 Sep-90 1500131 Sep-87 673770 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-11 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 5 Cycle 6 Cycle 7 Cycle 8 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-90 481701 May-92 657818 Nov-93 528099 May-95 1548130 Dec-90 2557248 Jun-92 2520702 Dec-93 2607610 Jun-95 2449671 Jan-91 2629501 Jul-92 2608804 Jan-94 2615603 Jul-95 2595094 Feb-91 2376421 Aug-92 2598660 Feb-94 2375405 Aug-95 2529270 Mar-91 2627949 Sep-92 2394003 Mar-94 2438287 Sep-95 2524783 Apr-91 2532714 Oct-92 2636341 Apr-94 2538220 Oct-95 1784097 May-91 2620742 Nov-92 2540736 May-94 2630033 Nov-95 2440443 Jun-91 2499859 Dec-92 2619556 Jun-94 2543816 Dec-95 2628282 Jul-91 2592727 Jan-93 2623796 Jul-94 2549178 Jan-96 2562335 Aug-91 2638549 Feb-93 2361386 Aug-94 2635385 Feb-96 2421446 Sep-91 2485628 Mar-93 2625029 Sep-94 2549205 Mar-96 2649287 Oct-91 2576841 Apr-93 2371166 Oct-94 2636622 Apr-96 2415714 Nov-91 2266100 May-93 2618825 Nov-94 2547712 May-96 2518985 Dec-91 2602283 Jun-93 2530244 Dec-94 2632215 Jun-96 2554157 Jan-92 2326738 Jul-93 2612147 Jan-95 2593618 Jul-96 2641740 Feb-92 2436718 Aug-93 2609724 Feb-95 2378690 Aug-96 2648027 Mar-92 1477011 Sep-93 2296400 Mar-95 1845718 Sep-96 2552943 Oct-93 712 Oct-96 829440 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-12 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 9 Cycle 10 Cycle 11 Cycle 12 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-96 1378918 May-98 2203216 Nov-99 1816209 May-01 592977 Dec-96 2240097 Jun-98 2565200 Dec-99 2637391 Jun-01 2559683 Jan-97 2649851 Jul-98 2649883 Jan-00 2645815 Jul-01 2646500 Feb-97 2290910 Aug-98 2648018 Feb-00 2327865 Aug-01 2650426 Mar-97 2647771 Sep-98 2559293 Mar-00 2640004 Sep-01 2563968 Apr-97 2557647 Oct-98 2520911 Apr-00 2552392 Oct-01 2653245 May-97 2634434 Nov-98 2412725 May-00 2634464 Nov-01 2562952 Jun-97 2562176 Dec-98 2460391 Jun-00 2551992 Dec-01 2296225 Jul-97 2555565 Jan-99 2632436 Jul-00 2643090 Jan-02 2595946 Aug-97 2268517 Feb-99 2384612 Aug-00 2649040 Feb-02 1081321 Sep-97 1743769 Mar-99 2648559 Sep-00 2562040 Mar-02 2594174 Oct-97 1855609 Apr-99 2557196 Oct-00 2652795 Apr-02 2552512 Nov-97 1789970 May-99 2640967 Nov-00 2561131 May-02 2563532 Dec-97 1907667 Jun-99 2561307 Dec-00 2649483 Jun-02 2563713 Jan-98 1993020 Jul-99 2645814 Jan-01 2649682 Jul-02 2648139 Feb-98 1914092 Aug-99 1813256 Feb-01 2390468 Aug-02 2649574 Mar-98 2221975 Sep-99 2435460 Mar-01 2171069 Sep-02 2562792 Apr-98 97613 Oct-99 33272 Apr-01 378935 Oct-02 1837500 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-13 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 13 Cycle 14 Cycle 15 Cycle 16 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-02 259329 Jun-04 1316596 Nov-05 878563 May-07 1776516 Dec-02 2234008 Jul-04 2651150 Dec-05 2648916 Jun-07 2562458 Jan-03 2650782 Aug-04 2644655 Jan-06 2651165 Jul-07 2650823 Feb-03 2386156 Sep-04 2563708 Feb-06 2393308 Aug-07 2650708 Mar-03 1796248 Oct-04 2649341 Mar-06 2650430 Sep-07 2563252 Apr-03 2379519 Nov-04 2560242 Apr-06 2561630 Oct-07 2649187 May-03 2649671 Dec-04 2646370 May-06 1265698 Nov-07 2567302 Jun-03 2565526 Jan-05 2446478 Jun-06 2277193 Dec-07 2647539 Jul-03 2649539 Feb-05 2394377 Jul-06 2650620 Jan-08 2647435 Aug-03 2650619 Mar-05 2144356 Aug-06 2650699 Feb-08 2478102 Sep-03 2552394 Apr-05 2379250 Sep-06 2565194 Mar-08 2596436 Oct-03 2318184 May-05 2650581 Oct-06 2653569 Apr-08 2554914 Nov-03 2550214 Jun-05 2360203 Nov-06 2565181 May-08 2649463 Dec-03 2650963 Jul-05 2648684 Dec-06 2650432 Jun-08 2564338 Jan-04 2492470 Aug-05 2649193 Jan-06 2650533 Jul-08 2629985 Feb-04 1203392 Sep-05 1352545 Feb-07 2393598 Aug-08 2649172 Mar-04 2650570 Mar-07 2271252 Sep-08 2564947 Apr-04 719857 Apr-07 67941 Oct-08 838215 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-14 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 17 Cycle 18 Cycle 19 Cycle 20 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-08 1704193 Jun-10 1499003 Nov-11 385273 May-13 178430 Dec-08 1639016 Jul-10 2650897 Dec-11 2650840 Jun-13 2564098 Jan-09 2651078 Aug-10 2650782 Jan-12 2650817 Jul-13 2146631 Feb-09 1570120 Sep-10 2564807 Feb-12 2479390 Aug-13 1034909 Mar-09 2403702 Oct-10 2650536 Mar-12 2647173 Sep-13 2565262 Apr-09 2331500 Nov-10 2568427 Apr-12 2544119 Oct-13 2650010 May-09 2650372 Dec-10 2649808 May-12 2650393 Nov-13 2568480 Jun-09 2483116 Jan-11 2588016 Jun-12 2565043 Dec-13 2650392 Jul-09 2650531 Feb-11 2393612 Jul-12 2650101 Jan-14 2649662 Aug-09 2650765 Mar-11 2644270 Aug-12 2650503 Feb-14 2393563 Sep-09 2565276 Apr-11 2565032 Sep-12 2564922 Mar-14 2646443 Oct-09 2649462 May-11 2647739 Oct-12 2649238 Apr-14 2565105 Nov-09 2568639 Jun-11 2564747 Nov-12 2568195 May-14 2649755 Dec-09 2649580 Jul-11 2650371 Dec-12 2650389 Jun-14 2564943 Jan-10 2650452 Aug-11 2650164 Jan-13 2649965 Jul-14 2648905 Feb-10 2393193 Sep-11 2521112 Feb-13 2393843 Aug-14 2650008 Mar-10 2646205 Oct-11 1195960 Mar-13 2583339 Sep-14 2564831 Apr-10 1366927 Apr-13 512141 Oct-14 854474 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-15 Table A-3 Surveillance Capsules U, Y, V, X, and W Fluence Rates for Cj Calculation, Core Midplane Elevation (E > 1.0 MeV) [n/cm2-s]

Cycle Fuel Cycle Length Capsule U Capsule Y Capsule V Capsule X Capsule W (EFPY) 1 1.05 9.42E+10 8.79E+10 8.79E+10 9.42E+10 9.34E+10 2 1.15 7.76E+10 7.76E+10 8.09E+10 8.01E+10 3 1.23 8.26E+10 8.26E+10 9.51E+10 9.45E+10 4 1.26 7.16E+10 7.16E+10 7.92E+10 7.86E+10 5 1.27 7.10E+10 7.71E+10 7.65E+10 6 1.32 7.13E+10 7.57E+10 7.51E+10 7 1.30 6.66E+10 7.11E+10 7.05E+10 8 1.35 6.58E+10 7.12E+10 7.06E+10 9 1.19 6.80E+10 6.74E+10 10 1.36 7.21E+10 7.15E+10 11 1.38 6.93E+10 12 1.35 7.09E+10 13 1.26 8.66E+10 14 1.22 7.29E+10 15 1.29 7.43E+10 16 1.42 6.73E+10 17 1.35 7.12E+10 18 1.33 6.56E+10 19 1.36 6.38E+10 20 1.30 5.98E+10 Average - 9.42E+10 7.96E+10 7.38E+10 7.81E+10 7.36E+10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-16 Table A-4 Surveillance Capsules U, Y, V, X, and W Cj Factors, Core Midplane Elevation Cj Cycle Fuel Cycle Length Capsule U Capsule Y Capsule V Capsule X Capsule W (EFPY) 1 1.05 1.00 1.10 1.19 1.21 1.27 2 1.15 0.97 1.05 1.04 1.09 3 1.23 1.04 1.12 1.22 1.28 4 1.26 0.90 0.97 1.02 1.07 5 1.27 0.96 0.99 1.04 6 1.32 0.97 0.97 1.02 7 1.30 0.90 0.91 0.96 8 1.35 0.89 0.91 0.96 9 1.19 0.87 0.92 10 1.36 0.92 0.97 11 1.38 0.94 12 1.35 0.96 13 1.26 1.18 14 1.22 0.99 15 1.29 1.01 16 1.42 0.91 17 1.35 0.97 18 1.33 0.89 19 1.36 0.87 20 1.30 0.81 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-17 Table A-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 4.78E+04 3.93E+05 6.00E-17 Cu-63 Co-60 4.57E+04 3.76E+05 5.73E-17 Cu-63 Co-60 4.29E+04 3.53E+05 5.38E-17 5.71E-17 5.71E-17 Fe-54 Mn-54 1.54E+06 3.87E+06 6.14E-15 Fe-54 Mn-54 1.44E+06 3.63E+06 5.76E-15 Fe-54 Mn-54 1.36E+06 3.43E+06 5.45E-15 5.78E-15 5.78E-15 Ni-58 Co-58 1.31E+07 6.03E+07 8.64E-15 Ni-58 Co-58 1.19E+07 5.50E+07 7.87E-15 Ni-58 Co-58 1.17E+07 5.38E+07 7.70E-15 8.07E-15 8.07E-15 U-238 Cs-137 1.47E+05 6.20E+06 4.07E-14 4.07E-14 3.43E-14 Np-237 Cs-137 1.08E+06 4.57E+07 2.91E-13 2.91E-13 2.88E-13 Co-59 Co-60 9.73E+06 8.00E+07 5.22E-12 Co-59 Co-60 9.94E+06 8.17E+07 5.33E-12 Co-59 Co-60 9.40E+06 7.73E+07 5.04E-12 5.20E-12 5.20E-12 Co-59(Cd) Co-60 4.98E+06 4.10E+07 2.67E-12 Co-59(Cd) Co-60 5.38E+06 4.43E+07 2.89E-12 Co-59(Cd) Co-60 4.90E+06 4.03E+07 2.63E-12 2.73E-12 2.73E-12 Note:

(a) Measured activity decay corrected to July 22, 1986 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-18 Table A-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 1.50E+05 1.50E+05 3.58E+05 5.46E-17 Cu-63 Co-60 1.34E+05 1.34E+05 3.20E+05 4.88E-17 Cu-63 Co-60 1.31E+05 1.31E+05 3.13E+05 4.77E-17 5.03E-17 Fe-54 Mn-54 2.12E+06 2.12E+06 3.12E+06 4.95E-15 Fe-54 Mn-54 1.90E+06 1.90E+06 2.80E+06 4.44E-15 Fe-54 Mn-54 1.88E+06 1.88E+06 2.77E+06 4.39E-15 4.59E-15 Ni-58 Co-58 1.89E+07 1.89E+07 5.04E+07 7.21E-15 Ni-58 Co-58 1.73E+07 1.73E+07 4.61E+07 6.60E-15 Ni-58 Co-58 1.70E+07 1.70E+07 4.53E+07 6.49E-15 6.77E-15 U-238 Cs-137 5.37E+05 5.37E+05 5.33E+06 3.50E-14 2.84E-14 Np-237 Cs-137 4.07E+06 4.07E+06 4.04E+07 2.58E-13 2.55E-13 Co-59 Co-60 2.53E+07 2.53E+07 6.04E+07 3.94E-12 Co-59 Co-60 2.45E+07 2.45E+07 5.84E+07 3.81E-12 Co-59 Co-60 2.54E+07 2.54E+07 6.06E+07 3.95E-12 3.90E-12 Co-59(Cd) Co-60 1.32E+07 1.32E+07 3.15E+07 2.05E-12 Co-59(Cd) Co-60 1.30E+07 1.30E+07 3.10E+07 2.02E-12 Co-59(Cd) Co-60 1.34E+07 1.34E+07 3.20E+07 2.09E-12 2.05E-12 Note:

(a) Measured activity decay corrected to December 12, 1990 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-19 Table A-7 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.09E+05 3.29E+05 5.01E-17 Cu-63 Co-60 1.87E+05 2.94E+05 4.48E-17 Cu-63 Co-60 1.83E+05 2.88E+05 4.39E-17 4.63E-17 4.63E-17 Fe-54 Mn-54 2.01E+06 2.89E+06 4.58E-15 Fe-54 Mn-54 1.83E+06 2.63E+06 4.17E-15 Fe-54 Mn-54 1.78E+06 2.56E+06 4.06E-15 4.27E-15 4.27E-15 Ni-58 Co-58 1.83E+07 4.49E+07 6.43E-15 Ni-58 Co-58 1.67E+07 4.10E+07 5.87E-15 Ni-58 Co-58 1.61E+07 3.95E+07 5.66E-15 5.98E-15 5.98E-15 U-238 Cs-137 1.02E+06 5.12E+06 3.37E-14 3.37E-14 2.60E-14 Np-237 Cs-137 7.16E+06 3.60E+07 2.30E-13 2.30E-13 2.27E-13 Co-59 Co-60 3.26E+07 5.12E+07 3.34E-12 Co-59 Co-60 3.24E+07 5.09E+07 3.32E-12 Co-59 Co-60 3.24E+07 5.09E+07 3.32E-12 3.33E-12 3.33E-12 Co-59(Cd) Co-60 1.67E+07 2.63E+07 1.71E-12 Co-59(Cd) Co-60 1.75E+07 2.75E+07 1.79E-12 Co-59(Cd) Co-60 1.75E+07 2.75E+07 1.79E-12 1.77E-12 1.77E-12 Note:

(a) Measured activity decay corrected to December 27, 1996 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-20 Table A-8 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.26E+05 3.32E+05 5.07E-17 Cu-63 Co-60 2.02E+05 2.97E+05 4.53E-17 Cu-63 Co-60 1.97E+05 2.90E+05 4.42E-17 4.67E-17 4.67E-17 Fe-54 Mn-54 2.02E+06 3.08E+06 4.89E-15 Fe-54 Mn-54 1.81E+06 2.76E+06 4.38E-15 Fe-54 Mn-54 1.77E+06 2.70E+06 4.28E-15 4.51E-15 4.51E-15 Ni-58 Co-58 1.47E+07 4.68E+07 6.70E-15 Ni-58 Co-58 1.34E+07 4.27E+07 6.11E-15 Ni-58 Co-58 1.24E+07 3.95E+07 5.65E-15 6.16E-15 6.16E-15 U-238 Cs-137 6.10E+05 2.53E+06 1.66E-14 1.66E-14 1.24E-14 Np-237 Cs-137 8.97E+06 3.72E+07 2.37E-13 2.37E-13 2.35E-13 Co-59 Co-60 3.62E+07 5.32E+07 3.47E-12 Co-59 Co-60 3.37E+07 4.95E+07 3.23E-12 Co-59 Co-60 3.29E+07 4.84E+07 3.16E-12 3.29E-12 3.29E-12 Co-59(Cd) Co-60 1.97E+07 2.90E+07 1.89E-12 Co-59(Cd) Co-60 2.04E+07 3.00E+07 1.96E-12 Co-59(Cd) Co-60 2.01E+07 2.95E+07 1.93E-12 1.92E-12 1.92E-12 Note:

(a) Measured activity decay corrected to January 10, 2000 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-21 Table A-9 Measured Sensor Activities and Reaction Rates for Surveillance Capsule W Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.49E+05 2.49E+05 3.22E+05 4.91E-17 Cu-63 Co-60 2.21E+05 2.21E+05 2.86E+05 4.36E-17 Cu-63 Co-60 2.17E+05 2.17E+05 2.81E+05 4.28E-17 4.52E-17 Fe-54 Mn-54 1.89E+06 1.89E+06 3.19E+06 5.06E-15 Fe-54 Mn-54 1.66E+06 1.66E+06 2.80E+06 4.45E-15 Fe-54 Mn-54 1.66E+06 1.66E+06 2.80E+06 4.45E-15 4.65E-15 Ni-58 Co-58 1.27E+07 1.27E+07 5.05E+07 7.23E-15 Ni-58 Co-58 1.15E+07 1.15E+07 4.57E+07 6.54E-15 Ni-58 Co-58 1.14E+07 1.14E+07 4.53E+07 6.49E-15 6.75E-15 U-238 Cs-137 1.40E+06 2.73E+06 6.47E+06 4.25E-14 2.82E-14 Np-237 Cs-137 8.27E+06 1.61E+07 3.82E+07 2.43E-13 2.41E-13 Co-59 Co-60 4.26E+07 4.26E+07 5.51E+07 3.60E-12 Co-59 Co-60 4.30E+07 4.30E+07 5.56E+07 3.63E-12 Co-59 Co-60 4.33E+07 4.33E+07 5.60E+07 3.65E-12 3.63E-12 Co-59(Cd) Co-60 2.19E+07 2.19E+07 2.83E+07 1.85E-12 Co-59(Cd) Co-60 2.26E+07 2.26E+07 2.92E+07 1.91E-12 Co-59(Cd) Co-60 2.33E+07 2.33E+07 3.01E+07 1.97E-12 1.91E-12 Note:

(a) Measured activity decay corrected to February 6, 2015 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-22 Table A-10 Least-Squares Evaluation of Dosimetry in Surveillance Capsule U (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycle 1 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 5.70E-17 4.77E-17 5.52E-17 1.20 1.03 1.16 Fe-54(n,p)Mn-54 5.78E-15 5.41E-15 5.89E-15 1.07 0.98 1.09 Ni-58(n,p)Co-58 8.07E-15 7.62E-15 8.23E-15 1.06 0.98 1.08 U-238(Cd)(n,f)Cs-137 3.43E-14 2.96E-14 3.16E-14 1.16 1.09 1.07 Np-237(Cd)(n,f)Cs-137 2.88E-13 2.96E-13 2.98E-13 0.97 0.97 1.01 Co-59(n,g)Co-60 5.20E-12 4.60E-12 5.15E-12 1.13 1.01 1.12 Co-59(Cd)(n,g)Co-60 2.73E-12 3.00E-12 2.76E-12 0.91 0.99 0.92 Average of Fast Energy Threshold Reactions 1.09 1.01 1.08

% standard deviation 8.3 5.0 5.0 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 9.47E+10 13 9.94E+10 6 1.05 (n/cm2-s)

Fluence rate E > 0.1 MeV 4.31E+11 - 4.43E+11 10 1.02 (n/cm2-s) dpa/s 1.85E-10 13 1.93E-10 8 1.04 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-23 Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule Y (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 4 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 5.03E-17 4.20E-17 4.80E-17 1.20 1.05 1.14 Fe-54(n,p)Mn-54 4.59E-15 4.64E-15 4.85E-15 0.99 0.95 1.04 Ni-58(n,p)Co-58 6.77E-15 6.52E-15 6.85E-15 1.04 0.99 1.05 U-238(Cd)(n,f)Cs-137 2.84E-14 2.51E-14 2.60E-14 1.13 1.09 1.03 Np-237(Cd)(n,f)Cs-137 2.55E-13 2.50E-13 2.55E-13 1.02 1.00 1.02 Co-59(n,g)Co-60 3.90E-12 3.80E-12 3.87E-12 1.03 1.01 1.02 Co-59(Cd)(n,g)Co-60 2.05E-12 2.51E-12 2.08E-12 0.82 0.99 0.83 Average of Fast Energy Threshold Reactions 1.08 1.02 1.06

% standard deviation 8.1 5.4 4.6 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 8.00E+10 13 8.19E+10 6 1.02 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.63E+11 - 3.71E+11 10 1.02 (n/cm2-s) dpa/s 1.56E-10 13 1.60E-10 8 1.02 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-24 Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 8 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 4.63E-17 3.97E-17 4.41E-17 1.17 1.05 1.11 Fe-54(n,p)Mn-54 4.27E-15 4.34E-15 4.43E-15 0.98 0.96 1.02 Ni-58(n,p)Co-58 5.98E-15 6.09E-15 6.17E-15 0.98 0.97 1.01 U-238(Cd)(n,f)Cs-137 2.60E-14 2.34E-14 2.35E-14 1.11 1.11 1.01 Np-237(Cd)(n,f)Cs-137 2.27E-13 2.31E-13 2.28E-13 0.98 1.00 0.99 Co-59(n,g)Co-60 3.33E-12 3.50E-12 3.31E-12 0.95 1.01 0.95 Co-59(Cd)(n,g)Co-60 1.77E-12 2.30E-12 1.79E-12 0.77 0.99 0.78 Average of Fast Energy Threshold Reactions 1.04 1.02 1.03

% standard deviation 8.6 6.1 4.6 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 7.42E+10 13 7.38E+10 6 0.99 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.35E+11 - 3.32E+11 10 0.99 (n/cm2-s) dpa/s 1.44E-10 13 1.44E-10 8 0.99 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-25 Table A-13 Least-Squares Evaluation of Dosimetry in Surveillance Capsule X (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 10 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 4.67E-17 4.17E-17 4.49E-17 1.12 1.04 1.08 Fe-54(n,p)Mn-54 4.51E-15 4.59E-15 4.56E-15 0.98 0.99 1.00 Ni-58(n,p)Co-58 6.16E-15 6.44E-15 6.32E-15 0.96 0.97 0.98 Np-237(Cd)(n,f)Cs-137 2.35E-13 2.44E-13 2.32E-13 0.96 1.01 0.95 Co-59(n,g)Co-60 3.29E-12 3.74E-12 3.28E-12 0.88 1.00 0.88 Co-59(Cd)(n,g)Co-60 1.92E-12 2.45E-12 1.94E-12 0.79 0.99 0.79 Average of Fast Energy Threshold Reactions 1.01 1.00 1.00

% standard deviation 7.7 3.0 5.5 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 7.84E+10 13 7.44E+10 7 0.94 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.53E+11 - 3.35E+11 10 0.94 (n/cm2-s) dpa/s 1.52E-10 13 1.45E-10 8 0.95 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-26 Table A-14 Least-Squares Evaluation of Dosimetry in Surveillance Capsule W (31.5° Azimuth, Core Midplane - Single Capsule Holder) Cycles 1 Through 20 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 4.52E-17 3.94E-17 4.45E-17 1.15 1.02 1.13 Fe-54(n,p)Mn-54 4.65E-15 4.32E-15 4.78E-15 1.08 0.97 1.11 Ni-58(n,p)Co-58 6.75E-15 6.06E-15 6.75E-15 1.11 1.00 1.11 U-238(Cd)(n,f)Cs-137 2.82E-14 2.32E-14 2.58E-14 1.21 1.09 1.11 Np-237(Cd)(n,f)Cs-137 2.41E-13 2.30E-13 2.47E-13 1.05 0.98 1.07 Co-59(n,g)Co-60 3.63E-12 3.20E-12 3.59E-12 1.13 1.01 1.12 Co-59(Cd)(n,g)Co-60 1.91E-12 2.13E-12 1.93E-12 0.90 0.99 0.91 Average of Fast Energy Threshold Reactions 1.12 1.01 1.11

% standard deviation 5.6 4.7 2.0 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 7.40E+10 13 8.17E+10 6 1.10 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.33E+11 - 3.60E+11 10 1.08 (n/cm2-s) dpa/s 1.43E-10 13 1.57E-10 8 1.09 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-27 Table A-15 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions M/C Capsule 63 54 58 238 237 Cu(n,) Fe(n,p) Ni(n,p) U(n,f) Np(n,f)

U 1.20 1.07 1.06 1.16 0.97 Y 1.20 0.99 1.04 1.13 1.02 V 1.17 0.98 0.98 1.11 0.98 X 1.12 0.98 0.96 - 0.96 W 1.15 1.08 1.11 1.21 1.05 Average 1.17 1.02 1.03 1.15 1.00

% Standard 2.9 5.0 5.9 3.8 3.8 Deviation Average 1.07

% Standard 7.9 Deviation Table A-16 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Capsule Neutron Fluence Rate Iron Atom Displacement Rate (E > 1.0 MeV)

U 1.05 1.04 Y 1.02 1.02 V 0.99 0.99 X 0.94 0.95 W 1.10 1.09 Average 1.02 1.02

% Standard deviation 5.9 5.2 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-28 A.2 REFERENCES A-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

A-2 Westinghouse Report WCAP-11374, Revision 1, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

A-3 Westinghouse Report WCAP-12946, Revision 0, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1991.

A-4 Westinghouse Report WCAP-14895, Revision 0, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, July 1997.

A-5 Westinghouse Report WCAP-15400, Revision 0, Analysis of Capsule X from the Ameren-UE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 2000.

