L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 1 of 3

WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 1 of 3

ML15154B077
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Site:	Saint Lucie
Issue date:	05/31/2015
From:	Alpan A, Long E J Westinghouse
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML15154B111	List: L-2015-160, St. Lucie, Unit 2 - Reactor Vessel Surveillance Capsule 97 Degree Report Submittal L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 1 of 3 L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 2 of 3 L-2015-160, WCAP-17939-NP, Revision 0, Analysis of Capsule 97 Degrees from the Florida Power & Light Company St. Lucie, Unit 2, Reactor Vessel Radiation Surveillance Program, Part 3 of 3 ML15154B075 ML15154B077 ML15154B079
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L-2015-160 WCAP-17939-NP, Rev. 0
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L-2015-160 Attachment Westinghouse Report WCAP-17939-NP, Rev 0, Analysis of Capsule 970 from the Florida Power & Light Company St. Lucie Unit 2 Reactor Vessel Radiation Surveillance Program (Following 253 Pages)

Westinghouse Non-Proprietary Class 3 WCAP-17939-NP Revision 0 May 2015 Analysis of Capsule 970 from the Florida Power & Light Company St. Lucie Unit 2 Reactor Vessel Radiation Surveillance Program fWestinghouse Westinghouse Non-Proprietary Class 3 WCAP-17939-NP Revision 0 Analysis of Capsule 970 from the Florida Power & Light Company St. Lucie Unit 2 Reactor Vessel Radiation Surveillance Program Elliot J. Long*Materials Center of Excellence Arzu Alpan*Radiation Engineering and Analysis May 2015 Reviewers:

Benjamin A. Rosier*Materials Center of Excellence Gregory A. Fischer*Radiation Engineering and Analysis Approved:

Frank C. Gift*, Manager Materials Center of Excellence 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© 2015 Westinghouse Electric Company LLC All Rights Reserved Westinghouse Non-Proprietary Class 3 ii TABLE OF CONTENTS L IST O F TA B L E S .......................................................................................................................................

iii L IST O F F IG U R E S ......................................................................................................................................

v EX EC U TIV E SU M M A RY ........................................................................................................................

viii 1 SU M M A R Y O F RESU LTS ..........................................................................................................

1-1 2 IN T R O D U C T IO N ........................................................................................................................

2-1 3 B A C K G R O U N D ..........................................................................................................................

3-1 4 DESCRIPTION OF PROGRAM .................................................................................................

4-1 5 TESTING OF SPECIMENS FROM CAPSULE 970 ...................................................................

5-1 5.1 O V E R V IE W .....................................................................................................................

5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS ...........................................................

5-2 5.3 TEN SILE TEST RESULTS .............................................................................................

5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

.................................................

6-1 6.1 IN T R O D U C TIO N ...........................................................................................................

6-1 6.2 DISCRETE ORDINATES ANALYSIS ...........................................................................

6-2 6.3 NEU TRO N D O SIM ETRY ..............................................................................................

6-4 6.4 CALCULATIONAL UNCERTAINTIES

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6-4 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE ............................................................

7-1 8 RE FE R E N C E S .............................................................................................................................

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 ST. LUCIE UNIT 2 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION..

D- 1 APPENDIX E ST. LUCIE UNIT 2 UPPER-SHELF ENERGY EVALUATION

...............................

E-1 WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 iii LIST OF TABLES Table 4-1 Chemical Composition (wt. %) of the St. Lucie Unit 2 Reactor Vessel Surveillance M aterials (U nirradiated)

...................................................................................................

4-3 Table 4-2 Arrangement of Encapsulated Test Specimens within St. Lucie Unit 2 Capsule 970 ...... 4-4 Table 5-1 Charpy V-notch Data for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 10 1 9 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation)

.................

5-5 Table 5-2 Charpy V-notch Data for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) (Transverse Orientation)

....................

5-6 Table 5-3 Charpy V-notch Data for the St. Lucie Unit 2 Surveillance Program Weld Metal (Heat #83637) Irradiated to a Fluence of 2.25 x 10'9 n/cm 2 (E > 1.0 MeV) ................................

5-7 Table 5-4 Charpy V-notch Data for the St. Lucie Unit 2 Heat-Affected Zone (HAZ) Material Irradiated to a Fluence of 2.25 x 10 1 9 n/cm 2 (E > 1.0 MeV) ............................................

5-8 Table 5-5 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV)(L ongitudinal O rientation)

...............................................................................................

5-9 Table 5-6 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 101 9 n/cm 2 (E > 1.0 MeV)(Transverse O rientation)

................................................................................................

5-10 Table 5-7 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Surveillance Program Weld Metal (Heat # 83637) Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV).......................

...................................................................................................................

5 -1 1 Table 5-8 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Heat-Affected Zone (HAZ) Material Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) ................

5-12 Table 5-9 Effect of Irradiation to 2.25 x 101 9 n/cm 2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the St. Lucie Unit 2 Reactor Vessel Surveillance Capsule 97'M aterials ........................................................................................................................

5-13 Table 5-10 Comparison of the St. Lucie Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper-Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, P red iction s .....................................................................................................................

5-14 Table 5-11 Tensile Properties of the St. Lucie Unit 2 Capsule 97' Reactor Vessel Surveillance Materials Irradiated to 2.25 x 1019 n/cm 2 (E > 1.0 MeV) ..............................................

