W3F1-2019-0022, Resubmittal of Reactor Vessel Material Surveillance Program Capsule Test Results

From kanterella
(Redirected from W3F1-2019-0022)
Jump to navigation Jump to search

Resubmittal of Reactor Vessel Material Surveillance Program Capsule Test Results
ML19073A302
Person / Time
Site: Waterford Entergy icon.png
Issue date: 03/14/2019
From: Signorelli J
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2019-0022
Download: ML19073A302 (259)
v • d • e


Text

t Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6032 John V. Signorelli Manager, Regulatory Assurance (Acting) 10 CFR 50 Appendix H W3F1 -201 9-0022 March 14, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Resubmittal of Reactor Vessel Material Surveillance Program Capsule Test Results Waterford Steam Electric Station, Unit 3 (Waterford 3)

NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38 The Westinghouse report, WCAP-17969-NP, Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Radiation Surveillance Program, was submitted to the NRC on August 6, 2015 as Attachment 1 to Entergy letter W3F1 -2015-0056 (ADAMS Accession No. ML15222A361). The report for Capsule 83° has been revised to include updated references. The revised WCAP-17969-NP report is attached.

There are no regulatory commitments contained in this correspondence.

If you have any questions or require additional information, please contact the Acting Regulatory Assurance Manager, John V. Signorelli, at (504) 739-6032.

Respectfully, John V. Signorelli JVS/ajh

Attachment:

WCAP-17969-NP Revision 2, Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Radiation Surveillance Program cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector Waterford Steam Electric Station, Unit 3 NRR Project Manager

ATTACHMENT W3F1-2019-0022 WCAP-17969-NP Revision 2, Analysis of Capsule 83° from the Entergy Operations, Inc.

Waterford Unit 3 Reactor Vessel Radiation Surveillance Program

Westinghouse Non-Proprietary Class 3 WCAP-17969-NP November 2017 Revision 2 Analysis of Capsule 83° from the Entergy Operations, Inc.

Waterford Unit 3 Reactor Vessel Radiation Surveillance Program

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 WCAP-17969-NP Revision 2 Analysis of Capsule 83° from the Entergy Operations, Inc.

Waterford Unit 3 Reactor Vessel Radiation Surveillance Program D. Brett Lynch

  • Structural Design & Analysis III Benjamin W. Amiri*

Radiation Engineering and Analysis November 2017 Verifiers: Benjamin E. Mays*

Structural Design & Analysis III Greg A. Fischer*

Radiation Engineering and Analysis Approved: Lynn A. Patterson*, Manager Structural Design & Analysis III Laurent P. Houssay*, Manager Radiation Engineering and Analysis

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

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

© 2017 Westinghouse Electric Company LLC All Rights Reserved

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iii LIST OF FIGURES ..................................................................................................................................... vi EXECUTIVE

SUMMARY

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

SUMMARY

OF RESULTS .......................................................................................................... 1-1 2 INTRODUCTION ........................................................................................................................ 2-1 3 BACKGROUND .......................................................................................................................... 3-1 4 DESCRIPTION OF PROGRAM .................................................................................................. 4-1 5 TESTING OF SPECIMENS FROM CAPSULE 83° ................................................................... 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 ........................................................................ 6-4 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE ............................................................ 7-1 8 REFERENCES ............................................................................................................................. 8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ............................................................................................. A-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 WATERFORD UNIT 3 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION

........................................................................................................................................ D-1 APPENDIX E WATERFORD UNIT 3 UPPER-SHELF ENERGY EVALUATION ............................. E-1 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii LIST OF TABLES Table 4-1 Chemical Composition (wt. %) of the Waterford Unit 3 Reactor Vessel Surveillance Materials (Unirradiated)................................................................................................... 4-3 Table 4-2 Arrangement of Encapsulated Test Specimens within Waterford Unit 3 Capsule 83° .... 4-4 Table 5-1 Charpy V-notch Data for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation) ..................... 5-5 Table 5-2 Charpy V-notch Data for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation) ........................ 5-6 Table 5-3 Charpy V-notch Data for the Waterford Unit 3 Surveillance Program Weld Material (Heat

  1. 88114) Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)............................. 5-7 Table 5-4 Charpy V-notch Data for the Waterford Unit 3 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV) ............................................ 5-8 Table 5-5 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)

(Longitudinal Orientation) ............................................................................................... 5-9 Table 5-6 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)

(Transverse Orientation) ................................................................................................ 5-10 Table 5-7 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114) Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)

....................................................................................................................................... 5-11 Table 5-8 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)................ 5-12 Table 5-9 Effect of Irradiation to 2.42 x 1019 n/cm2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Waterford Unit 3 Reactor Vessel Surveillance Capsule 83° Materials ........................................................................................................................ 5-13 Table 5-10 Comparison of the Waterford Unit 3 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 Waterford Unit 3 Capsule 83° Reactor Vessel Surveillance Materials Irradiated to 2.42 x 1019 n/cm2 (E > 1.0 MeV) .............................................. 5-15 Table 6-1 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 19 ............................................ 6-7 Table 6-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Future Projections ............................................................................. 6-7 Table 6-3 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 19 ....................... 6-8 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv Table 6-4 Calculated Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Future Projections ...................................................................................... 6-8 Table 6-5 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence Rates (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface ............................................................. 6-9 Table 6-6 Calculated Azimuthal Variation of Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface ..................................................................... 6-10 Table 6-7 Calculated Azimuthal Variation of Maximum Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface ..................................................................... 6-11 Table 6-8 Calculated Azimuthal Variation of Maximum Iron Atom Displacements at the Reactor Vessel Clad/Base Metal Interface .................................................................................. 6-12 Table 6-9 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Waterford Unit 3 ............................................................................................................................. 6-12 Table 6-10 Calculated Surveillance Capsule Lead Factors .............................................................. 6-13 Table 7-1 Surveillance Capsule Withdrawal Schedule .................................................................... 7-1 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors of Surveillance Capsule 97° ................................................................................................................................. A-10 Table A-2 Nuclear Parameters Used in the Evaluation of Neutron Sensors of Surveillance Capsules 263° and 83° .................................................................................................. A-10 Table A-3 Monthly Thermal Generation during the First 19 Fuel Cycles of the Waterford Unit 3 Reactor .......................................................................................................................... A-11 Table A-4 Surveillance Capsule 97° and 83° Fluence Rates for Cj Calculation, Core Midplane Elevation ....................................................................................................................... A-16 Table A-5 Surveillance Capsule 97° and 83° Cj Factors, Core Midplane Elevation ..................... A-17 Table A-6 Surveillance Capsule 263° Reaction Rates for Cj Calculation, Core Midplane Elevation

...................................................................................................................................... A-18 Table A-7 Surveillance Capsule 263° Cj Factors, Core Midplane Elevation ................................ A-19 Table A-8 Measured Sensor Activities and Reaction Rates for Surveillance Capsule 97° ............ A-20 Table A-9 Measured Sensor Activities and Reaction Rates for Surveillance Capsule 263° .......... A-21 Table A-10 Measured Sensor Activities and Reaction Rates for Surveillance Capsule 83° ............ A-22 Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 97° (7-Degree Azimuth, Core Midplane) Cycles 1 Through 4 Irradiation ........................................................... A-23 Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 263° (7-Degree Azimuth, Core Midplane) Cycles 1 Through 11 Irradiation ......................................................... A-24 Table A-13 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 83° (7-Degree Azimuth, Core Midplane) Cycles 1 Through 19 Irradiation ......................................................... A-25 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 v Table A-14 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions ....................................................................................... A-26 Table A-15 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... A-26 Table C-1 Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH ................................................ C-1 Table D-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation using Waterford Unit 3 Surveillance Capsule Data ................................................................................... D-4 Table D-2 Waterford Unit 3 Surveillance Capsule Data Scatter about the Best-Fit Line ................ D-5 Table D-3 Calculation of Residual vs. Fast Fluence for Waterford Unit 3 ...................................... D-6 Table E-1 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 32 EFPY....................... E-3 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the Waterford Unit 3 Reactor Vessel ............. 4-5 Figure 4-2 Original Surveillance Program Capsule in the Waterford Unit 3 Reactor Vessel ............ 4-6 Figure 4-3 Surveillance Capsule Charpy Impact Specimen Compartment Assembly in the Waterford Unit 3 Reactor Vessel ....................................................................................................... 4-7 Figure 4-4 Surveillance Capsule Tensile and Flux-Monitor Compartment Assembly in the Waterford Unit 3 Reactor Vessel ....................................................................................................... 4-8 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation) .............................................. 5-16 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation) .............................................. 5-17 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation) .............................................. 5-18 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) ................................................. 5-19 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) - Continued ............................. 5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) ................................................. 5-21 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) - Continued ............................. 5-22 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) ................................................. 5-23 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) - Continued ............................. 5-24 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) .................................................... 5-25 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) - Continued ............................... 5-26 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) ......................................... 5-27 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) - Continued .................... 5-28 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) .................................................... 5-29 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) - Continued ............................... 5-30 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vii Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material.......................................................................................... 5-31 Figure 5-10(a) Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material - Continued ..................................................................... 5-32 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material .............................................................................. 5-33 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material - Continued .......................................................... 5-34 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material.......................................................................................... 5-35 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material - Continued ..................................................................... 5-36 Figure 5-13 Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Standard Reference Material.......................................................................................... 5-37 Figure 5-14 Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Standard Reference Material .............................................................................. 5-38 Figure 5-15 Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Standard Reference Material.......................................................................................... 5-39 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation).......................................................... 5-40 Figure 5-17 Charpy Impact Specimen Fracture Surfaces for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) ............................................................. 5-41 Figure 5-18 Charpy Impact Specimen Fracture Surfaces for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) .................................................... 5-42 Figure 5-19 Charpy Impact Specimen Fracture Surfaces for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material .................................................................................................. 5-43 Figure 5-20 Tensile Properties for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) ................................................................................................ 5-44 Figure 5-21 Tensile Properties for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) ................................................................................................. 5-45 Figure 5-22 Tensile Properties for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material ...

.. .................................................................................................................................... 5-46 Figure 5-23 Fractured Tensile Specimens from Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) .................................................................................... 5-47 Figure 5-24 Fractured Tensile Specimens from the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) ......................................................................... 5-48 Figure 5-25 Fractured Tensile Specimens from the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material................................................................................................................. 5-49 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 viii Figure 5-26 Engineering Stress-Strain Curves for Waterford Unit 3 Lower Shell Plate M-1004-2 Tensile Specimens 2J3 and 2L7 (Transverse Orientation) ............................................. 5-50 Figure 5-27 Engineering Stress-Strain Curve for Waterford Unit 3 Lower Shell Plate M-1004-2 Tensile Specimen 2K4 (Transverse Orientation) ........................................................... 5-51 Figure 5-28 Engineering Stress-Strain Curves for Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114) Tensile Specimens 3K5 and 3KD ........................................... 5-52 Figure 5-29 Engineering Stress-Strain Curve for Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114) Tensile Specimen 3L3 ............................................................ 5-53 Figure 5-30 Engineering Stress-Strain Curves for Waterford Unit 3 Heat Affected Zone Material Tensile Specimens 4J3 and 4KB .................................................................................... 5-54 Figure 5-31 Engineering Stress-Strain Curve for Waterford Unit 3 Heat Affected Zone Material Tensile Specimen 4JC .................................................................................................... 5-55 Figure 6-1 Waterford Unit 3 r,,z Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules .................................................................................................... 6-14 Figure 6-2 Waterford Unit 3 r,,z Reactor Geometry Plan View at the Core Midplane with 7° and 14° Surveillance Capsules ............................................................................................. 6-15 Figure 6-3 Waterford Unit 3 r,,z Reactor Geometry Section View at 7° Azimuthal Angle ........... 6-16 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-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ix EXECUTIVE

SUMMARY

The purpose of this report is to document the testing results of surveillance Capsule 83° from Waterford Unit 3. Capsule 83° was removed at 24.66 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 83° received a fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV) after irradiation to 24.66 EFPY. The peak clad/base metal interface vessel fluence after 24.66 EFPY of plant operation was 2.02 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 and weld materials contained in Waterford Unit 3 Capsule 83° are less than the Regulatory Guide 1.99, Revision 2 [Ref. 1] predictions. 2) The Waterford Unit 3 surveillance plate data are judged to be non-credible. The Waterford Unit 3 surveillance weld (Heat # 88114) data are judged to be credible. This credibility evaluation can be found in Appendix D. 3) With consideration of 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 (32 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-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule 83°, the third capsule removed and tested from the Waterford Unit 3 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 83° and previous capsules, along with the program input data.

  • Capsule 83° received an average fast neutron fluence (E > 1.0 MeV) of 2.42 x 1019 n/cm2 after 24.66 effective full power years (EFPY) of plant operation.
  • Irradiation of the reactor vessel Lower Shell Plate M-1004-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation),

resulted in an irradiated 30 ft-lb transition temperature of 0.1°F and an irradiated 50 ft-lb transition temperature of 35.3°F. This results in a 30 ft-lb transition temperature increase of 13.6°F and a 50 ft-lb transition temperature increase of 23.6°F for the longitudinally oriented specimens.

  • Irradiation of the reactor vessel Lower Shell Plate M-1004-2 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 0.8°F and an irradiated 50 ft-lb transition temperature of 37.4°F. This results in a 30 ft-lb transition temperature increase of 25.3°F and a 50 ft-lb transition temperature increase of 34.6°F for the transversely oriented specimens.
  • Irradiation of the Surveillance Program Weld Material (Heat # 88114) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -65.4°F and an irradiated 50 ft-lb transition temperature of -44.0°F. This results in a 30 ft-lb transition temperature increase of 19.0°F and a 50 ft-lb transition temperature increase of 21.0°F.
  • Irradiation of the Heat Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -84.3°F and an irradiated 50 ft-lb transition temperature of -37.5°F.

This results in a 30 ft-lb transition temperature increase of 32.7°F and a 50 ft-lb transition temperature increase of 52.6°F.

  • The average upper-shelf energy of Lower Shell Plate M-1004-2 (longitudinal orientation) resulted in an average energy decrease of 12 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 158 ft-lb for the longitudinally oriented specimens.
  • The average upper-shelf energy of Lower Shell Plate M-1004-2 (transverse orientation) resulted in an average energy decrease of 3 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 138 ft-lb for the transversely oriented specimens.
  • The average upper-shelf energy of the Surveillance Program Weld Material (Heat # 88114) Charpy specimens resulted in an average energy decrease of 23 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 133 ft-lb for the weld metal specimens.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 12 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 158 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 Waterford Unit 3 reactor vessel surveillance materials are presented in Table 5-10.

Standard Reference Material (SRM) HSST 01 Charpy specimens were not included in the Waterford Unit 3 Capsule 83°. However, the SRM HSST 01 Charpy specimens were reanalyzed in this report. The SRM HSST 01 material was contained in Capsule 263°, which was irradiated to a neutron fluence of 1.45 x 1019 n/cm2 (E > 1.0 MeV). The results of the SRM HSST 01 reanalysis will be included in Table 5-10 and shown in Figures 5-13 through 5-15.