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

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

A-8 ASTM Standard E944-13, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA), 2014.

A-9 ASTM Standard E844-09, Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC), 2014.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS CLXX denotes Lower Shell Plate R2708-1, longitudinal orientation CTXX denotes Lower Shell Plate R2708-1, transverse orientation CWXX denotes weld material CHXX denotes heat affected zone material Note that the instrumented Charpy data is not required per ASTM Standards E185-82 or E23-12c.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-2 CL32: Tested at 0°F CL38: Tested at 20°F CL31: Tested at 35°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-3 CL45: Tested at 40°F CL39: Tested at 50°F CL40: Tested at 50°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-4 CL34: Tested at 72°F CL37: Tested at 90°F CL44: Tested at 100°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-5 CL36: Tested at 120°F CL41: Tested at 150°F CL33: Tested at 170°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-6 CL42: Tested at 200°F CL43: Tested at 230°F CL35: Tested at 250°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-7 CT31: Tested at 0°F CT42: Tested at 20°F CT36: Tested at 40°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-8 CT40: Tested at 50°F CT43: Tested at 60°F CT44: Tested at 60°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-9 CT38: Tested at 72°F CT32: Tested at 100°F CT34: Tested at 120°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-10 CT35: Tested at 150°F CT39: Tested at 170°F CT37: Tested at 200°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-11 CT45: Tested at 230°F CT41: Tested at 250°F CT33: Tested at 275°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-12 CW45: Tested at -25°F CW36: Tested at 0°F CW40: Tested at 10°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-13 CW32: Tested at 20°F CW34: Tested at 25°F CW42: Tested at 35°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-14 CW33: Tested at 40°F CW44: Tested at 50°F CW43: Tested at 60°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-15 CW38: Tested at 72°F CW37: Tested at 120°F CW31: Tested at 150°F WCAP-18001-NP September 2015 Revision 0

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Westinghouse Non-Proprietary Class 3 B-18 CH37: Tested at 0°F CH43: Tested at 10°F CH34: Tested at 20°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-19 CH40: Tested at 35°F CH33: Tested at 50°F CH39: Tested at 60°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-20 CH31: Tested at 70°F CH35: Tested at 72°F CH36: Tested at 100°F WCAP-18001-NP September 2015 Revision 0

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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD C.1 METHODOLOGY Contained in Table C-1 are the upper-shelf energy (USE) values that are used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 6.0. The definition for USE is given in ASTM E185-82 [Ref. C-1], Section 4.18, and reads as follows:

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

Westinghouse reports the average of all Charpy data ( 95% shear) as the USE, excluding any values that are deemed outliers using engineering judgment. Hence, the Capsule W USE values reported in Table C-1 were determined by applying this methodology to the Charpy data tabulated in Tables 5-1 through 5-4 of this report. USE values documented in Table C-1 for the unirradiated material, as well as Capsules U, Y, V and X were also determined by applying the methodology described above to the Charpy impact data reported in WCAP-9842, Revision 0 [Ref. C-2], WCAP-11374, Revision 1 [Ref. C-3], WCAP-12946, Revision 0 [Ref. C-4], WCAP-14895, Revision 0 [Ref. C-5] and WCAP-15400, Revision 0 [Ref. C-6]. Three data points differ from those previously reported, due to the inclusion of all Charpy data ( 95% shear) as the USE: Unirradiated material (weld material), Capsule U (lower shell plate material, longitudinal direction) and Capsule V (HAZ material). The USE values reported in Table C-1 were used in generation of the Charpy V-notch curves.

The lower-shelf energy values were fixed at 2.2 ft-lb for all cases. The lower-shelf lateral expansion (L.E.) values were fixed at 1.0 mils in order to be consistent with the previous capsule analysis [Ref. C-6].

Upper-shelf L.E. is not typically fixed in CVGRAPH; however, due to an inaccurate curve fit, the upper shelf L.E. value will be fixed in a summary plot, as documented in Section 5 of this report, for the T-L material in Capsule W. The individual L.E. plot for this material, as documented in this Appendix, will still allow the upper-shelf L.E. to float for comparison between the two methods. The fixed upper-shelf L.E. value was determined using the same Charpy V-notch test specimens that were used for the upper-shelf energy determination and is shown in Table C-2.

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Westinghouse Non-Proprietary Class 3 C-2 Table C-1 Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH Capsule Material Unirradiated U Y V X W Lower Shell Plate R2708-1 126 124 118 126 120 117 Longitudinal Orientation Lower Shell Plate R2708-1 104 93 91 101 99 83 Transverse Orientation Surveillance Program Weld Metal 107 101 97 104 104 103 (Heat # 90077)

Heat Affected Zone (HAZ) Material 106 120 91 106 115 101 Table C-2 Upper-Shelf L.E. Value (mils) Fixed in CVGRAPH Capsule Material W

Lower Shell Plate R2708-1 62 Transverse Orientation (see Figure 5-5 of this report)

CVGRAPH, Version 6.0 plots of all surveillance data are provided in this appendix, on the pages following the reference list.

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Westinghouse Non-Proprietary Class 3 C-3 C.2 REFERENCES C-1 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706(IF), ASTM, 1982.

C-2 Westinghouse Report WCAP-9842, Revision 0, Union Electric Company Callaway Unit No. 1 Reactor Vessel Radiation Surveillance Program, May 1981 C-3 Westinghouse Report WCAP-11374, Revision 1, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

C-4 Westinghouse Report WCAP-12946, Revision 0, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1991.

C-5 Westinghouse Report WCAP-14895, Revision 0, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, July 1997.

C-6 Westinghouse Report WCAP-15400, Revision 0, Analysis of Capsule X from the Ameren-UE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 2000.

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Westinghouse Non-Proprietary Class 3 C-4 C.3 CVGRAPH VERSION 6.0 INDIVIDUAL PLOTS WCAP-18001-NP September 2015 Revision 0

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Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D CALLAWAY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION D.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there have been five surveillance capsules removed and tested from the Callaway Unit 1 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Callaway Unit 1 reactor vessel surveillance data and determine if that surveillance data is credible.

D.2 EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, Fracture Toughness Requirements [Ref. D-2], as follows:

the region of 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.

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Westinghouse Non-Proprietary Class 3 D-2 The Callaway Unit 1 reactor vessel beltline region consists of the following materials:

1. Intermediate Shell Plates R2707-1, R2707-2 and R2707-3

2. Lower Shell Plates R2708-1, R2708-2 and R2708-3

3. Intermediate Shell to Lower Shell Circumferential Weld Seam 101-171 (Heat # 90077, Linde Flux Type 124)

4. Intermediate Shell Plate Longitudinal Weld Seams 101-124A, 101-124B and 101-124C (Heat # 90077, Linde Flux Type 0091)

5. Lower Shell Longitudinal Weld Seams 101-142A, 101-142B and 101-142C (Heat #

90077, Linde Flux Type 0091).

Per WCAP-9842, Revision 0 [Ref. D-3], the Callaway Unit 1 surveillance program was developed to the requirements of ASTM E185-73:

The base material and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime. The test material should be selected on the basis of initial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper (Cu) and phosphorus (P)) and neutron fluence.

At the time of the surveillance program development, copper and phosphorus content were considered the most limiting factors in embrittlement of reactor vessel steels. Since all plate materials had approximately the same copper and phosphorus content, initial RTNDT and USE were considered to select the representative plate. The lower shell plate R2708-1 had the highest initial RTNDT and the second lowest USE energy. Thus, the lower shell plate R2708-1 was selected as the plate material for the surveillance program. Note that lower shell plate R2708-1 has the same material heat number as the lower shell plate R2708-3; therefore, surveillance credibility conclusions and subsequent Position 2.1 chemistry factor determinations and Position 2.2 percent USE reduction calculations apply to this plate material as well.

The intermediate shell to lower shell circumferential weld seam, intermediate shell longitudinal weld seams and lower shell longitudinal weld seams all used the same weld material. This material was 3/16 inch Mil B-4 weld filler wire, heat number 90077. The surveillance weld material was also fabricated from this same heat of weld material. The intermediate to lower shell circumferential weld seam 101-171 was fabricated with Linde flux type 124, lot number 1061. The intermediate and lower shell longitudinal weld seams were fabricated with Linde flux type 0091, lot number 0842. The surveillance weld was fabricated with Linde flux type 124, lot number 1061. Thus, the surveillance material is representative of all the weld seams in the beltline region with respect to transition temperature shift, credibility conclusions and Position 2.1 chemistry factor determinations. However, it is only representative of the intermediate shell to lower shell circumferential weld with respect to Position 2.2 percent USE decrease evaluations because both the heat number and flux type (Linde 124) match for this weld material only.

Based on the discussion above, Criterion 1 is met for the Callaway Unit 1 surveillance program.

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Westinghouse Non-Proprietary Class 3 D-3 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, as documented in Appendix C of this report, is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Callaway Unit 1 surveillance materials unambiguously.

Hence, the Callaway Unit 1 surveillance program meets this criterion.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-4].

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 RTNDT values about this line is less than 28°F for welds and less than 17°F 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 [Ref. D-5]. 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 for the Callaway Unit 1 surveillance plate and weld material.

Case 1: Lower Shell Plate R2708-1 and Weld Material Heat # 90077 Following the NRC Case 1 guidelines, the Callaway Unit 1 surveillance plate and weld metal will be evaluated using the Callaway Unit 1 data. This evaluation is contained in Table D-1. Note that when evaluating the credibility of the surveillance weld data, the measured RTNDT values for the surveillance weld material do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld material measured shift values.

In addition, only Callaway Unit 1 data is being considered; therefore, no temperature adjustment is required.

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Westinghouse Non-Proprietary Class 3 D-4 Table D-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Callaway Unit 1 Surveillance Data Capsule Fluence(a) RTNDT FF*RTNDT Material Capsule FF(b) FF2 (x 1019 n/cm2, (°F) (°F)

E > 1.0 MeV)

U 0.313 0.681 -5.9(c) -4.02 0.464 Y 1.18 1.046 25.1 26.26 1.095 Lower Shell Plate R2708-1 V 2.32 1.227 16.3 20.01 1.506 (Longitudinal)

X 3.08 1.297 25.9 33.59 1.682 W 5.98 1.436 58.6 84.15 2.062 U 0.313 0.681 26.1 17.78 0.464 Y 1.18 1.046 46.7 48.86 1.095 Lower Shell Plate R2708-1 V 2.32 1.227 45.2 55.48 1.506 (Transverse)

X 3.08 1.297 30.8 39.95 1.682 W 5.98 1.436 95.2 136.7 2.062 SUM: 458.75 13.618 CF R2708-1 = (FF

• RTNDT) (FF2) = (458.75) (13.618) = 33.7F U 0.313 0.681 66.2 45.10 0.464 Y 1.18 1.046 35.0 36.62 1.095 Surveillance Weld Material V 2.32 1.227 46.2 56.70 1.506 (Heat #90077)

X 3.08 1.297 49.7 64.46 1.682 W 5.98 1.436 65.8 94.49 2.062 SUM: 297.37 6.809 CF Surv. Weld = (FF

• RTNDT) (FF2) = (297.37) (6.809) = 43.7F Notes:

(a) Updated in Capsule W dosimetry analysis; see Table 6-9.

(b) FF = fluence factor = f(0.28 - 0.10*log (f)).

(c) Even though a reduction should not occur, using the negative measured RTNDT value produces the most conservative results for this credibility evaluation (See Table D-2).

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Westinghouse Non-Proprietary Class 3 D-5 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2.

Table D-2 Callaway Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line CF Capsule Measured Predicted Scatter <17°F Material Capsule (Slopebest-fit) Fluence FF RTNDT RTNDT RTNDT(a) (Base Metal)

(°F) (x 1019 n/cm2) (°F) (°F) (°F) <28°F (Weld)

U 33.7 0.313 0.681 -5.9 23.0 28.9 No Y 33.7 1.18 1.046 25.1 35.3 10.2 Yes Lower Shell Plate R2708-1 V 33.7 2.32 1.227 16.3 41.4 25.1 No (Longitudinal)

X 33.7 3.08 1.297 25.9 43.7 17.8 No W 33.7 5.98 1.436 58.6 48.4 10.2 Yes U 33.7 0.313 0.681 26.1 23.0 3.1 Yes Y 33.7 1.18 1.046 46.7 35.3 11.4 Yes Lower Shell Plate R2708-1 V 33.7 2.32 1.227 45.2 41.4 3.8 Yes (Transverse)

X 33.7 3.08 1.297 30.8 43.7 12.9 Yes W 33.7 5.98 1.436 93.1 48.4 46.8 No U 43.7 0.313 0.681 66.2 29.8 36.4 No Y 43.7 1.18 1.046 35.0 45.7 10.7 Yes Surveillance Weld Material V 43.7 2.32 1.227 46.2 53.6 7.4 Yes (Heat # 90077)

X 43.7 3.08 1.297 49.7 56.7 7.0 Yes W 43.7 5.98 1.436 65.8 62.8 3.0 Yes Note:

(a) Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT].

From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table D-2 indicates that six of the ten surveillance data points fall inside the +/- 1 of 17F scatter band for surveillance base metals; therefore, the plate data is deemed non-credible per the third criterion.

Table D-2 indicates that four of the five surveillance data points fall inside the +/- 1 of 28F scatter band for surveillance weld materials; therefore, the surveillance weld data is deemed credible per the third criterion.

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Westinghouse Non-Proprietary Class 3 D-6 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The capsule specimens are located in the reactor between the neutron shielding pads and the vessel wall and are positioned opposite to the center of the core. The test capsules are located in guide tubes attached to the neutron shielding pads [Ref. D-3]. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F.

Hence, Criterion 4 is met for the Callaway Unit 1 surveillance program.

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 Callaway Unit 1 surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to Callaway Unit 1 surveillance program.

Hence, Criterion 5 is met for the Callaway Unit 1 surveillance program.

D.3 CONCLUSION Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B:

The Callaway Unit 1 surveillance plate data are deemed non-credible The Callaway Unit 1 surveillance weld data are deemed credible WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 D-7 D.4 REFERENCES D-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

D-2 10 CFR 50, Appendix G, Fracture Toughness Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.

D-3 Westinghouse Report WCAP-9842, Revision 0, Union Electric Company Callaway Unit No.1 Reactor Vessel Radiation Surveillance Program, May 1981.

D-4 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels E706 (IF), ASTM, 1982.

D-5 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Assessment Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E CALLAWAY UNIT 1 UPPER-SHELF ENERGY EVALUATION E.1 EVALUATION Per Regulatory Guide 1.99, Revision 2 [Ref. E-1], the Charpy upper-shelf energy (USE) is assumed to decrease as a function of fluence and copper content as indicated in Figure 2 of the Guide (Figure E-1 of this appendix) when surveillance data is not used. Linear interpolation is permitted. In addition, if surveillance data is to be used, the decrease in upper-shelf energy may be obtained by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure E-1 of this appendix) and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.

The 35 EFPY (end-of-license) and 54 EFPY (end-of-license renewal) upper-shelf energy of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the beltline materials and/or the results of the capsules tested to date using Figure 2 in Regulatory Guide 1.99, Revision 2. The maximum vessel clad/base metal interface fluence value was used to determine the corresponding 1/4T fluence value at 35 EFPY and 54 EFPY for the beltline materials. Likewise, the peak extended beltline fluence values at 35 and 54 EFPY was used for the extended beltline materials.

The Callaway Unit 1 reactor vessel beltline region minimum thickness is 8.63 inches. Calculation of the 1/4T vessel fluence values at 35 EFPY and 54 EFPY for the beltline and extended beltline materials are shown in Table E-1.

Table E-1 Callaway Unit 1 Pressure Vessel 1/4T Fast Neutron Fluence Calculation Fluence (n/cm2, E > 1.0 MeV)

Material 35 EFPY 54 EFPY Surface 1/4T Surface 1/4T Beltline 1.95E+19 1.16E+19 2.91E+19 1.73E+19 Extended 2.66E+17 1.58E+17 4.12E+17 2.45E+17 Beltline The following pages present the Callaway Unit 1 upper-shelf energy evaluation. Figure E-1, as indicated previously, is used in making predictions in accordance with Regulatory Guide 1.99, Revision 2.

Table E-2 provides the predicted EOL USE values for 35 EFPY.

Table E-3 provides the predicted EOLR USE values for 54 EFPY.

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Westinghouse Non-Proprietary Class 3 E-2 Figure E-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 E-3 Table E-2 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 35 EFPY 1/4T EOL Projected Projected Weight % Fluence(b) Unirradiated USE Material EOL USE of Cu (x 1019 n/cm2, USE (ft-lb) Decrease (ft-lb)

E > 1.0 MeV) (%)

Position 1.2(a)

Intermediate Shell Plate R2707-1 0.05 1.16 78 20 62 Intermediate Shell Plate R2707-2 0.06 1.16 100 20 80 Intermediate Shell Plate R2707-3 0.06 1.16 99 20 79 Lower Shell Plate R2708-1 0.07 1.16 82 20 66 Lower Shell Plate R2708-2 0.06 1.16 105 20 84 Lower Shell Plate R2708-3 0.08 1.16 101 20 81 Intermediate Shell Longitudinal Weld 0.04 1.16 143 20 114 Seams 101-124A, B &C (Heat # 90077)

Lower Longitudinal Weld 0.04 1.16 143 20 114 Seams 101-142A, B &C (Heat # 90077)

Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.16 112 20 90 101-171 (Heat # 90077)

Nozzle Shell Plate R2706-1(c) 0.045 (d) 102 7.5 94 Nozzle Shell Plate R2706-2(c) 0.055 (d) 87 7.5 80 Nozzle Shell Plate R2706-3(c) 0.075 (d) 101 7.5 93 Nozzle Shell to Intermediate Shell Weld 0.04 (d) 145 7.5 134 Seam 103-121 (Various/Multiple)(c)

Nozzle Shell Longitudinal Weld Seams 0.045 (d) 128 7.5 118 101-122A, B &C (Various/Multiple)(c)

Position 2.2(b)

Lower Shell Plate R2708-1 0.07 1.16 82 14 71 Lower Shell Plate R2708-3 0.08 1.16 101 14 87 Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.16 112 9 102 101-171 (Heat # 90077)

Notes:

(a) Calculated using the Cu wt. % values and 1/4T fluence value for each material and Regulatory Guide, Revision 2, Position 1.2. For the predicted USE decrease determinations, the base metal and weld Cu weight percentages were conservatively rounded up to the lowest line (Cu weight percent of 0.10 for base metal and 0.05 for weld) in Regulatory Guide 1.99, Revision 2, Figure 2.

(b) Calculated using surveillance capsule measured percent decrease in USE from Table 5-10 and Regulatory Guide 1.99, Revision 2, Position 2.2; see Figure E-1.

(c) Extended beltline materials with 35 EFPY projected fluence greater than 1 x 1017 n/cm2 (E > 1.0 MeV) were included.

(d) The minimum fluence value (2 x 1017 n/cm2) displayed on Figure 2 of Regulatory Guide 1.99, Revision 2 was conservatively used to determine the predicted USE decrease values for the nozzles shell plate and weld materials.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 E-4 Table E-3 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 54 EFPY 1/4T EOLR Projected Projected Weight % Fluence Unirradiated USE Material EOLR of Cu (x 1019 n/cm2, USE (ft-lb) Decrease USE (ft-lb)

E > 1.0 MeV) (%)

Position 1.2(a)

Intermediate Shell Plate R2707-1 0.05 1.73 78 22 61 Intermediate Shell Plate R2707-2 0.06 1.73 100 22 78 Intermediate Shell Plate R2707-3 0.06 1.73 99 22 77 Lower Shell Plate R2708-1 0.07 1.73 82 22 64 Lower Shell Plate R2708-2 0.06 1.73 105 22 82 Lower Shell Plate R2708-3 0.08 1.73 101 22 79 Intermediate Shell Longitudinal Weld 0.04 1.73 143 22 112 Seams 101-124A, B &C (Heat # 90077)

Lower Longitudinal Weld 0.04 1.73 143 22 112 Seams 101-142A, B &C (Heat # 90077)

Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.73 112 22 87 101-171 (Heat # 90077)

Nozzle Shell Plate R2706-1 0.045 0.0245 102 8 94 Nozzle Shell Plate R2706-2 0.055 0.0245 87 8 80 Nozzle Shell Plate R2706-3 0.075 0.0245 101 8 93 Inlet Nozzle R2702-1 0.16 (c) 135 0(c) 135 Inlet Nozzle R2702-2 0.16 (c) 137 0(c) 137 Inlet Nozzle R2702-3 0.16 (c) 137 0(c) 137 Inlet Nozzle R2702-4 0.16 (c) 134 0(c) 134 Outlet Nozzle R2703-1 0.16 (c) 90 0(c) 90 Outlet Nozzle R2703-2 0.16 (c) 114 0(c) 114 Outlet Nozzle R2703-3 0.16 (c) 113 0(c) 113 Outlet Nozzle R2703-4 0.16 (c) 118 0(c) 118 Nozzle Shell to Intermediate Shell 0.04 0.0245 145 8 133 Weld Seam 103-121 (Various/Multiple)

Inlet Nozzle to Shell Weld Seams 0.163 (c) 101 0(c) 101 105-121A, B, C &D (Various/Multiple)

Outlet Nozzle to Shell Weld Seams 0.163 (c) 99 0(c) 99 107-121A, B, C &D (Various/Multiple)

Nozzle Shell Longitudinal Weld Seams 0.045 0.0245 128 8 118 101-122A, B &C (Various/Multiple)

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 E-5 1/4T EOLR Projected Projected Weight % Fluence Unirradiated USE Material EOLR of Cu (x 1019 n/cm2, USE (ft-lb) Decrease USE (ft-lb)

E > 1.0 MeV) (%)

Position 2.2(b)

Lower Shell Plate R2708-1 0.07 1.73 82 16 69 Lower Shell Plate R2708-3 0.08 1.73 101 16 85 Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.73 112 10.2 101 101-171 (Heat # 90077)

Notes:

(a) Calculated using the Cu wt. % values and 1/4T fluence value for each material and Regulatory Guide, Revision 2, Position 1.2. For the predicted USE decrease determinations, the base metal and weld Cu weight percentages were conservatively rounded up to the nearest higher line in Regulatory Guide 1.99, Revision 2, Figure 2, unless otherwise noted.

(b) Calculated using surveillance capsule measured percent decrease in USE from Table 5-10 and Regulatory Guide 1.99, Revision 2, Position 2.2; see Figure E-1.

(c) Consistent with RIS 2014-1 [Ref. E-2], which established a formal NRC position on material inclusion and embrittlement requirements of extended beltline materials, the effects of neutron radiation must be considered for any locations that are predicted to experience a neutron fluence exposure greater than 1 x 1017 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period. Per Table 6-11 of this report, the inlet and outlet nozzle forging materials and the inlet and outlet nozzle to shell welds have a 54 EFPY fluence level below this threshold; therefore, percent USE reduction is not considered for these materials.

USE Conclusion As shown in Table E-2 and Table E-3, all of the Callaway Unit 1 reactor vessel beltline materials and extended beltline materials are projected to remain above the USE screening criterion of 50 ft-lbs (per 10 CFR 50, Appendix G) at 35 EPFY and 54 EFPY.

E.2 REFERENCES E-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

E-2 U. S. Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014.

WCAP-18001-NP September 2015 Revision 0

2/JIJlereII Cal laway Plant MISSOURI October 15, 2015 ULNRC-06259 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-000 1 10 CFR 50.61(b) 10 CFR 50 Appendix H Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

RENEWED FACILITY OPERATING LICENSE NPF-30 RESULTS OF ANALYSIS OF CAPSULE W FROM AMEREN MISSOURI (UNION ELECTRIC) CALLAWAY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Per 10 Cf R 50, Appendix H, this letter transmits the report, WCAP- 18001-NP, Revision 0, which provides the results of the testing of surveillance Capsule W from the Callaway Unit 1 reactor vessel.

Capsule W was removed during Refuel 20 (November 2014) at 25.75 Effective full-Power Years (EFPY). All surveillance capsules have now been removed from the Callaway Unit 1 reactor vessel, and ex-vessel dosimetry has been installed for future measurement of vessel fluence. Post irradiation mechanical testing of the Capsule W Charpy V-notch and tensile specimens was performed, and a fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDf/3-VT database.

A summary of results includes the calculated peak clad/base metal vessel fluence after 25.75 EFPY of plant operation was 1.49 x i nlcm2 (E> 1.0 MeV), and the fluence received by Capsule W was 5.98 x i nlcm2 (E> 1.0 MeV). The report also includes a brief summary of the Charpy V-notch testing results. Appendix E to the report presents an upper-shelf energy evaluation, using the screening criteria of 10 CFR 50 Appendix G, that documents a conclusion that with consideration of surveillance data, all beltline and extended beltline materials exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY).

P0 Box 620  : Fulton, MO 65251 : AmerenMissouri.com STARS Alliance

ULNRC- 06259 October 15, 2015 Page 2 With the receipt of WCAP- 18001-NP, Revision 0, the Callaway Pressure and Temperature Limits Report (PTLR) document will require revision. The PTLR revision is anticipated to be ready for transmittal to the NRC, in accordance with Technical Specification 5.6.6.c, during the first quarter of 2016.

This letter does not contain new commitments.

If there are any questions, please contact Jim Nurrenbem at 314-225-1908.