5-15 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance C apsule C enter .................................................................................................................

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

.................................................................

6-8 Table 6-3 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from St. L ucie U n it 2 ..............................................................................................................

6-12 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 iv Table 6-4 Calculated Surveillance Capsule Lead Factors ............................................................

6-13 Table 7-1 Surveillance Capsule Withdrawal Schedule ..............................................................

7-1 Table A- I Nuclear Parameters Used in the Evaluation of Neutron Sensors .............................

A-10 Table A-2 Monthly Thermal Generation during the First 20 Fuel Cycles of the St. Lucie Unit 2 R eactor ..........................................................................................................................

A -lI Table A-3 Surveillance Capsule Fluence Rate for Cj Factors Calculation, Core Midplane Elevation......................................................................................................................................

A -16 Table A-4a Measured Sensor Activities and Reaction Rates for Surveillance Capsule 830 ............

A-18 Table A-4b Measured Sensor Activities and Reaction Rates for Surveillance Capsule 2630 ..........

A-19 Table A-4c Measured Sensor Activities and Reaction Rates for Surveillance Capsule 970 ............

A-20 Table A-5 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 83' (7-Degree Azimuth, Core M idplane) Cycle 1 Irradiation

..............................................................................

A -21 Table A-6 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 2630 (7-Degree Azimuth, Core Midplane)

Cycles 1 Through 9 Irradiation

...........................................................

A-22 Table A-7 Least-Squares Evaluation of Dosimetry in Surveillance Capsule 970 (7-Degree Azimuth, Core Midplane)

Cycles 1 Through 20 Irradiation

.........................................................

A-23 Table A-8 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios for Fast Neutron T hreshold R eactions .....................................................................................................

A -24 Table A-9 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios .....................

A-24 Table C-I Upper-Shelf Energy Values (ft-lb) Fixed in CVGRAPH ................................................

C-2 Table C-2 Upper-Shelf L.E. Values (mils) Fixed in CVGRAPH Summary Plots ...........................

C-2 Table D-I Calculation of Interim Chemistry Factors for the Credibility Evaluation for St. Lucie Unit 2using A ll Available Surveillance Data ..........................................................................

D-4 Table D-2 St. Lucie Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line Using All A vailable Surveillance D ata ............................................................................................

D -5 Table D-3 Calculation of Interim Chemistry Factors for the Credibility Evaluation for St. Lucie Unit 2 Using Only Transverse Orientation Base Metal Surveillance Data .............................

D-6 Table D-4 St. Lucie Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line Using Transverse Orientation Base Metal Surveillance Data ...................................................

D-6 Table D-5 Calculation of Residual vs. Fast Fluence for St. Lucie Unit 2 ........................................

D-7 Table E- I Predicted Positions 1.2 and 2.2 Upper-Shelf Energy Values at 55 EFPY .......................

E-3 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 V LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the St. Lucie Unit 2 Reactor Vessel ...............

4-5 Figure 4-2 Original Surveillance Program Capsule in the St. Lucie Unit 2 Reactor Vessel ..............

4-6 Figure 4-3 Surveillance Capsule Charpy Impact Specimen Compartment Assembly in the St. Lucie U nit 2 R eactor V essel .......................................................................................................

4-7 Figure 4-4 Surveillance Capsule Tensile and Flux-Monitor Compartment Assembly in the St. Lucie U nit 2 R eactor V essel .......................................................................................................

4-8 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

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5-16 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

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5-17 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

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5-18 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

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5-19 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

-Continued

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5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

..........................................

5-21 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

-Continued

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5-22 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

..........................................

5-23 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

-Continued

.....................

5-24 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld M etal (Heat # 83637) ........................................................

5-25 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) -Continued

...................................

5-26 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) .............................................

5-27 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)- Continued

........................

5-28 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program W eld M etal (Heat # 83637) ........................................................

5-29 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) -Continued

...................................

5-30 WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 vi Figure 5-10 Figure 5-10(a)Figure 5-11 Figure 5-11 (a)Figure 5-12 Figure 5-12(a)Figure 5-13 Figure 5-14 Figure 5-15 Figure 5-16 Figure 5-17 Figure 5-18 Figure 5-19 Figure 5-20 Figure 5-21 Figure 5-22 Figure 5-23 Figure 5-24 Figure 5-25 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel H eat-A ffected Zone M aterial .........................................................................................

5-31 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material -Continued

....................................................................

5-32 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone M aterial ..............................................................................

5-33 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material -Continued

.........................................................

5-34 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel H eat-A ffected Zone M aterial .........................................................................................

5-35 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material -Continued

....................................................................

5-36 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Standard R eference M aterial ..........................................................................................

5-37 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Standard Reference M aterial ..............................................................................

5-38 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Standard R eference M aterial ..........................................................................................

5-39 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

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5-40 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

..........................................

5-41 Charpy Impact Specimen Fracture Surfaces for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) ........................................................

5-42 Charpy Impact Specimen Fracture Surfaces for the St. Lucie Unit 2 Reactor Vessel H eat-A ffected Zone M aterial .........................................................................................

5-43 Tensile Properties for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse O rientation)

................................................................................................

5-44 Tensile Properties for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld M etal (H eat # 83637) .....................................................................................................