  • Irradiation of the SRM HSST 01 Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 185.0°F and an irradiated 50 ft-lb transition temperature of 211.4°F. This results in a 30 ft-lb transition temperature increase of 150.5°F and a 50 ft-lb transition temperature increase of 151.3°F.
  • The average upper-shelf energy of the SRM HSST 01 Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 113 ft-lb.
  • Based on the credibility evaluation presented in Appendix D, the Waterford Unit 3 surveillance plate data is non-credible, and the surveillance weld (Heat # 88114) data is credible.
  • Based on the upper-shelf energy evaluation in Appendix E, all beltline materials contained in the Waterford Unit 3 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 (32 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2].
  • The maximum calculated 32 EFPY (end-of-license) neutron fluence (E > 1.0 MeV) for the Waterford Unit 3 reactor vessel beltline using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e.,

Equation #3 in the Guide) is as follows:

Calculated (32 EFPY): Vessel clad/base metal interface fluence* = 2.57 x 1019 n/cm2 Vessel 1/4 thickness fluence = 1.53 x 1019 n/cm2

  • This fluence value is documented in Table 6-6 WCAP-17969-NP November 2017 Revision 2
      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-1 2 INTRODUCTION This report presents the results of the examination of Capsule 83°, the third capsule removed and tested in the continuing surveillance program, which monitors the effects of neutron irradiation on the Entergy Operations, Inc. Waterford Unit 3 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Waterford Unit 3 reactor pressure vessel materials was designed and recommended by Combustion Engineering, Inc. A detailed description of the surveillance program is contained in TR-C-MCS-001 [Ref. 3], Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Waterford-Unit 3 Reactor Vessel Materials. The pre-irradiation mechanical properties of the reactor vessel materials are presented in TR-C-MCS-002

[Ref. 4]. The surveillance program is generally described in C-NLM-003, Revision 1 [Ref. 5]. It was originally planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73 [Ref. 6], Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels.

Capsule 83° was removed from the reactor after 24.66 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.

Revision 1 of this report updates the uncertainty analysis in Section 6.4 for consistency with Reference

25. No other changes are made, and the conclusions of the report are not affected. Revision 2 of this report updates the revision number of Reference 26. No other changes were made.

This report summarizes the testing and post-irradiation data obtained from surveillance Capsule 83° removed from the Waterford Unit 3 reactor vessel and discusses the analysis of the data.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 Waterford Unit 3 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. 7]. 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. 8]) or the temperature 60°F 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. 7]. 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 Waterford Unit 3 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.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Waterford Unit 3 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 core barrel 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 M-1004-2 (longitudinal orientation)
  • Lower Shell Plate M-1004-2 (transverse orientation)
  • Weld metal fabricated with weld wire Heat Number 88114, Linde Type 0091 flux, which is equivalent to the heat number and Flux Type used in the actual fabrication of the intermediate shell to lower shell circumferential weld seam
  • Weld heat affected zone (HAZ) material of Lower Shell Plate M-1004-2
  • Standard Reference Material (SRM) Heavy-Section Steel Technology (HSST)-01MY Plate 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. Weld test specimens were removed from the weld metal of a stress-relieved weldment joining Lower Shell Plate M-1004-1 and adjacent Lower Shell Plate M-1004-3.

All heat affected zone specimens were obtained from the weld heat affected zone of Lower Shell Plate M-1004-2.

Charpy V-notch impact specimens from Lower Shell Plate M-1004-2 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 M-1004-2 were machined in the transverse orientation only.

Tensile specimens from the weld metal were oriented perpendicular to the welding direction.

Some of the Waterford Unit 3 capsules, specifically the previously tested Capsule 263° and also Capsule 104°, which is still in the reactor vessel, contain SRM, which was supplied by the Oak Ridge National Laboratory, from plate materials used in the HSST Program. The material for the Waterford Unit 3 Capsules was obtained from an A533, Grade B Class 1 plate labeled HSST 01. The plate was produced by the Lukens Steel Company and heat treated by Combustion Engineering, Inc.

All six capsules contain flux monitor assemblies that include sulfur pellets, iron wire, titanium wire, nickel wire (cadmium-shielded), aluminum-cobalt wire (cadmium-shielded and unshielded), copper wire (cadmium-shielded) and uranium foil (cadmium-shielded and unshielded).

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-2 The capsules contain (12 total) thermal monitors made from four low-melting-point eutectic alloys, which were sealed in glass tubes. These thermal monitors were located in three different positions in the capsule. These thermal monitors are used to define the maximum temperature attained by the test specimens during irradiation. The composition of the four eutectic alloys and their melting points are as follows:

80.0% Au, 20.0% Sn Melting Point: 536°F (280°C) 5.0% Ag, 5.0% Sn, 90.0% Pb Melting Point: 558°F (292°C) 2.5% Ag, 97.5% Pb Melting Point: 580°F (304°C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 590°F (310°C)

The chemical composition and the arrangement of the various mechanical specimens in Capsule 83° 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, TR-C-MCS-001 [Ref. 3], Tables III and XX.

Capsule 83° was removed after 24.66 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch and tensile specimens, dosimeters, and thermal monitors. Figures 4-1 through 4-4 detail the arrangement of the surveillance capsules, an example of an original program surveillance capsule, a close-up of the Charpy impact specimen compartment, and the tensile and flux-monitor compartment assembly in the Waterford Unit 3 reactor vessel. Capsules 83°, 97°, 263° and 277° are radiologically equivalent to the 7° azimuth, while Capsules 104° and 284° are radiologically equivalent to the 14° azimuth.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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

Standard Reference Surveillance Weld Material Lower Shell Plate Element Material HSST Original CE Best-Estimate M-1004-2(a) 01MY Plate(b) Analysis(c) Analysis(d)

C 0.23 --- 0.23 ---

Mn 1.38 --- 1.35 ---

P 0.005 --- 0.008 ---

S 0.005 --- 0.006 ---

Si 0.23 --- 0.16 ---

Ni 0.58 0.66 0.22 0.16 Mo 0.57 --- 0.57 ---

Cr 0.01 --- 0.05 ---

Cu 0.03 0.18 0.04 0.05 Al 0.016 --- 0.016 ---

Co 0.009 --- 0.007 ---

Pb <0.001 --- <0.001 ---

W <0.01 --- <0.01 ---

Ti <0.01 --- <0.01 ---

Zr <0.001 --- <0.001 ---

V 0.002 --- 0.005 ---

Sn 0.002 --- 0.001 ---

As 0.018 --- 0.001 ---

Cb <0.01 --- <0.01 ---

Sb 0.0015 --- 0.0011 ---

N2 0.009 --- 0.009 ---

B <0.001 --- <0.001 ---

Notes:

(a) Data obtained from TR-C-MCS-001, Table III [Ref. 3]

(b) Data obtained from NUREG/CR-6413 [Ref. 9].

(c) Data obtained from TR-C-MCS-001, Table III [Ref. 3]. Weld Wire Heat Number 88114, Flux Type Linde 0091.

(d) Best-Estimate Cu and Ni wt. % values were taken from WCAP-16088-NP, Revision 2 [Ref. 10].

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-4 Table 4-2 Arrangement of Encapsulated Test Specimens within Waterford Unit 3 Capsule 83° Compartment Number Compartment Position(a) Specimen Numbers(a)

(Specimen Type and Material)(a) 1 E114 4J3, 4KB, 4JC (Tensile HAZ Specimens) 47M, 454, 42E, 41Y, E124 2 45A, 472, 44K, 45U, (Charpy Impact HAZ Specimens) 43J, 45L, 457, 46K E131 14E, 115, 15L, 12B, 3 (Charpy Impact Longitudinal 116, 15M, 145, 11C, Plate Specimens) 14J, 114, 12U, 13P E142 4 (Tensile Transverse 2J3, 2L7, 2K4 Plate Specimens)

E152 225, 261, 25Y, 23T, 5 (Charpy Impact Transverse 21A, 226, 25L, 231, Plate Specimens) 22K, 24L, 21Y, 222 325, 3A2, 37B, 337, 6 E163 312, 347, 31L, 35P, (Charpy Impact Weld Specimens) 371, 31P, 334, 34P 7 E173 3K5, 3KD, 3L3 (Tensile Weld Specimens)

Note:

(a) Data obtained from TR-C-MCS-001, Table XIX and/or Table XX [Ref. 3].

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-5 Figure 4-1 Arrangement of Surveillance Capsules in the Waterford Unit 3 Reactor Vessel WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-6 Figure 4-2 Original Surveillance Program Capsule in the Waterford Unit 3 Reactor Vessel WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-7 Figure 4-3 Surveillance Capsule Charpy Impact Specimen Compartment Assembly in the Waterford Unit 3 Reactor Vessel WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-8 Figure 4-4 Surveillance Capsule Tensile and Flux-Monitor Compartment Assembly in the Waterford Unit 3 Reactor Vessel WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-1 5 TESTING OF SPECIMENS FROM CAPSULE 83° 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. 11].

Capsule 83° 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 TR-C-MCS-001 [Ref. 3]. All of the items were in their proper locations.

Examination of the thermal monitors indicated that 6 of the 12 temperature monitors had melted, as described below:

  • Capsule compartment E114, the 536°F (280°C) and 558°F (292°C) temperature monitors melted
  • Capsule compartment E142, the 536°F (280°C) and 558°F (292°C) temperature monitors melted
  • Capsule compartment E173, the 536°F (280°C) and 558°F (292°C) temperature monitors melted Based on this examination, the maximum temperature to which the specimens were exposed was less than 580°F (304°C), but greater than 558°F (292°C).

The Charpy impact tests were performed per ASTM Specification E185-82 [Ref. 11] and E23-12c

[Ref. 12] 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. 13]. The temperature requirements in ASTM E23-12c [Ref. 12] were met.

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. 13].

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 25% of the calibrated dial energy values as required in ASTM E2298-13a [Ref. 13].

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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. 12] and A370-13 [Ref. 14]. 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. 11]. Testing met ASTM Specifications E8/E8M-13a [Ref. 15] for room temperature or E21-09 [Ref. 16] for elevated temperatures. Load was applied through a threaded connection. Strain measurements were made using an extensometer, which was attached to the 1.00 inch gage section of the tensile specimen. The strain rate obtained met the requirements of ASTM E8/E8M-13a [Ref. 15] and ASTM E21-09 [Ref. 16].

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with an 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 3.00 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 5 (Drawing CND-B-3654 Rev 2) of TR-C-MCS-001 [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. 15]. 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 83°, which received a fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV) in 24.66 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 TR-C-MCS-002, Revision 0 [Ref. 4], BAW-2177, Revision 01 [Ref. 17] and WCAP-16002, Revision 0 [Ref. 18]. The previous capsules, along with the original program unirradiated material input data, were updated using CVGRAPH, Version 6.0 from the hand-drawn plots presented in the earliest reports. 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 Waterford Unit 3 capsule reports.

The transition temperature increases and changes in upper-shelf energies for the Capsule 83° materials are summarized in Table 5-9 and led to the following results:

  • Irradiation of the reactor vessel Lower Shell Plate M-1004-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation),

resulted in an irradiated 30 ft-lb transition temperature of 0.1°F and an irradiated 50 ft-lb transition WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-3 temperature of 35.3°F. This results in a 30 ft-lb transition temperature increase of 13.6°F and a 50 ft-lb transition temperature increase of 23.6°F for the longitudinally oriented specimens.

  • Irradiation of the reactor vessel Lower Shell Plate M-1004-2 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 0.8°F and an irradiated 50 ft-lb transition temperature of 37.4°F. This results in a 30 ft-lb transition temperature increase of 25.3°F and a 50 ft-lb transition temperature increase of 34.6°F for the transversely oriented specimens.
  • Irradiation of the Surveillance Program Weld Material (Heat # 88114) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -65.4°F and an irradiated 50 ft-lb transition temperature of -44.0°F. This results in a 30 ft-lb transition temperature increase of 19.0°F and a 50 ft-lb transition temperature increase of 21.0°F.
  • Irradiation of the Heat Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -84.3°F and an irradiated 50 ft-lb transition temperature of -37.5°F.

This results in a 30 ft-lb transition temperature increase of 32.7°F and a 50 ft-lb transition temperature increase of 52.6°F.

  • The irradiated upper-shelf energy of Lower Shell Plate M-1004-2 (longitudinal orientation) resulted in an average energy decrease of 12 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 158 ft-lb for the longitudinally oriented specimens.
  • The average upper-shelf energy of Lower Shell Plate M-1004-2 (transverse orientation) resulted in an average energy decrease of 3 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 138 ft-lb for the transversely oriented specimens.
  • The average upper-shelf energy of the Surveillance Program Weld Material (Heat # 88114) Charpy specimens resulted in an average energy decrease of 23 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 133 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 12 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 158 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 Waterford Unit 3 reactor vessel surveillance materials are presented in Table 5-10.

Standard Reference Material (SRM) HSST 01 Charpy specimens were not included in the Waterford Unit 3 Capsule 83°. However, the SRM HSST 01 Charpy specimens were reanalyzed in this report. The SRM HSST 01 material was contained in Capsule 263°, which was irradiated to a neutron fluence of 1.45 x 1019 n/cm2 (E > 1.0 MeV). The results of the SRM HSST 01 reanalysis will be included in Table 5-10 and shown in Figures 5-13 through 5-15.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-4

  • Irradiation of the SRM HSST 01 Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 185.0°F and an irradiated 50 ft-lb transition temperature of 211.4°F. This results in a 30 ft-lb transition temperature increase of 150.5°F and a 50 ft-lb transition temperature increase of 151.3°F.
  • The average upper-shelf energy of the SRM HSST 01 Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation. This results in an irradiated average upper-shelf energy of 113 ft-lb.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-16 through 5-19. 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 (32 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 83° irradiated to 2.42 x 1019 n/cm2 (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-20 through 5-22.

The results of the tensile tests performed on the Lower Shell Plate M-1004-2 (transverse orientation) indicated that irradiation to 2.42 x 1019 n/cm2 (E > 1.0 MeV) caused increases (except in one instance there was a slight decrease) in the 0.2 percent offset yield strength, and consistently caused increases in the ultimate tensile strength when compared to unirradiated data [Ref. 4]. See Figure 5-20 and Table 5-11.

The results of the tensile tests performed on the Surveillance Program Weld Material (Heat # 88114) indicated that irradiation to 2.42 x 1019 n/cm2 (E > 1.0 MeV) caused increases (except in one instance there was a decrease) in the 0.2 percent offset yield strength, and caused an increase in the ultimate tensile strength for the single available data point when compared to unirradiated data [Ref. 4]. See Figure 5-21 and Table 5-11.

The results of the tensile tests performed on the Heat Affected Zone Material indicated that irradiation to 2.42 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. 4]. See Figure 5-22 and Table 5-11.