Sincerely, Scott A. Maglio Manager, Regulatory Affairs JPKI Attachment WCAP-18001-NP, Revision 0, Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program

ULNRC- 06259 October 15, 2015 Page 3 cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission

$201 NRC Road Steedman, MO 65077 Mr. L. John Klos Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 08H4 Washington, DC 20555-0001

ULNRC- 06259 October 15, 2015 Page 4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 6100 Western Place, Suite 1050 fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Other Situations ULNRC Distribution:

f. M. Diya D. W. Neterer L. H. Graessle T. E. Herrmann B. L. Cox M. A. McLachlan S. A. Maglio M. M. Hall T. B. Elwood M. G. Hoehn, III J. D. Nurrenbem Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. John ONeill (Pillsbury Winthrop Shaw Pittman LLP)

Missouri Public Service Commission

Attachment to ULNRC-06259 WCAP-1$OO 1-NP, Revision 0, Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program 310 Pages

Westinghouse Non-Proprietary Class 3 WCAP-18001-NP September 2015 Revision 0 Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program

Westinghouse Non-Proprietary Class 3 WCAP-18001-NP Revision 0 Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program Elaine M. Ruminski*

Materials Center of Excellence Arzu Alpan*

Nuclear Operations & Radiation Analysis September 2015 Reviewers: Elliot J. Long*

Materials Center of Excellence Eugene T. Hayes*

Nuclear Operations & Radiation Analysis Approved: David B. Love*, Acting Manager Materials Center of Excellence Laurent P. Houssay*, Manager Nuclear Operations & Radiation Analysis

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2015 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES .................................................................................................................................... vii EXECUTIVE

SUMMARY

........................................................................................................................... x 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 W ..................................................................... 5-1 5.1 OVERVIEW .................................................................................................................... 5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS........................................................... 5-2 5.3 TENSILE TEST RESULTS ............................................................................................. 5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ....................................................... 6-1

6.1 INTRODUCTION

........................................................................................................... 6-1 6.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 6-2 6.3 NEUTRON DOSIMETRY .............................................................................................. 6-4 6.4 CALCULATIONAL UNCERTAINTIES ........................................................................ 6-5 7 SURVEILLANCE CAPSULE REMOVAL

SUMMARY

............................................................ 7-1 8 REFERENCES ............................................................................................................................. 8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ............................................................................................. A-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS .................................... B-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD ........................................................ C-1 APPENDIX D CALLAWAY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION ....

........................................................................................................................................ D-1 APPENDIX E CALLAWAY UNIT 1 UPPER-SHELF ENERGY EVALUATION................................ E-1 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 4-1 Chemical Composition (wt. %) of the Callaway Unit 1 Reactor Vessel Surveillance Materials (Unirradiated)................................................................................................... 4-3 Table 4-2 Heat Treatment History of the Callaway Unit 1 Reactor Vessel Surveillance Materials . 4-4 Table 5-1 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation) ........................ 5-5 Table 5-2 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation) ........................... 5-6 Table 5-3 Charpy V-Notch Data for the Callaway Unit 1 Surveillance Program Weld Metal (Heat #

90077) Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)................................ 5-7 Table 5-4 Charpy V-Notch Data for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) ............................................ 5-8 Table 5-5 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

(Longitudinal Orientation) ............................................................................................... 5-9 Table 5-6 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

(Transverse Orientation) ................................................................................................ 5-10 Table 5-7 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Surveillance Program Weld Metal (Heat #90077) Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) .....

....................................................................................................................................... 5-11 Table 5-8 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)................ 5-12 Table 5-9 Effect of Irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Callaway Unit 1 Reactor Vessel Surveillance Capsule W Materials ........................................................................................................................ 5-13 Table 5-10 Comparison of the Callaway Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper-Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions ..................................................................................................................... 5-14 Table 5-11 Tensile Properties of the Callaway Unit 1 Capsule W Reactor Vessel Surveillance Materials Irradiated to 5.98 x 1019 n/cm2 (E > 1.0 MeV) .............................................. 5-15 Table 6-1 Calculated Fast Neutron Fluence Rate (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 ..................................................................... 6-7 Table 6-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections .................................... 6-8 Table 6-3 Calculated Iron Atom Displacement Rate at the Surveillance Capsule Center and at Core Midplane for Cycles 1 Through 20 .................................................................................. 6-9 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 v Table 6-4 Calculated Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections ........................................................... 6-10 Table 6-5 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence Rates (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface ........................................................... 6-11 Table 6-6 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface ..................................................................... 6-12 Table 6-7 Calculated Azimuthal Variation of Maximum Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface ..................................................................... 6-13 Table 6-8 Calculated Azimuthal Variation of Maximum Iron Atom Displacements at the Reactor Vessel Clad/Base Metal Interface .................................................................................. 6-14 Table 6-9 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Callaway Unit 1 ............................................................................................................................. 6-15 Table 6-10 Calculated Surveillance Capsule Lead Factors .............................................................. 6-15 Table 6-11 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface.............................................................................................. 6-16 Table 6-12 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface ......................................................................................................................... 6-19 Table 7-1 Surveillance Capsule Withdrawal Summary.................................................................... 7-1 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors .................................... A-9 Table A-2 Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor .......................................................................................................................... A-10 Table A-3 Surveillance Capsules U, Y, V, X, and W Fluence Rates for Cj Calculation, Core Midplane Elevation ....................................................................................................................... A-15 Table A-4 Surveillance Capsules U, Y, V, X, and W Cj Factors, Core Midplane Elevation .......... A-16 Table A-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U .............. A-17 Table A-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y............... A-18 Table A-7 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V............... A-19 Table A-8 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X .............. A-20 Table A-9 Measured Sensor Activities and Reaction Rates for Surveillance Capsule W .............. A-21 Table A-10 Least-Squares Evaluation of Dosimetry in Surveillance Capsule U (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycle 1 Irradiation ................................................. A-22 Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule Y (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 4 Irradiation .............................. A-23 Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 8 Irradiation .............................. A-24 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 vi Table A-13 Least-Squares Evaluation of Dosimetry in Surveillance Capsule X (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 10 Irradiation ............................ A-25 Table A-14 Least-Squares Evaluation of Dosimetry in Surveillance Capsule W (31.5° Azimuth, Core Midplane - Single Capsule Holder) Cycles 1 Through 20 Irradiation ......................... A-26 Table A-15 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions ..................................................................................................... A-27 Table A-16 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... A-27 Table C-1 Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH ................................................ C-2 Table C-2 Upper-Shelf L.E. Value (mils) Fixed in CVGRAPH ...................................................... C-2 Table D-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Callaway Unit 1 Surveillance Data ................................................................................................. D-4 Table D-2 Callaway Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line ................. D-5 Table E-1 Callaway Unit 1 Pressure Vessel 1/4T Fast Neutron Fluence Calculation ..................... E-1 Table E-2 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 35 EFPY ....................... E-3 Table E-3 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 54 EFPY ....................... E-4 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 vii LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Callaway Unit 1 Reactor Vessel .............. 4-5 Figure 4-2 Capsule W Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeter .................................................................................................................. 4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ................................................ 5-16 Figure 5-1(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued ............................ 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ................................................ 5-18 Figure 5-2(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued ............................ 5-19 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ................................................ 5-20 Figure 5-3(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued ............................ 5-21 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) .................................................... 5-22 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued ............................... 5-23 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) .................................................... 5-24 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued ............................... 5-25 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) .................................................... 5-26 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued ............................... 5-27 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ........................................................ 5-28 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued ................................... 5-29 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ............................................. 5-30 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued ........................ 5-31 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 viii Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ........................................................ 5-32 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued ................................... 5-33 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material ......................................................................................... 5-34 Figure 5-10(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued ..................................................................... 5-35 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material .............................................................................. 5-36 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued .......................................................... 5-37 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material ......................................................................................... 5-38 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued ..................................................................... 5-39 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) ............................................................ 5-40 Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) ............................................................... 5-41 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ........................................................ 5-42 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material .................................................................................................. 5-43 Figure 5-17 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV) ... 5-44 Figure 5-18 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV) ....... 5-45 Figure 5-19 Tensile Properties for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV) ........... 5-46 Figure 5-20 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) .............................................................................. 5-47 Figure 5-21 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) ................................................................................. 5-48 Figure 5-22 Fractured Tensile Specimens from the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) ............................................................................. 5-49 Figure 5-23 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Longitudinal Orientation) ........................................................................... 5-50 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 ix Figure 5-24 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Transverse Orientation) .............................................................................. 5-51 Figure 5-25 Engineering Stress-Strain Curves for Callaway Unit 1 Program Weld Metal (Heat #

90077) Tensile Specimens ............................................................................................. 5-52 Figure 6-1 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules and 12.5° Neutron Pad Configuration ....................................... 6-22 Figure 6-2 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Single Capsule Holder and 20.0° Neutron Pad Configuration .................................................. 6-23 Figure 6-3 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Dual Capsule Holder and 22.5° Neutron Pad Configuration .................................................. 6-24 Figure 6-4 Callaway Unit 1 r,,z Reactor Geometry Section View at 31.5° Azimuthal Angle with Surveillance Capsule...................................................................................................... 6-25 Figure E-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence ..................................................................................... E-2 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 x EXECUTIVE

SUMMARY

The purpose of this report is to document the testing results of surveillance Capsule W from Callaway Unit 1. Capsule W was removed at 25.75 Effective Full Power Years (EFPY) and post-irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI database. Capsule W received a fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) after irradiation to 25.75 EFPY. The peak clad/base metal interface vessel fluence after 25.75 EFPY of plant operation was 1.49 x 1019 n/cm2 (E > 1.0 MeV).

This evaluation led to the following conclusions: 1) The measured percent decreases in upper-shelf energy for the surveillance plate (longitudinal and transverse orientations) and weld materials contained in Callaway Unit 1 Capsule W are less than the Regulatory Guide 1.99, Revision 2 [Ref. 1] predictions.

2) The Callaway Unit 1 surveillance plate is judged to be non-credible and the surveillance weld (Heat #

90077) is judged to be credible. This credibility evaluation can be found in Appendix D. 3) With consideration of surveillance data, all beltline and extended beltline materials exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2]. The upper-shelf energy evaluation is presented in Appendix E.

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

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Westinghouse Non-Proprietary Class 3 1-1 1

SUMMARY

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

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

Appendix C presents the CVGRAPH, Version 6.0, Charpy V-notch plots for Capsule W and previous capsules, along with the program input data.

Capsule W received an average fast neutron fluence (E > 1.0 MeV) of 5.98 x 1019 n/cm2 after 25.75 effective full power years (EFPY) of plant operation.

Irradiation of the reactor vessel Lower Shell Plate R2708-1 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 55.9F and an irradiated 50 ft-lb transition temperature of 92.3F. This results in a 30 ft-lb transition temperature increase of 58.6F and a 50 ft-lb transition temperature increase of 69.1F for the longitudinally oriented specimens.

Irradiation of the reactor vessel Lower Shell Plate R2708-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 77.1F and an irradiated 50 ft-lb transition temperature of 139.3F. This results in a 30 ft-lb transition temperature increase of 95.2F and a 50 ft-lb transition temperature increase of 111.2F for the transversely oriented specimens.

Irradiation of the Surveillance Program Weld Metal (Heat # 90077) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 12.8F and an irradiated 50 ft-lb transition temperature of 42.0F. This results in a 30 ft-lb transition temperature increase of 65.8F and a 50 ft-lb transition temperature increase of 60.7F.

Irradiation of the Heat Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -8.6F and an irradiated 50 ft-lb transition temperature of 30.6F.

This results in a 30 ft-lb transition temperature increase of 91.6F and a 50 ft-lb transition temperature increase of 84.3F.

The average upper-shelf energy (USE) of Lower Shell Plate R2708-1 (longitudinal orientation) resulted in an average energy decrease of 9 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 117 ft-lb for the longitudinally oriented specimens.

The average upper-shelf energy of Lower Shell Plate R2708-1 (transverse orientation) resulted in an average energy decrease of 21 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 83 ft-lb for the transversely oriented specimens.

The average upper-shelf energy of the Surveillance Program Weld Metal (Heat # 90077) Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 103 ft-lb for the weld metal specimens.

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Westinghouse Non-Proprietary Class 3 1-2 The average upper-shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 5 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 101 ft-lb for the HAZ Material.

Comparisons of the measured 30 ft-lb shift in transition temperature values and upper-shelf energy decreases to those predicted by Regulatory Guide 1.99, Revision 2 [Ref. 1] for the Callaway Unit 1 reactor vessel surveillance materials are presented in Table 5-10.

Based on the credibility evaluation presented in Appendix D, the Callaway Unit 1 surveillance plate is non-credible and the surveillance weld material (Heat # 90077) is credible.

Based on the upper-shelf energy evaluation in Appendix E, all beltline and extended beltline materials contained in the Callaway Unit 1 reactor vessel exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2].

The maximum calculated 54 EFPY (end-of-license renewal) neutron fluence (E > 1.0 MeV) for the Callaway Unit 1 reactor vessel beltline using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the Guide) is as follows:

Calculated (54 EFPY): Vessel clad/base metal interface fluence* = 2.91 x 1019 n/cm2 Vessel 1/4 thickness fluence = 1.73 x 1019 n/cm2

• This fluence value is documented in Table 6-6.

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Westinghouse Non-Proprietary Class 3 2-1 2 INTRODUCTION This report presents the results of the examination of Capsule W, the fifth capsule removed and tested in the continuing surveillance program, which monitors the effects of neutron irradiation on the Ameren Missouri Callaway Unit 1 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Callaway Unit 1 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Company. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are contained in WCAP-9842 [Ref.

3], Union Electric Company Callaway Unit No. 1 Reactor Vessel Radiation Surveillance Program. The surveillance program was originally planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73 [Ref. 4], Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Capsule W was removed from the reactor after 25.75 EFPY of exposure and shipped to the Westinghouse Materials Center of Excellence 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 and post-irradiation data obtained from surveillance Capsule W removed from the Callaway Unit 1 reactor vessel and discusses the analysis of the data.

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Westinghouse Non-Proprietary Class 3 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 Callaway Unit 1 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 [Ref. 5]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop-weight nil-ductility transition temperature (NDTT per ASTM E208 [Ref. 6]) or the temperature 60F 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 Section XI of the ASME Code

[Ref. 5]. The KIc curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIc curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Callaway Unit 1 reactor vessel radiation surveillance program, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 30 ft-lb temperature (RTNDT) 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 (initial RTNDT + M + RTNDT) 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.

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Westinghouse Non-Proprietary Class 3 4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Callaway Unit 1 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel, as shown in Figure 4-1, between the neutron shielding pads and the vessel wall, at various azimuthal locations. The vertical center of the capsules is opposite the vertical center of the core. The capsules contain specimens made from the following:

Lower Shell Plate R2708-1 (longitudinal orientation)

Lower Shell Plate R2708-1 (transverse orientation)

Weld metal fabricated with weld wire Heat Number 90077, Linde Type 124 flux, Lot Number 1061, which is identical to that used in the actual fabrication of the intermediate to lower shell circumferential weld and is the equivalent heat number used in the actual fabrication of the intermediate shell and lower shell longitudinal weld seams; however, these vessel welds used Linde Type 0091 flux in their fabrication Weld heat affected zone (HAZ) material of Lower Shell Plate R2708-1 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 relieving treatment on the test material. Test specimens were also removed from weld metal of a stress relieved weldment joining Lower Shell Plate R2708-1 and adjacent Intermediate Shell Plate R2707-1.

All heat-affected zone specimens were obtained from the weld heat affected zone of Lower Shell Plate R2708-1.

Charpy V-notch impact specimens from Lower Shell Plate R2708-1 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major rolling direction).

The core-region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular (normal) 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 R2708-1 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented perpendicular to the welding direction.

Compact Tension (CT) specimens from Lower Shell Plate R2708-1 were machined in both the longitudinal and transverse orientations. CT specimens from the weld metal were machined with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399 [Ref. 7].

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Westinghouse Non-Proprietary Class 3 4-2 All six capsules contain 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 (237Np) and Uranium (238U) were placed in the capsules to measure the integrated flux at specific neutron energy levels.

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

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

The chemical composition and the heat treatment of the unirradiated surveillance materials in Capsule W are presented in Tables 4-1 and 4-2, respectively. The data in Tables 4-1 and 4-2 was obtained from the original surveillance program report, WCAP-9842 [Ref. 3].

Capsule W was removed after 25.75 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile and compact tension specimens, dosimeters, and thermal monitors.

Figures 4-1 and 4-2 detail the arrangement of the surveillance capsules in the reactor and the placement of specimens within Capsule W, respectively.

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Westinghouse Non-Proprietary Class 3 4-3 Table 4-1 Chemical Composition (wt. %) of the Callaway Unit 1 Reactor Vessel Surveillance Materials (Unirradiated)

Lower Shell Plate R2708-1(a) Weld Metal(a)

Element WCAP-9842 WCAP-11374 WCAP-9842 WCAP-11374 Analysis Analysis Analysis Analysis

[Ref. 3] [Ref. 14] [Ref. 3] [Ref. 14]

C 0.220 0.230 0.150 0.110 Mn 1.470 1.320 1.370 1.280 P 0.006 <0.005 0.005 0.005 S 0.014 0.016 0.008 0.016 Si 0.250 0.220 0.440 0.470 Ni 0.590 0.550 0.070 0.060 Mo 0.570 0.490 0.540 0.500 Cr 0.050 0.060 0.040 0.040 Cu 0.070 0.060 0.060 0.030 Al 0.025 -- 0.003 --

Co 0.013 0.006 0.011 0.004 Pb -- -- <0.001 --

W <0.010 -- <0.010 --

Ti <0.010 <0.001 <0.010 <0.001 Zr <0.001 -- <0.001 --

V 0.003 0.010 0.004 <0.010 Sn 0.002 -- 0.003 --

As 0.001 -- <0.001 --

Cb <0.010 -- <0.010 --

N 0.008 -- 0.007 --

Sb -- -- 0.0015 --

B <0.001 -- 0.001 --

Note:

(a) Data obtained from WCAP-15400, Revision 0 [Ref. 17].

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Westinghouse Non-Proprietary Class 3 4-4 Table 4-2 Heat Treatment History of the Callaway Unit 1 Reactor Vessel Surveillance Materials Material(a) Temperature(a) Time(a) Coolant(a)

Austenitized @ 1600 +/- 25 4 hrs. Water-Quenched (871°C)

Lower Shell Plate R2708-1 Tempered @ 1225 +/- 25 (663°C) 4 hrs. Air-Cooled Stress Relieved @ 1150 +/- 50 13 hrs. Furnace-Cooled (621°C)

Surveillance Program Weld Stress Relieved @ 1150 +/- 50 7 hrs. 45 min. Furnace-Cooled Metal (Heat # 90077) (621°C)

Note:

(a) Data obtained from WCAP-9842 [Ref. 3].

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Westinghouse Non-Proprietary Class 3 4-5 Figure 4-1 Arrangement of Surveillance Capsules in the Callaway Unit 1 Reactor Vessel WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 4-6 Figure 4-2 Capsule W Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeter WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-1 5 TESTING OF SPECIMENS FROM CAPSULE W 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed at the Westinghouse Materials Center of Excellence Hot Cell Facility. Testing was performed in accordance with 10 CFR 50, Appendix H [Ref. 2] and ASTM Specification E185-82 [Ref.

8].

Capsule W was opened upon receipt 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-9842 [Ref. 3]. All of the items were in their proper locations.

Examination of the thermal monitors indicated that 1 of the 3 temperature monitors had melted, as described below:

Top Thermal Monitor, the 579°F (304°C) was melted.

Middle Thermal Monitor, the 590°F (310°C) was not melted.

Bottom Thermal Monitor, the 579°F (304°C) was not melted.

Based on this examination, the maximum temperature to which the specimens were exposed was less than 590°F (310°C), but greater than 579°F (304°C).

The Charpy impact tests were performed per ASTM Specification E185-82 [Ref. 8] and E23-12c [Ref. 9]

on a Tinius-Olsen Model 74, 358J machine. The Charpy machine striker was instrumented with an Instron Impulse system. Instrumented testing and calibration were performed to ASTM E2298-13a [Ref.

10].

The instrumented striker load signal data acquisition rate was 819 kHz with data acquired for 10 ms.

From the load-time curve, the load of general yielding (Fgy), the maximum load (Fm) and the time to maximum load were 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 brittle fracture load (Fbf). The termination load after the fast load drop is identified as the arrest load (Fa). Fgy, Fm, Fbf, and Fa were determined per the guidance in ASTM Standard E2298-13a [Ref. 10].

The energy at maximum load (Wm) was determined by integrating the load-time record to the maximum load point. 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 (WP) is the difference between the total energy (Wt) and the energy at maximum load (Wm). Wt is compared to the dial energy (KV). Wt derived from the instrumented striker were all within 15% of the calibrated dial energy values as required in ASTM E2298-13a [Ref. 10].

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Westinghouse Non-Proprietary Class 3 5-2 Percent shear was determined from post-fracture photographs using the ratio-of-areas method in compliance with ASTM E23-12c [Ref. 9] and A370-13 [Ref. 11]. The lateral expansion was measured using a dial gage rig similar to that shown in the same specifications.

Tensile tests were performed on a 250 KN Instron screw driven tensile machine (Model 5985) per ASTM E185-82 [Ref. 8]. Testing met ASTM Specifications E8/E8M-13a [Ref. 12] or E21-09 [Ref. 13]. Load was applied through a clevis and pin connection. The strain rate obtained met the requirements of ASTM E8/E8M-13a [Ref. 12] and ASTM E21-09 [Ref. 13].

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 11-inch hot zone. Tensile specimens were soaked at temperature (+/-5ºF) for a minimum of 20 minutes before testing. All tests were conducted in air.

The tensile specimens were, nominally, 4.230 inches long with a 1.00 inch gage section and a reduced section of 1.50 inches long by 0.250 inch in diameter, as documented in Figure 2-2 of WCAP-9842 [Ref.

3]. The yield load, ultimate load, fracture load, uniform elongation and elongation at fracture were determined directly from the load-extension curve. The yield strength (0.2% offset method), ultimate tensile strength and fracture strength were calculated using the original cross-sectional area. Yield point elongation (YPE) was calculated as the difference in strain between the upper yield strength and the onset of uniform strain hardening using the methodology described in E8/E8M-13a [Ref. 12]. The final diameter and final gage length were determined from post-fracture photographs. This final diameter measurement was used to calculate the fracture stress (true stress at fracture) and the percent reduction in area. The final and original gage lengths were used to calculate total elongation after fracture.

5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule W, which received a fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) in 25.75 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with the unirradiated and previously withdrawn capsule results as shown in Figures 5-1 through 5-12. The unirradiated and previously withdrawn capsule results were taken from WCAP-9842 [Ref. 3], WCAP-11374 [Ref. 14], WCAP-12946 [Ref. 15], WCAP-14895 [Ref. 16] and WCAP-15400 [Ref. 17]. The previous capsules, along with the original program unirradiated material input data, were updated using CVGRAPH, Version 6.0. This accounts for the differences in measured values of 30 ft-lb and 50 ft-lb transition temperature between the results documented in this report and those shown in prior Callaway Unit 1 capsule reports.

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

Irradiation of the reactor vessel Lower Shell Plate R2708-1 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 55.9F and an irradiated 50 ft-lb transition temperature of 92.3F. This results in a 30 ft-lb transition temperature increase of 58.6F and a 50 ft-lb transition temperature increase of 69.1F for the longitudinally oriented specimens.

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Westinghouse Non-Proprietary Class 3 5-3 Irradiation of the reactor vessel Lower Shell Plate R2708-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 77.1F and an irradiated 50 ft-lb transition temperature of 139.3F. This results in a 30 ft-lb transition temperature increase of 95.2F and a 50 ft-lb transition temperature increase of 111.2F for the transversely oriented specimens.

Irradiation of the Surveillance Program Weld Metal (Heat # 90077) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 12.8F and an irradiated 50 ft-lb transition temperature of 42.0F. This results in a 30 ft-lb transition temperature increase of 65.8F and a 50 ft-lb transition temperature increase of 60.7F.

Irradiation of the Heat Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -8.6F and an irradiated 50 ft-lb transition temperature of 30.6F.

This results in a 30 ft-lb transition temperature increase of 91.6F and a 50 ft-lb transition temperature increase of 84.3F.

The irradiated upper-shelf energy of Lower Shell Plate R2708-1 (longitudinal orientation) resulted in an average energy decrease of 9 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 117 ft-lb for the longitudinally oriented specimens.

The average upper-shelf energy of Lower Shell Plate R2708-1 (transverse orientation) resulted in an average energy decrease of 21 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 83 ft-lb for the transversely oriented specimens.

The average upper-shelf energy of the Surveillance Program Weld Metal (Heat #90077) Charpy specimens resulted in an average energy decrease of 4 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 103 ft-lb for the weld metal specimens.

The average upper-shelf energy of the HAZ Material Charpy specimens resulted in an average energy decrease of 5 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 101 ft-lb for the HAZ Material.

Comparisons of the measured 30 ft-lb shift in transition temperature values and upper-shelf energy decreases to those predicted by Regulatory Guide 1.99, Revision 2 [Ref. 1] for the Callaway Unit 1 reactor vessel surveillance materials are presented in Table 5-10.

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Westinghouse Non-Proprietary Class 3 5-4 The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-13 through 5-16. The fractures show an increasingly ductile or tougher appearance with increasing test temperature. Load-time records for the individual instrumented Charpy specimens are contained in Appendix B.