5-45 Tensile Properties for the St. Lucie Unit 2 Reactor Vessel Heat Affected Zone Material ...........................................................................................................................................

5 -4 6 Fractured Tensile Specimens from St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M -605-1 (Transverse Orientation)

........................................................................

5-47 Fractured Tensile Specimens from the St. Lucie Unit 2 Reactor Vessel Surveillance Program W eld M etal (Heat # 83637) .............................................................................

5-48 Fractured Tensile Specimens from the St. Lucie Unit 2 Reactor Vessel Heat Affected Z one M aterial .................................................................................................................

5-49 WCAP- 1 7939-NP May 2015 WCAP- I17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 vii Figure 5-26 Engineering Stress-Strain Curves for St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Tensile Specimens 2L5 and 2KU (Transverse Orientation)

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5-50 Figure 5-27 Engineering Stress-Strain Curve for St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Tensile Specimen 2JM (Transverse Orientation)

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5-51 Figure 5-28 Engineering Stress-Strain Curves for St. Lucie Unit 2 Surveillance Program Weld Metal (Heat # 83637) Tensile Specimens 3J4 and 3K7 ...........................................................

5-52 Figure 5-29 Engineering Stress-Strain Curve for St. Lucie Unit 2 Surveillance Program Weld Metal (Heat # 83637) Tensile Specim en 3L5 ...........................................................................

5-53 Figure 5-30 Engineering Stress-Strain Curves for St. Lucie Unit 2 Heat Affected Zone Material Tensile Specim ens 4K 5 and 4J5 ....................................................................................

5-54 Figure 5-31 Engineering Stress-Strain Curve for St. Lucie Unit 2 Heat Affected Zone Material Tensile Specimen 4JK ..................................................................................................

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5-55 Figure 6-1 St. Lucie Unit 2 r,0,z Reactor Geometry r,0 Plan View without Surveillance Capsules ..............................................................................................................................................

6 -14 Figure 6-2 St. Lucie Unit 2 r,0,z Reactor Geometry r,0 Plan View with 70 and 140 Surveillance C ap su les .........................................................................................................................

6-15 Figure 6-3 St. Lucie Unit 2 r,0,z Reactor Geometry rz Axial View .....................

6-16 Figure E- I Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence .....................................................................................

E-2 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 viii EXECUTIVE

SUMMARY

The purpose of this report is to document the testing results of surveillance Capsule 970 from St. Lucie Unit 2. Capsule 970 was removed at 25.55 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 970 received a fluence of 2.25 x 101 9 n/cm 2 (E > 1.0 MeV) after irradiation to 25.55 EFPY. The peak clad/base metal interface vessel fluence after 25.55 EFPY of plant operation was 1.73 x 1019 n/cm 2 (E > 1.0 MeV).This evaluation led to the following conclusions:

1) The measured percent decreases in upper-shelf energy for the surveillance plate (longitudinal orientation) and weld materials contained in St. Lucie Unit 2 Capsule 97' are less than the Regulatory Guide 1.99, Revision 2 [Ref. 1] predictions.

The measured decrease for the surveillance plate (transverse orientation) material is equivalent to the Regulatory Guide 1.99, Revision 2 [Ref. 1] prediction.

2) The St. Lucie Unit 2 surveillance plate, with consideration of all data or considering only the transverse orientation Charpy data points, and weld (Heat # 83637) data are judged to be credible.

It is standard to use all surveillance plate data in subsequent reactor vessel integrity evaluations.

However, a transverse orientation only credibility evaluation is presented in this report as an additional analysis of the data. 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 (55 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- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule 970, the third capsule removed and tested from the St. Lucie Unit 2 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 970 and previous capsules, along with the program input data.* Capsule 97' received an average fast neutron fluence (E > 1.0 MeV) of 2.25 x 1019 n/cm 2 after 25.55 effective full power years (EFPY) of plant operation." Irradiation of the reactor vessel Intermediate Shell Plate M-605-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 124.3 0 F and an irradiated 50 ft-lb transition temperature of 173.3'F. This results in a 30 ft-lb transition temperature increase of 132.7°F and a 50 ft-lb transition temperature increase of 140.1°F for the longitudinally oriented specimens.

• Irradiation of the reactor vessel Intermediate Shell Plate M-605-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 158.0°F and an irradiated 50 ft-lb transition temperature of 219.8°F. This results in a 30 ft-lb transition temperature increase of 127.6°F and a 50 ft-lb transition temperature increase of 148.3°F for the transversely oriented specimens." Irradiation of the Surveillance Program Weld Metal (Heat # 83637) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -25.7°F and an irradiated 50 ft-lb transition temperature of 16.2°F. This results in a 30 ft-lb transition temperature increase of 24.8°F and a 50 ft-lb transition temperature increase of 28.7°F.* Irradiation-of the Heat-Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -137.0°F and an irradiated 50 ft-lb transition temperature of 64.1°F.This results in a 30 ft-lb transition temperature increase of -103.9°F and a 50 ft-lb transition temperature increase of 46.2°F. It is noted that the scatter in the HAZ data was significant and the resulting CVGRAPH, Version 6.0 Charpy V-notch symmetric hyperbolic tangent curve-fit plots are not an ideal rendition of the data. However, this is inconsequential since HAZ material is not considered limiting as compared to the base and weld materials." The average upper-shelf energy of Intermediate Shell Plate M-605-1 (longitudinal orientation) resulted in an average energy decrease of 26 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 108 ft-lb for the longitudinally oriented specimens." The average upper-shelf energy of Intermediate Shell Plate M-605-1 (transverse orientation) resulted in an average energy decrease of 25 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 78 ft-lb for the transversely oriented specimens.