The fractured tensile specimens for the Lower Shell Plate M-1004-2 (transverse orientation) material are shown in Figure 5-23, the fractured tensile specimens for the Surveillance Program Weld Material (Heat # 88114) are shown in Figure 5-24, and the fractured tensile specimens for the Heat Affected Zone Material are shown in Figure 5-25. The engineering stress-strain curves for the tensile tests are shown in Figures 5-26 through 5-31.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-5 Table 5-1 Charpy V-notch Data for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 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  %

145 -25 -32 11 15 7 0.18 5 15M 0 -18 13 18 13 0.33 15 15L 5 -15 32 43 27 0.69 20 14E 10 -12 35 47 25 0.64 20 13P 20 -7 43 58 33 0.84 20 12B 30 -1 53 72 38 0.97 20 12U 40 4 62 84 45 1.14 25 114 100 38 104 141 74 1.88 60 14J 200 93 121 164 84 2.13 85 11C 230 110 154 209 89 2.26 100 116 250 121 162 220 87 2.21 100 115 300 149 158 214 93 2.36 100 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-6 Table 5-2 Charpy V-notch Data for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 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  %

231 -50 -46 18 24 15 0.38 5 21Y -25 -32 19 26 17 0.43 5 22K -10 -23 13 18 12 0.30 10 25L 0 -18 33 45 29 0.74 20 24L 10 -12 33 45 26 0.66 20 25Y 25 -4 44 60 36 0.91 20 23T 40 4 62 84 46 1.17 25 261 100 38 84 114 67 1.70 50 222 150 66 120 163 82 2.08 85 21A 200 93 124 168 81 2.06 100 225 250 121 145 197 71 1.80 100 226 300 149 145 197 91 2.31 100 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-7 Table 5-3 Charpy V-notch Data for the Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114) Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)

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

337 -90 -68 10 14 13 0.33 10 3A2 -70 -57 28 38 20 0.51 25 31L -65 -54 29 39 21 0.53 20 31P -60 -51 28 38 23 0.58 30 34P -55 -48 41 56 33 0.84 35 325 -50 -46 48 65 35 0.89 40 334 -30 -34 70 95 54 1.37 50 35P 0 -18 96 130 66 1.68 60 347 69 21 117 159 84 2.13 95 371 100 38 137 186 90 2.29 98 37B 150 66 140 190 97 2.46 100 312 200 93 136 184 90 2.29 100 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-8 Table 5-4 Charpy V-notch Data for the Waterford Unit 3 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 2.42 x 1019 n/cm2 (E > 1.0 MeV)

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

47M -125 -87 21 28 13 0.33 5 44K -90 -68 23 31 13 0.33 10 42E -80 -62 14 19 10 0.25 15 45U -75 -59 31 42 23 0.58 15 454 -70 -57 39 53 29 0.74 25 45L -60 -51 49 66 34 0.86 35 45A -50 -46 59 80 36 0.91 45 41Y 0 -18 71 96 53 1.35 65 43J 69 21 99 134 66 1.68 70 46K 150 66 131 178 84 2.13 100 457 200 93 163 221 88 2.24 100 472 250 121 179 243 85 2.16 100 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 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) 145 -25 11 10 9 4.3 4400 0.11 3400 3700 0 15M 0 13 11 15 3.5 4000 0.09 3400 3700 0 15L 5 32 30 6 27.8 4200 0.48 3200 4200 0 14E 10 35 31 11 29.1 4200 0.50 3200 4100 0 13P 20 43 39 9 35.5 4400 0.60 3100 4300 500 12B 30 53 47 11 45.5 4300 0.76 3200 4300 500 12U 40 62 56 10 51.8 4300 0.87 3000 4100 600 114 100 104 103 1 44.0 4200 0.77 3000 3200 2100 14J 200 121 118 3 43.1 4000 0.79 2700 2900 2000 11C 230 154 148 4 51.5 4000 0.95 2600 0 0 116 250 162 157 3 44.0 4200 0.83 2600 0 0 115 300 158 153 3 51.6 3900 0.95 2500 0 0 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Lower Shell Plate M-1004-2 Irradiated to a Fluence of 2.42 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) 231 -50 18 18 0 16.2 4100 0.29 3500 4100 0 21Y -25 19 18 5 3.1 4200 0.09 3300 4000 0 22K -10 13 11 15 3.5 4100 0.09 3300 3700 0 25L 0 33 31 6 29.6 4200 0.51 3300 4200 0 24L 10 33 30* 9 24.3 4100 0.43 3200 4000 300 25Y 25 44 39 11 35 4200 0.61 3200 4100 400 23T 40 62 58 7 36.2 4300 0.62 3200 4200 300 261 100 84 80 5 33.2 4100 0.60 2900 3300 1600 222 150 120 117 3 43.1 4000 0.79 2700 2400 1600 21A 200 124 121 2 32 4000 0.60 2800 0 0 225 250 145 140 3 52.6 4000 0.94 2600 0 0 226 300 145 140 3 42.1 4000 0.80 2800 0 0

  • Note: In accordance with Reference 13, an adjustment was made to this value to include additional absorbed energy after the load crossed zero.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114)

Irradiated to a Fluence of 2.42 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) 337 -90 10 10 0 4.5 5200 0.12 3800 3900 0 3A2 -70 28 25 11 5.1 4400 0.11 3900 4100 300 31L -65 29 27 7 3.5 4600 0.09 3500 4400 0 31P -60 28 25 11 4.4 5100 0.12 3900 4600 400 34P -55 41 36 12 3.5 4500 0.09 3400 4200 700 325 -50 48 43 10 3.7 4600 0.09 3800 4400 1400 334 -30 70 65 7 38.0 4400 0.60 3700 4200 1900 35P 0 96 93 3 37.2 4400 0.60 3500 3500 1700 347 69 117 116 1 35.0 4200 0.60 3200 2800 2600 371 100 137 133 3 34.5 4200 0.60 3200 2300 2000 37B 150 140 136 3 34.7 4100 0.63 3100 0 0 312 200 136 134 2 34.9 4100 0.63 3000 0 0 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Waterford Unit 3 Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 2.42 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) 47M -125 21 21 0 4.8 5200 0.11 4300 4900 0 44K -90 23 23 0 3.8 4600 0.09 3500 4500 0 42E -80 14 12 14 3.8 4600 0.09 4100 4300 0 45U -75 31 31 0 3.9 4800 0.09 3800 4500 0 454 -70 39 36 8 32.3 4600 0.50 3600 4400 0 45L -60 49 47 4 40.4 4600 0.61 3700 4500 0 45A -50 59 54 9 3.7 4600 0.09 3700 4400 600 41Y 0 71 70 1 30.5 4400 0.50 3600 4200 2500 43J 69 99 93 6 36.4 4400 0.61 3300 3300 2200 46K 150 131 128 2 4.0 4900 0.13 3200 0 0 457 200 163 159 3 45.5 4200 0.79 3100 0 0 472 250 179 174 3 56.0 4200 0.95 3000 0 0 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-13 Table 5-9 Effect of Irradiation to 2.42 x 1019 n/cm2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the Waterford Unit 3 Reactor Vessel Surveillance Capsule 83° 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(a) (ft-lb)

Unirradiated Irradiated T Unirradiated Irradiated T Unirradiated Irradiated T Unirradiated Irradiated E Lower Shell Plate M-1004-2 -13.5 0.1 13.6 5.3 26.6 21.3 11.7 35.3 23.6 170 158 -12 (Longitudinal)

Lower Shell Plate M-1004-2 -24.5 0.8 25.3 -6.7 23 29.7 2.8 37.4 34.6 141 138 -3 (Transverse)

Surveillance Weld Material -84.4 -65.4 19.0 -68.2 -48.1 20.1 -65.0 -44.0 21.0 156 133 -23 (Heat # 88114)

Heat Affected

-117.0 -84.3 32.7 -89.7 -40.5 49.2 -90.1 -37.5 52.6 170 158 -12 Zone Material Note:

(a) Average value is determined by CVGRAPH, Version 6.0 (see Appendix C).

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-14 Table 5-10 Comparison of the Waterford Unit 3 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, Predicted(a) Measured(b) Predicted(a) Measured(b)

E > 1.0 MeV) (°F) (°F) (%) (%)

Lower Shell Plate 97° 0.631 17.4 6.1 17 9 M-1004-2 (Longitudinal) 83° 2.42 24.8 13.6 23 7 97° 0.631 17.4 28.0 17 12 Lower Shell Plate 263° 1.45 22.1 -9.1 21 7 M-1004-2 (Transverse) 83° 2.42 24.8 25.3 23 2 97° 0.631 38.7 23.5 17 1 Surveillance Weld Material 263° 1.45 49.0 6.6 21 7 (Heat # 88114) 83° 2.42 55.0 19.0 23 15 97° 0.631 --- 13.5 --- 8 Heat Affected Zone Material 263° 1.45 --- 25.8 --- 4 83° 2.42 --- 32.7 --- 7 Standard Reference Material 263° 1.45 --- 150.5 --- 15 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).

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-15 Table 5-11 Tensile Properties of the Waterford Unit 3 Capsule 83° Reactor Vessel Surveillance Materials Irradiated to 2.42 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) 2J3 69 74.7 94.6 3.11 63.3 188 11.9 26.5 66 Lower Shell Plate M-1004-2 2L7 250 68.2 86.7 2.74 55.9 161 9.7 23.7 65 (Transverse) 2K4 550 64.4 89.3 3.15 64.2 147 9.9 19.8 56 3K5 71 85.4 * * * * * *

  • Surveillance Weld Material 3KD 250 76.4 89.5 2.71 55.3 179 8.1 21.4 69 (Heat # 88114) 3L3 550 76.2 * * * * * *
  • 4J3 71 73.3 96.6 2.92 59.4 184 7.1 19.9 68 Heat Affected 4KB 250 67.4 88.5 2.73 55.7 203 5.2 18.0 73 Zone Material 4JC 550 69.7 90.9 3.05 62.1 339 4.9 15.8 82
  • Note: For specimens 3K5 and 3L3, the specimens broke outside of the gage section; as a result, the tensile results may not reflect the weld behavior and are therefore not reported.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-16 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-20 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-21 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-22 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-23 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-24 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation) - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-25 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-26 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-27 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-28 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-29 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-30 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114) - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-31 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-32 Figure 5-10(a) Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-33 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-34 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-35 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-36 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material - Continued WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-37 Figure 5-13 Charpy V-Notch Impact Energy vs. Temperature for the Waterford Unit 3 Reactor Vessel Standard Reference Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-38 Figure 5-14 Charpy V-Notch Lateral Expansion vs. Temperature for the Waterford Unit 3 Reactor Vessel Standard Reference Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-39 Figure 5-15 Charpy V-Notch Percent Shear vs. Temperature for the Waterford Unit 3 Reactor Vessel Standard Reference Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-40 145, -25°F 15M, 0°F 15L, 5°F 14E, 10°F 13P, 20°F 12B, 30°F 12U, 40°F 114, 100°F 14J, 200°F 11C, 230°F 116, 250°F 115, 300°F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Longitudinal Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-41 231, -50°F 21Y, -25°F 22K, -10°F 25L, 0°F 24L, 10°F 25Y, 25°F 23T, 40°F 261, 100°F 222, 150°F 21A, 200°F 225, 250°F 226, 300°F Figure 5-17 Charpy Impact Specimen Fracture Surfaces for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-42 337, -90°F 3A2, -70°F 31L, -65°F 31P, -60°F 34P, -55°F 325, -50°F 334, -30°F 35P, 0°F 347, 69°F 371, 100°F 37B, 150°F 312, 200°F Figure 5-18 Charpy Impact Specimen Fracture Surfaces for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-43 47M, -125°F 44K, -90°F 42E, -80°F 45U, -75°F 454, -70°F 45L, -60°F 45A, -50°F 41Y, 0°F 43J, 69°F 46K, 150°F 457, 200°F 472, 250°F Figure 5-19 Charpy Impact Specimen Fracture Surfaces for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-44 100.0 Ultimate Tensile Strength 90.0 80.0 70.0 0.2% Yield Strength 60.0 Stress (ksi) 50.0 40.0 30.0 20.0 10.0 0.0 0 100 200 300 400 500 600 Temperature (°F)

Legend:and and are unirradiated and and are irradiated to 2.42 x 1019 n/cm2 (E > 1.0 MeV) 80 70 Area Reduction 60 50 Ductility (%)

40 30 Total Elongation 20 10 Uniform Elongation 0

0 100 200 300 400 500 600 Temperature (°F)

Figure 5-20 Tensile Properties for Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-45 100.0 Ultimate Tensile Strength 90.0 80.0 0.2% Yield Strength 70.0 60.0 Stress (ksi) 50.0 40.0 30.0 20.0 10.0 0.0 0 100 200 300 400 500 600 Temperature (°F)

Legend:and and are unirradiated and and are irradiated to 2.42 x 1019 n/cm2 (E > 1.0 MeV) 80 70 Area Reduction 60 50 Ductility (%)

40 30 Total Elongation 20 10 Uniform Elongation 0

0 100 200 300 400 500 600 Temperature (°F)

  • Note: Irradiated weld specimens 3K5 and 3L3 broke outside of the gage section. Thus, limited data is available for these specimens as seen in the plots above. See Table 5-11 for more information.

Figure 5-21 Tensile Properties for the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-46 120.0 100.0 Ultimate Tensile Strength 80.0 Stress (ksi) 0.2% Yield Strength 60.0 40.0 20.0 0.0 0 100 200 300 400 500 600 Temperature (°F)

Legend:and and are unirradiated and and are irradiated to 2.42 x 1019 n/cm2 (E > 1.0 MeV) 90 80 Area Reduction 70 60 Ductility (%)

50 40 30 Total Elongation 20 10 Uniform Elongation 0

0 100 200 300 400 500 600 Temperature (°F)

Figure 5-22 Tensile Properties for the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-47 2J3 - Tested at 69°F 2L7 - Tested at 250°F 2K4 - Tested at 550°F Figure 5-23 Fractured Tensile Specimens from Waterford Unit 3 Reactor Vessel Lower Shell Plate M-1004-2 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-48 3K5 - Tested at 71°F 3KD - Tested at 250°F 3L3 - Tested at 550°F Figure 5-24 Fractured Tensile Specimens from the Waterford Unit 3 Reactor Vessel Surveillance Program Weld Material (Heat # 88114)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-49 4J3 - Tested at 71°F 4KB - Tested at 250°F 4JC - Tested at 550°F Figure 5-25 Fractured Tensile Specimens from the Waterford Unit 3 Reactor Vessel Heat Affected Zone Material WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-50 Tensile Specimen 2J3 Tested at 69°F Tensile Specimen 2L7 Tested at 250°F Figure 5-26 Engineering Stress-Strain Curves for Waterford Unit 3 Lower Shell Plate M-1004-2 Tensile Specimens 2J3 and 2L7 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-51 Tensile Specimen 2K4 Tested at 550°F Figure 5-27 Engineering Stress-Strain Curve for Waterford Unit 3 Lower Shell Plate M-1004-2 Tensile Specimen 2K4 (Transverse Orientation)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-52 Tensile Specimen 3K5 Tested at 71°F

  • Note: Irradiated weld specimen 3K5 broke outside of the gage section. As a result, the stress-strain curve is atypical and may not reflect the weld behavior.