With consideration of the surveillance data, all beltline materials exhibit adequate upper-shelf energy levels for continued safe plant operation and are predicted to maintain an upper-shelf energy greater than 50 ft-lb through end-of-license (35 EFPY) and end-of-license renewal (54 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2]. This evaluation can be found in Appendix E.

5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule W irradiated to 5.98 x 1019 n/cm2 (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-17 through 5-19.

The results of the tensile tests performed on the Lower Shell Plate R2708-1 (longitudinal orientation) indicated that irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) caused increases in the 0.2 percent offset yield strength and the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-17 and Table 5-11.

The results of the tensile tests performed on the Lower Shell Plate R2708-1 (transverse orientation) indicated that irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) caused increases in the 0.2 percent offset yield strength and the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-18 and Table 5-11.

The results of the tensile tests performed on the Surveillance Program Weld Metal (Heat # 90077) indicated that irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) caused increases in the 0.2 percent offset yield strength and the ultimate tensile strength when compared to unirradiated data [Ref. 3]. See Figure 5-19 and Table 5-11.

The fractured tensile specimens for the Lower Shell Plate R2708-1 (longitudinal orientation) material are shown in Figure 5-20; the fractured tensile specimens for the Lower Shell Plate R2708-1 (transverse orientation) material are shown in Figure 5-21; the fractured tensile specimens for the Surveillance Program Weld Metal (Heat # 90077) are shown in Figure 5-22. The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-25.

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Westinghouse Non-Proprietary Class 3 5-5 Table 5-1 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)

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

CL32 0 -18 7 9 6 0.2 5 CL38 20 -7 14 19 9 0.2 10 CL31 35 2 10 14 7 0.2 15 CL45 40 4 42 57 30 0.8 25 CL39 50 10 38 52 27 0.7 25 CL40 50 10 33 45 23 0.6 20 CL34 72 22 41 56 31 0.8 30 CL37 90 32 38 52 29 0.7 40 CL44 100 38 52 71 38 1.0 40 CL36 120 49 62 84 49 1.2 60 CL41 150 66 77 104 58 1.5 70 CL33 170 77 101 137 70 1.8 85 CL42 200 93 102 138 67 1.7 95 CL43 230 110 122 165 76 1.9 100 CL35 250 121 126 171 76 1.9 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-6 Table 5-2 Charpy V-Notch Data for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation)

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

CT31 0 -18 17 23 13 0.3 5 CT42 20 -7 11 15 6 0.2 5 CT36 40 4 20 27 15 0.4 10 CT40 50 10 21 28 14 0.4 10 CT43 60 16 32 43 22 0.6 20 CT44 60 16 28 38 18 0.5 15 CT38 72 22 32 43 22 0.6 25 CT32 100 38 40 54 29 0.7 25 CT34 120 49 38 52 30 0.8 25 CT35 150 66 46 62 34 0.9 45 CT39 170 77 50 68 40 1.0 55 CT37 200 93 67 91 46 1.2 75 CT45 230 110 74 100 55 1.4 100 CT41 250 121 92 125 68 1.7 100 CT33 275 135 82 111 62 1.6 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-7 Table 5-3 Charpy V-Notch Data for the Callaway Unit 1 Surveillance Program Weld Metal (Heat # 90077) Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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

CW45 -25 -32 5 7 2 0.1 5 CW36 0 -18 25 34 18 0.5 25 CW40 10 -12 15 20 14 0.4 30 CW32 20 -7 36 49 26 0.7 35 CW34 25 -4 55 75 43 1.1 55 CW42 35 2 26 35 22 0.6 45 CW33 40 4 60 81 46 1.2 60 CW44 50 10 69 94 52 1.3 65 CW43 60 16 51 69 43 1.1 60 CW38 72 22 77 104 56 1.4 75 CW37 120 49 89 121 76 1.9 90 CW31 150 66 87 118 76 1.9 90 CW35 170 77 99 134 73 1.9 95 CW39 200 93 96 130 77 2.0 100 CW41 250 121 115 156 81 2.1 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-8 Table 5-4 Charpy V-Notch Data for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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

CH41 -50 -46 14 19 9 0.2 10 CH44 -40 -40 17 23 11 0.3 15 CH42 -25 -32 33 45 24 0.6 30 CH37 0 -18 30 41 33 0.8 25 CH43 10 -12 50 68 34 0.9 50 CH34 20 -7 45 61 30 0.8 55 CH40 35 2 56 76 39 1.0 60 CH33 50 10 53 72 40 1.0 70 CH39 60 16 41 56 31 0.8 75 CH31 70 21 49 66 38 1.0 75 CH35 72 22 94 127 61 1.5 70 CH36 100 38 100 136 72 1.8 90 CH38 150 66 107 145 78 2.0 100 CH32 170 77 93 126 68 1.7 100 CH45 200 93 104 141 65 1.7 100 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)

Total Energy to General Total Dial Fracture Test Instrumented Difference, Max Maximum Time to Yield Arrest Sample Energy, Load, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fa Number KV Fbf

(°F) Wt (%) Wm (lb) (msec) Fgy (lb)

(ft-lb) (lb)

(ft-lb) (ft-lb) (lb)

CL32 0 7 6.04 14 3.62 4100 0.09 3500 3700 0 CL38 20 14 11.66* 17* 9.45* 3900* 0.20* 3400* 3800* 0*

CL31 35 10 8.09* 19* 3.41* 4000* 0.09* 3500* 3400* 0*

CL45 40 42 39.5 6 35.91 4400 0.6 3200 4200 0 CL39 50 38 34.97 8 27.30 4200 0.48 3100 4000 300 CL40 50 33 29.89 9 27.45 4200 0.48 3100 4200 0 CL34 72 41 37.13 9 35.29 4300 0.60 3000 4200 300 CL37 90 38 31.64* 17* 23.28* 4100* 0.43* 3000* 3900* 600*

CL44 100 52 45.70 12 34.43 4200 0.61 3000 4000 1000 CL36 120 62 56.08 10 34.12 4200 0.60 3000 4000 1200 CL41 150 77 72.07 6 33.25 4100 0.60 2900 3800 2300 CL33 170 101 96.47 4 33.16 4100 0.61 2900 3000 1900 CL42 200 102 99.18 3 33.02 4100 0.61 2800 3200 2900 CL43 230 122 118.63 3 32.75 4000 0.60 2700 0 0 CL35 250 126 121.98 3 39.34 4100 0.72 2700 0 0

• The difference between instrumented Charpy and dial values was greater than 15%. The values were not adjusted per Reference 10.

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Westinghouse Non-Proprietary Class 3 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Lower Shell Plate R2708-1 Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation)

Total Energy to General Total Dial Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Sample Energy, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa Number KV

(°F) Wt (%) Wm (lb) (msec) Fgy (lb) (lb)

(ft-lb)

(ft-lb) (ft-lb) (lb)

CT31 0 17 16.72 2 3.66 4200 0.09 3400 4100 0 CT42 20 11 10.40 5 3.47 4100 0.09 3500 3900 0 CT36 40 20 18.71 6 17.13 4100 0.31 3300 4100 0 CT40 50 21 19.90 5 16.73 4100 0.31 3200 4000 0 CT43 60 32 29.73 7 24.57 4200 0.43 3300 4100 0 CT44 60 28 25.02 11 19.64 4200 0.35 3300 4100 0 CT38 72 32 28.79 10 25.20 4200 0.43 3200 4200 300 CT32 100 40 35.79 11 23.33 4100 0.43 3100 3800 800 CT34 120 38 33.78 11 26.50 4000 0.48 3000 4000 1000 CT35 150 46 42.48 8 26.00 4000 0.48 2900 3800 1500 CT39 170 50 46.78 6 22.50 3900 0.43 2900 3900 2300 CT37 200 67 64.51 4 22.55 4000 0.43 3100 3600 2300 CT45 230 74 71.86 3 2.23 4100 0.09 2600 0 0 CT41 250 92 89.35 3 28.67 4300 0.56 2800 0 0 CT33 275 82 80.20 2 31.79 3900 0.60 2500 0 0 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Surveillance Program Weld Metal (Heat #90077)

Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

Total Energy to General Total Dial Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Sample Energy, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa Number KV

(°F) Wt (%) Wm (lb) (msec) Fgy (lb) (lb)

(ft-lb)

(ft-lb) (ft-lb) (lb)

CW45 -25 5 3.40* 32* 2.25* 3900* 0.08* 3500* 3600* 0*

CW36 0 25 23.11 8 3.56 4100 0.09 3600 3100 600 CW40 10 15 10.97* 27* 3.66* 4100* 0.09* 3300* 3700* 400*

CW32 20 36 32.30 10 3.26 4000 0.09 3000 3800 400 CW34 25 55 52.28 5 33.92 4100 0.60 3100 3800 1800 CW42 35 26 24.25 7 12.88 3800 0.26 3200 3600 1500 CW33 40 60 56.77 5 33.68 4100 0.60 3100 3900 2600 CW44 50 69 66.04 4 33.55 4100 0.60 2900 3700 2000 CW43 60 51 49.13 4 33.23 4000 0.60 2900 3800 1700 CW38 72 77 73.97 4 35.29 4200 0.60 3100 4000 2500 CW37 120 89 87.14 2 32.60 3900 0.61 2800 2000 1500 CW31 150 87 84.34 3 31.49 3800 0.60 2800 2700 2200 CW35 170 99 95.12 4 31.18 3800 0.60 2700 2100 1800 CW39 200 96 93.08 3 31.24 3800 0.60 2700 0 0 CW41 250 115 110.20 4 42.51 3700 0.83 2300 0 0

• The difference between instrumented Charpy and dial values was greater than 25%. The values were not discarded as required by Reference 10.

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Westinghouse Non-Proprietary Class 3 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Callaway Unit 1 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 5.98 x 1019 n/cm2 (E > 1.0 MeV)

Total Energy to General Total Dial Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Sample Energy, Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa Number KV

(°F) Wt (%) Wm (lb) (msec) Fgy (lb) (lb)

(ft-lb)

(ft-lb) (ft-lb) (lb)

CH41 -50 14 12.48 11 3.70 4500 0.09 3600 3900 0 CH44 -40 17 15.84 7 3.77 4500 0.09 3500 4100 0 CH42 -25 33 30.46 8 3.96 4500 0.09 3500 4300 0 CH37 0 30 27.84 7 3.52 4100 0.09 3200 3900 900 CH43 10 50 45.98 8 35.62 4300 0.60 3100 4200 1600 CH34 20 45 41.11 9 3.38 4200 0.09 3200 3800 1500 CH40 35 56 52.88 6 35.79 4100 0.61 3400 3900 1100 CH33 50 53 50.56 5 34.93 4100 0.61 3200 4100 2200 CH39 60 41 39.69 3 3.52 4000 0.09 3200 3600 2500 CH31 70 49 47.13 4 14.67 3900 0.29 3000 3200 2200 CH35 72 94 87.72 7 36.59 4400 0.60 3400 3700 1600 CH36 100 100 97.91 2 34.32 4200 0.61 3000 2000 1700 CH38 150 107 102.46 4 32.29 4000 0.61 2700 0 0 CH32 170 93 90.96 2 31.60 3800 0.60 2600 0 0 CH45 200 104 101.06 3 32.27 4000 0.60 2900 0 0 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-13 Table 5-9 Effect of Irradiation to 5.98 x 1019 n/cm2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Callaway Unit 1 Reactor Vessel Surveillance Capsule W Materials Average 30 ft-lb Transition Average 35 mil Lateral Expansion Average 50 ft-lb Transition Average Energy Absorption at Material Temperature(a) (°F) Temperature(a) (°F) Temperature(a) (°F) Full Shear(b) (ft-lb)

Unirradiated Irradiated T Unirradiated Irradiated T Unirradiated Irradiated T Unirradiated Irradiated E Lower Shell Plate R2708-1 -2.7 55.9 58.6 26.7 88.0 61.3 23.2 92.3 69.1 126 117 -9 (Longitudinal)

Lower Shell Plate R2708-1 -18.1 77.1 95.2 28.9 132.5(c) 103.6 28.1 139.3 111.2 104 83 -21 (Transverse)

Surveillance Weld Material -53.0 12.8 65.8 -21.7 34.7 56.4 -18.7 42.0 60.7 107 103 -4 (Heat #90077)

Heat Affected

-100.2 -8.6 91.6 -42.1 27.1 69.2 -53.7 30.6 84.3 106 101 -5 Zone Material Notes:

(a) Average value is determined by CVGRAPH (see Appendix A).

(b) Irradiated value is a calculated average of Capsule W test data at greater than or equal to 95% shear.

(c) The upper shelf lateral expansion for Capsule W was fixed, based on the average of three (3) data points at 100% shear.

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Westinghouse Non-Proprietary Class 3 5-14 Table 5-10 Comparison of the Callaway Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper-Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Capsule 30 ft-lb Transition Upper-Shelf Energy Fluence Temperature Shift Decrease Material Capsule (x1019 n/cm2, E Predicted(a) Measured(b) Predicted(a) Measured(b)

> 1.0 MeV) (°F) (°F) (%) (%)

U 0.313 30.0 0(c) 14.5 2 Lower Shell Plate Y 1.18 46.0 25.1 20.0 6 R2708-1 (Longitudinal) V 2.32 54.0 16.3 23.0 0 X 3.08 57.1 25.9 25.0 5 W 5.98 63.2 58.6 29.0 7 U 0.313 30.0 26.1 14.5 11 Y 1.18 46.0 46.7 20.0 13 Lower Shell Plate V 2.32 54.0 45.2 23.0 3 R2708-1 (Transverse)

X 3.08 57.1 30.8 25.0 5 W 5.98 63.2 95.2 29.0 20 U 0.313 21.7 66.2 14.5 6 Y 1.18 33.3 35.0 20.0 9 Surveillance Weld V 2.32 39.0 46.2 23.0 3 Material (Heat # 90077)

X 3.08 41.2 49.7 25.0 3 W 5.98 45.7 65.8 29.0 4 U 0.313 --- 66.3 --- 0(d)

Y 1.18 --- 56.5 --- 14 Heat Affected Zone V 2.32 --- 56.8 --- 0 Material X 3.08 --- 42.3 --- 0(d)

W 5.98 --- 91.6 --- 5 Notes:

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

(b) Calculated by CVGRAPH, Version 6.0 using measured Charpy data (See Appendix C).

(c) This RTNDT was calculated to be negative (-5.9°F). Physically, this should not occur; therefore, a conservative value of zero degrees F is shown in this table.

(d) USE values were calculated to have increased. Physically, this should not occur; therefore, conservative values of zero percent are shown in this table.

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Westinghouse Non-Proprietary Class 3 5-15 Table 5-11 Tensile Properties of the Callaway Unit 1 Capsule W Reactor Vessel Surveillance Materials Irradiated to 5.98 x 1019 n/cm2 (E > 1.0 MeV) 0.2% Fracture Test Ultimate Fracture Fracture Uniform Total Reduction Sample Yield True Material Temp. Strength Load Strength Elongation Elongation in Area Number Strength Stress

(°F) (ksi) (kip) (ksi) (%) (%) (%)

(ksi) (ksi)

CL7 72 81.2 101.6 3.4 68.8 182.5 9.1 22.3 62.1 Lower Shell Plate R2708-1 CL8 175 77.8 96.3 3.2 64.3 186.0 9.5 22.6 65.0 (Longitudinal)

CL9 550 73.0 94.8 3.5 70.4 171.9 8.2 18.3 58.5 CT7 72 80.8 101.8 3.5 71.7 165.7 8.8 21.1 57.0 Lower Shell Plate R2708-1 CT8 175 77.1 96.4 2.8 57.9 134.2 9.1 20.7 57.5 (Transverse)

CT9 550 72.3 94.2 3.5 70.7 145.5 7.1 15.6 51.0 CW7 72 79.1 94.8 3.0 60.9 179.2 9.1 25.9 65.9 Surveillance Weld Material CW8 175 76.6 90.8 2.8 57.8 181.9 10.0 25.5 68.6 (Heat # 90077)

CW9 550 74.1 91.8 3.2 64.6 178.7 6.8 19.2 63.5 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-16 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-17 Figure 5-1(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-18 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-19 Figure 5-2(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-20 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-21 Figure 5-3(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-22 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-23 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-24 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-25 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-26 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-27 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-28 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-29 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-30 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-31 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-32 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-33 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077) - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-34 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-35 Figure 5-10(a) Charpy V-Notch Impact Energy vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-36 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-37 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-38 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-39 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material - Continued WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-40 CL32, 0°F CL38, 20°F CL31, 35°F CL45, 40°F CL39, 50°F CL40, 50°F CL34, 72°F CL37, 90°F CL44, 100°F CL36, 120°F CL41, 150°F CL33, 170°F CL42, 200°F CL43, 230°F CL35, 250°F Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-41 CT31, 0°F CT42, 20°F CT36, 40°F CT40, 50°F CT43, 60°F CT44, 60°F CT38, 72°F CT32, 100°F CT34, 120°F CT35, 150°F CT39, 170°F CT37, 200°F CT45, 230°F CT41, 250°F CT33, 275°F Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-42 CW45, -25°F CW36, 0°F CW40, 10°F CW32, 20°F CW34, 25°F CW42, 35°F CW33, 40°F CW44, 50°F CW43, 60°F CW38, 72°F CW37, 120°F CW31, 150°F CW35, 170°F CW39, 200°F CW41, 250°F Figure 5-15 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-43 CH41, -50°F CH44, -40°F CH42, -25°F CH37, 0°F CH43, 10°F CH34, 20°F CH40, 35°F CH33, 50°F CH39, 60°F CH31, 70°F CH35, 72°F CH36, 100°F CH38, 150°F CH32, 170°F CH45, 200°F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for the Callaway Unit 1 Reactor Vessel Heat Affected Zone Material WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 5-44 Figure 5-17 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 5-45 Figure 5-18 Tensile Properties for Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 5-46 Figure 5-19 Tensile Properties for the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077), Capsule W Fluence = 5.98 x 1019 n/cm2 (E > 1.0 MeV)

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Westinghouse Non-Proprietary Class 3 5-47 CL7 - Tested at 72°F CL8 - Tested at 175°F CL9 - Tested at 550°F Figure 5-20 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-48 CT7 - Tested at 72°F CT8 - Tested at 175°F CT9 - Tested at 550°F Figure 5-21 Fractured Tensile Specimens from Callaway Unit 1 Reactor Vessel Lower Shell Plate R2708-1 (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-49 CW7 - Tested at 72°F CW8 - Tested at 175°F CW9 - Tested at 550°F Figure 5-22 Fractured Tensile Specimens from the Callaway Unit 1 Reactor Vessel Surveillance Program Weld Metal (Heat # 90077)

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Westinghouse Non-Proprietary Class 3 5-50 Tensile Specimen CL7, Tested at 72°F Tensile Specimen CL8, Tested at 175°F Tensile Specimen CL9, Tested at 550°F Figure 5-23 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Longitudinal Orientation)

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Westinghouse Non-Proprietary Class 3 5-51 Tensile Specimen CT7, Tested at 72°F Tensile Specimen CT8, Tested at 175°F Tensile Specimen CT9, Tested at 550°F Figure 5-24 Engineering Stress-Strain Curves for Callaway Unit 1 Lower Shell Plate R2708-1 Tensile Specimens (Transverse Orientation)

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Westinghouse Non-Proprietary Class 3 5-52 Tensile Specimen CW7, Tested at 72°F Tensile Specimen CW8, Tested at 175°F Tensile Specimen CW9, Tested at 550°F Figure 5-25 Engineering Stress-Strain Curves for Callaway Unit 1 Program Weld Metal (Heat #

90077) Tensile Specimens WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates (Sn) transport analysis performed for the Callaway Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant- and fuel-cycle-specific basis. An evaluation of the most recent dosimetry sensor set from Capsule W, withdrawn at the end of the 20th plant operating cycle, is provided. In addition, the sensor sets from the previously withdrawn and analyzed capsules (U, Y, V, and X) were re-analyzed and are presented. 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 form the basis for projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY) at 3565 MWt.

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.

However, in recent years, 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 E853-13, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, [Ref. 18] 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 E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom [Ref. 19]. 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 [Ref. 1].

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on nuclear cross-section data derived from ENDF/B-VI and used the latest available calculational tools.

Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. 20]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040, Revision 4, [Ref. 21]. As an improvement, instead of the fluence rate synthesis technique, three-dimensional transport calculations were performed.

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Westinghouse Non-Proprietary Class 3 6-2 6.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Callaway Unit 1 reactor vessel is shown in Figure 4-

1. Six irradiation capsules attached to the neutron pad are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.5°, 61.0°,

121.5°, 238.5°, 241.0°, and 301.5°, as shown in Figure 4-1. These full-core positions correspond to the octant symmetric locations shown in Figures 6-2 and 6-3: 29.0° from the core cardinal axes (for the 61.0° and 241.0° capsules) and 31.5° from the core cardinal axes (for the 58.5°, 121.5°, 238.5°, and 301.5° capsules). The stainless steel specimen containers are 1.6-inch by 1.25-inch and are approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the approximate 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 significant effect on both the spatial distribution of neutron fluence rate and the neutron spectrum in the vicinity of the capsules. However, the capsules are far enough apart that they do not interfere with one another. 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 Callaway Unit 1 reactor vessel and surveillance capsules, a series of fuel-cycle-specific forward transport calculations were carried out using a three-dimensional geometrical reactor model. For the Callaway Unit 1 transport calculations, the r,,z models depicted (given as r, plan view) in Figures 6-1 (12.5° neutron pad configuration), 6-2 (20.0° neutron pad configuration), and 6-3 (22.5° neutron pad configuration) were utilized since, with the exception of the neutron pads, capsules, and associated support structures, the reactor is octant symmetric.

The r,z section view depicted in Figure 6-4 shows the model having an axial span from an elevation approximately 6 feet below the bottom of the active fuel to 5 feet above the top of the active fuel. These r,,z models include the core, the reactor internals, the neutron pads including explicit representations of octants not containing surveillance capsules and octants with surveillance capsules at 29° and/or 31.5°,

the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing these analytical models, nominal design dimensions were employed for the various structural components with one exception; the minimum pressure vessel thickness was used. Water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,,z reactor models for the 20.0° and 22.5° neutron pad configurations consisted of 183 radial, 110 azimuthal, and 206 axial intervals. The geometric mesh description of the r,,z reactor model for the 12.5° neutron pad configuration consisted of 183 radial, 110 azimuthal, and 204 axial intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,,z calculations was set at a value of 0.001.

The core power distributions used in the plant-specific transport analysis for each of the first 20 fuel cycles at Callaway Unit 1 included cycle-dependent fuel assembly initial enrichments, burnups, and axial WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-3 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 fluence rate, 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 RAPTOR-M3G discrete ordinates code [Ref. 22] and the BUGLE-96 cross-section library [Ref. 23]. The BUGLE-96 library provides a coupled 47-neutron, 20-gamma-group cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion, and angular discretization was modeled with an S16 order of angular quadrature.

Energy- and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-12. In Tables 6-1 and 6-2, the calculated exposure rates and integral exposures expressed in terms of fast neutron fluence rate (E > 1.0 MeV) and fast neutron fluence (E > 1.0 MeV), respectively, are given at the radial and azimuthal center of the surveillance capsule positions, i.e., for the 29.0° and 31.5° dual capsule holder locations and 31.5° single capsule holder location. In Tables 6-3 and 6-4, the calculated exposure rates and integral exposures expressed in terms of iron atom displacement rate (dpa/s) and iron atom displacements (dpa), respectively, are given at the radial and azimuthal center of the surveillance capsule positions, i.e., for the 29.0° and 31.5° dual capsule holder locations and 31.5° single capsule holder location. In Tables 6-2 and 6-4, the calculated integral exposures expressed for future projections, in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements, respectively, are given at the radial and azimuthal center of the surveillance capsule positions. These results, representative of the average axial exposure of the material specimens, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projections into the future.

Similar information, in terms of calculated fast neutron fluence rate (E > 1.0 MeV), fast neutron fluence (E > 1.0 MeV), dpa/s, and dpa, are provided in Tables 6-5 through 6-8, for the reactor vessel inner radius at four azimuthal locations, as well as the maximum exposure observed within the octant. The vessel data given in Tables 6-5 through 6-8 were taken at the clad/base metal interface and represent maximum calculated exposure levels on the vessel. From the data provided in Table 6-6, it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the 20th fuel cycle (i.e., after 25.75 EFPY of plant operation) was 1.49E+19 n/cm2.

These data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end of the 20th fuel cycle, as well as future projections to 32, 35, 40, 48, 54, and 60 EFPY. The calculations account for the uprate from 3411 MWt to 3565 MWt that occurred prior to Cycle 3. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 21 is representative of future plant operation. The future projections are based on the current reactor power level of 3565 MWt.

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Westinghouse Non-Proprietary Class 3 6-4 The calculated fast neutron exposures for the six surveillance capsules withdrawn from the Callaway Unit 1 reactor are provided in Table 6-9. These neutron exposure levels are based on the plant- and fuel-cycle-specific neutron transport calculations performed for the Callaway Unit 1 reactor. From the data provided in Table 6-9, Capsule W received a fast neutron fluence (E > 1.0 MeV) of 5.98E+19 n/cm2 after exposure through the end of the 20th fuel cycle (i.e., after 25.75 EFPY).

Updated lead factors for the Callaway Unit 1 surveillance capsules are provided in Table 6-10. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric radial and azimuthal center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel clad/base metal interface. In Table 6-10, the lead factors for capsules that have been withdrawn from the reactor (Capsules U, Y, V, X, Z, and W) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules.