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 1-2" The average upper-shelf energy of the Surveillance Program Weld Metal (Heat # 83637) Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 95 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 93 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 St. Lucie Unit 2 reactor vessel surveillance materials are presented in Table 5-10.Standard Reference Material (SRM) HSST 01 Charpy specimens were not included in the St.Lucie Unit 2 Capsule 97'. However, the SRM HSST 01 Charpy specimens were reanalyzed in this report. The SRM HSST 01 material was contained in Capsule 2630, which was irradiated to a neutron fluence of 1.00 x 1019 n/cm 2 (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 157.1°F and an irradiated 50 ft-lb transition temperature of 200.9°F. This results in a 30 ft-lb transition temperature increase of 131.2'F and a 50 ft-lb transition temperature increase of 147.7°F.The average upper-shelf energy of the SRM HSST 01 Charpy specimens resulted in an average energy decrease of 36 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 86 ft-lb." Based on the credibility evaluation presented in Appendix D, the St. Lucie Unit 2 surveillance plate, with consideration of all data or considering only the transverse orientation Charpy data points, and weld (Heat # 83637) data are both credible.* Based on the upper-shelf energy evaluation in Appendix E, all beltline materials contained in the St.Lucie Unit 2 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 (55 EFPY) as required by 10 CFR 50, Appendix G [Ref. 2].* The maximum calculated 55 EFPY (end-of-license) neutron fluence (E > 1.0 MeV) for the St. Lucie Unit 2 reactor vessel beltline using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the Guide) is as follows: Calculated (55 EFPY): Vessel clad/base metal interface fluence* = 4.53 x 101 9 n/cm 2 Vessel 1/4 thickness fluence = 2.700 x 10" 9 n/cm 2*This fluence value is documented in Table 6-2 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 2-1 2 INTRODUCTION This report presents the results of the examination of Capsule 970, the third capsule removed and tested in the continuing surveillance program, which monitors the effects of neutron irradiation on the Florida Power & Light Company St. Lucie Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the St. Lucie Unit 2 reactor pressure vessel materials was designed and recommended by Combustion Engineering, Inc. A description of the surveillance program is contained in TR-L-MCM-001

[Ref. 3], "Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of St. Lucie No. 2 Reactor Vessel Materials." The pre-irradiation mechanical properties of the reactor vessel materials are presented in BAW- 1880 [Ref. 4]. The surveillance program was originally planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73 [Ref. 5], "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels." Capsule 970 was removed from the reactor after 25.55 EFPY of exposure and shipped to the Westinghouse Materials Center of Excellence Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing and post-irradiation data obtained from surveillance Capsule 97'removed from the St. Lucie Unit 2 reactor vessel and discusses the analysis of the data.WCAP- 17939-NP May 2015 Revision 0 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 St. Lucie Unit 2 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. 6]. 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. 7]) or the temperature 607F 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 (Ki, curve) which appears in Appendix G to Section XI of the ASME Code[Ref. 6]. The Klc 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 St. Lucie Unit 2 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 (ARTNDT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement.

This ART (initial RTNDT + M + ARTNDT) is used to index the material to the K 1 c 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- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-1 4 DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the St. Lucie Unit 2 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:

• Intermediate Shell Plate M-605-1 (longitudinal orientation)

• Intermediate Shell Plate M-605-1 (transverse orientation)

• Weld metal fabricated with weld wire Heat Number 83637, Linde Type 124 flux, Lot Number 0951, which is equivalent to the heat number used in the actual fabrication of the intermediate shell longitudinal weld seam repair and the lower shell longitudinal weld seams; however, these vessel welds used Linde Type 0091 flux in their fabrication

• Weld heat-affected zone (HAZ) material of Intermediate Shell Plate M-605-1* Standard Reference Material (SRM) Heavy-Section Steel Technology (HSST)-O1MY Plate Test material obtained from the intermediate 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 1/414 thickness location of the plate after performing a simulated post-weld stress-relieving treatment on the test material.

Test specimens were also removed from weld metal of a stress-relieved weldment joining Intermediate Shell Plate M-605-2 and adjacent Intermediate Shell Plate M-605-3. All heat-affected zone specimens were obtained from the weld heat-affected zone of Intermediate Shell Plate M-605-1.Charpy V-notch impact specimens from Intermediate Shell Plate M-605-1 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major rolling direction).

The core-region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular_(normal) to the weld- direction.

The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.

Tensile specimens from Intermediate Shell Plate M-605-1 were machined in the transverse orientation only. Tensile specimens from the weld metal were oriented perpendicular to the welding direction.

Some of the St. Lucie Unit 2 capsules, specifically the previously tested Capsule 2630 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 St. Lucie Unit 2 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.WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-2 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).