Tensile Specimen 3KD Tested at 250°F Figure 5-28 Engineering Stress-Strain Curves for Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114) Tensile Specimens 3K5 and 3KD WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-53 Tensile Specimen 3L3 Tested at 550°F

  • Note: Irradiated weld specimen 3L3 broke outside of the gage section. As a result, the stress-strain curve is atypical and may not reflect the weld behavior.

Figure 5-29 Engineering Stress-Strain Curve for Waterford Unit 3 Surveillance Program Weld Material (Heat # 88114) Tensile Specimen 3L3 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-54 Tensile Specimen 4J3 Tested at 71°F Tensile Specimen 4KB Tested at 250°F Figure 5-30 Engineering Stress-Strain Curves for Waterford Unit 3 Heat Affected Zone Material Tensile Specimens 4J3 and 4KB WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-55 Tensile Specimen 4JC Tested at 550°F Figure 5-31 Engineering Stress-Strain Curve for Waterford Unit 3 Heat Affected Zone Material Tensile Specimen 4JC WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 Waterford Unit 3 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 83°, withdrawn at the end of the nineteenth plant operating cycle, is provided. In addition, the sensor sets from the previously withdrawn capsules (97° and 263°) 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 3716 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. 19] recommends reporting displacements per iron atom 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. 20]. 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. 21]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004 [Ref. 22]. As an improvement, instead of the fluence rate synthesis technique, three-dimensional transport calculations were performed.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-2 6.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Waterford Unit 3 reactor vessel is shown in Figure 4-1. Six irradiation capsules attached to the pressure vessel inside wall are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 83°, 97°, 104°, 263°, 277°, and 284° as shown in Figure 4-1. These full-core positions correspond to the following octant symmetric locations represented in Figure 6-1: 7° from the core cardinal axes (for the 83°, 97°, 263° and 277° capsules) and 14° from the core cardinal axes (for the 104° and 284° capsules). The stainless steel specimen containers are 1.402-inch by 0.652-inch and are approximately 98 inches in height. The containers are positioned axially such that the test specimens are centered 6.25 inches above the core midplane, thus spanning the approximate central eight feet of the 12.5-foot-high reactor core.

From a neutronic standpoint, the surveillance capsules and capsule holders 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 Waterford Unit 3 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 Waterford Unit 3 transport calculations, the r,,z models depicted (given as r, plan view) in Figures 6-1 and 6-2 were utilized since, with the exception of the capsules, the reactor is octant symmetric. The r,z section view depicted in Figure 6-3 shows the model having an axial span from an elevation 5.5 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 surveillance capsules, 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 a few exceptions.

The radius to the center of the surveillance capsule holder and the radius to the pressure vessel were taken from as-built drawings for the Waterford Unit 3 reactor for key differences between the nominal and as-built dimensions. For the reactor pressure vessel, the minimum vessel thickness was used. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,,z reactor models consisted of 160 radial by 121 azimuthal by 247 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 19 fuel cycles at Waterford Unit 3 included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-3 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. 23] and the BUGLE-96 cross-section library [Ref. 24]. 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 6-10. In Tables 6-1 and 6-3, 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), and iron atom displacement rate (dpa/s) and iron atom displacements, respectively, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 7° capsule and 14° capsule. 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 octant symmetric surveillance capsule positions, i.e., for the 7° capsule and 14° capsule. 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 in 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 nineteenth fuel cycle (i.e., after 24.66 EFPY at 3716 MWt of plant operation) was 2.02E+19 n/cm2.

These data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end of the nineteenth fuel cycle, as well as future projections to 32, 36, 40, 48, 55, and 60 EFPY at 3716 MWt. The calculations account for uprates from 3390 MWt to 3441 MWt that occurred prior to Cycle 12, and from 3441 MWt to 3716 MWt that occurred prior to Cycle 14. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 17, 18, and 19 are representative of future plant operation. The future projections are based on the current reactor power level of 3716 MWt and include a 5% positive bias applied to the power generated in the peripheral fuel assemblies.

The calculated fast neutron exposures for the three surveillance capsules withdrawn from the Waterford Unit 3 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 Waterford Unit 3 reactor. From the data WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-4 provided in Table 6-9, Capsule 83° received a fluence (E > 1.0 MeV) of 2.42E+19 n/cm2 after exposure through the end of the nineteenth fuel cycle (i.e., after 24.66 EFPY).

Updated lead factors for the Waterford Unit 3 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 (97°, 263°, and 83°) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (104°, 277°, and 284°), the lead factor corresponds to the calculated fluence values at the end of Cycle 19, the last completed fuel cycle for Waterford Unit 3.

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 83°, which was withdrawn from Waterford Unit 3 at the end of the nineteenth fuel cycle, is summarized below.

Reaction Rate (rps/atom)

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

Cu-63(n,)Co-60 4.85E-17 4.60E-17 1.05 Ti-46(n,p)Sc-46 8.55E-16 7.19E-16 1.19 Fe-54(n,p)Mn-54 4.50E-15 4.07E-15 1.10 Ni-58(n,p)Co-58 6.24E-15 5.32E-15 1.17 Average 1.13

% standard deviation 5.7 The measured-to-calculated (M/C) reaction rate ratios for the Capsule 83° threshold reactions range from 1.05 to 1.19, and the average M/C ratio is 1.13 +/- 5.7% (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 Waterford Unit 3.

6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Waterford Unit 3 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-5 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 Waterford Unit 3 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 Waterford Unit 3 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Waterford Unit 3 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 Waterford Unit 3 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 25 and 26.

Description Capsule and Vessel IR PCA Comparisons 5%

H. B. Robinson Comparisons 7%

Analytical Sensitivity Studies 10%

Additional Uncertainty for Factors not Explicitly 5%

Net Calculational Uncertainty 14%

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-6 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 Waterford Unit 3.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-7 Table 6-1 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 19 Cycle Total Fluence Rate (n/cm2-s) Fluence (n/cm2)

Cycle Length Time (EFPY) (EFPY) 7-Degree 14-Degree 7-Degree 14-Degree 1 1.04 1.04 5.63E+10 3.92E+10 1.85E+18 1.28E+18 2 1.01 2.05 4.41E+10 3.04E+10 3.25E+18 2.25E+18 3 1.15 3.20 4.38E+10 2.92E+10 4.84E+18 3.31E+18 4 1.21 4.41 3.84E+10 2.64E+10 6.31E+18 4.32E+18 5 1.25 5.66 3.98E+10 2.74E+10 7.87E+18 5.40E+18 6 1.30 6.95 3.90E+10 2.31E+10 9.47E+18 6.35E+18 7 1.35 8.30 2.13E+10 1.65E+10 1.04E+19 7.05E+18 8 1.35 9.66 2.65E+10 1.82E+10 1.15E+19 7.83E+18 9 1.44 11.10 2.55E+10 1.88E+10 1.27E+19 8.68E+18 10 1.40 12.50 2.39E+10 1.78E+10 1.37E+19 9.47E+18 11 1.33 13.83 1.87E+10 1.35E+10 1.45E+19 1.00E+19 12 1.48 15.31 2.43E+10 1.72E+10 1.56E+19 1.08E+19 13 1.40 16.70 2.56E+10 1.89E+10 1.68E+19 1.17E+19 14 1.40 18.10 3.20E+10 2.20E+10 1.82E+19 1.26E+19 15 1.29 19.39 3.26E+10 2.16E+10 1.95E+19 1.35E+19 16 1.35 20.74 2.99E+10 2.05E+10 2.08E+19 1.44E+19 17 1.31 22.05 2.80E+10 1.92E+10 2.19E+19 1.52E+19 18 1.40 23.46 3.15E+10 2.12E+10 2.33E+19 1.61E+19 19 1.20 24.66 2.26E+10 1.77E+10 2.42E+19 1.68E+19 Table 6-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center at Core Midplane for Future Projections Fluence (n/cm2)

Total Time (EFPY) 7-Degree 14-Degree 32.00 3.08E+19 2.15E+19 36.00 3.44E+19 2.40E+19 40.00 3.80E+19 2.66E+19 48.00 4.51E+19 3.16E+19 55.00 5.14E+19 3.61E+19 60.00 5.59E+19 3.93E+19 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-8 Table 6-3 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Cycles 1 Through 19 Cycle Total dpa/s dpa Cycle Length Time (EFPY) (EFPY) 7-Degree 14-Degree 7-Degree 14-Degree 1 1.04 1.04 8.21E-11 5.74E-11 2.69E-03 1.88E-03 2 1.01 2.05 6.44E-11 4.46E-11 4.74E-03 3.30E-03 3 1.15 3.20 6.39E-11 4.29E-11 7.07E-03 4.87E-03 4 1.21 4.41 5.61E-11 3.89E-11 9.21E-03 6.35E-03 5 1.25 5.66 5.82E-11 4.03E-11 1.15E-02 7.93E-03 6 1.30 6.95 5.70E-11 3.40E-11 1.38E-02 9.33E-03 7 1.35 8.30 3.11E-11 2.42E-11 1.52E-02 1.04E-02 8 1.35 9.66 3.87E-11 2.67E-11 1.68E-02 1.15E-02 9 1.44 11.10 3.73E-11 2.76E-11 1.85E-02 1.28E-02 10 1.40 12.50 3.50E-11 2.63E-11 2.01E-02 1.39E-02 11 1.33 13.83 2.73E-11 1.98E-11 2.12E-02 1.48E-02 12 1.48 15.31 3.56E-11 2.54E-11 2.29E-02 1.59E-02 13 1.40 16.70 3.74E-11 2.77E-11 2.45E-02 1.72E-02 14 1.40 18.10 4.67E-11 3.24E-11 2.66E-02 1.86E-02 15 1.29 19.39 4.77E-11 3.18E-11 2.85E-02 1.99E-02 16 1.35 20.74 4.38E-11 3.01E-11 3.04E-02 2.12E-02 17 1.31 22.05 4.10E-11 2.83E-11 3.21E-02 2.23E-02 18 1.40 23.46 4.60E-11 3.12E-11 3.41E-02 2.37E-02 19 1.20 24.66 3.31E-11 2.60E-11 3.54E-02 2.47E-02 Table 6-4 Calculated Iron Atom Displacements at the Surveillance Capsule Center at Core Midplane for Future Projections Total Time dpa (EFPY) 7-Degree 14-Degree 32.00 4.50E-02 3.16E-02 36.00 5.03E-02 3.53E-02 40.00 5.55E-02 3.91E-02 48.00 6.60E-02 4.66E-02 55.00 7.52E-02 5.31E-02 60.00 8.18E-02 5.78E-02 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-9 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 1.04 1.04 4.49E+10 2.76E+10 2.40E+10 1.80E+10 4.49E+10 2 1.01 2.05 3.80E+10 2.24E+10 2.06E+10 1.44E+10 3.80E+10 3 1.15 3.20 3.79E+10 2.12E+10 1.75E+10 1.22E+10 3.79E+10 4 1.21 4.41 3.34E+10 1.95E+10 1.78E+10 1.28E+10 3.34E+10 5 1.25 5.66 3.43E+10 2.00E+10 1.73E+10 1.26E+10 3.43E+10 6 1.30 6.95 3.48E+10 1.63E+10 1.16E+10 9.84E+09 3.48E+10 7 1.35 8.30 1.71E+10 1.21E+10 1.25E+10 9.59E+09 1.71E+10 8 1.35 9.66 2.22E+10 1.31E+10 1.05E+10 8.87E+09 2.22E+10 9 1.44 11.10 2.05E+10 1.36E+10 1.04E+10 7.32E+09 2.05E+10 10 1.40 12.50 1.91E+10 1.29E+10 1.24E+10 9.59E+09 1.91E+10 11 1.33 13.83 1.54E+10 9.78E+09 9.18E+09 8.10E+09 1.54E+10 12 1.48 15.31 2.00E+10 1.25E+10 1.05E+10 8.26E+09 2.00E+10 13 1.40 16.70 2.04E+10 1.36E+10 1.07E+10 8.52E+09 2.04E+10 14 1.40 18.10 2.66E+10 1.58E+10 1.31E+10 1.00E+10 2.66E+10 15 1.29 19.39 2.80E+10 1.56E+10 1.41E+10 1.20E+10 2.80E+10 16 1.35 20.74 2.59E+10 1.52E+10 1.46E+10 1.38E+10 2.59E+10 17 1.31 22.05 2.52E+10 1.48E+10 1.56E+10 1.22E+10 2.52E+10 18 1.40 23.46 2.76E+10 1.58E+10 1.54E+10 1.15E+10 2.76E+10 19 1.20 24.66 1.86E+10 1.34E+10 1.66E+10 1.47E+10 1.86E+10 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-10 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 1.04 1.04 1.47E+18 9.04E+17 7.88E+17 5.89E+17 1.47E+18 2 1.01 2.05 2.66E+18 1.60E+18 1.43E+18 1.04E+18 2.66E+18 3 1.15 3.20 4.04E+18 2.38E+18 2.07E+18 1.48E+18 4.04E+18 4 1.21 4.41 5.32E+18 3.12E+18 2.75E+18 1.97E+18 5.32E+18 5 1.25 5.66 6.66E+18 3.91E+18 3.42E+18 2.47E+18 6.66E+18 6 1.30 6.95 8.09E+18 4.57E+18 3.90E+18 2.87E+18 8.09E+18 7 1.35 8.30 8.82E+18 5.09E+18 4.43E+18 3.28E+18 8.82E+18 8 1.35 9.66 9.76E+18 5.65E+18 4.88E+18 3.66E+18 9.76E+18 9 1.44 11.10 1.07E+19 6.27E+18 5.35E+18 3.99E+18 1.07E+19 10 1.40 12.50 1.15E+19 6.84E+18 5.90E+18 4.41E+18 1.15E+19 11 1.33 13.83 1.22E+19 7.25E+18 6.29E+18 4.75E+18 1.22E+19 12 1.48 15.31 1.31E+19 7.82E+18 6.77E+18 5.13E+18 1.31E+19 13 1.40 16.70 1.40E+19 8.43E+18 7.24E+18 5.51E+18 1.40E+19 14 1.40 18.10 1.52E+19 9.12E+18 7.82E+18 5.95E+18 1.52E+19 15 1.29 19.39 1.63E+19 9.76E+18 8.39E+18 6.44E+18 1.63E+19 16 1.35 20.74 1.74E+19 1.04E+19 9.00E+18 7.01E+18 1.74E+19 17 1.31 22.05 1.84E+19 1.10E+19 9.60E+18 7.48E+18 1.84E+19 18 1.40 23.46 1.95E+19 1.16E+19 1.03E+19 7.98E+18 1.96E+19 19 1.20 24.66 2.02E+19 1.21E+19 1.09E+19 8.52E+18 2.02E+19 Future 32.00 2.57E+19 1.55E+19 1.46E+19 1.15E+19 2.57E+19 Future 36.00 2.86E+19 1.73E+19 1.66E+19 1.32E+19 2.86E+19 Future 40.00 3.16E+19 1.92E+19 1.87E+19 1.49E+19 3.16E+19 Future 48.00 3.78E+19 2.30E+19 2.29E+19 1.83E+19 3.78E+19 Future 55.00 4.32E+19 2.63E+19 2.65E+19 2.12E+19 4.32E+19 Future 60.00 4.70E+19 2.87E+19 2.92E+19 2.34E+19 4.70E+19 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-11 Table 6-7 Calculated Azimuthal Variation of Maximum Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface Cycle Total dpa/s Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.04 1.04 6.84E-11 4.24E-11 3.67E-11 2.76E-11 6.84E-11 2 1.01 2.05 5.78E-11 3.45E-11 3.14E-11 2.22E-11 5.78E-11 3 1.15 3.20 5.77E-11 3.26E-11 2.67E-11 1.88E-11 5.77E-11 4 1.21 4.41 5.08E-11 3.00E-11 2.71E-11 1.98E-11 5.08E-11 5 1.25 5.66 5.21E-11 3.09E-11 2.64E-11 1.94E-11 5.21E-11 6 1.30 6.95 5.28E-11 2.51E-11 1.78E-11 1.52E-11 5.28E-11 7 1.35 8.30 2.62E-11 1.87E-11 1.92E-11 1.48E-11 2.62E-11 8 1.35 9.66 3.38E-11 2.02E-11 1.60E-11 1.37E-11 3.38E-11 9 1.44 11.10 3.13E-11 2.09E-11 1.59E-11 1.13E-11 3.13E-11 10 1.40 12.50 2.93E-11 2.00E-11 1.91E-11 1.48E-11 2.93E-11 11 1.33 13.83 2.35E-11 1.51E-11 1.41E-11 1.25E-11 2.35E-11 12 1.48 15.31 3.06E-11 1.92E-11 1.61E-11 1.28E-11 3.06E-11 13 1.40 16.70 3.11E-11 2.10E-11 1.64E-11 1.32E-11 3.11E-11 14 1.40 18.10 4.05E-11 2.44E-11 2.01E-11 1.55E-11 4.05E-11 15 1.29 19.39 4.27E-11 2.40E-11 2.15E-11 1.85E-11 4.27E-11 16 1.35 20.74 3.95E-11 2.34E-11 2.24E-11 2.13E-11 3.95E-11 17 1.31 22.05 3.85E-11 2.29E-11 2.39E-11 1.88E-11 3.85E-11 18 1.40 23.46 4.20E-11 2.44E-11 2.35E-11 1.78E-11 4.20E-11 19 1.20 24.66 2.84E-11 2.07E-11 2.53E-11 2.26E-11 2.84E-11 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-12 Table 6-8 Calculated Azimuthal Variation of Maximum Iron Atom Displacements at the Reactor Vessel Clad/Base Metal Interface Cycle Total dpa Cycle Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.04 1.04 2.24E-03 1.39E-03 1.20E-03 9.06E-04 2.24E-03 2 1.01 2.05 4.05E-03 2.47E-03 2.19E-03 1.60E-03 4.05E-03 3 1.15 3.20 6.15E-03 3.65E-03 3.16E-03 2.29E-03 6.15E-03 4 1.21 4.41 8.09E-03 4.80E-03 4.19E-03 3.04E-03 8.09E-03 5 1.25 5.66 1.01E-02 6.01E-03 5.23E-03 3.80E-03 1.01E-02 6 1.30 6.95 1.23E-02 7.04E-03 5.96E-03 4.42E-03 1.23E-02 7 1.35 8.30 1.34E-02 7.84E-03 6.78E-03 5.05E-03 1.34E-02 8 1.35 9.66 1.49E-02 8.70E-03 7.46E-03 5.64E-03 1.49E-02 9 1.44 11.10 1.63E-02 9.65E-03 8.19E-03 6.15E-03 1.63E-02 10 1.40 12.50 1.76E-02 1.05E-02 9.03E-03 6.81E-03 1.76E-02 11 1.33 13.83 1.86E-02 1.12E-02 9.62E-03 7.33E-03 1.86E-02 12 1.48 15.31 2.00E-02 1.20E-02 1.04E-02 7.92E-03 2.00E-02 13 1.40 16.70 2.13E-02 1.30E-02 1.11E-02 8.50E-03 2.13E-02 14 1.40 18.10 2.31E-02 1.40E-02 1.20E-02 9.18E-03 2.31E-02 15 1.29 19.39 2.49E-02 1.50E-02 1.28E-02 9.93E-03 2.49E-02 16 1.35 20.74 2.65E-02 1.60E-02 1.38E-02 1.08E-02 2.65E-02 17 1.31 22.05 2.80E-02 1.69E-02 1.47E-02 1.15E-02 2.80E-02 18 1.40 23.46 2.98E-02 1.79E-02 1.57E-02 1.23E-02 2.98E-02 19 1.20 24.66 3.08E-02 1.87E-02 1.66E-02 1.31E-02 3.08E-02 Future 32.00 3.91E-02 2.39E-02 2.23E-02 1.78E-02 3.91E-02 Future 36.00 4.37E-02 2.67E-02 2.54E-02 2.03E-02 4.37E-02 Future 40.00 4.82E-02 2.96E-02 2.86E-02 2.29E-02 4.82E-02 Future 48.00 5.76E-02 3.55E-02 3.50E-02 2.82E-02 5.76E-02 Future 55.00 6.58E-02 4.06E-02 4.06E-02 3.27E-02 6.58E-02 Future 60.00 7.17E-02 4.43E-02 4.46E-02 3.60E-02 7.17E-02 Table 6-9 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Waterford Unit 3 Fluence Iron Atom Irradiation Irradiation Time Capsule (E > 1.0 MeV) Displacements Cycles (EFPY)