Table 6-11 presents the maximum fast neutron fluences (E > 1.0 MeV) and Table 6-12 presents the maximum dpa for pressure vessel materials.

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 serve to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least-squares evaluation comparisons is documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule W, which was withdrawn from Callaway Unit 1 at the end of the 20th fuel cycle, is summarized below.

Reaction Rate (rps/atom)

Reaction M/C Measured (M) Calculated (C)

Cu-63(n,)Co-60 4.52E-17 3.94E-17 1.15 Fe-54(n,p)Mn-54 4.65E-15 4.32E-15 1.08 Ni-58(n,p)Co-58 6.75E-15 6.06E-15 1.11 U-238(Cd)(n,f)Cs-137 2.82E-14 2.32E-14 1.21 Np-237(Cd)(n,f)Cs-137 2.41E-13 2.30E-13 1.05 Average 1.12

% standard deviation 5.6 The measured-to-calculated (M/C) reaction rate ratios for the Capsule W threshold reactions range from 1.05 to 1.21, and the average M/C ratio is 1.12 5.6% (1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Callaway Unit 1.

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Westinghouse Non-Proprietary Class 3 6-5 6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Callaway Unit 1 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

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

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

4. Comparisons of the plant-specific calculations with all available dosimetry results from the Callaway Unit 1 surveillance program.

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

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

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

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

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Westinghouse Non-Proprietary Class 3 6-6 Description Capsule and Vessel IR PCA Comparisons 3%

H. B. Robinson Comparisons 3%

Analytical Sensitivity Studies 11%

Additional Uncertainty for Factors not Explicitly 5%

Net Calculational Uncertainty 13%

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

Therefore, the resultant uncertainty was treated as random, and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons described in Appendix A support these uncertainty assessments for Callaway Unit 1.

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Westinghouse Non-Proprietary Class 3 6-7 Table 6-1 Calculated Fast Neutron Fluence Rate (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 Fluence Rate (n/cm2-s)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 8.79E+10 9.42E+10 9.34E+10 2 1.15 2.21 7.76E+10 8.09E+10 8.01E+10 3 1.23 3.43 8.26E+10 9.51E+10 9.45E+10 4 1.26 4.70 7.16E+10 7.92E+10 7.86E+10 5 1.27 5.97 7.10E+10 7.71E+10 7.65E+10 6 1.32 7.29 7.13E+10 7.57E+10 7.51E+10 7 1.30 8.59 6.66E+10 7.11E+10 7.05E+10 8 1.35 9.94 6.58E+10 7.12E+10 7.06E+10 9 1.19 11.13 6.37E+10 6.80E+10 6.74E+10 10 1.36 12.49 6.79E+10 7.21E+10 7.15E+10 11 1.38 13.87 6.42E+10 6.99E+10 6.93E+10 12 1.35 15.22 6.57E+10 7.15E+10 7.09E+10 13 1.26 16.48 7.80E+10 8.73E+10 8.66E+10 14 1.22 17.70 6.76E+10 7.35E+10 7.29E+10 15 1.29 18.99 6.81E+10 7.49E+10 7.43E+10 16 1.42 20.41 6.44E+10 6.80E+10 6.73E+10 17 1.35 21.76 6.65E+10 7.18E+10 7.12E+10 18 1.33 23.09 6.10E+10 6.61E+10 6.56E+10 19 1.36 24.45 6.09E+10 6.44E+10 6.38E+10 20 1.30 25.75 5.72E+10 6.03E+10 5.98E+10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-8 Table 6-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections Fluence (n/cm2)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 2.92E+18 3.13E+18 3.10E+18 2 1.15 2.21 5.75E+18 6.07E+18 6.02E+18 3 1.23 3.43 8.95E+18 9.76E+18 9.68E+18 4 1.26 4.70 1.18E+19 1.29E+19 1.28E+19 5 1.27 5.97 1.46E+19 1.60E+19 1.59E+19 6 1.32 7.29 1.76E+19 1.92E+19 1.90E+19 7 1.30 8.59 2.03E+19 2.21E+19 2.19E+19 8 1.35 9.94 2.32E+19 2.51E+19 2.49E+19 9 1.19 11.13 2.56E+19 2.77E+19 2.75E+19 10 1.36 12.49 2.85E+19 3.08E+19 3.05E+19 11 1.38 13.87 3.13E+19 3.38E+19 3.35E+19 12 1.35 15.22 3.41E+19 3.69E+19 3.66E+19 13 1.26 16.48 3.72E+19 4.03E+19 4.00E+19 14 1.22 17.70 3.98E+19 4.32E+19 4.28E+19 15 1.29 18.99 4.25E+19 4.62E+19 4.58E+19 16 1.42 20.41 4.54E+19 4.92E+19 4.88E+19 17 1.35 21.76 4.82E+19 5.23E+19 5.19E+19 18 1.33 23.09 5.08E+19 5.51E+19 5.46E+19 19 1.36 24.45 5.34E+19 5.79E+19 5.74E+19 20 1.30 25.75 5.58E+19 6.03E+19 5.98E+19 Future 32.00 6.78E+19 7.33E+19 7.27E+19 Future 35.00 7.36E+19 7.95E+19 7.89E+19 Future 40.00 8.32E+19 8.99E+19 8.92E+19 Future 48.00 9.87E+19 1.07E+20 1.06E+20 Future 54.00 1.10E+20 1.19E+20 1.18E+20 Future 60.00 1.22E+20 1.31E+20 1.30E+20 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-9 Table 6-3 Calculated Iron Atom Displacement Rate at the Surveillance Capsule Center and at Core Midplane for Cycles 1 Through 20 Iron Atom Displacement Rate (dpa/s)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 1.75E-10 1.88E-10 1.86E-10 2 1.15 2.21 1.54E-10 1.60E-10 1.59E-10 3 1.23 3.43 1.65E-10 1.90E-10 1.89E-10 4 1.26 4.70 1.42E-10 1.57E-10 1.56E-10 5 1.27 5.97 1.41E-10 1.53E-10 1.51E-10 6 1.32 7.29 1.41E-10 1.50E-10 1.48E-10 7 1.30 8.59 1.32E-10 1.41E-10 1.39E-10 8 1.35 9.94 1.30E-10 1.41E-10 1.40E-10 9 1.19 11.13 1.26E-10 1.34E-10 1.33E-10 10 1.36 12.49 1.34E-10 1.43E-10 1.41E-10 11 1.38 13.87 1.27E-10 1.38E-10 1.37E-10 12 1.35 15.22 1.30E-10 1.42E-10 1.40E-10 13 1.26 16.48 1.54E-10 1.73E-10 1.71E-10 14 1.22 17.70 1.34E-10 1.45E-10 1.44E-10 15 1.29 18.99 1.35E-10 1.48E-10 1.47E-10 16 1.42 20.41 1.27E-10 1.34E-10 1.33E-10 17 1.35 21.76 1.31E-10 1.42E-10 1.40E-10 18 1.33 23.09 1.21E-10 1.31E-10 1.29E-10 19 1.36 24.45 1.20E-10 1.27E-10 1.26E-10 20 1.30 25.75 1.13E-10 1.19E-10 1.18E-10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-10 Table 6-4 Calculated Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 20 and Future Projections Iron Atom Displacements (dpa)

Cycle Total Single Cycle Length Time Dual Capsule Holder Capsule (EFPY) (EFPY) Holder 29.0° 31.5° 31.5° 1 1.05 1.05 5.82E-03 6.24E-03 6.18E-03 2 1.15 2.21 1.14E-02 1.21E-02 1.19E-02 3 1.23 3.43 1.78E-02 1.94E-02 1.93E-02 4 1.26 4.70 2.35E-02 2.57E-02 2.55E-02 5 1.27 5.97 2.91E-02 3.18E-02 3.15E-02 6 1.32 7.29 3.50E-02 3.80E-02 3.77E-02 7 1.30 8.59 4.04E-02 4.38E-02 4.34E-02 8 1.35 9.94 4.59E-02 4.98E-02 4.94E-02 9 1.19 11.13 5.07E-02 5.49E-02 5.44E-02 10 1.36 12.49 5.64E-02 6.10E-02 6.04E-02 11 1.38 13.87 6.20E-02 6.70E-02 6.64E-02 12 1.35 15.22 6.75E-02 7.30E-02 7.24E-02 13 1.26 16.48 7.37E-02 7.99E-02 7.92E-02 14 1.22 17.70 7.88E-02 8.55E-02 8.47E-02 15 1.29 18.99 8.43E-02 9.15E-02 9.07E-02 16 1.42 20.41 9.00E-02 9.75E-02 9.67E-02 17 1.35 21.76 9.56E-02 1.04E-01 1.03E-01 18 1.33 23.09 1.01E-01 1.09E-01 1.08E-01 19 1.36 24.45 1.06E-01 1.15E-01 1.14E-01 20 1.30 25.75 1.10E-01 1.19E-01 1.18E-01 Future 32.00 1.34E-01 1.45E-01 1.44E-01 Future 35.00 1.46E-01 1.57E-01 1.56E-01 Future 40.00 1.65E-01 1.78E-01 1.76E-01 Future 48.00 1.95E-01 2.11E-01 2.09E-01 Future 54.00 2.18E-01 2.35E-01 2.33E-01 Future 60.00 2.41E-01 2.60E-01 2.57E-01 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-11 Table 6-5 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence Rates (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface Cycle Total Fluence Rate (n/cm2-s)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 1.25E+10 1.93E+10 2.17E+10 2.32E+10 2.32E+10 2 1.15 2.21 1.22E+10 1.88E+10 1.91E+10 1.86E+10 2.17E+10 3 1.23 3.43 1.23E+10 1.87E+10 2.07E+10 2.81E+10 2.81E+10 4 1.26 4.70 1.06E+10 1.51E+10 1.78E+10 1.99E+10 1.99E+10 5 1.27 5.97 9.89E+09 1.49E+10 1.76E+10 1.93E+10 1.93E+10 6 1.32 7.29 9.05E+09 1.53E+10 1.76E+10 1.80E+10 1.86E+10 7 1.30 8.59 9.26E+09 1.47E+10 1.70E+10 1.77E+10 1.78E+10 8 1.35 9.94 9.44E+09 1.44E+10 1.67E+10 1.82E+10 1.82E+10 9 1.19 11.13 9.93E+09 1.44E+10 1.58E+10 1.56E+10 1.67E+10 10 1.36 12.49 9.54E+09 1.48E+10 1.68E+10 1.74E+10 1.78E+10 11 1.38 13.87 9.22E+09 1.40E+10 1.67E+10 1.86E+10 1.86E+10 12 1.35 15.22 9.24E+09 1.45E+10 1.72E+10 1.90E+10 1.90E+10 13 1.26 16.48 9.13E+09 1.48E+10 2.00E+10 2.24E+10 2.24E+10 14 1.22 17.70 8.47E+09 1.43E+10 1.72E+10 1.91E+10 1.91E+10 15 1.29 18.99 9.39E+09 1.46E+10 1.73E+10 1.97E+10 1.97E+10 16 1.42 20.41 8.61E+09 1.41E+10 1.61E+10 1.62E+10 1.71E+10 17 1.35 21.76 9.52E+09 1.43E+10 1.65E+10 1.74E+10 1.74E+10 18 1.33 23.09 8.79E+09 1.37E+10 1.55E+10 1.66E+10 1.66E+10 19 1.36 24.45 8.70E+09 1.37E+10 1.54E+10 1.49E+10 1.63E+10 20 1.30 25.75 8.86E+09 1.29E+10 1.44E+10 1.35E+10 1.52E+10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-12 Table 6-6 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface Cycle Total Fluence (n/cm2)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 4.15E+17 6.40E+17 7.20E+17 7.71E+17 7.71E+17 2 1.15 2.21 8.47E+17 1.30E+18 1.40E+18 1.42E+18 1.53E+18 3 1.23 3.43 1.32E+18 2.02E+18 2.19E+18 2.51E+18 2.51E+18 4 1.26 4.70 1.74E+18 2.62E+18 2.90E+18 3.30E+18 3.30E+18 5 1.27 5.97 2.14E+18 3.22E+18 3.61E+18 4.07E+18 4.07E+18 6 1.32 7.29 2.51E+18 3.86E+18 4.33E+18 4.82E+18 4.82E+18 7 1.30 8.59 2.89E+18 4.45E+18 5.02E+18 5.53E+18 5.53E+18 8 1.35 9.94 3.29E+18 5.06E+18 5.72E+18 6.30E+18 6.30E+18 9 1.19 11.13 3.66E+18 5.60E+18 6.32E+18 6.89E+18 6.89E+18 10 1.36 12.49 4.07E+18 6.23E+18 7.03E+18 7.63E+18 7.63E+18 11 1.38 13.87 4.46E+18 6.83E+18 7.74E+18 8.41E+18 8.41E+18 12 1.35 15.22 4.84E+18 7.42E+18 8.44E+18 9.19E+18 9.19E+18 13 1.26 16.48 5.19E+18 8.00E+18 9.22E+18 1.01E+19 1.01E+19 14 1.22 17.70 5.51E+18 8.54E+18 9.87E+18 1.08E+19 1.08E+19 15 1.29 18.99 5.89E+18 9.12E+18 1.06E+19 1.16E+19 1.16E+19 16 1.42 20.41 6.27E+18 9.75E+18 1.13E+19 1.23E+19 1.23E+19 17 1.35 21.76 6.68E+18 1.04E+19 1.20E+19 1.30E+19 1.30E+19 18 1.33 23.09 7.04E+18 1.09E+19 1.26E+19 1.37E+19 1.37E+19 19 1.36 24.45 7.41E+18 1.15E+19 1.33E+19 1.44E+19 1.44E+19 20 1.30 25.75 7.77E+18 1.20E+19 1.39E+19 1.49E+19 1.49E+19 Future 32.00 9.45E+18 1.47E+19 1.68E+19 1.80E+19 1.80E+19 Future 35.00 1.03E+19 1.59E+19 1.83E+19 1.95E+19 1.95E+19 Future 40.00 1.16E+19 1.80E+19 2.07E+19 2.20E+19 2.20E+19 Future 48.00 1.37E+19 2.14E+19 2.45E+19 2.60E+19 2.60E+19 Future 54.00 1.54E+19 2.39E+19 2.73E+19 2.91E+19 2.91E+19 Future 60.00 1.70E+19 2.64E+19 3.02E+19 3.21E+19 3.21E+19 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-13 Table 6-7 Calculated Azimuthal Variation of Maximum Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface Cycle Total Iron Atom Displacement Rate (dpa/s)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 1.93E-11 2.95E-11 3.33E-11 3.66E-11 3.66E-11 2 1.15 2.21 1.88E-11 2.87E-11 2.94E-11 2.93E-11 3.30E-11 3 1.23 3.43 1.90E-11 2.86E-11 3.19E-11 4.43E-11 4.43E-11 4 1.26 4.70 1.64E-11 2.32E-11 2.74E-11 3.13E-11 3.13E-11 5 1.27 5.97 1.53E-11 2.28E-11 2.70E-11 3.05E-11 3.05E-11 6 1.32 7.29 1.40E-11 2.34E-11 2.70E-11 2.84E-11 2.84E-11 7 1.30 8.59 1.43E-11 2.26E-11 2.61E-11 2.79E-11 2.79E-11 8 1.35 9.94 1.46E-11 2.21E-11 2.56E-11 2.87E-11 2.87E-11 9 1.19 11.13 1.54E-11 2.21E-11 2.44E-11 2.47E-11 2.55E-11 10 1.36 12.49 1.48E-11 2.27E-11 2.58E-11 2.74E-11 2.74E-11 11 1.38 13.87 1.43E-11 2.15E-11 2.57E-11 2.92E-11 2.92E-11 12 1.35 15.22 1.43E-11 2.23E-11 2.65E-11 3.00E-11 3.00E-11 13 1.26 16.48 1.41E-11 2.27E-11 3.08E-11 3.53E-11 3.53E-11 14 1.22 17.70 1.31E-11 2.19E-11 2.64E-11 3.01E-11 3.01E-11 15 1.29 18.99 1.45E-11 2.23E-11 2.67E-11 3.11E-11 3.11E-11 16 1.42 20.41 1.33E-11 2.16E-11 2.48E-11 2.56E-11 2.62E-11 17 1.35 21.76 1.47E-11 2.20E-11 2.55E-11 2.75E-11 2.75E-11 18 1.33 23.09 1.36E-11 2.10E-11 2.39E-11 2.62E-11 2.62E-11 19 1.36 24.45 1.35E-11 2.10E-11 2.36E-11 2.35E-11 2.49E-11 20 1.30 25.75 1.37E-11 1.97E-11 2.21E-11 2.13E-11 2.32E-11 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-14 Table 6-8 Calculated Azimuthal Variation of Maximum Iron Atom Displacements at the Reactor Vessel Clad/Base Metal Interface Cycle Total Iron Atom Displacements (dpa)

Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.05 1.05 6.41E-04 9.79E-04 1.11E-03 1.22E-03 1.22E-03 2 1.15 2.21 1.31E-03 1.99E-03 2.15E-03 2.25E-03 2.34E-03 3 1.23 3.43 2.04E-03 3.10E-03 3.38E-03 3.96E-03 3.96E-03 4 1.26 4.70 2.69E-03 4.02E-03 4.47E-03 5.20E-03 5.20E-03 5 1.27 5.97 3.31E-03 4.93E-03 5.55E-03 6.43E-03 6.43E-03 6 1.32 7.29 3.89E-03 5.91E-03 6.67E-03 7.60E-03 7.60E-03 7 1.30 8.59 4.47E-03 6.81E-03 7.72E-03 8.72E-03 8.72E-03 8 1.35 9.94 5.09E-03 7.75E-03 8.81E-03 9.94E-03 9.94E-03 9 1.19 11.13 5.67E-03 8.58E-03 9.73E-03 1.09E-02 1.09E-02 10 1.36 12.49 6.30E-03 9.55E-03 1.08E-02 1.20E-02 1.20E-02 11 1.38 13.87 6.90E-03 1.05E-02 1.19E-02 1.33E-02 1.33E-02 12 1.35 15.22 7.49E-03 1.14E-02 1.30E-02 1.45E-02 1.45E-02 13 1.26 16.48 8.04E-03 1.23E-02 1.42E-02 1.59E-02 1.59E-02 14 1.22 17.70 8.54E-03 1.31E-02 1.52E-02 1.70E-02 1.70E-02 15 1.29 18.99 9.13E-03 1.40E-02 1.63E-02 1.82E-02 1.82E-02 16 1.42 20.41 9.72E-03 1.49E-02 1.74E-02 1.94E-02 1.94E-02 17 1.35 21.76 1.03E-02 1.59E-02 1.84E-02 2.05E-02 2.05E-02 18 1.33 23.09 1.09E-02 1.67E-02 1.94E-02 2.16E-02 2.16E-02 19 1.36 24.45 1.15E-02 1.76E-02 2.04E-02 2.26E-02 2.26E-02 20 1.30 25.75 1.20E-02 1.84E-02 2.13E-02 2.35E-02 2.35E-02 Future 32.00 1.46E-02 2.25E-02 2.59E-02 2.84E-02 2.84E-02 Future 35.00 1.59E-02 2.44E-02 2.81E-02 3.08E-02 3.08E-02 Future 40.00 1.80E-02 2.76E-02 3.18E-02 3.47E-02 3.47E-02 Future 48.00 2.13E-02 3.28E-02 3.77E-02 4.11E-02 4.11E-02 Future 54.00 2.38E-02 3.66E-02 4.21E-02 4.58E-02 4.58E-02 Future 60.00 2.63E-02 4.05E-02 4.65E-02 5.05E-02 5.05E-02 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-15 Table 6-9 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Callaway Unit 1 Irradiation Fluence Iron Atom Irradiation Capsule Time (E > 1.0 MeV) Displacements Cycle(s)

(EFPY) (n/cm2) (dpa)

U 1 1.05 3.13E+18 6.24E-03 Y 1-4 4.70 1.18E+19 2.35E-02 V 1-8 9.94 2.32E+19 4.59E-02 X 1-10 12.49 3.08E+19 6.10E-02 Z(a) 1-13 16.48 4.00E+19 7.92E-02 W 1-20 25.75 5.98E+19 1.18E-01 Note:

(a) Capsule Z was placed in storage.

Table 6-10 Calculated Surveillance Capsule Lead Factors Capsule Location Status Lead Factor 58.5º (Capsule U) Withdrawn EOC 1 4.06 241° (Capsule Y) Withdrawn EOC 4 3.58 61° (Capsule V) Withdrawn EOC 8 3.67 238.5° (Capsule X) Withdrawn EOC 10 4.03 301.5° (Capsule Z)(a) Withdrawn EOC 13 3.98 121.5° (Capsule W) Withdrawn EOC 20 4.01 Note:

(a) Capsule Z was placed in storage.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-16 Table 6-11 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Outlet Inlet Total Nozzle to Nozzle to Upper Shell Cycle Cycle Upper Upper Upper Upper Shell Longitudinal Cycle Length Time Shell Shell Shell Longitudinal Welds at (EFPY)

(EFPY) Welds - Welds - Plates Weld at 90° 210° and Lowest Lowest 330° Extent Extent 1 1.05 1.05 1.11E+15 1.26E+15 8.60E+15 4.84E+15 7.13E+15 2 1.15 2.21 2.38E+15 2.71E+15 1.94E+16 1.16E+16 1.65E+16 3 1.23 3.43 3.77E+15 4.29E+15 3.10E+16 1.75E+16 2.52E+16 4 1.26 4.70 4.96E+15 5.64E+15 4.08E+16 2.29E+16 3.32E+16 5 1.27 5.97 6.06E+15 6.90E+15 4.91E+16 2.74E+16 3.99E+16 6 1.32 7.29 7.25E+15 8.25E+15 5.87E+16 3.24E+16 4.80E+16 7 1.30 8.59 8.30E+15 9.43E+15 6.50E+16 3.60E+16 5.34E+16 8 1.35 9.94 9.45E+15 1.07E+16 7.50E+16 4.14E+16 6.15E+16 9 1.19 11.13 1.05E+16 1.19E+16 8.24E+16 4.61E+16 6.80E+16 10 1.36 12.49 1.17E+16 1.33E+16 9.36E+16 5.22E+16 7.73E+16 11 1.38 13.87 1.29E+16 1.46E+16 1.03E+17 5.72E+16 8.48E+16 12 1.35 15.22 1.41E+16 1.60E+16 1.13E+17 6.27E+16 9.34E+16 13 1.26 16.48 1.54E+16 1.75E+16 1.24E+17 6.77E+16 1.02E+17 14 1.22 17.70 1.65E+16 1.87E+16 1.34E+17 7.25E+16 1.10E+17 15 1.29 18.99 1.77E+16 2.01E+16 1.44E+17 7.81E+16 1.19E+17 16 1.42 20.41 1.89E+16 2.15E+16 1.55E+17 8.40E+16 1.28E+17 17 1.35 21.76 2.02E+16 2.30E+16 1.66E+17 9.01E+16 1.37E+17 18 1.33 23.09 2.14E+16 2.43E+16 1.76E+17 9.59E+16 1.45E+17 19 1.36 24.45 2.25E+16 2.57E+16 1.86E+17 1.02E+17 1.54E+17 20 1.30 25.75 2.36E+16 2.69E+16 1.95E+17 1.07E+17 1.62E+17 Future 32.00 2.91E+16 3.31E+16 2.43E+17 1.34E+17 2.02E+17 Future 35.00 3.17E+16 3.61E+16 2.66E+17 1.47E+17 2.22E+17 Future 40.00 3.61E+16 4.11E+16 3.05E+17 1.69E+17 2.54E+17 Future 48.00 4.31E+16 4.91E+16 3.66E+17 2.03E+17 3.06E+17 Future 54.00 4.84E+16 5.51E+16 4.12E+17 2.29E+17 3.45E+17 Future 60.00 5.36E+16 6.12E+16 4.59E+17 2.55E+17 3.83E+17 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-17 Table 6-11 (cont.) Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Total Upper Int. Shell Int. Shell Cycle Int. Shell Cycle Shell to Long. to Lower Cycle Length Int. Shell Long.

Time Int. Shell Welds at Shell (EFPY) Plates Weld at (EFPY) Circ. 210° and Circ.