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 5.0% Ag, 5.0% Sn, 90.0% Pb 2.5% Ag, 97.5% Pb 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point: 536 0 F (280 0 C)Melting Point: 558 0 F (292 0 C)Melting Point: 580'F (304'C)Melting Point: 590'F (31 0 0 C)The chemical composition and the arrangement of the various mechanical specimens in Capsule 970 is 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-L-MCM-001

[Ref. 3], Tables III and XX.Capsule 970 was removed after 25.55 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 St. Lucie Unit 2 reactor vessel. Capsules 830, 970, 2630 and 2770 are radiologically equivalent to the 7' azimuth, while Capsules 104' and 2840 are radiologically equivalent to the 140 azimuth.WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-3 Table 4-1 Chemical Composition (wt. %) of the St. Lucie Unit 2 Reactor Vessel Surveillance Materials (Unirradiated)

Intermediate Shell Plate Standard Reference Surveillance Weld Metal Element M605_l~a Material HSST Original CE Best-Estimate 01MY Plate(b) Analysis(c)

Analysis(d)

C 0.23 ---0.12 ---Mn 1.37 ---1.55 ---P 0.004 --- 0.003 ---S 0.010 --- 0.011 ---Si 0.23 --- 0.38 ---Ni 0.61 0.66 0.07 0.066 Mo 0.57 --- 0.59 -- -Cr 0.07 --- 0.04 ---Cu 0.11 0.18 0.05 0.048 A] 0.022 --- 0.002 ---Co 0.010 ---0.006 -- -Pb <0.001 ---<0.001 ---W <0.01 ---<0.01 --Ti <0.01 --- <0.01 --Zr <0.001 --- 0.001 ---V 0.003 ---0.005 ---Sn 0.009 --- 0.002 ---As 0.002 --- <0.001 ---Cb <0.01 ---<0.01 ---Sb 0.0024 --- 0.0030 ---N 2 0.009 --- 0.003 13 <0.001 --- 0.001 ---Notes: (a) Data obtained from TR-L-MCM-001, Table III [Ref. 3](b) Data obtained from NUREG/CR-6413

[Ref. 8].(c) Data obtained from TR-L-MCM-001, Table III [Ref. 3]. Weld Wire Heat Number 83637, Flux Type Linde 124, and Flux Lot Number 0951.(d) Best-Estimate Cu and Ni wt. % values were taken from CE-NPSD-1039, Revision 2 [Ref. 9].WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-4 Table 4-2 Arrangement of Encapsulated Test Specimens within St. Lucie Unit 2 Capsule 970 tPosition (Compartment Number Compartment (Specimen Type and Material)(a)

Specimen Numbers~a) 1 K214 (Tensile HAZ Specimens) 4K5, 415, 4JK K224 41P, 456, 42D, 46Y, 2(Charpy Impact HAZ Specimens) 425, 47D, 42C, 47L, 43K, 473, 43D, 46A K231 124, 13K, 14M, 12B, 3 (Charpy Impact Longitudinal 14D, 127, 117, 15D, Plate Specimens) 13C, 1 ID, 143, 15J K242 4 (Tensile Transverse 2L5, 2KU, 2JM Plate Specimens)

K252 21B, 25D, 23A, 263, 5 (Charpy Impact Transverse 22K, 25M, 21D, 24M, Plate Specimens) 21U, 24P, 26E, 21Y K263 34U, 32A, 37Y, 31A, 6(Charpy Impact Weld Specimens) 36E, 32B, 355, 32L, 37P, 32E, 354, 311 K273 (Tensile Weld Specimens) 3J4, 3K7, 3L5 Note: (a) Data obtained from TR-L-MCM-001, Table XIX and/or Table XX [Ref. 3].WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-5 Westinghouse Non-Proprietary Class 3 4-5 S Surveillance capsule Core Support BAM l.I 1evadon ErnRed Plan 'View VIew Figure 4-1 Arrangement of Surveillance Capsules in the St. Lucie Unit 2 Reactor Vessel WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-6 Figure 4-2 Original Surveillance Program Capsule in the St. Lucie Unit 2 Reactor Vessel WCAP- 17939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-7 CouP&C -End Cap chaawwrMPRAe Spaca=Recungular Tubing Coupling -End Cap Figure 4-3 Surveillance Capsule Charpy Impact Specimen Compartment Assembly in the St.Lucie Unit 2 Reactor Vessel WCAP- 17939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 4-8 Wedge Coupling -End Cap Stainless Steel Tubing Threshold.

Detector .'Flux Spectrum Monitor Cadmium Shielded-Stainless Steel Tubing%Cadmium Sfiield Threshold Detector.-Quartz Tubing o-Weght~Low .Melting Alloy Housing Tensile Specimen S pl it S pace r L Tensile Specimen Housing-Rectangular Tubing Figure 4-4 Surveillance Capsule Tensile and Flux-Monitor Compartment Assembly in the St.Lucie Unit 2 Reactor Vessel WCAP-1 7939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-1 5 TESTING OF SPECIMENS FROM CAPSULE 970 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.10].Capsule 970 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-L-MCM-001 [Ref. 3]. All of the items were in their proper locations.