(n/cm2) (dpa) 97° 1-4 4.41 6.31E+18 9.21E-03 263° 1-11 13.83 1.45E+19 2.12E-02 83° 1-19 24.66 2.42E+19 3.54E-02 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-13 Table 6-10 Calculated Surveillance Capsule Lead Factors Capsule Location Status Lead Factor 97º Withdrawn EOC 4 1.19 263° Withdrawn EOC 11 1.19 83° Withdrawn EOC 19 1.20 104° In Reactor(1) 0.83 277° In Reactor(1) 1.20 284° In Reactor(1) 0.83 Note:

1. Lead factors are based on the cumulative exposures from Cycles 1 through 19.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-14 Figure 6-1 Waterford Unit 3 r,,z Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-15 Figure 6-2 Waterford Unit 3 r,,z Reactor Geometry Plan View at the Core Midplane with 7° and 14° Surveillance Capsules WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-16 Figure 6-3 Waterford Unit 3 r,,z Reactor Geometry Section View at 7° Azimuthal Angle WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule (Table 7-1) meets the requirements of ASTM E185-82 [Ref. 11]. Note that it is recommended for future capsule(s) to be removed from the Waterford Unit 3 reactor vessel.

Table 7-1 Surveillance Capsule Withdrawal Schedule Capsule ID and Capsule Lead Withdrawal Capsule Fluence Status(a)

Location Factor(a) EFPY(b, c) (n/cm2, E > 1.0 MeV)(c) 97° Withdrawn (EOC 4) 1.19 4.41 6.31E+18 263° Withdrawn (EOC 11) 1.19 13.83 1.45E+19 83° Withdrawn (EOC 19) 1.20 24.66 2.42E+19 277° In Reactor 1.20 48(d) 4.51E+19(d) 104° In Reactor 0.83 (e) (e) 284° In Reactor 0.83 (e) (e)

Notes:

(a) Updated in Capsule 83° dosimetry analysis; see Table 6-10.

(b) EFPY from plant startup.

(c) Updated in Capsule 83° dosimetry analysis; see Table 6-9.

(d) Capsule 277° should be withdrawn at the vessel refueling outage nearest to 48 EFPY of plant operation, which is when the fluence on the capsule will have reached the projected 60-year (55 EFPY) peak vessel fluence.

(e) Capsules 104° and 284° currently have a lead factor of less than one. If additional metallurgical data is needed for Waterford Unit 3, such as in support of a second license renewal to 80 total years of operation, relocation of one or both of these capsules to a higher lead factor location will be required. Since it is not known when or if Waterford Unit 3 will apply for a second license extension, and given that many cycles of irradiation will be required for Capsules 104° and 284° to accumulate fluence greater than the 80-year vessel wall fluence, it is suggested that a potential relocation decision be implemented prior to 40 total years of operation.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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. TR-C-MCS-001, Revision 0, Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Waterford-Unit 3 Reactor Vessel Materials, December 1977.
4. TR-C-MCS-002, Revision 0, Louisiana Power & Light Waterford Steam Electric Station Unit No. 3 Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program, October 1977.
5. C-NLM-003, Revision 1, Program for Irradiation Surveillance of Waterford Unit Three Reactor Vessel Materials, October 1974.
6. ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, ASTM, 1973.
7. Appendix G of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.
8. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM.
9. NUREG/CR-6413; ORNL/TM-13133, Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials, April 1996.
10. WCAP-16088-NP, Revision 2, Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation, June 2012.
11. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.
12. ASTM E23-12c, Standard Test Methods for Notched Bar Impact Testing of Metallic Materials, ASTM, 2012.
13. ASTM E2298-13a, Standard Test Method for Instrumented Impact Testing of Metallic Materials, ASTM, 2013.
14. ASTM A370-13, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 2013.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-2

15. ASTM E8/E8M-13a, Standard Test Methods for Tension Testing of Metallic Materials, ASTM, 2013.
16. ASTM E21-09, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, ASTM, 2009.
17. BAW-2177, Revision 01, Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station, Unit No. 3, February 2004.
18. WCAP-16002, Revision 0, Analysis of Capsule 263° from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program, March 2003.
19. ASTM E853-13, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, ASTM, 2013
20. 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.
21. 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.
22. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.
23. WCAP-16083-NP, Revision 1, Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry, April 2013.
24. 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, March 1996.
25. WCAP-18060-NP, Revision 1, Response to RAIs Concerning the Use of RAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate Fluence Evaluations, November 2015.
26. LTR-REA-17-75-NP, Revision 2, Evaluation per NEI 96-07 Section 4.3.8.2 for Changing from the use of the DORT Code to the RAPTOR-M3G Code at Waterford Unit 3, November 10, 2017.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 Waterford Unit 3 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 three surveillance capsules analyzed to date as part of the Waterford Unit 3 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 (EFPY)

Location 97° End of Cycle 4 4.41 263° End of Cycle 11 13.83 83° End of Cycle 19 24.66 The passive neutron sensors included in the evaluations of surveillance Capsules 97°, 263°, and 83° are summarized as follows:

Reaction Of Capsule Capsule Capsule Sensor Material Interest 97° 263° 83° 63 Copper (Cd) Cu(n,)60Co X X X 46 Titanium Ti(n,p)46Sc X X X 54 54 Iron Fe(n,p) Mn X X X 58 Nickel (Cd) Ni(n,p)58Co X X X 238 Uranium-238* U(n,f)FP X X X 59 60 Cobalt-Aluminum* Co(n,) Co X X X Note:

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

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-2 The capsules also contained sulfur monitors, which were not analyzed because of the short half-life of the activation product isotope (32P, 14.3 days). Pertinent physical and nuclear characteristics of the passive neutron sensors analyzed are listed in Table A-1 for Capsule 97°, and Table A-2 for Capsules 263° and 83°.

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.

The radiometric counting of the sensors from Capsule 83° 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 97°, 263°, and 83° was based on the monthly power generation of Waterford Unit 3 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 97°,

263°, and 83° is given in Table 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 ]

Pref 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.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-3 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 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-4 and A-5, respectively, for Capsules 97° and 83°. These fluence rates represent the capsule- and cycle-dependent results at the radial and azimuthal center of the respective capsules at core midplane. For Capsule 263°, which was removed at the conclusion of Cycle 11, it was noticed that the peripheral assembly closest to the cardinal axis for Cycle 11 had an average relative power significantly lower compared to the other cycles. It was decided to split Cycle 11 into beginning-of-cycle (BOC), middle-of-cycle (MOC), and end-of-cycle (EOC) segments. This has an impact on the calculations performed via the Cj ratios that account for changes in the sensor reaction rates due to variations in the flux level induced by changes in the spatial power distribution. The Cj terms were based on the individual reaction rates determined from Cycles 1 though the end of Cycle 11 at the 263° surveillance capsule location.

Reaction rates in the 263° capsule location, and the Cj terms from Cycles 1 through 11 are given in Tables A-6 and A-7, respectively. These reaction rates represent the capsule- and cycle-dependent results at the radial and azimuthal center of the respective capsules at core midplane.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-4 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 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 Waterford Unit 3 fission sensor reaction rates are summarized as follows:

Correction Capsule 97° Capsule Capsule 83° 263° 235 U Impurity/Pu Build-0.8599 0.8273 0.7955 in 238 U(,f) 0.8705 0.8739 0.8745 Net 238U Correction 0.7485 0.7230 0.6957 The correction factors for Capsules 97°, 263° and 83° 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 97°, 263°, and 83°, are given in Tables A-8 through A-10. In Tables A-8 through A-10, 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 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 Waterford Unit 3 surveillance capsule dosimetry, the FERRET code [Ref. A-2] was employed to combine the results of the plant-specific neutron transport calculations WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-5 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 three 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 Waterford Unit 3 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-3].

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, Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance [Ref. A-4].

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Waterford Unit 3 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:

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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

46 Ti(n,p)46Sc 5%

54 54 Fe(n,p) Mn 5%

58 Ni(n,p)58Co 5%

238 U(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 Waterford Unit 3 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%

46 Ti(n,p)46Sc 4.50-4.87%

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

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

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

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-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 Waterford Unit 3 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 ()

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-8 (E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 A.1.3 Comparisons of Measurements and Calculations Results of the least-squares evaluations of the dosimetry from the Waterford Unit 3 surveillance capsules withdrawn to date are provided in Tables A-11, A-12, and A-13 for Capsules 97°, 263°, and 83°,

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.

For Capsule 97°, the titanium monitor was discarded. For all three capsules, both bare and cadmium-covered uranium monitors were discarded. These dosimetry data were discarded because they were outside the expected values.

The data comparisons provided in Tables A-11 through A-13 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-14 and A-15.

These comparisons are given on two levels. In Table A-14, 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-15, calculations of fast neutron exposure rates in terms of 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 1.07 to 1.17 in the data set. The overall average M/C ratio for the entire set of Waterford Unit 3 data is 1.11 with an associated standard deviation of 7.0%.

In the comparisons of best-estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 1.04 to 1.12 for neutron fluence rate (E > 1.0 MeV) and from 1.05 to 1.11 for iron atom displacement rate. The overall average BE/C ratios WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-9 for neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are 1.08 with a standard deviation of 3.7% and 1.08 with a standard deviation of 2.8%, 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 Waterford Unit 3 reactor pressure vessel.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-10 Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors of Surveillance Capsule 97° 90%

Reaction Target Product Fission Atomic Weight Response of Atom Half-life Yield (g/g-atom) Range(a)

Interest Fraction (days) (%)

(MeV) 63 Cu (n,) 60Co 62.9296 1.0 1925.5 n/a 5.0 - 12.0 46 46 Ti (n,p) Sc 45.9526 1.0 83.79 n/a 4.1 - 10.5 54 54 Fe (n,p) Mn 53.9396 1.0 312.11 n/a 2.4 - 8.8 58 58 Ni (n,p) Co 57.9353 1.0 70.82 n/a 2.1 - 8.7 238 137 U (n,f) Cs 238.051 1.0 10983.07 6.02 1.5 - 7.9 59 60 Co (n,) Co 58.933 1.0 1925.5 n/a non-threshold Note:

(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Waterford Unit 3 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.