90° Weld 330° Weld 1 1.05 1.05 8.60E+15 7.70E+17 4.15E+17 7.20E+17 7.70E+17 2 1.15 2.21 1.94E+16 1.53E+18 8.47E+17 1.40E+18 1.53E+18 3 1.23 3.43 3.10E+16 2.51E+18 1.32E+18 2.19E+18 2.51E+18 4 1.26 4.70 4.08E+16 3.30E+18 1.74E+18 2.90E+18 3.30E+18 5 1.27 5.97 4.91E+16 4.08E+18 2.14E+18 3.61E+18 4.08E+18 6 1.32 7.29 5.87E+16 4.82E+18 2.51E+18 4.33E+18 4.82E+18 7 1.30 8.59 6.50E+16 5.53E+18 2.88E+18 5.02E+18 5.53E+18 8 1.35 9.94 7.50E+16 6.30E+18 3.29E+18 5.72E+18 6.30E+18 9 1.19 11.13 8.24E+16 6.89E+18 3.66E+18 6.32E+18 6.89E+18 10 1.36 12.49 9.36E+16 7.63E+18 4.07E+18 7.03E+18 7.63E+18 11 1.38 13.87 1.03E+17 8.41E+18 4.45E+18 7.73E+18 8.41E+18 12 1.35 15.22 1.13E+17 9.18E+18 4.83E+18 8.44E+18 9.18E+18 13 1.26 16.48 1.24E+17 1.00E+19 5.19E+18 9.21E+18 1.00E+19 14 1.22 17.70 1.34E+17 1.08E+19 5.51E+18 9.86E+18 1.08E+19 15 1.29 18.99 1.44E+17 1.16E+19 5.89E+18 1.06E+19 1.16E+19 16 1.42 20.41 1.55E+17 1.23E+19 6.27E+18 1.13E+19 1.23E+19 17 1.35 21.76 1.66E+17 1.30E+19 6.67E+18 1.20E+19 1.30E+19 18 1.33 23.09 1.76E+17 1.37E+19 7.03E+18 1.26E+19 1.37E+19 19 1.36 24.45 1.86E+17 1.43E+19 7.40E+18 1.33E+19 1.43E+19 20 1.30 25.75 1.95E+17 1.49E+19 7.76E+18 1.38E+19 1.49E+19 Future 32.00 2.43E+17 1.80E+19 9.44E+18 1.68E+19 1.80E+19 Future 35.00 2.66E+17 1.95E+19 1.02E+19 1.83E+19 1.95E+19 Future 40.00 3.05E+17 2.20E+19 1.16E+19 2.06E+19 2.20E+19 Future 48.00 3.66E+17 2.60E+19 1.37E+19 2.44E+19 2.60E+19 Future 54.00 4.12E+17 2.90E+19 1.53E+19 2.73E+19 2.90E+19 Future 60.00 4.59E+17 3.20E+19 1.70E+19 3.02E+19 3.20E+19 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-18 Table 6-11 (cont.) Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Lower Total Lower Cycle Lower Shell to Cycle Shell Cycle Length Lower Shell Lower Time Long.

(EFPY) Shell Long. Vessel (EFPY) Weld at Plates Weld at Head 210° and 90° Circ.

330° Weld 1 1.05 1.05 7.71E+17 4.15E+17 7.20E+17 9.07E+14 2 1.15 2.21 1.53E+18 8.44E+17 1.39E+18 1.86E+15 3 1.23 3.43 2.51E+18 1.32E+18 2.19E+18 3.08E+15 4 1.26 4.70 3.30E+18 1.74E+18 2.90E+18 4.06E+15 5 1.27 5.97 4.08E+18 2.14E+18 3.61E+18 4.99E+15 6 1.32 7.29 4.82E+18 2.51E+18 4.33E+18 5.92E+15 7 1.30 8.59 5.53E+18 2.89E+18 5.02E+18 6.89E+15 8 1.35 9.94 6.30E+18 3.29E+18 5.72E+18 7.91E+15 9 1.19 11.13 6.89E+18 3.66E+18 6.32E+18 8.73E+15 10 1.36 12.49 7.63E+18 4.07E+18 7.03E+18 9.79E+15 11 1.38 13.87 8.41E+18 4.46E+18 7.74E+18 1.08E+16 12 1.35 15.22 9.19E+18 4.84E+18 8.44E+18 1.19E+16 13 1.26 16.48 1.01E+19 5.19E+18 9.22E+18 1.30E+16 14 1.22 17.70 1.08E+19 5.51E+18 9.87E+18 1.39E+16 15 1.29 18.99 1.16E+19 5.89E+18 1.06E+19 1.49E+16 16 1.42 20.41 1.23E+19 6.27E+18 1.13E+19 1.59E+16 17 1.35 21.76 1.30E+19 6.68E+18 1.20E+19 1.69E+16 18 1.33 23.09 1.37E+19 7.04E+18 1.26E+19 1.78E+16 19 1.36 24.45 1.44E+19 7.41E+18 1.33E+19 1.88E+16 20 1.30 25.75 1.49E+19 7.77E+18 1.39E+19 1.96E+16 Future 32.00 1.80E+19 9.45E+18 1.68E+19 2.38E+16 Future 35.00 1.95E+19 1.03E+19 1.83E+19 2.58E+16 Future 40.00 2.20E+19 1.16E+19 2.07E+19 2.92E+16 Future 48.00 2.60E+19 1.37E+19 2.45E+19 3.46E+16 Future 54.00 2.90E+19 1.54E+19 2.73E+19 3.87E+16 Future 60.00 3.21E+19 1.70E+19 3.02E+19 4.27E+16 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-19 Table 6-12 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements (dpa)

Outlet Inlet Cycle Total Nozzle to Nozzle to Upper Shell Cycle Length Time Upper Upper Upper Shell Longitudinal Upper (EFPY) (EFPY) Shell Shell Longitudinal Welds at Shell Plates Welds - Welds - Weld at 90° 210° and Lowest Lowest 330° Extent Extent 1 1.05 1.05 6.96E-06 7.89E-06 1.52E-05 1.09E-05 1.28E-05 2 1.15 2.21 1.48E-05 1.68E-05 3.39E-05 2.37E-05 2.95E-05 3 1.23 3.43 2.36E-05 2.68E-05 5.45E-05 3.72E-05 4.52E-05 4 1.26 4.70 3.10E-05 3.52E-05 7.17E-05 4.88E-05 5.94E-05 5 1.27 5.97 3.79E-05 4.30E-05 8.64E-05 5.95E-05 7.17E-05 6 1.32 7.29 4.53E-05 5.14E-05 1.03E-04 7.09E-05 8.61E-05 7 1.30 8.59 5.18E-05 5.88E-05 1.15E-04 8.10E-05 9.59E-05 8 1.35 9.94 5.89E-05 6.69E-05 1.32E-04 9.22E-05 1.10E-04 9 1.19 11.13 6.51E-05 7.39E-05 1.45E-04 1.02E-04 1.22E-04 10 1.36 12.49 7.27E-05 8.25E-05 1.65E-04 1.14E-04 1.38E-04 11 1.38 13.87 8.01E-05 9.10E-05 1.81E-04 1.26E-04 1.52E-04 12 1.35 15.22 8.78E-05 9.97E-05 1.99E-04 1.37E-04 1.67E-04 13 1.26 16.48 9.57E-05 1.09E-04 2.18E-04 1.49E-04 1.83E-04 14 1.22 17.70 1.03E-04 1.16E-04 2.35E-04 1.60E-04 1.97E-04 15 1.29 18.99 1.10E-04 1.25E-04 2.54E-04 1.72E-04 2.13E-04 16 1.42 20.41 1.18E-04 1.34E-04 2.73E-04 1.84E-04 2.29E-04 17 1.35 21.76 1.26E-04 1.43E-04 2.92E-04 1.96E-04 2.45E-04 18 1.33 23.09 1.33E-04 1.51E-04 3.10E-04 2.08E-04 2.61E-04 19 1.36 24.45 1.40E-04 1.59E-04 3.27E-04 2.19E-04 2.76E-04 20 1.30 25.75 1.47E-04 1.67E-04 3.43E-04 2.30E-04 2.90E-04 Future 32.00 1.81E-04 2.05E-04 4.27E-04 2.83E-04 3.62E-04 Future 35.00 1.97E-04 2.24E-04 4.67E-04 3.09E-04 3.96E-04 Future 40.00 2.24E-04 2.55E-04 5.35E-04 3.52E-04 4.54E-04 Future 48.00 2.67E-04 3.04E-04 6.42E-04 4.21E-04 5.46E-04 Future 54.00 3.00E-04 3.41E-04 7.23E-04 4.72E-04 6.15E-04 Future 60.00 3.32E-04 3.78E-04 8.04E-04 5.24E-04 6.84E-04 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-20 Table 6-12 (cont.) Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements (dpa)

Cycle Total Int. Shell Cycle Length Time Upper Shell to Int. Shell Int. Shell to Int. Shell Long. Welds (EFPY) (EFPY) Int. Shell Circ. Long. Weld Lower Shell Plates at 210° and Weld at 90° Circ. Weld 330° 1 1.05 1.05 1.52E-05 1.22E-03 6.41E-04 1.11E-03 1.22E-03 2 1.15 2.21 3.39E-05 2.34E-03 1.31E-03 2.15E-03 2.33E-03 3 1.23 3.43 5.45E-05 3.96E-03 2.04E-03 3.38E-03 3.96E-03 4 1.26 4.70 7.17E-05 5.20E-03 2.69E-03 4.47E-03 5.20E-03 5 1.27 5.97 8.64E-05 6.43E-03 3.31E-03 5.55E-03 6.43E-03 6 1.32 7.29 1.03E-04 7.60E-03 3.89E-03 6.67E-03 7.60E-03 7 1.30 8.59 1.15E-04 8.72E-03 4.46E-03 7.72E-03 8.72E-03 8 1.35 9.94 1.32E-04 9.94E-03 5.09E-03 8.81E-03 9.94E-03 9 1.19 11.13 1.45E-04 1.09E-02 5.66E-03 9.73E-03 1.09E-02 10 1.36 12.49 1.65E-04 1.20E-02 6.29E-03 1.08E-02 1.20E-02 11 1.38 13.87 1.81E-04 1.33E-02 6.90E-03 1.19E-02 1.33E-02 12 1.35 15.22 1.99E-04 1.45E-02 7.48E-03 1.30E-02 1.45E-02 13 1.26 16.48 2.18E-04 1.59E-02 8.03E-03 1.42E-02 1.59E-02 14 1.22 17.70 2.35E-04 1.70E-02 8.53E-03 1.52E-02 1.70E-02 15 1.29 18.99 2.54E-04 1.82E-02 9.12E-03 1.63E-02 1.82E-02 16 1.42 20.41 2.73E-04 1.94E-02 9.71E-03 1.74E-02 1.94E-02 17 1.35 21.76 2.92E-04 2.05E-02 1.03E-02 1.84E-02 2.05E-02 18 1.33 23.09 3.10E-04 2.16E-02 1.09E-02 1.94E-02 2.16E-02 19 1.36 24.45 3.27E-04 2.26E-02 1.15E-02 2.04E-02 2.26E-02 20 1.30 25.75 3.43E-04 2.35E-02 1.20E-02 2.13E-02 2.35E-02 Future 32.00 4.27E-04 2.84E-02 1.46E-02 2.59E-02 2.84E-02 Future 35.00 4.67E-04 3.08E-02 1.59E-02 2.81E-02 3.08E-02 Future 40.00 5.35E-04 3.47E-02 1.80E-02 3.18E-02 3.47E-02 Future 48.00 6.42E-04 4.10E-02 2.13E-02 3.76E-02 4.10E-02 Future 54.00 7.23E-04 4.57E-02 2.38E-02 4.20E-02 4.57E-02 Future 60.00 8.04E-04 5.05E-02 2.63E-02 4.64E-02 5.05E-02 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-21 Table 6-12 (cont.) Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements (dpa)

Cycle Total Lower Shell Lower Shell Cycle Length Time Lower Shell Lower Shell Long. Weld to Lower (EFPY) (EFPY) Long. Weld Plates at 210° and Vessel Head at 90° 330° Circ. Weld 1 1.05 1.05 1.22E-03 6.41E-04 1.11E-03 5.76E-06 2 1.15 2.21 2.33E-03 1.30E-03 2.14E-03 1.17E-05 3 1.23 3.43 3.96E-03 2.04E-03 3.38E-03 1.96E-05 4 1.26 4.70 5.20E-03 2.69E-03 4.47E-03 2.57E-05 5 1.27 5.97 6.43E-03 3.31E-03 5.55E-03 3.16E-05 6 1.32 7.29 7.60E-03 3.89E-03 6.67E-03 3.74E-05 7 1.30 8.59 8.72E-03 4.47E-03 7.72E-03 4.35E-05 8 1.35 9.94 9.94E-03 5.09E-03 8.81E-03 4.99E-05 9 1.19 11.13 1.09E-02 5.67E-03 9.73E-03 5.50E-05 10 1.36 12.49 1.20E-02 6.30E-03 1.08E-02 6.17E-05 11 1.38 13.87 1.33E-02 6.90E-03 1.19E-02 6.81E-05 12 1.35 15.22 1.45E-02 7.49E-03 1.30E-02 7.49E-05 13 1.26 16.48 1.59E-02 8.04E-03 1.42E-02 8.20E-05 14 1.22 17.70 1.70E-02 8.54E-03 1.52E-02 8.78E-05 15 1.29 18.99 1.82E-02 9.13E-03 1.63E-02 9.41E-05 16 1.42 20.41 1.94E-02 9.72E-03 1.74E-02 1.00E-04 17 1.35 21.76 2.05E-02 1.03E-02 1.84E-02 1.06E-04 18 1.33 23.09 2.16E-02 1.09E-02 1.94E-02 1.12E-04 19 1.36 24.45 2.26E-02 1.15E-02 2.04E-02 1.18E-04 20 1.30 25.75 2.35E-02 1.20E-02 2.13E-02 1.23E-04 Future 32.00 2.84E-02 1.46E-02 2.59E-02 1.50E-04 Future 35.00 3.08E-02 1.59E-02 2.81E-02 1.62E-04 Future 40.00 3.47E-02 1.80E-02 3.18E-02 1.84E-04 Future 48.00 4.10E-02 2.13E-02 3.77E-02 2.17E-04 Future 54.00 4.58E-02 2.38E-02 4.21E-02 2.43E-04 Future 60.00 5.05E-02 2.63E-02 4.65E-02 2.68E-04 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-22 Figure 6-1 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules and 12.5° Neutron Pad Configuration WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-23 Figure 6-2 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Single Capsule Holder and 20.0° Neutron Pad Configuration WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-24 Figure 6-3 Callaway Unit 1 r,,z Reactor Geometry Plan View at the Core Midplane with a Dual Capsule Holder and 22.5° Neutron Pad Configuration WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 6-25 Figure 6-4 Callaway Unit 1 r,,z Reactor Geometry Section View at 31.5° Azimuthal Angle with Surveillance Capsule WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 7-1 7 SURVEILLANCE CAPSULE REMOVAL

SUMMARY

The following surveillance capsule removal summary (Table 7-1) meets the requirements of ASTM E185-82 [Ref. 8]. Note that all capsules have been removed from the Callaway Unit 1 reactor vessel.

Table 7-1 Surveillance Capsule Withdrawal Summary Capsule Fluence Capsule Capsule Lead Withdrawal Capsule ID Status(a) (n/cm2, E > 1.0 Location Factor(a) EFPY(b, c)

MeV)(c)

Withdrawn U 58.5º 4.07 1.05 0.313 (EOC 1)

Withdrawn Y 241º 3.58 4.70 1.18 (EOC 4)

Withdrawn V 61º 3.68 9.94 2.32 (EOC 8)

Withdrawn X(d) 238.5º 4.03 12.49 3.08 (EOC 10)

Withdrawn W(e) 121.5º 4.02 25.75 5.98 (EOC 20)

Withdrawn Z(f) 301.5º 3.98 16.48 4.00 (EOC 13)

Notes:

(a) Updated in Capsule W dosimetry analysis; see Table 6-10.

(b) EFPY from plant startup.

(c) Updated in Capsule W dosimetry analysis; see Table 6-9.

(d) Capsule X satisfies the 60-year EOLR requirements for 54 EFPY of operation.

(e) Capsule W satisfies the 80, 100, 120 year potential future EOL requirements for 72, 90 and 108 EFPY terms of operation, respectively. The Capsule W fluence value is equal to the projected peak vessel fluence at approximately 115.4 EFPY.

(f) Capsule Z has been placed in the Callaway Unit 1 spent fuel pool per MRP-326 [Ref. 24]. The fluence at the time of withdrawal and the lead factor were updated as part of the Capsule W dosimetry analysis. No specific recommendations are provided herein for this capsule. It should remain in the spent fuel pool for potential testing or reinsertion for further irradiation and/or additional metallurgical testing, if required.

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Westinghouse Non-Proprietary Class 3 8-1 8 REFERENCES

1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

2. 10 CFR 50, Appendix G, Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.

3. Westinghouse Report WCAP-9842, Revision 0, Union Electric Company Callaway Unit No. 1 Reactor Vessel Radiation Surveillance Program, May 1981.

4. ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, ASTM, 1973.

5. Appendix G of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

6. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM.

7. ASTM E399, Test Method for Plane-Strain Fracture Toughness of Metallic Materials, ASTM.

8. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.

9. ASTM E23-12c, Standard Test Methods for Notched Bar Impact Testing of Metallic Materials, ASTM, 2012.

10. ASTM E2298-13a, Standard Test Method for Instrumented Impact Testing of Metallic Materials, ASTM, 2013.

11. ASTM A370-13, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 2013.

12. ASTM E8/E8M-13a, Standard Test Methods for Tension Testing of Metallic Materials, ASTM, 2013.

13. ASTM E21-09, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, ASTM, 2009.

14. Westinghouse Report WCAP-11374, Revision 1, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

15. Westinghouse Report WCAP-12946, Revision 0, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1991.

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Westinghouse Non-Proprietary Class 3 8-2

16. Westinghouse Report WCAP-14895, Revision 0, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, July 1997.

17. Westinghouse Report WCAP-15400, Revision 0, Analysis of Capsule X from the Ameren-UE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 2000.

18. ASTM E853-13, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, ASTM, 2014

19. ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706 (ID), ASTM, 1994.

20. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

21. WCAP-14040, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

22. WCAP-16083-NP, Revision 1, Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry, April 2013.

23. RSICC Data Library Collection DLC-185, BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, July 1999.

24. Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)

Guidelines (MRP-326). EPRI, Palo Alto, CA: 2011. 1022871.

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS A.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn and analyzed to date at Callaway Unit 1 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. A-1]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least-squares adjusted values to within 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report.

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

Capsule Azimuthal Withdrawal Time Irradiation Time Location (EFPY) 58.5º (Capsule U) End of Cycle 1 1.05 241° (Capsule Y) End of Cycle 4 4.70 61° (Capsule V) End of Cycle 8 9.94 238.5° (Capsule X) End of Cycle 10 12.49 121.5° (Capsule W) End of Cycle 20 25.75 The passive neutron sensors included in the evaluations of surveillance Capsules U, Y, V, X, and W are summarized as follows:

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Westinghouse Non-Proprietary Class 3 A-2 Reaction Of Capsule Capsule Capsule Capsule Capsule Sensor Material Interest U Y V X W 63 Copper Cu(n,)60Co X X X X X 54 Iron Fe(n,p)54Mn X X X X X 58 58 Nickel Ni(n,p) Co X X X X X 238 Uranium-238(Cd) U(n,f)FP X X X X X 237 Neptunium-237(Cd) Np(n,f)FP X X X X X Cobalt-Aluminum(1) 59 Co(n,)60Co X X X X X Note:

1. The cobalt-aluminum and uranium monitors for this plant include both bare and cadmium-covered sensors.

Pertinent physical and nuclear characteristics of the passive neutron sensors analyzed are listed in Table A-1 for Capsules U, Y, V, X, and W.

The use of passive monitors such as those listed above do not yield a direct measure of the energy-dependent neutron fluence rate 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 fluence rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron fluence rate level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

the measured specific activity of each monitor, the physical characteristics of each monitor, the operating history of the reactor, the energy response of each monitor, and the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules U, Y, V, and X are documented in References A-2, A-3, A-4, and A-5, respectively. The radiometric counting of the sensors from Capsule W was carried out by Pace Analytical Services, Inc. The radiometric counting followed established ASTM procedures.

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

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

A R

Pj - t j - t d, j N0 F Y C j [1 - e ] [e ]

P ref where:

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

A = Measured specific activity (dps/g).

N0 = Number of target element atoms per gram of sensor.

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

Y = Number of product atoms produced per reaction.

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

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

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

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

tj = Length of irradiation period j (sec).

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

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

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in fluence rate level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, the additional Cj term should be employed. The impact of changing fluence rate 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 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-4 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 fluence rates and the computed values for Cj are listed in Tables A-3 and A-4, respectively, for Capsules U, Y, V, X, and W. These fluence rates represent the capsule- and cycle-dependent results at the radial and azimuthal center of the respective capsules at core 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 cadmium-covered measurements to account for the presence of 235U impurities in the sensors, as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the capsule irradiations. The correction factors corresponding to the Callaway Unit 1 fission sensor reaction rates are summarized as follows:

Correction Capsule U Capsule Y Capsule V Capsule X Capsule W 235 U Impurity/Pu Build-in 0.8720 0.8385 0.7989 0.7737 0.6840 238 U(,f) 0.9652 0.9662 0.9665 0.9659 0.9689 238 Net U Correction 0.8417 0.8102 0.7721 0.7473 0.6627 238 Np(,f) Correction 0.9900 0.9902 0.9903 0.9901 0.9910 The correction factors for Capsules U, Y, V, X, and W, were applied in a multiplicative fashion to the decay-corrected cadmium-covered uranium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules U, Y, V, X, and W, are given in Tables A-5 through A-9. In Tables A-5 through A-9, the measured specific activities, decay-corrected saturated specific activities, and computed reaction rates for each sensor are listed. The cadmium-covered fission sensor reaction rates are listed both with and without the applied corrections for 235U impurities, plutonium build-in, and gamma-ray-induced fission effects.

A.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best-estimates for key exposure parameters such as fluence rate (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 i R i ( ig ig )( g g )

g WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-5 relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross-sections, ig, each with an uncertainty . 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 Callaway Unit 1 surveillance capsule dosimetry, the FERRET code [Ref. A-6] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (fluence rate (E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.

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

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

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

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

For the Callaway Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A.1.1.

The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [Ref. A-7].

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance [Ref. A-8].

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Callaway Unit 1 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 ensured 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:

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Westinghouse Non-Proprietary Class 3 A-6 Reaction Uncertainty 63 Cu(n,)60Co 5%

54 Fe(n,p)54Mn 5%

58 Ni(n,p)58Co 5%

238 U(n,f)FP 10%

237 Np(n,f)FP 10%

59 Co(n,)60Co 5%

These uncertainties are given at the 1 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 recent cross-section evaluations, and they have been tested for 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 Callaway Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

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

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

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

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

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

59 Co(n,)60Co 0.79-3.59%

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

Calculated Neutron Spectrum The neutron spectra inputs 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 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-7 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:

M gg' R 2n R g

• R g'

• 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:

Pgg = [1 - ] gg + e-H where (g g' ) 2 H

2 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term). The value of is 1.0 when g = g, and is 0.0 otherwise.

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

Fluence Rate Normalization Uncertainty (Rn) 15%

Fluence Rate Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Fluence Rate Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-8 A.1.3 Comparisons of Measurements and Calculations Results of the least-squares evaluations of the dosimetry from the Callaway Unit 1 surveillance capsules withdrawn to date are provided in Tables A-10, A-11, A-12, A-13, and A-14 for Capsules U, Y, V, X, and W, respectively. In these tables, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in these tabulations 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. Additionally, comparisons of the calculated and best-estimate values of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the BE/C ratios observed for each of the capsules. Note that for Capsule X, the cadmium-covered uranium monitor was discarded because it was outside the expected values.

The data comparisons provided in Tables A-10 through A-14 show that the adjustments to the calculated spectra are relatively small and 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, the calculational uncertainty is specified as 13% at the 1 level.

Further comparisons of the measurement results with calculations are given in Tables A-15 and A-16.

These comparisons are given on two levels. In Table A-15, 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-16, calculations of fast neutron exposure rates in terms of fast neutron fluence rate (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 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, for the individual threshold foils considered in the least-squares analysis, the average M/C comparisons for fast neutron reactions range from 1.00 to 1.17 in the data set. The overall average M/C ratio for the entire set of Callaway Unit 1 data is 1.07 with an associated standard deviation of 7.9%.

In the comparisons of best-estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.94 to 1.10 for neutron fluence rate (E > 1.0 MeV) and from 0.95 to 1.09 for iron atom displacement rate. The overall average BE/C ratios for neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are 1.02 with a standard deviation of 5.9% and 1.02 with a standard deviation of 5.2%, respectively.

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

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Westinghouse Non-Proprietary Class 3 A-9 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors 90%

Reaction Atomic Target Product Fission

Response

of Weight Atom Half-life Yield Range(a)

Interest (g/g-atom) Fraction (days) (%)

(MeV) 63 Cu (n,) 60Co 63.546 0.6917 1925.5 n/a 5 - 12 54 Fe (n,p) 54Mn 55.845 0.05845 312.11 n/a 2-9 58 Ni (n,p) 58Co 58.693 0.68077 70.82 n/a 2-8 238 U (n,f) 137Cs 238.051 0.9996 10983.07 6.02 1-7 237 Np (n,f) 137Cs 237.048 1.0 10983.07 6.17 0.3 - 4 59 Co (n,) 60Co 58.933 0.0015 1925.5 n/a non-threshold Note:

(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Callaway Unit 1 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 [Ref. A-9].