Examination of the thermal monitors indicated that 4 of the 12 temperature monitors had melted, as described below:* Capsule compartment K214, the 536°F (280"C) temperature monitor melted* Capsule compartment K242, the 536"F (280°C) and 558"F (292*C) temperature monitors melted" Capsule compartment K273, the 536°F (280"C) temperature monitor melted.Based on this examination, the maximum temperature to which the specimens were exposed was less than 580OF (304 0 C), but greater than 558'F (292 0 C).The Charpy impact tests were performed per ASTM Specification E185-82 [Ref. 10] and E23-12c[Ref. 11] 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. 12].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 (F.y), 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, F., Fbf, and Fa were determined per the guidance in ASTM Standard E2298-13a

[Ref. 12].The energy at maximum load (Wi) 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 (W.). W, is compared to the dial energy (KV). W, derived from the instrumented striker were all within 15% of the calibrated dial energy values as required in ASTM E2298-13a

[Ref. 12].WCAP- 17939-NP May 2015 Revision 0 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. 11] and A370-13 [Ref. 13]. 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. 10]. Testing met ASTM Specifications E8/E8M-13a

[Ref. 14] or E21-09 [Ref. 15]. Load was applied through a threaded connection.

The strain rate obtained met the requirements of ASTM E8/E8M-13a

[Ref. 14] and ASTM E21-09 [Ref. 15].Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 10-inch hot zone. Tensile specimens were soaked at temperature

(+/-5°F) 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 6 (Drawing CND-B-3654 Rev 2) of TR-L-MCM-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. 14]. 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 970, which received a fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) in 25.55 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 BAW-1880 [Ref. 4] and WCAP-15040, Revision 1 [Ref. 16]. 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 St. Lucie Unit 2 capsule reports.The transition temperature increases and changes in upper-shelf energies for the Capsule 970 materials are summarized in Table 5-9 and led to the following results: Irradiation of the reactor vessel Intermediate Shell Plate M-605-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of 124.3'F and an irradiated 50 ft-lb transition temperature of 173.3'F. This results in a 30 ft-lb transition temperature increase of 132.7°F and a 50 ft-lb transition temperature increase of 140.1°F for the longitudinally oriented specimens.

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-3" Irradiation of the reactor vessel Intermediate Shell Plate M-605-1 Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 158.0'F and an irradiated 50 ft-lb transition temperature of 219.8°F. This results in a 30 ft-lb transition temperature increase of 127.6°F and a 50 ft-lb transition temperature increase of 148.3'F for the transversely oriented specimens.

• Irradiation of the Surveillance Program Weld Metal (Heat # 83637) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-25.7'F and an irradiated 50 ft-lb transition temperature of 16.2'F. This results in a 30 ft-lb transition temperature increase of 24.8°F and a 50 ft-lb transition temperature increase of 28.7'F.* Irradiation of the Heat-Affected Zone (HAZ) Material Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-137.0'F and an irradiated 50 ft-lb transition temperature of 64.1'F.This results in a 30 ft-lb transition temperature increase of -103.9'F and a 50 ft-lb transition temperature increase of 46.2°F. It is noted that the scatter in the HAZ data was significant and the resulting CVGRAPH, Version 6.0 Charpy V-notch symmetric hyperbolic tangent curve-fit plots are not an ideal rendition of the data. However, this is inconsequential since HAZ material is not considered limiting as compared to the base and weld materials." The irradiated upper-shelf energy of Intermediate Shell Plate M-605-1 (longitudinal orientation) resulted in an average energy decrease of 26 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 108 ft-lb for the longitudinally oriented specimens." The average upper-shelf energy of Intermediate Shell Plate M-605-1 (transverse orientation) resulted in an average energy decrease of 25 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 78 ft-lb for the transversely oriented specimens." The average upper-shelf energy of the Surveillance Program Weld Metal (Heat # 83637) Charpy specimens resulted in an average energy decrease of 20 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 95 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 93 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 St. Lucie Unit 2 reactor vessel surveillance materials are presented in Table 5-10.Standard Reference Material (SRM) HSST 01 Charpy specimens were not included in the St.Lucie Unit 2 Capsule 970. However, the SRM HSST 01 Charpy specimens were reanalyzed in this report. The SRM HSST 01 material was contained in Capsule 2630, which was irradiated to a neutron fluence of 1.00 x 101 9 n/cm 2 (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- 1 7939-NP May 2015 Revision 0 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 157.1°F and an irradiated 50 ft-lb transition temperature of 200.9'F. This results in a 30 ft-lb transition temperature increase of 131.2'F and a 50 ft-lb transition temperature increase of 147.7'F.The average upper-shelf energy of the SRM HSST 01 Charpy specimens resulted in an average energy decrease of 36 ft-lb after irradiation.