Table A-2 Nuclear Parameters Used in the Evaluation of Neutron Sensors of Surveillance Capsules 263° and 83° 90%

Reaction Target Product Fission Atomic Weight Response of Atom Half-life Yield (g/g-atom) Range(a)

Interest Fraction (days) (%)

(MeV) 63 Cu (n,) 60Co 63.546 0.6917 1925.5 n/a 5.0 - 12.0 46 46 Ti (n,p) Sc 47.867 0.0825 83.79 n/a 4.1 - 10.5 54 54 Fe (n,p) Mn 55.845 0.05845 312.11 n/a 2.4 - 8.8 58 58 Ni (n,p) Co 58.693 0.68077 70.82 n/a 2.1 - 8.8 238 137 U (n,f) Cs 238.051 1.0 10983.07 6.02 1.5 - 8.0 59 60 Co (n,) Co 58.933 0.0017 1925.5 n/a non-threshold Note:

(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Waterford Unit 3 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.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-11 Table A-3 Monthly Thermal Generation during the First 19 Fuel Cycles of the Waterford Unit 3 Reactor Cycle 1 Cycle 2 Cycle 3 Cycle 4 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Mar-85 42830 Dec-86 0 May-88 3254 Oct-89 0 Apr-85 613673 Jan-87 0 Jun-88 1934985 Nov-89 640027 May-85 809488 Feb-87 1379459 Jul-88 2304929 Dec-89 2444128 Jun-85 198642 Mar-87 2149043 Aug-88 2509305 Jan-90 1787064 Jul-85 846176 Apr-87 2257170 Sep-88 2107712 Feb-90 1612051 Aug-85 0 May-87 2278405 Oct-88 1509724 Mar-90 2271449 Sep-85 317589 Jun-87 2426725 Nov-88 1052050 Apr-90 2433315 Oct-85 1581557 Jul-87 2489860 Dec-88 2282880 May-90 2499444 Nov-85 2319818 Aug-87 2332754 Jan-89 2231876 Jun-90 2434503 Dec-85 1423475 Sep-87 1406064 Feb-89 2264143 Jul-90 2508532 Jan-86 2313146 Oct-87 1734107 Mar-89 2458113 Aug-90 2333299 Feb-86 2236098 Nov-87 2418833 Apr-89 2418581 Sep-90 2363915 Mar-86 551214 Dec-87 2224789 May-89 2509484 Oct-90 1728843 Apr-86 2380512 Jan-88 2143917 Jun-89 2390707 Nov-90 2433754 May-86 2337147 Feb-88 2339100 Jul-89 2292538 Dec-90 2516099 Jun-86 2345202 Mar-88 2298501 Aug-89 2249799 Jan-91 2515171 Jul-86 1646157 Apr-88 76267 Sep-89 1783533 Feb-91 2211690 Aug-86 2497833 Mar-91 1196293 Sep-86 2198347 Oct-86 2206239 Nov-86 2009429 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-12 Table A-3 (Continued) Monthly Thermal Generation during the First 19 Fuel Cycles of the Waterford Unit 3 Reactor Cycle 5 Cycle 6 Cycle 7 Cycle 8 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Apr-91 0 Oct-92 0 Apr-94 433275 Oct-95 0 May-91 163566 Nov-92 1581907 May-94 2515334 Nov-95 1929688 Jun-91 2275493 Dec-92 2505327 Jun-94 2306353 Dec-95 2517767 Jul-91 2453834 Jan-93 2514707 Jul-94 2518426 Jan-96 2517086 Aug-91 2389958 Feb-93 2271376 Aug-94 2514659 Feb-96 2357937 Sep-91 2435284 Mar-93 2373995 Sep-94 2430858 Mar-96 2516925 Oct-91 2522290 Apr-93 2433762 Oct-94 2518011 Apr-96 2436647 Nov-91 2256967 May-93 2518857 Nov-94 2439425 May-96 2326692 Dec-91 2500136 Jun-93 2310136 Dec-94 2515464 Jun-96 2440018 Jan-92 2517213 Jul-93 2513308 Jan-95 2517937 Jul-96 1247706 Feb-92 1649550 Aug-93 2517897 Feb-95 2273361 Aug-96 2105079 Mar-92 2284044 Sep-93 2431720 Mar-95 2516977 Sep-96 2396700 Apr-92 2429922 Oct-93 2519890 Apr-95 2436285 Oct-96 2471709 May-92 2418613 Nov-93 2437790 May-95 2518637 Nov-96 2295324 Jun-92 2436179 Dec-93 2517986 Jun-95 839448 Dec-96 2449101 Jul-92 2400380 Jan-94 2517230 Jul-95 2521005 Jan-97 2520469 Aug-92 2476631 Feb-94 2273776 Aug-95 2521249 Feb-97 2273932 Sep-92 1417812 Mar-94 318321 Sep-95 1774478 Mar-97 2520600 Apr-97 886717 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-13 Table A-3 (Continued) Monthly Thermal Generation during the First 19 Fuel Cycles of the Waterford Unit 3 Reactor Cycle 9 Cycle 10 Cycle 11 Cycle 12 Month MWt-h Month MWt-h Month MWt-h Month MWt-h May-97 0 Mar-99 0 Nov-00 1030289 Apr-02 980490 Jun-97 0 Apr-99 2183602 Dec-00 2517757 May-02 2508388 Jul-97 99594 May-99 2513051 Jan-01 2521026 Jun-02 2475037 Aug-97 2505653 Jun-99 2244073 Feb-01 2121310 Jul-02 2554918 Sep-97 2439155 Jul-99 2521175 Mar-01 2520656 Aug-02 2558472 Oct-97 2523657 Aug-99 1808465 Apr-01 2436332 Sep-02 2472369 Nov-97 2433511 Sep-99 880564 May-01 2518333 Oct-02 2561743 Dec-97 2433511 Oct-99 2524076 Jun-01 2367725 Nov-02 2471286 Jan-98 2513765 Nov-99 2080135 Jul-01 2521008 Dec-02 2554752 Feb-98 2276816 Dec-99 2370546 Aug-01 2521122 Jan-03 2556532 Mar-98 2520985 Jan-00 2513887 Sep-01 2434974 Feb-03 1972402 Apr-98 2436390 Feb-00 2358614 Oct-01 2520999 Mar-03 2558500 May-98 2503708 Mar-00 2346459 Nov-01 2439796 Apr-03 2470264 Jun-98 2439738 Apr-00 2436564 Dec-01 2520997 May-03 2558445 Jul-98 2283997 May-00 2521322 Jan-02 2518017 Jun-03 2470519 Aug-98 2521163 Jun-00 1873406 Feb-02 2277045 Jul-03 2558736 Sep-98 1029804 Jul-00 2521097 Mar-02 1704398 Aug-03 2558749 Oct-98 2456652 Aug-00 2521192 Sep-03 2406613 Nov-98 1396144 Sep-00 2436861 Oct-03 1282988 Dec-98 2262325 Oct-00 1051538 Jan-99 2486638 Feb-99 1291559 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-14 Table A-3 (Continued) Monthly Thermal Generation during the First 19 Fuel Cycles of the Waterford Unit 3 Reactor Cycle 13 Cycle 14 Cycle 15 Cycle 16 Month MWt-h Month MWt-h Month MWt-h Month MWt-h Nov-03 680404 May-05 0 Dec-06 306512 May-08 0 Dec-03 2558425 Jun-05 1636426 Jan-07 2762347 Jun-08 2504571 Jan-04 2558662 Jul-05 2762118 Feb-07 2494923 Jul-08 2760325 Feb-04 2346194 Aug-05 2450061 Mar-07 2758724 Aug-08 2752042 Mar-04 2558479 Sep-05 1517824 Apr-07 2673470 Sep-08 1805293 Apr-04 2472647 Oct-05 2766811 May-07 2762294 Oct-08 2762806 May-04 2556502 Nov-05 2409150 Jun-07 2673734 Nov-08 2677040 Jun-04 2476175 Dec-05 2762652 Jul-07 2762850 Dec-08 2761940 Jul-04 2558700 Jan-06 2755253 Aug-07 2762946 Jan-09 2762120 Aug-04 2558734 Feb-06 2490376 Sep-07 2673647 Feb-09 2489307 Sep-04 2474747 Mar-06 2762655 Oct-07 1484298 Mar-09 2757710 Oct-04 2561708 Apr-06 2669816 Nov-07 2674092 Apr-09 2673058 Nov-04 2474159 May-06 2759553 Dec-07 2761656 May-09 2761585 Dec-04 2558590 Jun-06 2673604 Jan-08 2762651 Jun-09 2673378 Jan-05 2558462 Jul-06 2762880 Feb-08 2584461 Jul-09 2762385 Feb-05 2309099 Aug-06 2736791 Mar-08 2758941 Aug-09 2762163 Mar-05 2558504 Sep-06 2675386 Apr-08 2314440 Sep-09 2664994 Apr-05 1316652 Oct-06 2766864 Oct-09 1624218 Nov-06 2211054 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-15 Table A-3 (Continued) Monthly Thermal Generation during the First 19 Fuel Cycles of the Waterford Unit 3 Reactor Cycle 17 Cycle 18 Cycle 19 Month MWt-h Month MWt-h Month MWt-h Nov-09 0 May-11 1577376 Nov-12 0 Dec-09 2384615 Jun-11 2673136 Dec-12 0 Jan-10 2762121 Jul-11 2762028 Jan-13 838638 Feb-10 2495012 Aug-11 2760555 Feb-13 2493392 Mar-10 2758464 Sep-11 2672753 Mar-13 2759429 Apr-10 2673197 Oct-11 2762209 Apr-13 2319009 May-10 2759349 Nov-11 2676675 May-13 2557806 Jun-10 2672667 Dec-11 2760601 Jun-13 2673766 Jul-10 2762032 Jan-12 2651616 Jul-13 2753495 Aug-10 2761860 Feb-12 2582768 Aug-13 2761633 Sep-10 2672841 Mar-12 2754454 Sep-13 2666734 Oct-10 2762116 Apr-12 2626514 Oct-13 2754529 Nov-10 2676410 May-12 2761137 Nov-13 2675437 Dec-10 2760245 Jun-12 2659648 Dec-13 2759168 Jan-11 2762027 Jul-12 2761798 Jan-14 2764348 Feb-11 2377290 Aug-12 2422281 Feb-14 2494353 Mar-11 2350801 Sep-12 2413466 Mar-14 2757639 Apr-11 378111 Oct-12 1403317 Apr-14 1067533 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-16 Table A-4 Surveillance Capsule 97° and 83° Fluence Rates for Cj Calculation, Core Midplane Elevation Flux (E > 1.0 MeV) [n/cm2-s]

Cycle Fuel Cycle Length Capsule 97° Capsule 83° (EFPY) 1 1.04 5.63E+10 5.63E+10 2 1.01 4.41E+10 4.41E+10 3 1.15 4.38E+10 4.38E+10 4 1.21 3.84E+10 3.84E+10 5 1.25 3.98E+10 6 1.30 3.90E+10 7 1.35 2.13E+10 8 1.35 2.65E+10 9 1.44 2.55E+10 10 1.40 2.39E+10 11 1.33 1.87E+10 12 1.48 2.43E+10 13 1.40 2.56E+10 14 1.40 3.20E+10 15 1.29 3.26E+10 16 1.35 2.99E+10 17 1.31 2.80E+10 18 1.40 3.15E+10 19 1.20 2.26E+10 Average - 4.53E+10 3.11E+10 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-17 Table A-5 Surveillance Capsule 97° and 83° Cj Factors, Core Midplane Elevation Cycle Cj Length Fuel Cycle (EFPY) Capsule 97° Capsule 83° 1 1.04 1.24 1.81 2 1.01 0.97 1.42 3 1.15 0.97 1.41 4 1.21 0.85 1.24 5 1.25 1.28 6 1.30 1.25 7 1.35 0.68 8 1.35 0.85 9 1.44 0.82 10 1.40 0.77 11 1.33 0.60 12 1.48 0.78 13 1.40 0.82 14 1.40 1.03 15 1.29 1.05 16 1.35 0.96 17 1.31 0.90 18 1.40 1.01 19 1.20 0.73 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-18 Table A-6 Surveillance Capsule 263° Reaction Rates for Cj Calculation, Core Midplane Elevation Reaction Rate (rps/atom)

Cycle 63 46 54 58 238 59 59 Cu(n,) Ti(n,p) Fe(n,p) Ni(n,p) U(n,f) Co(n,) Co(Cd)(n,)

1 7.75E+07 1.31E+09 7.16E+09 9.36E+09 2.49E+10 3.02E+12 6.22E+11 2 6.26E+07 1.05E+09 5.69E+09 7.42E+09 1.96E+10 2.32E+12 4.82E+11 3 6.24E+07 1.05E+09 5.65E+09 7.37E+09 1.95E+10 2.30E+12 4.78E+11 4 5.50E+07 9.23E+08 4.97E+09 6.48E+09 1.71E+10 2.01E+12 4.19E+11 5 5.69E+07 9.56E+08 5.15E+09 6.72E+09 1.77E+10 2.09E+12 4.35E+11 6 5.62E+07 9.43E+08 5.07E+09 6.61E+09 1.74E+10 2.02E+12 4.21E+11 7 3.25E+07 5.37E+08 2.83E+09 3.68E+09 9.54E+09 1.07E+12 2.25E+11 8 4.00E+07 6.63E+08 3.50E+09 4.57E+09 1.19E+10 1.34E+12 2.81E+11 9 3.90E+07 6.45E+08 3.39E+09 4.42E+09 1.14E+10 1.28E+12 2.70E+11 10 3.69E+07 6.09E+08 3.20E+09 4.16E+09 1.07E+10 1.20E+12 2.52E+11 11 BOC 2.78E+07 4.58E+08 2.40E+09 3.12E+09 8.04E+09 8.91E+11 1.88E+11 11 2.90E+07 4.77E+08 2.50E+09 3.25E+09 8.39E+09 9.33E+11 1.97E+11 11 EOC 3.28E+07 5.40E+08 2.84E+09 3.70E+09 9.56E+09 1.07E+12 2.25E+11 Average 4.68E+07 7.82E+08 4.18E+09 5.45E+09 1.43E+10 1.66E+12 3.46E+11 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-19 Table A-7 Surveillance Capsule 263° Cj Factors, Core Midplane Elevation Cj Cycle 63 46 54 58 238 59 59 Cu(n,) Ti(n,p) Fe(n,p) Ni(n,p) U(n,f) Co(n,) Co(Cd)(n,)