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Westinghouse Non-Proprietary Class 3 A-10 Table A-2 Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 1 Cycle 2 Cycle 3 Cycle 4 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Oct-84 73090 Apr-86 558152 Nov-87 776234 May-89 459251 Nov-84 878649 May-86 1628231 Dec-87 2498378 Jun-89 2230340 Dec-84 1001525 Jun-86 1891846 Jan-88 2400761 Jul-89 2646374 Jan-85 1825824 Jul-86 2279207 Feb-88 1944633 Aug-89 2567249 Feb-85 1997026 Aug-86 2236572 Mar-88 2534237 Sep-89 2346488 Mar-85 2093839 Sep-86 2395748 Apr-88 1401495 Oct-89 2607932 Apr-85 1476721 Oct-86 2478466 May-88 2444952 Nov-89 2529924 May-85 2351261 Nov-86 2000621 Jun-88 2547644 Dec-89 2644439 Jun-85 1990891 Dec-86 2476793 Jul-88 2624336 Jan-90 2595499 Jul-85 2204437 Jan-87 2310037 Aug-88 2590033 Feb-90 2367162 Aug-85 2226770 Feb-87 2278365 Sep-88 2295718 Mar-90 2604203 Sep-85 2399444 Mar-87 2310093 Oct-88 2526306 Apr-90 2557687 Oct-85 2206396 Apr-87 99068 Nov-88 2355533 May-90 2394320 Nov-85 1940798 May-87 1527462 Dec-88 1913183 Jun-90 2193115 Dec-85 2212562 Jun-87 2442877 Jan-89 2640944 Jul-90 2587861 Jan-86 2445738 Jul-87 2426722 Feb-89 2391524 Aug-90 2614245 Feb-86 2161005 Aug-87 2466901 Mar-89 2483347 Sep-90 1500131 Sep-87 673770 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-11 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 5 Cycle 6 Cycle 7 Cycle 8 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-90 481701 May-92 657818 Nov-93 528099 May-95 1548130 Dec-90 2557248 Jun-92 2520702 Dec-93 2607610 Jun-95 2449671 Jan-91 2629501 Jul-92 2608804 Jan-94 2615603 Jul-95 2595094 Feb-91 2376421 Aug-92 2598660 Feb-94 2375405 Aug-95 2529270 Mar-91 2627949 Sep-92 2394003 Mar-94 2438287 Sep-95 2524783 Apr-91 2532714 Oct-92 2636341 Apr-94 2538220 Oct-95 1784097 May-91 2620742 Nov-92 2540736 May-94 2630033 Nov-95 2440443 Jun-91 2499859 Dec-92 2619556 Jun-94 2543816 Dec-95 2628282 Jul-91 2592727 Jan-93 2623796 Jul-94 2549178 Jan-96 2562335 Aug-91 2638549 Feb-93 2361386 Aug-94 2635385 Feb-96 2421446 Sep-91 2485628 Mar-93 2625029 Sep-94 2549205 Mar-96 2649287 Oct-91 2576841 Apr-93 2371166 Oct-94 2636622 Apr-96 2415714 Nov-91 2266100 May-93 2618825 Nov-94 2547712 May-96 2518985 Dec-91 2602283 Jun-93 2530244 Dec-94 2632215 Jun-96 2554157 Jan-92 2326738 Jul-93 2612147 Jan-95 2593618 Jul-96 2641740 Feb-92 2436718 Aug-93 2609724 Feb-95 2378690 Aug-96 2648027 Mar-92 1477011 Sep-93 2296400 Mar-95 1845718 Sep-96 2552943 Oct-93 712 Oct-96 829440 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-12 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 9 Cycle 10 Cycle 11 Cycle 12 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-96 1378918 May-98 2203216 Nov-99 1816209 May-01 592977 Dec-96 2240097 Jun-98 2565200 Dec-99 2637391 Jun-01 2559683 Jan-97 2649851 Jul-98 2649883 Jan-00 2645815 Jul-01 2646500 Feb-97 2290910 Aug-98 2648018 Feb-00 2327865 Aug-01 2650426 Mar-97 2647771 Sep-98 2559293 Mar-00 2640004 Sep-01 2563968 Apr-97 2557647 Oct-98 2520911 Apr-00 2552392 Oct-01 2653245 May-97 2634434 Nov-98 2412725 May-00 2634464 Nov-01 2562952 Jun-97 2562176 Dec-98 2460391 Jun-00 2551992 Dec-01 2296225 Jul-97 2555565 Jan-99 2632436 Jul-00 2643090 Jan-02 2595946 Aug-97 2268517 Feb-99 2384612 Aug-00 2649040 Feb-02 1081321 Sep-97 1743769 Mar-99 2648559 Sep-00 2562040 Mar-02 2594174 Oct-97 1855609 Apr-99 2557196 Oct-00 2652795 Apr-02 2552512 Nov-97 1789970 May-99 2640967 Nov-00 2561131 May-02 2563532 Dec-97 1907667 Jun-99 2561307 Dec-00 2649483 Jun-02 2563713 Jan-98 1993020 Jul-99 2645814 Jan-01 2649682 Jul-02 2648139 Feb-98 1914092 Aug-99 1813256 Feb-01 2390468 Aug-02 2649574 Mar-98 2221975 Sep-99 2435460 Mar-01 2171069 Sep-02 2562792 Apr-98 97613 Oct-99 33272 Apr-01 378935 Oct-02 1837500 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-13 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 13 Cycle 14 Cycle 15 Cycle 16 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-02 259329 Jun-04 1316596 Nov-05 878563 May-07 1776516 Dec-02 2234008 Jul-04 2651150 Dec-05 2648916 Jun-07 2562458 Jan-03 2650782 Aug-04 2644655 Jan-06 2651165 Jul-07 2650823 Feb-03 2386156 Sep-04 2563708 Feb-06 2393308 Aug-07 2650708 Mar-03 1796248 Oct-04 2649341 Mar-06 2650430 Sep-07 2563252 Apr-03 2379519 Nov-04 2560242 Apr-06 2561630 Oct-07 2649187 May-03 2649671 Dec-04 2646370 May-06 1265698 Nov-07 2567302 Jun-03 2565526 Jan-05 2446478 Jun-06 2277193 Dec-07 2647539 Jul-03 2649539 Feb-05 2394377 Jul-06 2650620 Jan-08 2647435 Aug-03 2650619 Mar-05 2144356 Aug-06 2650699 Feb-08 2478102 Sep-03 2552394 Apr-05 2379250 Sep-06 2565194 Mar-08 2596436 Oct-03 2318184 May-05 2650581 Oct-06 2653569 Apr-08 2554914 Nov-03 2550214 Jun-05 2360203 Nov-06 2565181 May-08 2649463 Dec-03 2650963 Jul-05 2648684 Dec-06 2650432 Jun-08 2564338 Jan-04 2492470 Aug-05 2649193 Jan-06 2650533 Jul-08 2629985 Feb-04 1203392 Sep-05 1352545 Feb-07 2393598 Aug-08 2649172 Mar-04 2650570 Mar-07 2271252 Sep-08 2564947 Apr-04 719857 Apr-07 67941 Oct-08 838215 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-14 Table A-2 (cont.) Monthly Thermal Generation during the First 20 Fuel Cycles of the Callaway Unit 1 Reactor Cycle 17 Cycle 18 Cycle 19 Cycle 20 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-08 1704193 Jun-10 1499003 Nov-11 385273 May-13 178430 Dec-08 1639016 Jul-10 2650897 Dec-11 2650840 Jun-13 2564098 Jan-09 2651078 Aug-10 2650782 Jan-12 2650817 Jul-13 2146631 Feb-09 1570120 Sep-10 2564807 Feb-12 2479390 Aug-13 1034909 Mar-09 2403702 Oct-10 2650536 Mar-12 2647173 Sep-13 2565262 Apr-09 2331500 Nov-10 2568427 Apr-12 2544119 Oct-13 2650010 May-09 2650372 Dec-10 2649808 May-12 2650393 Nov-13 2568480 Jun-09 2483116 Jan-11 2588016 Jun-12 2565043 Dec-13 2650392 Jul-09 2650531 Feb-11 2393612 Jul-12 2650101 Jan-14 2649662 Aug-09 2650765 Mar-11 2644270 Aug-12 2650503 Feb-14 2393563 Sep-09 2565276 Apr-11 2565032 Sep-12 2564922 Mar-14 2646443 Oct-09 2649462 May-11 2647739 Oct-12 2649238 Apr-14 2565105 Nov-09 2568639 Jun-11 2564747 Nov-12 2568195 May-14 2649755 Dec-09 2649580 Jul-11 2650371 Dec-12 2650389 Jun-14 2564943 Jan-10 2650452 Aug-11 2650164 Jan-13 2649965 Jul-14 2648905 Feb-10 2393193 Sep-11 2521112 Feb-13 2393843 Aug-14 2650008 Mar-10 2646205 Oct-11 1195960 Mar-13 2583339 Sep-14 2564831 Apr-10 1366927 Apr-13 512141 Oct-14 854474 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-15 Table A-3 Surveillance Capsules U, Y, V, X, and W Fluence Rates for Cj Calculation, Core Midplane Elevation (E > 1.0 MeV) [n/cm2-s]

Cycle Fuel Cycle Length Capsule U Capsule Y Capsule V Capsule X Capsule W (EFPY) 1 1.05 9.42E+10 8.79E+10 8.79E+10 9.42E+10 9.34E+10 2 1.15 7.76E+10 7.76E+10 8.09E+10 8.01E+10 3 1.23 8.26E+10 8.26E+10 9.51E+10 9.45E+10 4 1.26 7.16E+10 7.16E+10 7.92E+10 7.86E+10 5 1.27 7.10E+10 7.71E+10 7.65E+10 6 1.32 7.13E+10 7.57E+10 7.51E+10 7 1.30 6.66E+10 7.11E+10 7.05E+10 8 1.35 6.58E+10 7.12E+10 7.06E+10 9 1.19 6.80E+10 6.74E+10 10 1.36 7.21E+10 7.15E+10 11 1.38 6.93E+10 12 1.35 7.09E+10 13 1.26 8.66E+10 14 1.22 7.29E+10 15 1.29 7.43E+10 16 1.42 6.73E+10 17 1.35 7.12E+10 18 1.33 6.56E+10 19 1.36 6.38E+10 20 1.30 5.98E+10 Average - 9.42E+10 7.96E+10 7.38E+10 7.81E+10 7.36E+10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-16 Table A-4 Surveillance Capsules U, Y, V, X, and W Cj Factors, Core Midplane Elevation Cj Cycle Fuel Cycle Length Capsule U Capsule Y Capsule V Capsule X Capsule W (EFPY) 1 1.05 1.00 1.10 1.19 1.21 1.27 2 1.15 0.97 1.05 1.04 1.09 3 1.23 1.04 1.12 1.22 1.28 4 1.26 0.90 0.97 1.02 1.07 5 1.27 0.96 0.99 1.04 6 1.32 0.97 0.97 1.02 7 1.30 0.90 0.91 0.96 8 1.35 0.89 0.91 0.96 9 1.19 0.87 0.92 10 1.36 0.92 0.97 11 1.38 0.94 12 1.35 0.96 13 1.26 1.18 14 1.22 0.99 15 1.29 1.01 16 1.42 0.91 17 1.35 0.97 18 1.33 0.89 19 1.36 0.87 20 1.30 0.81 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-17 Table A-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 4.78E+04 3.93E+05 6.00E-17 Cu-63 Co-60 4.57E+04 3.76E+05 5.73E-17 Cu-63 Co-60 4.29E+04 3.53E+05 5.38E-17 5.71E-17 5.71E-17 Fe-54 Mn-54 1.54E+06 3.87E+06 6.14E-15 Fe-54 Mn-54 1.44E+06 3.63E+06 5.76E-15 Fe-54 Mn-54 1.36E+06 3.43E+06 5.45E-15 5.78E-15 5.78E-15 Ni-58 Co-58 1.31E+07 6.03E+07 8.64E-15 Ni-58 Co-58 1.19E+07 5.50E+07 7.87E-15 Ni-58 Co-58 1.17E+07 5.38E+07 7.70E-15 8.07E-15 8.07E-15 U-238 Cs-137 1.47E+05 6.20E+06 4.07E-14 4.07E-14 3.43E-14 Np-237 Cs-137 1.08E+06 4.57E+07 2.91E-13 2.91E-13 2.88E-13 Co-59 Co-60 9.73E+06 8.00E+07 5.22E-12 Co-59 Co-60 9.94E+06 8.17E+07 5.33E-12 Co-59 Co-60 9.40E+06 7.73E+07 5.04E-12 5.20E-12 5.20E-12 Co-59(Cd) Co-60 4.98E+06 4.10E+07 2.67E-12 Co-59(Cd) Co-60 5.38E+06 4.43E+07 2.89E-12 Co-59(Cd) Co-60 4.90E+06 4.03E+07 2.63E-12 2.73E-12 2.73E-12 Note:

(a) Measured activity decay corrected to July 22, 1986 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-18 Table A-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 1.50E+05 1.50E+05 3.58E+05 5.46E-17 Cu-63 Co-60 1.34E+05 1.34E+05 3.20E+05 4.88E-17 Cu-63 Co-60 1.31E+05 1.31E+05 3.13E+05 4.77E-17 5.03E-17 Fe-54 Mn-54 2.12E+06 2.12E+06 3.12E+06 4.95E-15 Fe-54 Mn-54 1.90E+06 1.90E+06 2.80E+06 4.44E-15 Fe-54 Mn-54 1.88E+06 1.88E+06 2.77E+06 4.39E-15 4.59E-15 Ni-58 Co-58 1.89E+07 1.89E+07 5.04E+07 7.21E-15 Ni-58 Co-58 1.73E+07 1.73E+07 4.61E+07 6.60E-15 Ni-58 Co-58 1.70E+07 1.70E+07 4.53E+07 6.49E-15 6.77E-15 U-238 Cs-137 5.37E+05 5.37E+05 5.33E+06 3.50E-14 2.84E-14 Np-237 Cs-137 4.07E+06 4.07E+06 4.04E+07 2.58E-13 2.55E-13 Co-59 Co-60 2.53E+07 2.53E+07 6.04E+07 3.94E-12 Co-59 Co-60 2.45E+07 2.45E+07 5.84E+07 3.81E-12 Co-59 Co-60 2.54E+07 2.54E+07 6.06E+07 3.95E-12 3.90E-12 Co-59(Cd) Co-60 1.32E+07 1.32E+07 3.15E+07 2.05E-12 Co-59(Cd) Co-60 1.30E+07 1.30E+07 3.10E+07 2.02E-12 Co-59(Cd) Co-60 1.34E+07 1.34E+07 3.20E+07 2.09E-12 2.05E-12 Note:

(a) Measured activity decay corrected to December 12, 1990 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-19 Table A-7 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.09E+05 3.29E+05 5.01E-17 Cu-63 Co-60 1.87E+05 2.94E+05 4.48E-17 Cu-63 Co-60 1.83E+05 2.88E+05 4.39E-17 4.63E-17 4.63E-17 Fe-54 Mn-54 2.01E+06 2.89E+06 4.58E-15 Fe-54 Mn-54 1.83E+06 2.63E+06 4.17E-15 Fe-54 Mn-54 1.78E+06 2.56E+06 4.06E-15 4.27E-15 4.27E-15 Ni-58 Co-58 1.83E+07 4.49E+07 6.43E-15 Ni-58 Co-58 1.67E+07 4.10E+07 5.87E-15 Ni-58 Co-58 1.61E+07 3.95E+07 5.66E-15 5.98E-15 5.98E-15 U-238 Cs-137 1.02E+06 5.12E+06 3.37E-14 3.37E-14 2.60E-14 Np-237 Cs-137 7.16E+06 3.60E+07 2.30E-13 2.30E-13 2.27E-13 Co-59 Co-60 3.26E+07 5.12E+07 3.34E-12 Co-59 Co-60 3.24E+07 5.09E+07 3.32E-12 Co-59 Co-60 3.24E+07 5.09E+07 3.32E-12 3.33E-12 3.33E-12 Co-59(Cd) Co-60 1.67E+07 2.63E+07 1.71E-12 Co-59(Cd) Co-60 1.75E+07 2.75E+07 1.79E-12 Co-59(Cd) Co-60 1.75E+07 2.75E+07 1.79E-12 1.77E-12 1.77E-12 Note:

(a) Measured activity decay corrected to December 27, 1996 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-20 Table A-8 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.26E+05 3.32E+05 5.07E-17 Cu-63 Co-60 2.02E+05 2.97E+05 4.53E-17 Cu-63 Co-60 1.97E+05 2.90E+05 4.42E-17 4.67E-17 4.67E-17 Fe-54 Mn-54 2.02E+06 3.08E+06 4.89E-15 Fe-54 Mn-54 1.81E+06 2.76E+06 4.38E-15 Fe-54 Mn-54 1.77E+06 2.70E+06 4.28E-15 4.51E-15 4.51E-15 Ni-58 Co-58 1.47E+07 4.68E+07 6.70E-15 Ni-58 Co-58 1.34E+07 4.27E+07 6.11E-15 Ni-58 Co-58 1.24E+07 3.95E+07 5.65E-15 6.16E-15 6.16E-15 U-238 Cs-137 6.10E+05 2.53E+06 1.66E-14 1.66E-14 1.24E-14 Np-237 Cs-137 8.97E+06 3.72E+07 2.37E-13 2.37E-13 2.35E-13 Co-59 Co-60 3.62E+07 5.32E+07 3.47E-12 Co-59 Co-60 3.37E+07 4.95E+07 3.23E-12 Co-59 Co-60 3.29E+07 4.84E+07 3.16E-12 3.29E-12 3.29E-12 Co-59(Cd) Co-60 1.97E+07 2.90E+07 1.89E-12 Co-59(Cd) Co-60 2.04E+07 3.00E+07 1.96E-12 Co-59(Cd) Co-60 2.01E+07 2.95E+07 1.93E-12 1.92E-12 1.92E-12 Note:

(a) Measured activity decay corrected to January 10, 2000 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-21 Table A-9 Measured Sensor Activities and Reaction Rates for Surveillance Capsule W Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(a) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.49E+05 2.49E+05 3.22E+05 4.91E-17 Cu-63 Co-60 2.21E+05 2.21E+05 2.86E+05 4.36E-17 Cu-63 Co-60 2.17E+05 2.17E+05 2.81E+05 4.28E-17 4.52E-17 Fe-54 Mn-54 1.89E+06 1.89E+06 3.19E+06 5.06E-15 Fe-54 Mn-54 1.66E+06 1.66E+06 2.80E+06 4.45E-15 Fe-54 Mn-54 1.66E+06 1.66E+06 2.80E+06 4.45E-15 4.65E-15 Ni-58 Co-58 1.27E+07 1.27E+07 5.05E+07 7.23E-15 Ni-58 Co-58 1.15E+07 1.15E+07 4.57E+07 6.54E-15 Ni-58 Co-58 1.14E+07 1.14E+07 4.53E+07 6.49E-15 6.75E-15 U-238 Cs-137 1.40E+06 2.73E+06 6.47E+06 4.25E-14 2.82E-14 Np-237 Cs-137 8.27E+06 1.61E+07 3.82E+07 2.43E-13 2.41E-13 Co-59 Co-60 4.26E+07 4.26E+07 5.51E+07 3.60E-12 Co-59 Co-60 4.30E+07 4.30E+07 5.56E+07 3.63E-12 Co-59 Co-60 4.33E+07 4.33E+07 5.60E+07 3.65E-12 3.63E-12 Co-59(Cd) Co-60 2.19E+07 2.19E+07 2.83E+07 1.85E-12 Co-59(Cd) Co-60 2.26E+07 2.26E+07 2.92E+07 1.91E-12 Co-59(Cd) Co-60 2.33E+07 2.33E+07 3.01E+07 1.97E-12 1.91E-12 Note:

(a) Measured activity decay corrected to February 6, 2015 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-22 Table A-10 Least-Squares Evaluation of Dosimetry in Surveillance Capsule U (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycle 1 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 5.70E-17 4.77E-17 5.52E-17 1.20 1.03 1.16 Fe-54(n,p)Mn-54 5.78E-15 5.41E-15 5.89E-15 1.07 0.98 1.09 Ni-58(n,p)Co-58 8.07E-15 7.62E-15 8.23E-15 1.06 0.98 1.08 U-238(Cd)(n,f)Cs-137 3.43E-14 2.96E-14 3.16E-14 1.16 1.09 1.07 Np-237(Cd)(n,f)Cs-137 2.88E-13 2.96E-13 2.98E-13 0.97 0.97 1.01 Co-59(n,g)Co-60 5.20E-12 4.60E-12 5.15E-12 1.13 1.01 1.12 Co-59(Cd)(n,g)Co-60 2.73E-12 3.00E-12 2.76E-12 0.91 0.99 0.92 Average of Fast Energy Threshold Reactions 1.09 1.01 1.08

% standard deviation 8.3 5.0 5.0 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 9.47E+10 13 9.94E+10 6 1.05 (n/cm2-s)

Fluence rate E > 0.1 MeV 4.31E+11 - 4.43E+11 10 1.02 (n/cm2-s) dpa/s 1.85E-10 13 1.93E-10 8 1.04 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-23 Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule Y (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 4 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 5.03E-17 4.20E-17 4.80E-17 1.20 1.05 1.14 Fe-54(n,p)Mn-54 4.59E-15 4.64E-15 4.85E-15 0.99 0.95 1.04 Ni-58(n,p)Co-58 6.77E-15 6.52E-15 6.85E-15 1.04 0.99 1.05 U-238(Cd)(n,f)Cs-137 2.84E-14 2.51E-14 2.60E-14 1.13 1.09 1.03 Np-237(Cd)(n,f)Cs-137 2.55E-13 2.50E-13 2.55E-13 1.02 1.00 1.02 Co-59(n,g)Co-60 3.90E-12 3.80E-12 3.87E-12 1.03 1.01 1.02 Co-59(Cd)(n,g)Co-60 2.05E-12 2.51E-12 2.08E-12 0.82 0.99 0.83 Average of Fast Energy Threshold Reactions 1.08 1.02 1.06

% standard deviation 8.1 5.4 4.6 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 8.00E+10 13 8.19E+10 6 1.02 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.63E+11 - 3.71E+11 10 1.02 (n/cm2-s) dpa/s 1.56E-10 13 1.60E-10 8 1.02 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-24 Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (29.0° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 8 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 4.63E-17 3.97E-17 4.41E-17 1.17 1.05 1.11 Fe-54(n,p)Mn-54 4.27E-15 4.34E-15 4.43E-15 0.98 0.96 1.02 Ni-58(n,p)Co-58 5.98E-15 6.09E-15 6.17E-15 0.98 0.97 1.01 U-238(Cd)(n,f)Cs-137 2.60E-14 2.34E-14 2.35E-14 1.11 1.11 1.01 Np-237(Cd)(n,f)Cs-137 2.27E-13 2.31E-13 2.28E-13 0.98 1.00 0.99 Co-59(n,g)Co-60 3.33E-12 3.50E-12 3.31E-12 0.95 1.01 0.95 Co-59(Cd)(n,g)Co-60 1.77E-12 2.30E-12 1.79E-12 0.77 0.99 0.78 Average of Fast Energy Threshold Reactions 1.04 1.02 1.03

% standard deviation 8.6 6.1 4.6 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 7.42E+10 13 7.38E+10 6 0.99 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.35E+11 - 3.32E+11 10 0.99 (n/cm2-s) dpa/s 1.44E-10 13 1.44E-10 8 0.99 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-25 Table A-13 Least-Squares Evaluation of Dosimetry in Surveillance Capsule X (31.5° Azimuth, Core Midplane - Dual Capsule Holder) Cycles 1 Through 10 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 4.67E-17 4.17E-17 4.49E-17 1.12 1.04 1.08 Fe-54(n,p)Mn-54 4.51E-15 4.59E-15 4.56E-15 0.98 0.99 1.00 Ni-58(n,p)Co-58 6.16E-15 6.44E-15 6.32E-15 0.96 0.97 0.98 Np-237(Cd)(n,f)Cs-137 2.35E-13 2.44E-13 2.32E-13 0.96 1.01 0.95 Co-59(n,g)Co-60 3.29E-12 3.74E-12 3.28E-12 0.88 1.00 0.88 Co-59(Cd)(n,g)Co-60 1.92E-12 2.45E-12 1.94E-12 0.79 0.99 0.79 Average of Fast Energy Threshold Reactions 1.01 1.00 1.00

% standard deviation 7.7 3.0 5.5 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 7.84E+10 13 7.44E+10 7 0.94 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.53E+11 - 3.35E+11 10 0.94 (n/cm2-s) dpa/s 1.52E-10 13 1.45E-10 8 0.95 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-26 Table A-14 Least-Squares Evaluation of Dosimetry in Surveillance Capsule W (31.5° Azimuth, Core Midplane - Single Capsule Holder) Cycles 1 Through 20 Irradiation Reaction Rate (rps/atom)

Reaction Best- M/C M/BE BE/C Measured Calculated Estimate (M) (C)

(BE)

Cu-63(n,)Co-60 4.52E-17 3.94E-17 4.45E-17 1.15 1.02 1.13 Fe-54(n,p)Mn-54 4.65E-15 4.32E-15 4.78E-15 1.08 0.97 1.11 Ni-58(n,p)Co-58 6.75E-15 6.06E-15 6.75E-15 1.11 1.00 1.11 U-238(Cd)(n,f)Cs-137 2.82E-14 2.32E-14 2.58E-14 1.21 1.09 1.11 Np-237(Cd)(n,f)Cs-137 2.41E-13 2.30E-13 2.47E-13 1.05 0.98 1.07 Co-59(n,g)Co-60 3.63E-12 3.20E-12 3.59E-12 1.13 1.01 1.12 Co-59(Cd)(n,g)Co-60 1.91E-12 2.13E-12 1.93E-12 0.90 0.99 0.91 Average of Fast Energy Threshold Reactions 1.12 1.01 1.11

% standard deviation 5.6 4.7 2.0 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 7.40E+10 13 8.17E+10 6 1.10 (n/cm2-s)

Fluence rate E > 0.1 MeV 3.33E+11 - 3.60E+11 10 1.08 (n/cm2-s) dpa/s 1.43E-10 13 1.57E-10 8 1.09 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-27 Table A-15 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions M/C Capsule 63 54 58 238 237 Cu(n,) Fe(n,p) Ni(n,p) U(n,f) Np(n,f)

U 1.20 1.07 1.06 1.16 0.97 Y 1.20 0.99 1.04 1.13 1.02 V 1.17 0.98 0.98 1.11 0.98 X 1.12 0.98 0.96 - 0.96 W 1.15 1.08 1.11 1.21 1.05 Average 1.17 1.02 1.03 1.15 1.00

% Standard 2.9 5.0 5.9 3.8 3.8 Deviation Average 1.07

% Standard 7.9 Deviation Table A-16 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Capsule Neutron Fluence Rate Iron Atom Displacement Rate (E > 1.0 MeV)

U 1.05 1.04 Y 1.02 1.02 V 0.99 0.99 X 0.94 0.95 W 1.10 1.09 Average 1.02 1.02

% Standard deviation 5.9 5.2 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 A-28 A.2 REFERENCES A-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

A-2 Westinghouse Report WCAP-11374, Revision 1, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

A-3 Westinghouse Report WCAP-12946, Revision 0, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1991.