This results in an irradiated average upper-shelf energy of 86 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 (55 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 970 irradiated to 2.25 x 10'9 n/cm 2 (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 Intermediate Shell Plate M-605-1 (transverse orientation) indicated that irradiation to 2.25 x 1019 n/cm 2 (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 5-20 and Table 5-11.The results of the tensile tests performed on the Surveillance Program Weld Metal (Heat # 83637)indicated that irradiation to 2.25 x 101 9 n/cm 2 (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 5-21 and Table 5-11.The results of the tensile tests performed on the Heat Affected Zone Material indicated that irradiation to 2.25 x 1019 n/cm 2 (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 Intermediate Shell Plate M-605-1 (transverse orientation) material are shown in Figure 5-23, the fractured tensile specimens for the Surveillance Program Weld Metal (Heat# 83637) 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- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-5 Table 5-1 Charpy V-notch Data for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF °C ft-lbs Joules mils mm %15D 70 21 19 26 15 0.38 20 14M 95 35 35 47 30 0.76 25 124 110 43 28 38 27 0.69 25 11ID 120 49 43 58 36 0.91 30 14D 140 60 24 32 25 0.64 25 13K 150 66 28 38 29 0.74 30 117 170 77 67 91 59 1.50 60 13C 200 93 41 56 32 0.81 45 143 200 93 42 57 38 0.96 40 15J 260 127 111 150 89 2.26 100 12B 300 149 100 136 82 2.08 100 127 375 191 113 153 89 2.26 100 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-6 Table 5-2 Charpy V-notch Data for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) (Transverse Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number OF 0C ft-lbs Joules mils mm %21B 70 21 18 24 14 0.36 15 22K 95 35 16 22 2 0.05 20 24M 120 49 27 37 24 0.61 25 25D 140 60 22 30 22 0.56 25 24P 150 66 37 50 34 0.86 45 26E 170 77 36 49 37 0.94 45 21U 200 93 25 34 23 0.58 35 21Y 200 93 29 39 23 0.58 45 25M 230 110 54 73 47 1.19 70 23A 270 132 80 108 60 1.52 100 21D 300 149 78 106 65 1.65 100 263 375 191 77 104 68 1.73 100 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-7 Table 5-3 Charpy V-notch Data for the St. Lucie Unit 2 Surveillance Program Weld Metal (Heat # 83637) Irradiated to a Fluence of 2.25 x 10'9 n/cm 2 (E > 1.0 MeV)Sample Temperature Impact Energy Lateral Expansion Shear Number OF °C ft-lbs Joules mils mm %36E 51 17 23 14 0.36 15 34U 40 25 34 21 0.53 25 32E 34 22 30 15 0.38 20 31A 32 24 32 25 0.64 25 355 29 45 61 36 0.91 30 37P 0 -18 49 *66 45 1.14 50 32A 20 -7 49 66 35 0.89 60 32B 70 21 61 83 60 1.52 80 354 120 49 93 126 84 2.13 95 37Y 170 77 113 153 84 2.13 98 311 250 121 82 111 72 1.83 100 32L 300 149 93 126 82 2.08 100 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-8 Table 5-4 Charpy V-notch Data for the St. Lucie Unit 2 Heat-Affected Zone (HAZ) Material Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV)Sample Temperature Impact Energy Lateral Expansion Shear Number OF 0 C ft-lbs Joules mils mm %425 50 10 106 144 58 1.47 60 43D 70 21 17 23 18 0.46 30 41P 100 38 42 57 37 0.94 55 456 120 49 34 46 29 0.74 50 46Y 130 54 24 32 22 0.56 50 43K 140 60 120 163 82 2.08 95 47L 150 66 21 28 20 0.51 45 473 170 77 22 30 27 0.68 45 42C 180 82 86 117 53 1.35 85 42D 250 121 123 167 80 2.03 100 46A 300 149 59 80 63 1.60 100 47D 375 191 68 92 57 1.45 100 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation)

Total Dial Total Energy to General Fracture Sample tEnergy, Instrumented Difference, Max Maximum Time to Yield Lo Arrest Numbe Temp Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, a, Load, Fa Nube Temp wt(%) Wm (lb) (msec) Fay (lb) (lb)(0 F) (ft-lb) (ft-lb) (ft-lb) (Ib)15D 70 19 16.9 11.0 3.9 4600 0.13 3200 3700 N/A 14M 95 35 33.3 4.9 22.2 3900 0.43 3100 3800 800 124 110 28 26.9 3.9 21.6 3800 0.43 2900 3500 700 11D 120 43 41.0 4.6 32.0 4000 0.60 2800 3900 1100 14D 140 24 23.4 4.4 3.71 4100 0.13 2700 3600 400 13K 150 28 26.2 6.4 3.35 4100 0.16 2800 3900 900 117 170 67 61.5 8.2 3.6 4200 0.14 2500 3600 1800 13C 200 41 39.4 3.9 2.2 4100 0.10 3000 3800 1500 143 200 42 39.1 6.9 3.8 4100 0.14 2700 3400 1500 15J 260 111 106.8 3.8 29.9 3700 0.60 2500 N/A N/A 12B 300 100 96.2 3.8 29.9 3800 0.60 2600 N/A N/A 127 375 113 108.3 4.2 39.9 3900 0.79 2500 N/A N/A WCAP-1 7939-NP May 2015 Revision 0 Westinghouse Non-Proprieta Class 3 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV) (Transverse Orientation)