1 1.66 1.68 1.71 1.72 1.74 1.82 1.80 2 1.34 1.35 1.36 1.36 1.37 1.40 1.40 3 1.33 1.34 1.35 1.35 1.36 1.39 1.38 4 1.18 1.18 1.19 1.19 1.20 1.21 1.21 5 1.22 1.22 1.23 1.23 1.24 1.26 1.26 6 1.20 1.21 1.21 1.21 1.22 1.22 1.22 7 0.69 0.69 0.68 0.68 0.67 0.65 0.65 8 0.85 0.85 0.84 0.84 0.83 0.81 0.81 9 0.83 0.82 0.81 0.81 0.80 0.77 0.78 10 0.79 0.78 0.76 0.76 0.75 0.72 0.73 11 BOC 0.59 0.59 0.57 0.57 0.56 0.54 0.54 11 0.62 0.61 0.60 0.60 0.59 0.56 0.57 11 EOC 0.70 0.69 0.68 0.68 0.67 0.64 0.65 Average 1.00 1.00 1.00 1.00 1.00 1.00 1.00 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-20 Table A-8 Measured Sensor Activities and Reaction Rates for Surveillance Capsule 97° Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(1) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 3.15E+05 7.77E+05 8.12E-17 Cu-63 Co-60 2.93E+05 7.23E+05 7.55E-17 Cu-63 Co-60 3.22E+05 7.94E+05 8.30E-17 7.99E-17 7.99E-17 Ti-46 Sc-46 1.23E+07 1.53E+07 1.17E-15 Ti-46 Sc-46 1.31E+07 1.63E+07 1.24E-15 Ti-46 Sc-46 1.52E+07 1.89E+07 1.44E-15 1.28E-15 1.28E-15 Fe-54 Mn-54 5.41E+07 7.15E+07 6.40E-15 Fe-54 Mn-54 5.08E+07 6.71E+07 6.01E-15 Fe-54 Mn-54 5.35E+07 7.07E+07 6.33E-15 6.25E-15 6.25E-15 Ni-58 Co-58 6.69E+07 8.26E+07 7.95E-15 Ni-58 Co-58 6.54E+07 8.08E+07 7.77E-15 Ni-58 Co-58 7.14E+07 8.82E+07 8.48E-15 8.07E-15 8.07E-15 U-238 Cs-137 2.94E+05 3.09E+06 2.03E-14 U-238 Cs-137 2.94E+05 3.09E+06 2.03E-14 U-238 Cs-137 3.12E+05 3.28E+06 2.15E-14 2.07E-14 1.55E-14 Co-59 Co-60 1.61E+10 3.97E+10 3.89E-12 Co-59 Co-60 1.81E+10 4.47E+10 4.37E-12 Co-59 Co-60 1.44E+10 3.55E+10 3.48E-12 3.91E-12 3.91E-12 Co-59(Cd) Co-60 1.91E+09 4.71E+09 4.61E-13 Co-59(Cd) Co-60 1.64E+09 4.05E+09 3.96E-13 Co-59(Cd) Co-60 1.84E+09 4.54E+09 4.44E-13 4.34E-13 4.34E-13 Note:

1. Measured activity decay corrected to March 15, 1991 WCAP-17969-NP November 2017 Revision 2
      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-21 Table A-9 Measured Sensor Activities and Reaction Rates for Surveillance Capsule 263° Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(1) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.36E+05 3.66E+05 5.58E-17 Cu-63 Co-60 1.69E+05 2.62E+05 4.00E-17 Cu-63 Co-60 1.98E+05 3.07E+05 4.68E-17 4.76E-17 4.76E-17 Ti-46 Sc-46 2.22E+05 9.43E+05 9.08E-16 Ti-46 Sc-46 2.12E+05 9.00E+05 8.67E-16 Ti-46 Sc-46 2.06E+05 8.75E+05 8.43E-16 8.73E-16 8.73E-16 Fe-54 Mn-54 1.42E+06 2.98E+06 4.73E-15 Fe-54 Mn-54 1.36E+06 2.86E+06 4.53E-15 Fe-54 Mn-54 1.33E+06 2.79E+06 4.43E-15 4.56E-15 4.56E-15 Ni-58 Co-58 8.83E+06 4.57E+07 6.55E-15 Ni-58 Co-58 8.27E+06 4.28E+07 6.13E-15 Ni-58 Co-58 8.31E+06 4.30E+07 6.16E-15 6.28E-15 6.28E-15 U-238 Cs-137 2.45E+05 9.21E+05 6.05E-15 U-238 Cs-137 6.76E+04 2.54E+05 1.67E-15 U-238 Cs-137 1.28E+05 4.81E+05 3.16E-15 3.63E-15 2.62E-15 Co-59 Co-60 2.45E+07 3.93E+07 2.26E-12 Co-59 Co-60 2.41E+07 3.87E+07 2.23E-12 Co-59 Co-60 1.96E+07 3.14E+07 1.81E-12 2.10E-12 2.10E-12 Co-59(Cd) Co-60 3.04E+06 4.85E+06 2.79E-13 Co-59(Cd) Co-60 3.11E+06 4.97E+06 2.86E-13 Co-59(Cd) Co-60 3.04E+06 4.85E+06 2.79E-13 2.82E-13 2.82E-13 Note:

1. Measured activity decay corrected to July 25, 2002 WCAP-17969-NP November 2017 Revision 2
      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-22 Table A-10 Measured Sensor Activities and Reaction Rates for Surveillance Capsule 83° Radially Corrected Average Measured Corrected Reaction Average Target Product Reaction Activity Saturated Rate Reaction Isotope Isotope Rate (dps/g)(1) Activity (rps/atom) Rate (rps/atom)

(dps/g) (rps/atom)

Cu-63 Co-60 2.21E+05 3.08E+05 4.70E-17 Cu-63 Co-60 2.00E+05 2.79E+05 4.25E-17 Cu-63 Co-60 2.63E+05 3.67E+05 5.59E-17 4.85E-17 4.85E-17 Ti-46 Sc-46 8.70E+04 9.37E+05 9.03E-16 Ti-46 Sc-46 7.85E+04 8.46E+05 8.15E-16 Ti-46 Sc-46 8.17E+04 8.80E+05 8.48E-16 8.55E-16 8.55E-16 Fe-54 Mn-54 1.25E+06 2.97E+06 4.71E-15 Fe-54 Mn-54 1.14E+06 2.71E+06 4.30E-15 Fe-54 Mn-54 1.19E+06 2.83E+06 4.49E-15 4.50E-15 4.50E-15 Ni-58 Co-58 2.90E+06 4.52E+07 6.47E-15 Ni-58 Co-58 2.66E+06 4.15E+07 5.94E-15 Ni-58 Co-58 2.83E+06 4.41E+07 6.32E-15 6.24E-15 6.24E-15 U-238 Cs-137 6.82E+05 1.70E+06 1.12E-14 U-238 Cs-137 1.63E+06 4.07E+06 2.67E-14 U-238 Cs-137 5.14E+05 1.28E+06 8.42E-15 1.54E-14 1.07E-14 Co-59 Co-60 2.61E+07 3.64E+07 2.10E-12 Co-59 Co-60 2.34E+07 3.26E+07 1.88E-12 Co-59 Co-60 1.99E+07 2.77E+07 1.60E-12 1.86E-12 1.86E-12 Co-59(Cd) Co-60 3.09E+06 4.31E+06 2.48E-13 Co-59(Cd) Co-60 3.24E+06 4.52E+06 2.60E-13 Co-59(Cd) Co-60 3.18E+06 4.43E+06 2.55E-13 2.54E-13 2.54E-13 Notes:

1. Measured activity decay corrected to December 15, 2014 WCAP-17969-NP November 2017 Revision 2
      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-23 Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 97° (7-Degree Azimuth, Core Midplane) 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 7.99E-17 6.40E-17 7.61E-17 1.25 1.05 1.19 Fe-54(n,p)Mn-54 6.25E-15 5.82E-15 6.36E-15 1.07 0.98 1.09 Ni-58(n,p)Co-58 8.07E-15 7.61E-15 8.26E-15 1.06 0.98 1.09 Co-59(n,)Co-60 3.91E-12 2.74E-12 3.89E-12 1.42 1.00 1.42 Co-59(Cd)(n,)Co-60 4.34E-13 4.74E-13 4.37E-13 0.91 0.99 0.92 Average 1.13 1.00 1.12

% standard deviation 9.5 4.0 5.1 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 4.53E+10 13 4.74E+10 7 1.04 (n/cm2-s)

Fluence rate E > 0.1 MeV 8.64E+10 - 8.91E+10 9 1.03 (n/cm2-s) dpa/s 6.54E-11 13 6.92E-11 6 1.05 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-24 Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 263° (7-Degree Azimuth, Core Midplane) Cycles 1 Through 11 Irradiation Reaction Rate (rps/atom)

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

(BE)

Cu-63(n,)Co-60 4.75E-17 4.85E-17 4.99E-17 0.98 1.05 1.03 Ti-46(n,p)Sc-46 8.73E-16 7.60E-16 8.30E-16 1.15 0.95 1.09 Fe-54(n,p)Mn-54 4.56E-15 4.33E-15 4.65E-15 1.05 1.02 1.07 Ni-58(n,p)Co-58 6.28E-15 5.65E-15 6.15E-15 1.11 0.98 1.09 Co-59(n,)Co-60 2.10E-12 1.97E-12 2.10E-12 1.06 1.00 1.06 Co-59(Cd)(n,)Co-60 2.81E-13 3.43E-13 2.84E-13 0.82 1.01 0.83 Average 1.07 1.00 1.07

% standard deviation 6.9 4.4 2.6 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 3.32E+10 13 3.60E+10 7 1.08 (n/cm2-s)

Fluence rate E > 0.1 MeV 6.31E+10 - 6.70E+10 9 1.06 (n/cm2-s) dpa/s 4.80E-11 13 5.16E-11 6 1.07 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-25 Table A-13 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 83° (7-Degree Azimuth, Core Midplane) Cycles 1 Through 19 Irradiation Reaction Rate (rps/atom)

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

(BE)

Cu-63(n,)Co-60 4.85E-17 4.60E-17 5.02E-17 1.05 0.96 1.09 Ti-46(n,p)Sc-46 8.55E-16 7.19E-16 8.22E-16 1.19 1.04 1.14 Fe-54(n,p)Mn-54 4.50E-15 4.07E-15 4.59E-15 1.10 0.98 1.13 Ni-58(n,p)Co-58 6.24E-15 5.32E-15 6.08E-15 1.17 1.03 1.14 Co-59(n,)Co-60 1.86E-12 1.83E-12 1.86E-12 1.02 1.00 1.02 Co-59(Cd)(n,)Co-60 2.54E-13 3.19E-13 2.57E-13 0.80 0.99 0.81 Average 1.13 1.00 1.13

% standard deviation 5.7 3.9 2.1 Best-Calculated Integral Quantity  % Unc. Estimate  % Unc. BE/C (C)

(BE)

Fluence rate E > 1.0 MeV 3.11E+10 13 3.50E+10 7 1.12 (n/cm2-s)

Fluence rate E > 0.1 MeV 5.89E+10 - 6.46E+10 9 1.09 (n/cm2-s) dpa/s 4.49E-11 13 5.02E-11 6 1.11 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-26 Table A-14 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron Threshold Reactions M/C Capsule 63 46 54 58 238 Cu(n,) Ti(n,p) Fe(n,p) Ni(n,p) U(n,f) 97° 1.25 - 1.07 1.06 -

263° 0.98 1.15 1.05 1.11 -

83° 1.05 1.19 1.10 1.17 -

Average 1.09 1.17 1.07 1.11 -

% Standard 12.8 2.4 2.3 4.9 -

Deviation Average 1.11

% Standard 7.0 Deviation Table A-15 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Capsule Neutron Fluence Rate Iron Atom Displacement Rate (E > 1.0 MeV) 97° 1.04 1.05 263° 1.08 1.07 83° 1.12 1.11 Average 1.08 1.08

% Standard deviation 3.7 2.8 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-27 A.1 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 A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-3 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.

A-4 ASTM Standard E944-13, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA), 2013.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS

  • 1XX denotes Lower Shell Plate M-1004-2, longitudinal orientation
  • 2XX denotes Lower Shell Plate M-1004-2, transverse orientation
  • 3XX denotes weld material
  • 4XX denotes heat affected zone material Note that the instrumented Charpy data is not required per ASTM Standards E185-82 or E23-12c.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-2 145: Tested at -25°F 15M: Tested at 0°F 15L: Tested at 5°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-3 14E: Tested at 10°F 13P: Tested at 20°F 12B: Tested at 30°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-4 12U: Tested at 40°F 114: Tested at 100°F 14J: Tested at 200°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-5 11C: Tested at 230°F 116: Tested at 250°F 115: Tested at 300°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-6 231: Tested at -50°F 21Y: Tested at -25°F 22K: Tested at -10°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-7 25L: Tested at 0°F 24L: Tested at 10°F 25Y: Tested at 25°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-8 23T: Tested at 40°F 261: Tested at 100°F 222: Tested at 150°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-9 21A: Tested at 200°F 225: Tested at 250°F 226: Tested at 300°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-10 337: Tested at -90°F 3A2: Tested at -70°F 31L: Tested at -65°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-11 31P: Tested at -60°F 34P: Tested at -55°F 325: Tested at -50°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-12 334: Tested at -30°F 35P: Tested at 0°F 347: Tested at 69°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-13 371: Tested at 100°F 37B: Tested at 150°F 312: Tested at 200°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-14 47M: Tested at -125°F 44K: Tested at -90°F 42E: Tested at -80°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-15 45U: Tested at -75°F 454: Tested at -70°F 45L: Tested at -60°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-16 45A: Tested at -50°F 41Y: Tested at 0°F 43J: Tested at 69°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-17 46K: Tested at 150°F 457: Tested at 200°F 472: Tested at 250°F WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 with 95% shear as the USE, excluding any values that are deemed outliers using engineering judgment. Hence, the Capsule 83° 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 97° and 263°, were also determined by applying the methodology described above to the Charpy impact data reported in TR-C-MCS-002, Revision 0 [Ref. C-2], BAW-2177, Revision 01

[Ref. C-3] and WCAP-16002, Revision 0 [Ref. C-4]. 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 values were fixed at 1.0 mils in order to be consistent with the previous capsule analysis [Ref. C-4].

Table C-1 Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH Capsule Material Unirradiated 97° 263° 83° Lower Shell Plate M-1004-2 170 155 --- 158 Longitudinal Orientation Lower Shell Plate M-1004-2 141 124 131 138 Transverse Orientation Surveillance Program Weld Material 156 154 145 133 (Heat # 88114)

Heat Affected Zone (HAZ) Material 170 156 163 158 Standard Reference Material (SRM) 133 --- 113 ---

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

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-2 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 TR-C-MCS-002, Revision 0, Louisiana Power & Light Waterford Steam Electric Station Unit No. 3 Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program, October 1977.

C-3 BAW-2177, Revision 01, Analysis of Capsule W-97 Entergy Operations, Inc. Waterford Generating Station, Unit No. 3, February 2004.