A-4 Westinghouse Report WCAP-14895, Revision 0, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, July 1997.

A-5 Westinghouse Report WCAP-15400, Revision 0, Analysis of Capsule X from the Ameren-UE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 2000.

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

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

A-8 ASTM Standard E944-13, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA), 2014.

A-9 ASTM Standard E844-09, Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC), 2014.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS CLXX denotes Lower Shell Plate R2708-1, longitudinal orientation CTXX denotes Lower Shell Plate R2708-1, transverse orientation CWXX denotes weld material CHXX denotes heat affected zone material Note that the instrumented Charpy data is not required per ASTM Standards E185-82 or E23-12c.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-2 CL32: Tested at 0°F CL38: Tested at 20°F CL31: Tested at 35°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-3 CL45: Tested at 40°F CL39: Tested at 50°F CL40: Tested at 50°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-4 CL34: Tested at 72°F CL37: Tested at 90°F CL44: Tested at 100°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-5 CL36: Tested at 120°F CL41: Tested at 150°F CL33: Tested at 170°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-6 CL42: Tested at 200°F CL43: Tested at 230°F CL35: Tested at 250°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-7 CT31: Tested at 0°F CT42: Tested at 20°F CT36: Tested at 40°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-8 CT40: Tested at 50°F CT43: Tested at 60°F CT44: Tested at 60°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-9 CT38: Tested at 72°F CT32: Tested at 100°F CT34: Tested at 120°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-10 CT35: Tested at 150°F CT39: Tested at 170°F CT37: Tested at 200°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-11 CT45: Tested at 230°F CT41: Tested at 250°F CT33: Tested at 275°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-12 CW45: Tested at -25°F CW36: Tested at 0°F CW40: Tested at 10°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-13 CW32: Tested at 20°F CW34: Tested at 25°F CW42: Tested at 35°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-14 CW33: Tested at 40°F CW44: Tested at 50°F CW43: Tested at 60°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-15 CW38: Tested at 72°F CW37: Tested at 120°F CW31: Tested at 150°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-16 CW35: Tested at 170°F CW39: Tested at 200°F CW41: Tested at 250°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-17 CH41: Tested at -50°F CH44: Tested at -40°F CH42: Tested at -25°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-18 CH37: Tested at 0°F CH43: Tested at 10°F CH34: Tested at 20°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-19 CH40: Tested at 35°F CH33: Tested at 50°F CH39: Tested at 60°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-20 CH31: Tested at 70°F CH35: Tested at 72°F CH36: Tested at 100°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 B-21 CH38: Tested at 150°F CH32: Tested at 170°F CH45: Tested at 200°F WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD C.1 METHODOLOGY Contained in Table C-1 are the upper-shelf energy (USE) values that are used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 6.0. The definition for USE is given in ASTM E185-82 [Ref. C-1], Section 4.18, and reads as follows:

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

Westinghouse reports the average of all Charpy data ( 95% shear) as the USE, excluding any values that are deemed outliers using engineering judgment. Hence, the Capsule W USE values reported in Table C-1 were determined by applying this methodology to the Charpy data tabulated in Tables 5-1 through 5-4 of this report. USE values documented in Table C-1 for the unirradiated material, as well as Capsules U, Y, V and X were also determined by applying the methodology described above to the Charpy impact data reported in WCAP-9842, Revision 0 [Ref. C-2], WCAP-11374, Revision 1 [Ref. C-3], WCAP-12946, Revision 0 [Ref. C-4], WCAP-14895, Revision 0 [Ref. C-5] and WCAP-15400, Revision 0 [Ref. C-6]. Three data points differ from those previously reported, due to the inclusion of all Charpy data ( 95% shear) as the USE: Unirradiated material (weld material), Capsule U (lower shell plate material, longitudinal direction) and Capsule V (HAZ material). The USE values reported in Table C-1 were used in generation of the Charpy V-notch curves.

The lower-shelf energy values were fixed at 2.2 ft-lb for all cases. The lower-shelf lateral expansion (L.E.) values were fixed at 1.0 mils in order to be consistent with the previous capsule analysis [Ref. C-6].

Upper-shelf L.E. is not typically fixed in CVGRAPH; however, due to an inaccurate curve fit, the upper shelf L.E. value will be fixed in a summary plot, as documented in Section 5 of this report, for the T-L material in Capsule W. The individual L.E. plot for this material, as documented in this Appendix, will still allow the upper-shelf L.E. to float for comparison between the two methods. The fixed upper-shelf L.E. value was determined using the same Charpy V-notch test specimens that were used for the upper-shelf energy determination and is shown in Table C-2.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-2 Table C-1 Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH Capsule Material Unirradiated U Y V X W Lower Shell Plate R2708-1 126 124 118 126 120 117 Longitudinal Orientation Lower Shell Plate R2708-1 104 93 91 101 99 83 Transverse Orientation Surveillance Program Weld Metal 107 101 97 104 104 103 (Heat # 90077)

Heat Affected Zone (HAZ) Material 106 120 91 106 115 101 Table C-2 Upper-Shelf L.E. Value (mils) Fixed in CVGRAPH Capsule Material W

Lower Shell Plate R2708-1 62 Transverse Orientation (see Figure 5-5 of this report)

CVGRAPH, Version 6.0 plots of all surveillance data are provided in this appendix, on the pages following the reference list.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-3 C.2 REFERENCES C-1 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706(IF), ASTM, 1982.

C-2 Westinghouse Report WCAP-9842, Revision 0, Union Electric Company Callaway Unit No. 1 Reactor Vessel Radiation Surveillance Program, May 1981 C-3 Westinghouse Report WCAP-11374, Revision 1, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

C-4 Westinghouse Report WCAP-12946, Revision 0, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 1991.

C-5 Westinghouse Report WCAP-14895, Revision 0, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, July 1997.

C-6 Westinghouse Report WCAP-15400, Revision 0, Analysis of Capsule X from the Ameren-UE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program, June 2000.

WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-4 C.3 CVGRAPH VERSION 6.0 INDIVIDUAL PLOTS WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-5 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-6 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-7 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-8 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-9 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-10 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-11 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-12 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-13 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-14 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-15 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-16 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-17 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-18 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-19 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-20 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-21 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-22 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-23 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-24 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-25 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-26 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-27 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-28 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-29 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-30 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-31 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-32 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-33 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-34 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-35 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-36 WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 C-37 WCAP-18001-NP September 2015 Revision 0

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Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D CALLAWAY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION D.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there have been five surveillance capsules removed and tested from the Callaway Unit 1 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Callaway Unit 1 reactor vessel surveillance data and determine if that surveillance data is credible.

D.2 EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, Fracture Toughness Requirements [Ref. D-2], as follows:

the region of 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.

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Westinghouse Non-Proprietary Class 3 D-2 The Callaway Unit 1 reactor vessel beltline region consists of the following materials:

1. Intermediate Shell Plates R2707-1, R2707-2 and R2707-3

2. Lower Shell Plates R2708-1, R2708-2 and R2708-3

3. Intermediate Shell to Lower Shell Circumferential Weld Seam 101-171 (Heat # 90077, Linde Flux Type 124)

4. Intermediate Shell Plate Longitudinal Weld Seams 101-124A, 101-124B and 101-124C (Heat # 90077, Linde Flux Type 0091)

5. Lower Shell Longitudinal Weld Seams 101-142A, 101-142B and 101-142C (Heat #

90077, Linde Flux Type 0091).

Per WCAP-9842, Revision 0 [Ref. D-3], the Callaway Unit 1 surveillance program was developed to the requirements of ASTM E185-73:

The base material and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime. The test material should be selected on the basis of initial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper (Cu) and phosphorus (P)) and neutron fluence.

At the time of the surveillance program development, copper and phosphorus content were considered the most limiting factors in embrittlement of reactor vessel steels. Since all plate materials had approximately the same copper and phosphorus content, initial RTNDT and USE were considered to select the representative plate. The lower shell plate R2708-1 had the highest initial RTNDT and the second lowest USE energy. Thus, the lower shell plate R2708-1 was selected as the plate material for the surveillance program. Note that lower shell plate R2708-1 has the same material heat number as the lower shell plate R2708-3; therefore, surveillance credibility conclusions and subsequent Position 2.1 chemistry factor determinations and Position 2.2 percent USE reduction calculations apply to this plate material as well.

The intermediate shell to lower shell circumferential weld seam, intermediate shell longitudinal weld seams and lower shell longitudinal weld seams all used the same weld material. This material was 3/16 inch Mil B-4 weld filler wire, heat number 90077. The surveillance weld material was also fabricated from this same heat of weld material. The intermediate to lower shell circumferential weld seam 101-171 was fabricated with Linde flux type 124, lot number 1061. The intermediate and lower shell longitudinal weld seams were fabricated with Linde flux type 0091, lot number 0842. The surveillance weld was fabricated with Linde flux type 124, lot number 1061. Thus, the surveillance material is representative of all the weld seams in the beltline region with respect to transition temperature shift, credibility conclusions and Position 2.1 chemistry factor determinations. However, it is only representative of the intermediate shell to lower shell circumferential weld with respect to Position 2.2 percent USE decrease evaluations because both the heat number and flux type (Linde 124) match for this weld material only.

Based on the discussion above, Criterion 1 is met for the Callaway Unit 1 surveillance program.

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Westinghouse Non-Proprietary Class 3 D-3 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, as documented in Appendix C of this report, is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Callaway Unit 1 surveillance materials unambiguously.

Hence, the Callaway Unit 1 surveillance program meets this criterion.

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-4].

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 RTNDT values about this line is less than 28°F for welds and less than 17°F 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 [Ref. D-5]. 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 for the Callaway Unit 1 surveillance plate and weld material.

Case 1: Lower Shell Plate R2708-1 and Weld Material Heat # 90077 Following the NRC Case 1 guidelines, the Callaway Unit 1 surveillance plate and weld metal will be evaluated using the Callaway Unit 1 data. This evaluation is contained in Table D-1. Note that when evaluating the credibility of the surveillance weld data, the measured RTNDT values for the surveillance weld material do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld material measured shift values.

In addition, only Callaway Unit 1 data is being considered; therefore, no temperature adjustment is required.

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Westinghouse Non-Proprietary Class 3 D-4 Table D-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Callaway Unit 1 Surveillance Data Capsule Fluence(a) RTNDT FF*RTNDT Material Capsule FF(b) FF2 (x 1019 n/cm2, (°F) (°F)

E > 1.0 MeV)

U 0.313 0.681 -5.9(c) -4.02 0.464 Y 1.18 1.046 25.1 26.26 1.095 Lower Shell Plate R2708-1 V 2.32 1.227 16.3 20.01 1.506 (Longitudinal)

X 3.08 1.297 25.9 33.59 1.682 W 5.98 1.436 58.6 84.15 2.062 U 0.313 0.681 26.1 17.78 0.464 Y 1.18 1.046 46.7 48.86 1.095 Lower Shell Plate R2708-1 V 2.32 1.227 45.2 55.48 1.506 (Transverse)

X 3.08 1.297 30.8 39.95 1.682 W 5.98 1.436 95.2 136.7 2.062 SUM: 458.75 13.618 CF R2708-1 = (FF

• RTNDT) (FF2) = (458.75) (13.618) = 33.7F U 0.313 0.681 66.2 45.10 0.464 Y 1.18 1.046 35.0 36.62 1.095 Surveillance Weld Material V 2.32 1.227 46.2 56.70 1.506 (Heat #90077)

X 3.08 1.297 49.7 64.46 1.682 W 5.98 1.436 65.8 94.49 2.062 SUM: 297.37 6.809 CF Surv. Weld = (FF

• RTNDT) (FF2) = (297.37) (6.809) = 43.7F Notes:

(a) Updated in Capsule W dosimetry analysis; see Table 6-9.

(b) FF = fluence factor = f(0.28 - 0.10*log (f)).

(c) Even though a reduction should not occur, using the negative measured RTNDT value produces the most conservative results for this credibility evaluation (See Table D-2).

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Westinghouse Non-Proprietary Class 3 D-5 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2.

Table D-2 Callaway Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line CF Capsule Measured Predicted Scatter <17°F Material Capsule (Slopebest-fit) Fluence FF RTNDT RTNDT RTNDT(a) (Base Metal)

(°F) (x 1019 n/cm2) (°F) (°F) (°F) <28°F (Weld)

U 33.7 0.313 0.681 -5.9 23.0 28.9 No Y 33.7 1.18 1.046 25.1 35.3 10.2 Yes Lower Shell Plate R2708-1 V 33.7 2.32 1.227 16.3 41.4 25.1 No (Longitudinal)

X 33.7 3.08 1.297 25.9 43.7 17.8 No W 33.7 5.98 1.436 58.6 48.4 10.2 Yes U 33.7 0.313 0.681 26.1 23.0 3.1 Yes Y 33.7 1.18 1.046 46.7 35.3 11.4 Yes Lower Shell Plate R2708-1 V 33.7 2.32 1.227 45.2 41.4 3.8 Yes (Transverse)

X 33.7 3.08 1.297 30.8 43.7 12.9 Yes W 33.7 5.98 1.436 93.1 48.4 46.8 No U 43.7 0.313 0.681 66.2 29.8 36.4 No Y 43.7 1.18 1.046 35.0 45.7 10.7 Yes Surveillance Weld Material V 43.7 2.32 1.227 46.2 53.6 7.4 Yes (Heat # 90077)

X 43.7 3.08 1.297 49.7 56.7 7.0 Yes W 43.7 5.98 1.436 65.8 62.8 3.0 Yes Note:

(a) Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT].

From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table D-2 indicates that six of the ten surveillance data points fall inside the +/- 1 of 17F scatter band for surveillance base metals; therefore, the plate data is deemed non-credible per the third criterion.

Table D-2 indicates that four of the five surveillance data points fall inside the +/- 1 of 28F scatter band for surveillance weld materials; therefore, the surveillance weld data is deemed credible per the third criterion.

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Westinghouse Non-Proprietary Class 3 D-6 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The capsule specimens are located in the reactor between the neutron shielding pads and the vessel wall and are positioned opposite to the center of the core. The test capsules are located in guide tubes attached to the neutron shielding pads [Ref. D-3]. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F.

Hence, Criterion 4 is met for the Callaway Unit 1 surveillance program.

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 Callaway Unit 1 surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to Callaway Unit 1 surveillance program.

Hence, Criterion 5 is met for the Callaway Unit 1 surveillance program.

D.3 CONCLUSION Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B:

The Callaway Unit 1 surveillance plate data are deemed non-credible The Callaway Unit 1 surveillance weld data are deemed credible WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 D-7 D.4 REFERENCES D-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

D-2 10 CFR 50, Appendix G, Fracture Toughness Requirements, Federal Register, Volume 60, No. 243, December 19, 1995.

D-3 Westinghouse Report WCAP-9842, Revision 0, Union Electric Company Callaway Unit No.1 Reactor Vessel Radiation Surveillance Program, May 1981.

D-4 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels E706 (IF), ASTM, 1982.

D-5 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Assessment Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

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Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E CALLAWAY UNIT 1 UPPER-SHELF ENERGY EVALUATION E.1 EVALUATION Per Regulatory Guide 1.99, Revision 2 [Ref. E-1], the Charpy upper-shelf energy (USE) is assumed to decrease as a function of fluence and copper content as indicated in Figure 2 of the Guide (Figure E-1 of this appendix) when surveillance data is not used. Linear interpolation is permitted. In addition, if surveillance data is to be used, the decrease in upper-shelf energy may be obtained by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure E-1 of this appendix) and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.

The 35 EFPY (end-of-license) and 54 EFPY (end-of-license renewal) upper-shelf energy of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the beltline materials and/or the results of the capsules tested to date using Figure 2 in Regulatory Guide 1.99, Revision 2. The maximum vessel clad/base metal interface fluence value was used to determine the corresponding 1/4T fluence value at 35 EFPY and 54 EFPY for the beltline materials. Likewise, the peak extended beltline fluence values at 35 and 54 EFPY was used for the extended beltline materials.

The Callaway Unit 1 reactor vessel beltline region minimum thickness is 8.63 inches. Calculation of the 1/4T vessel fluence values at 35 EFPY and 54 EFPY for the beltline and extended beltline materials are shown in Table E-1.

Table E-1 Callaway Unit 1 Pressure Vessel 1/4T Fast Neutron Fluence Calculation Fluence (n/cm2, E > 1.0 MeV)

Material 35 EFPY 54 EFPY Surface 1/4T Surface 1/4T Beltline 1.95E+19 1.16E+19 2.91E+19 1.73E+19 Extended 2.66E+17 1.58E+17 4.12E+17 2.45E+17 Beltline The following pages present the Callaway Unit 1 upper-shelf energy evaluation. Figure E-1, as indicated previously, is used in making predictions in accordance with Regulatory Guide 1.99, Revision 2.

Table E-2 provides the predicted EOL USE values for 35 EFPY.

Table E-3 provides the predicted EOLR USE values for 54 EFPY.

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Westinghouse Non-Proprietary Class 3 E-2 Figure E-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence WCAP-18001-NP September 2015 Revision 0

Westinghouse Non-Proprietary Class 3 E-3 Table E-2 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 35 EFPY 1/4T EOL Projected Projected Weight % Fluence(b) Unirradiated USE Material EOL USE of Cu (x 1019 n/cm2, USE (ft-lb) Decrease (ft-lb)

E > 1.0 MeV) (%)

Position 1.2(a)

Intermediate Shell Plate R2707-1 0.05 1.16 78 20 62 Intermediate Shell Plate R2707-2 0.06 1.16 100 20 80 Intermediate Shell Plate R2707-3 0.06 1.16 99 20 79 Lower Shell Plate R2708-1 0.07 1.16 82 20 66 Lower Shell Plate R2708-2 0.06 1.16 105 20 84 Lower Shell Plate R2708-3 0.08 1.16 101 20 81 Intermediate Shell Longitudinal Weld 0.04 1.16 143 20 114 Seams 101-124A, B &C (Heat # 90077)

Lower Longitudinal Weld 0.04 1.16 143 20 114 Seams 101-142A, B &C (Heat # 90077)

Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.16 112 20 90 101-171 (Heat # 90077)

Nozzle Shell Plate R2706-1(c) 0.045 (d) 102 7.5 94 Nozzle Shell Plate R2706-2(c) 0.055 (d) 87 7.5 80 Nozzle Shell Plate R2706-3(c) 0.075 (d) 101 7.5 93 Nozzle Shell to Intermediate Shell Weld 0.04 (d) 145 7.5 134 Seam 103-121 (Various/Multiple)(c)

Nozzle Shell Longitudinal Weld Seams 0.045 (d) 128 7.5 118 101-122A, B &C (Various/Multiple)(c)

Position 2.2(b)

Lower Shell Plate R2708-1 0.07 1.16 82 14 71 Lower Shell Plate R2708-3 0.08 1.16 101 14 87 Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.16 112 9 102 101-171 (Heat # 90077)

Notes:

(a) Calculated using the Cu wt. % values and 1/4T fluence value for each material and Regulatory Guide, Revision 2, Position 1.2. For the predicted USE decrease determinations, the base metal and weld Cu weight percentages were conservatively rounded up to the lowest line (Cu weight percent of 0.10 for base metal and 0.05 for weld) in Regulatory Guide 1.99, Revision 2, Figure 2.

(b) Calculated using surveillance capsule measured percent decrease in USE from Table 5-10 and Regulatory Guide 1.99, Revision 2, Position 2.2; see Figure E-1.

(c) Extended beltline materials with 35 EFPY projected fluence greater than 1 x 1017 n/cm2 (E > 1.0 MeV) were included.

(d) The minimum fluence value (2 x 1017 n/cm2) displayed on Figure 2 of Regulatory Guide 1.99, Revision 2 was conservatively used to determine the predicted USE decrease values for the nozzles shell plate and weld materials.

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Westinghouse Non-Proprietary Class 3 E-4 Table E-3 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 54 EFPY 1/4T EOLR Projected Projected Weight % Fluence Unirradiated USE Material EOLR of Cu (x 1019 n/cm2, USE (ft-lb) Decrease USE (ft-lb)

E > 1.0 MeV) (%)

Position 1.2(a)

Intermediate Shell Plate R2707-1 0.05 1.73 78 22 61 Intermediate Shell Plate R2707-2 0.06 1.73 100 22 78 Intermediate Shell Plate R2707-3 0.06 1.73 99 22 77 Lower Shell Plate R2708-1 0.07 1.73 82 22 64 Lower Shell Plate R2708-2 0.06 1.73 105 22 82 Lower Shell Plate R2708-3 0.08 1.73 101 22 79 Intermediate Shell Longitudinal Weld 0.04 1.73 143 22 112 Seams 101-124A, B &C (Heat # 90077)

Lower Longitudinal Weld 0.04 1.73 143 22 112 Seams 101-142A, B &C (Heat # 90077)

Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.73 112 22 87 101-171 (Heat # 90077)

Nozzle Shell Plate R2706-1 0.045 0.0245 102 8 94 Nozzle Shell Plate R2706-2 0.055 0.0245 87 8 80 Nozzle Shell Plate R2706-3 0.075 0.0245 101 8 93 Inlet Nozzle R2702-1 0.16 (c) 135 0(c) 135 Inlet Nozzle R2702-2 0.16 (c) 137 0(c) 137 Inlet Nozzle R2702-3 0.16 (c) 137 0(c) 137 Inlet Nozzle R2702-4 0.16 (c) 134 0(c) 134 Outlet Nozzle R2703-1 0.16 (c) 90 0(c) 90 Outlet Nozzle R2703-2 0.16 (c) 114 0(c) 114 Outlet Nozzle R2703-3 0.16 (c) 113 0(c) 113 Outlet Nozzle R2703-4 0.16 (c) 118 0(c) 118 Nozzle Shell to Intermediate Shell 0.04 0.0245 145 8 133 Weld Seam 103-121 (Various/Multiple)

Inlet Nozzle to Shell Weld Seams 0.163 (c) 101 0(c) 101 105-121A, B, C &D (Various/Multiple)

Outlet Nozzle to Shell Weld Seams 0.163 (c) 99 0(c) 99 107-121A, B, C &D (Various/Multiple)

Nozzle Shell Longitudinal Weld Seams 0.045 0.0245 128 8 118 101-122A, B &C (Various/Multiple)

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Westinghouse Non-Proprietary Class 3 E-5 1/4T EOLR Projected Projected Weight % Fluence Unirradiated USE Material EOLR of Cu (x 1019 n/cm2, USE (ft-lb) Decrease USE (ft-lb)

E > 1.0 MeV) (%)

Position 2.2(b)

Lower Shell Plate R2708-1 0.07 1.73 82 16 69 Lower Shell Plate R2708-3 0.08 1.73 101 16 85 Intermediate Shell to Lower Shell Circumferential Weld Seam 0.04 1.73 112 10.2 101 101-171 (Heat # 90077)

Notes:

(a) Calculated using the Cu wt. % values and 1/4T fluence value for each material and Regulatory Guide, Revision 2, Position 1.2. For the predicted USE decrease determinations, the base metal and weld Cu weight percentages were conservatively rounded up to the nearest higher line in Regulatory Guide 1.99, Revision 2, Figure 2, unless otherwise noted.

(b) Calculated using surveillance capsule measured percent decrease in USE from Table 5-10 and Regulatory Guide 1.99, Revision 2, Position 2.2; see Figure E-1.

(c) Consistent with RIS 2014-1 [Ref. E-2], which established a formal NRC position on material inclusion and embrittlement requirements of extended beltline materials, the effects of neutron radiation must be considered for any locations that are predicted to experience a neutron fluence exposure greater than 1 x 1017 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period. Per Table 6-11 of this report, the inlet and outlet nozzle forging materials and the inlet and outlet nozzle to shell welds have a 54 EFPY fluence level below this threshold; therefore, percent USE reduction is not considered for these materials.

USE Conclusion As shown in Table E-2 and Table E-3, all of the Callaway Unit 1 reactor vessel beltline materials and extended beltline materials are projected to remain above the USE screening criterion of 50 ft-lbs (per 10 CFR 50, Appendix G) at 35 EPFY and 54 EFPY.

E.2 REFERENCES E-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

E-2 U. S. Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014.

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