C-4 WCAP-16002, Revision 0, Analysis of Capsule 263° from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program, March 2003.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-3 C.3 CVGRAPH VERSION 6.0 INDIVIDUAL PLOTS WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-4 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-5 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-6 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-7 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-8 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-9 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-10 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-11 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-12 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-13 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-14 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-15 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-16 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-17 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-18 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-19 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-20 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-21 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-22 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-23 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-24 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-25 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-26 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-27 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-28 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-29 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-30 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-31 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-32 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-33 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-34 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-35 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-36 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-37 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-38 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-39 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-40 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-41 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-42 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-43 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-44 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-45 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-46 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-47 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-48 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-49 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-50 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-51 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-52 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-53 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-54 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-55 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-56 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-57 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-58 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-59 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-60 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-61 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-62 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-63 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-64 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-65 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-66 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-67 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-68 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-69 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-70 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-71 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-72 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-73 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-74 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-75 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-76 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-77 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-78 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-79 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-80 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-81 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-82 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-83 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-84 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-85 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-86 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-87 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-88 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-89 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-90 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-91 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-92 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-93 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-94 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-95 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-96 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-97 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-98 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-99 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-100 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-101 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-102 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-103 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 C-104 WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D WATERFORD UNIT 3 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 three surveillance capsules removed and tested from the Waterford Unit 3 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 Waterford Unit 3 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.

The Waterford Unit 3 reactor vessel beltline region consists of the following materials:

1. Intermediate Shell Plates M-1003-1, 2, and 3
2. Lower Shell Plates M-1004-1, 2, and 3
3. Intermediate Shell Longitudinal Welds (Heat # BOLA & HODA)
4. Lower Shell Longitudinal Welds (Heat # 83653, Flux Type Linde 0091)
5. Intermediate to Lower Shell Circumferential Weld (Heat # 88114, Flux Type Linde 0091)

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-2 Per WCAP-16002, Revision 0 [Ref. D-3], the Waterford Unit 3 surveillance program was developed to the requirements of ASTM E185-73. At the time of the surveillance program development, all of the beltline plates were considered in terms of irradiation embrittlement through end of life. Of all the beltline plates, Lower Shell Plate M-1004-2 was foreseen to be the most limiting plate. This is largely due to its initial RTNDT that is significantly greater than the other beltline plates. The chemistry values (Cu and Ni weight percent) and initial upper-shelf energy values for the beltline plates are relatively consistent. No plate is clearly differentiated from the rest by its high copper or nickel content or low upper-shelf energy. Therefore, Lower Shell Plate M-1004-2 was selected as the plate material for the surveillance program.

The beltline welds all have low copper content. Since Intermediate to Lower Shell Circumferential Weld 101-171 (Heat # 88114, Flux Type Linde 0091) has the highest copper content in comparison to the other beltline welds, it was selected for the surveillance program. Lastly, selection of the beltline circumferential weld is consistent with the general practice for Combustion Engineering surveillance programs because it was considered representative material.

Based on the discussion above, Criterion 1 is met for the Waterford Unit 3 surveillance program.

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

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the USE of the Waterford Unit 3 surveillance materials unambiguously.

Hence, Criterion 2 is met for the Waterford Unit 3 surveillance program.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-3 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 should normally 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 USE 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 Waterford Unit 3 surveillance plate and weld material.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-4 Case 1: Lower Shell Plate M-1004-2 and Weld Heat # 88114 Following the NRC Case 1 guidelines, the Waterford Unit 3 surveillance plate and weld metal (Heat # 88114) will be evaluated using the Waterford Unit 3 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 Waterford Unit 3 data is being considered; therefore, no temperature adjustment is required.

Table D-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation using Waterford Unit 3 Surveillance Capsule Data Capsule Fluence RTNDT FF*RTNDT Material Capsule FF FF2 (x 1019 n/cm2, E > 1.0 MeV) (°F) (°F)

Lower Shell Plate 97° 0.631 0.871 6.1 5.31 0.759 M-1004-2 (Longitudinal) 83° 2.42 1.238 13.6 16.84 1.533 97° 0.631 0.871 28.0 24.39 0.759 Lower Shell Plate M-1004-2 263° 1.45 1.103 -9.1(a) -10.04 1.217 (Transverse) 83° 2.42 1.238 25.3 31.32 1.533 SUM: 67.82 5.799 CF M-1004-2 = (FF

  • RTNDT) ÷ (FF2) = (67.82) ÷ (5.799) = 11.7°F 97° 0.631 0.871 23.5 20.47 0.759 Surveillance Weld Material 263° 1.45 1.103 6.6 7.28 1.217 (Heat # 88114) 83° 2.42 1.238 19.0 23.52 1.533 SUM: 51.27 3.508 CF Surv. Weld = (FF
  • RTNDT) ÷ (FF2) = (51.27) ÷ (3.508) = 14.6°F Note for Table D-1:

(a) 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).

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 Waterford Unit 3 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 (Base Metal)

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

Lower Shell Plate 97° 11.7 0.631 0.871 6.1 10.2 4.1 Yes M-1004-2 (Longitudinal) 83° 11.7 2.42 1.238 13.6 14.5 0.9 Yes 97° 11.7 0.631 0.871 28.0 10.2 17.8 No Lower Shell Plate M-1004-2 263° 11.7 1.45 1.103 -9.1 12.9 22.0 No (Transverse) 83° 11.7 2.42 1.238 25.3 14.5 10.8 Yes 97° 14.6 0.631 0.871 23.5 12.7 10.8 Yes Surveillance Weld Material 263° 14.6 1.45 1.103 6.6 16.1 9.5 Yes (Heat # 88114) 83° 14.6 2.42 1.238 19.0 18.1 0.9 Yes From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table D-2 indicates that only three of the five surveillance data points fall inside the +/- 1 of 17°F scatter band for surveillance base metals; therefore, the plate data is deemed non-credible per the third criterion.

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

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

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 surveillance materials are contained in capsules positioned near the reactor vessel inside wall so that the irradiation conditions (fluence, flux spectrum, temperature) of the test specimens resemble, as closely as possible, the irradiation conditions of the reactor vessel. The capsules are bisected by the midplane of the core and are placed in capsule holders positioned circumferentially about the core at locations near the regions of maximum flux. 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 Waterford Unit 3 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 Waterford Unit 3 surveillance program does contain Standard Reference Material (SRM). The material was obtained from an A533 Grade B, Class 1 plate (HSST Plate 01). NUREG/CR-6413, ORNL/TM-13133 [Ref. D-6] contains a plot of Residual vs. Fast Fluence for the SRM (Figure 11 in the report). This Figure shows a 2 uncertainty of 50°F. The data used for this plot is contained in Table 14 in the report. However, the NUREG Report does not consider the recalculated fluence and RTNDT values for Capsule 263°. Thus, Table D-3 contains an updated calculation of Residual vs. Fast Fluence, considering the recalculated capsule fluence and RTNDT values for Capsule 263°.

Table D-3 Calculation of Residual vs. Fast Fluence for Waterford Unit 3 Capsule f Measured RG 1.99, Rev. 2 Residual(c)

Capsule (x1019 n/cm2, FF Shift(a) (F) Shift(b) (F) (F)

E > 1.0 MeV) 263° 1.45 1.103 150.5 150.1 0.4 Notes for Table D-3:

(a) Measured T30 value for the SRM was taken from Figure 5-13 of this report.

(b) Per NUREG/CR-6413, ORNL/TM-13133, the Cu and Ni values for the SRM (HSST Plate 01) are 0.18 and 0.66, respectively. This equates to a chemistry factor value of 136.1F based on Regulatory Guide 1.99, Revision 2, Position 1.1. The calculated shift is thus equal to CF

(c) Residual = Absolute Value [Measured Shift - RG 1.99 Shift].

Table D-3 shows a 2 uncertainty of less than 50F, which is the allowable scatter in NUREG/CR-6413, ORNL/TM-13133.

Hence, Criterion 5 is met for the Waterford Unit 3 surveillance program.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 D-7 D.3 CONCLUSION Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B:

  • The Waterford Unit 3 surveillance plate data are deemed non-credible
  • The Waterford Unit 3 surveillance weld data are deemed credible D.4 REFERENCES D-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1998.

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

D-3 Westinghouse Report WCAP-16002, Revision 0, Analysis of Capsule 263° from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program, March 2003.

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

D-6 NUREG/CR-6413; ORNL/TM-13133, Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials, J. A. Wang, Oak Ridge National Laboratory, Oak Ridge, TN, April 1996.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E WATERFORD UNIT 3 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 32 EFPY (end-of-license) 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 32 EFPY.

The Waterford Unit 3 reactor vessel beltline region minimum thickness is 8.625 inches. Calculation of the 1/4T vessel fluence values at 32 EFPY for the beltline materials is shown as follows:

Maximum Vessel Fluence @ 32 EFPY = 2.57 x 1019 n/cm2 (E > 1.0 MeV) 1/4T Fluence @ 32 EFPY = (2.57 x 1019 n/cm2)

  • e(-0.24 * (8.625 / 4))

= 1.53 x 1019 n/cm2 (E > 1.0 MeV)

The following pages present the Waterford Unit 3 upper-shelf energy evaluation. Figure E-1, as indicated above, is used in making predictions in accordance with Regulatory Guide 1.99, Revision 2. Table E-1 provides the predicted upper-shelf energy values for 32 EFPY (EOL).

Finally, the initial USE values have been updated in this report from the values documented in WCAP-16088-NP, Revision 2 [Ref. E-2], which were based on longitudinal Charpy data reduced by 65%. The updated values herein reflect actual measured transverse Charpy data for each of the five, non-surveillance, reactor vessel beltline plate materials. The initial USE values for the surveillance plate material, Lower Shell Plate M-1004-2, and all of the reactor vessel beltline weld materials remain unchanged from those documented in WCAP-16088-NP, Revision 2. This change was undertaken to better reflect the actual Charpy test results of the Waterford Unit 3 reactor vessel beltline plate materials.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-2 Limiting Plate Percent USE Decrease 12% from Capsule 97° Limiting Weld Percent USE Decrease (transverse-orientation) 15% from Capsule 83° 100.0

% Copper Base Metal Weld 0.35 0.30 0.30 0.25 0.25 0.20 Upper Limit 0.20 0.15 0.15 0.10 Plate 0.10 0.05 Line Percentage Drop in USE Weld Line 10.0 Surveillance Material: LS Plate M-1004-2 Surveillance Material:

Weld Heat # 88114 32 EFPY 1/4T Fluence = 1.53 x 1019 n/cm2 1.0 1.00E+17 1.00E+18 1.00E+19 1.00E+20 Neutron Fluence, n/cm2 (E > 1 MeV)

Figure E-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-3 Table E-1 Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 32 EFPY 1/4T EOL Unirradiated Projected Wt % Fluence Projected USE Material USE EOL USE(c)

Cu (x1019 n/cm2, Decrease (%)

(ft-lb) (ft-lb)

E > 1.0 MeV)

Position 1.2(a)

Intermediate Shell Plate M-1003-1 0.02 1.53 108 21 85 Intermediate Shell Plate M-1003-2 0.02 1.53 132 21 104 Intermediate Shell Plate M-1003-3 0.02 1.53 111 21 88 Lower Shell Plate M-1004-1 0.03 1.53 135 21 107 Lower Shell Plate M-1004-2 0.03 1.53 141 21 111 Lower Shell Plate M-1004-3 0.03 1.53 118 21 93 Intermediate Shell Longitudinal Weld 0.02 1.53 106 21 84 101-124A (Heat # BOLA & HODA)

Intermediate Shell Longitudinal Welds 0.02 1.53 131 21 103 101-124B & C (Heat # HODA)

Lower Shell Longitudinal Welds 0.03 1.53 129 21 102 101-142A, B & C (Heat # 83653)

Intermediate to Lower Shell Circumferential Weld 101-171 0.05 1.53 156 21 123 (Heat # 88114)

Position 2.2(b)

Lower Shell Plate M-1004-2 0.03 1.53 141 16 118 Intermediate to Lower Shell Circumferential Weld 101-171 0.05 1.53 156 14 134 (Heat # 88114)

Notes:

(a) Calculated using the Cu wt. % value and 1/4T fluence value for each material and Regulatory Guide 1.99, Revision 2, Position 1.2. In calculating the Position 1.2 percent USE decreases, the base metal and weld Cu weight percentages were conservatively rounded up to the lowest line (Cu weight % 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) The initial USE values for the five non-surveillance reactor vessel beltline plate materials have been updated from those documented in WCAP-16088-NP, Revision 2, which were based on longitudinal Charpy data reduced by 65%. The updated values herein reflect actual measured transverse Charpy data. The initial USE values for the surveillance plate material, Lower Shell Plate M-1004-2, and all of the reactor vessel beltline weld materials remain unchanged from those documented in WCAP-16088-NP, Revision 2.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-4 USE Conclusion As shown in Table E-1, all of the Waterford Unit 3 reactor vessel beltline materials are projected to remain above the USE screening criterion of 50 ft-lbs (per 10 CFR 50, Appendix G) at 32 EFPY.

E.2 REFERENCES E-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1998.

E-2 Westinghouse Report WCAP-16088-NP, Revision 2, Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation, June 2012.

WCAP-17969-NP November 2017 Revision 2

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

WCAP-17969-NP Revision 2 Proprietary Class 3

    • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Amiri Benjamin W Nov-13-2017 08:56:49 Author Approval Lynch Donald Nov-13-2017 12:40:33 Verifier Approval Mays Benjamin E Nov-13-2017 13:11:26 Verifier Approval Fischer Greg A Nov-14-2017 09:51:58 Manager Approval Patterson Lynn Nov-14-2017 10:48:38 Manager Approval Houssay Laurent Nov-14-2017 14:35:45 Final Approval Amiri Benjamin W Nov-15-2017 09:14:49 Files approved on Nov-15-2017

      • This record was final approved on 11/15/2017 9:14:49 AM. ( This statement was added by the PRIME system upon its validation)

ATTACHMENT 9.1 VENDOR DOCUMENT REVIEW STATUS Sheet 1 of 1 Entergy ENTERGY NUCLEAR MANAGEMENT MANUAL EN-DC-i 49 VENDOR DOCUMENT REVIEW STATUS FOR ACCEPTANCE LI FOR IN FORMATION LI IPEC LI PLP LI PNPS LI ANO LI GGNS LI RBS W3 LI NP Document No.: WCAP-17969-NP Rev. No. 2 Document

Title:

Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Surveillance Program EC No.: 71527 Purchase Order No.

(N/A or NP)

STATUS NO:

1. ACCEPTED, WORK MAY PROCEED
2. LI ACCEPTED AS NOTED RESUBMITTAL NOT REQUIRED, WORK MAY PROCEED
3. LI ACCEPTED AS NOTED RESUBMITTAL REQUIRED
4. LI NOT ACCEPTED Acceptance does not constitute approval of design details, calculations, analyses, test methods, or materials developed or selected by the supplier and does not relieve the supplier from full compliance with contractual negotiations.

Responsible Engineer Paul Hymel I --j, rz I)

(7 Print Name Signature Date Engineering Supervisor Nicholas Petit I Print Name Signature Date EN-DC-i 49 REV 12

Retrieved from "https://kanterella.com/w/index.php?title=W3F1-2019-0022,_Resubmittal_of_Reactor_Vessel_Material_Surveillance_Program_Capsule_Test_Results&oldid=2395479"