Total Energy to General Tet Total Dial Toa Sample Test Energy, Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Number Temp KV Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbr Load, FB (OF) (ft-b) (%) Wm (Ib) (msec) F9V (lb) (lb)(ft-lb) (ft-lb) (Ib)21B 70 18 16.7 7.2 3.8 3500 0.13 3000 3400 500 22K 95 16 16.1 0.6 2.1 3600 0.37 2800 3600 400 24M 120 27 25.1 7.0 19.1 4200 0.36 3400 4100 900 25D 140 22 21.0 4.5 13.7 3400 0.30 2700 3200 1100 24P 150 37 35.8 3.2 21.0 3700 0.43 2800 3600 1900 26E 170 36 34.5 4.2 20.8 3600 0.43 2700 3400 1500 21U 200 25 24.6 1.6 2.1 3700 0.15 3200 3600 1100 21 Y 200 29 27.3 5.9 4.4 4300 0.40 2800 3600 1200 25M 230 54 49.8 7.8 29.0 3600 0.60 2600 3500 2400 23A 270 80 75.9 5.2 28.0 4400 0.56 2800 N/A N/A 21D 300 78 74.6 4.4 30.2 3600 0.63 2500 N/A N/A 263 375 77 72.7 5.5 24.1 3700 0.56 2500 N/A N/A WCAP- 17939-NP May 2015 WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Surveillance Program Weld Metal (Heat # 83637)Irradiated to a Fluence of 2.25 x 10'9 n/cm 2 (E > 1.0 MeV)Total Energy to General Tet Total Dial Toa Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest NuSample Temp Energy,KV Energy, (KV-Wt)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa (O) (t )Wt'(% Wm (lb) (msec) Fgy (lb) (lb)(ft-lb) (ft-lb) (ft-lb) (ib)36E -60 17 16.2 4.7 3.1 4100 0.09 3500 4000 200 34U -40 25 24.4 2.4 3.2 4200 0.09 3500 3800 600 32E -30 22 21.3 3.2 3.2 4100 0.09 3500 4000 600 31A -25 24 23.0 4.2 3.3 4100 0.09 3200 4000 400 355 -20 45 41.9 6.7 2.8 4200 0.09 3300 3900 600 37P 0 49 45.2 7.8 33.6 4100 0.60 3100 3900 1300 32A 20 49 46.4 5.2 33.6 3900 0.60 3100 3700 1900 32B 70 61 57.3 6.1 31.8 3800 0.60 2900 3200 1800 354 120 93 89.6 3.6 30.8 3800 0.60 2700 2500 2100 37Y 170 113 108.3 4.2 33.7 4100 0.73 2600 N/A N/A 311 250 82 79.4 3.2 29.5 3600 0.60 2600 N/A N/A 32L 300 93 90.4 2.8 28.9 3600 0.60 2400 N/A N/A WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprieta Class 3 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the St. Lucie Unit 2 Heat-Affected Zone (HAZ) Material Irradiated to a Fluence of 2.25 x 1019 n/cm 2 (E > 1.0 MeV)Total Energy to General Tet Total Dial Toa Test Instrumented Difference, Max Maximum Time to Yield Fracture Arrest Number T KV Energy, (KV-W,)/KV Load, Load, Fm Fm Load, Load, Fbf Load, Fa (NF) (ft-lb) W, (%) Wm (ib) (msec) FPV (lb) (Ib)(ft-lb) (ft-lb) (lb)425 50 106 101.0 4.7 47.4 4400 0.79 3200 3100 800 43D 70 17 16.1 5.3 2.1 4300 0.29 2600 3000 1500 41P 100 42 40.2 4.4 3.1 3900 0.09 2800 3800 1900 456 120 34 32. 5.6 2.4 3900 0.09 3000 3700 2300 46Y 130 24 22.0 8.4 11.7 3500 0.26 2900 3400 1800 43K 140 120 115.3 3.9 44.6 4200 0.79 3000 3100 3000 47L 150 21 19.0 9.5 4.1 3900 0.15 2800 3400 1100 473 170 22 21.4 2.7 1.9 4600 0.24 2600 3400 1400 42C 180 86 81.6 5.0 43.3 4000 0.79 2900 3800 2700 42D 250 123 117.6 4.4 2.0 4600 0.12 2300 N/A N/A 46A 300 59 56.0 5.1 28.7 3400 0.60 2400 N/A N/A 47D 375 68 65.2 4.1 20.3 3700 0.47 2600 N/A N/A WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-13 Table 5-9 Effect of Irradiation to 2.25 X 1019 n/cm 2 (E > 1.0 MeV) on the Charpy V-Notch Toughness Properties of the St. Lucie Unit 2 Reactor Vessel Surveillance Capsule 970 Materials Average 30 ft-lb Transition Average 35 mil Lateral Expansion Average 50 ft-lb Transition Average Energy Absorption at Material Temperature(a) (OF) Temperature(a) (OF) Temperature(a) (OF) Full Shear(a) (ft-lb)Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Intermediate Shell Plate-8.4 124.3 132.7 13.1 146.3 133.2 33.2 173.3 140.1 134 108 26 M-605-1 (Longitudinal)

Intermediate Shell Plate 30.4 158.0 127.6 38.2 190.7 152.5 71.5 219.8 148.3 103 78 25 M-605-1 (Transverse)

Surveillance Weld Material -50.5 -25.7 24.8 -27.1 -2.0 25.1 -12.5 16.2 28.7 115 95 20 (Heat # 83637)Heat Affected-33.1 -137.0 0 (b) 8.5 81.6 73.1 17.9 64.1 46.2 105 93 12 Zone Material Notes: (a) Average value is determined by CVGraph, Version 6.0 (see Appendix C).(b) The St. Lucie Unit 2 Heat Affected Zone Material 30 ft-lb transition temperature shift was calculated to be a negative value. Physically, this should not occur; therefore, a conservative value of zero degrees F is shown in this table.WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-14 Table 5-10 Comparison of the St. Lucie Unit 2 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 (xlO'9 n/cm 2 , E Measured(b)

Predicted(a)

Measured(b)

> 1.0 MeV) (OF) (OF) (%) (%)Intermediate Shell Plate 830 0.140 36.2 45.1 12.5 11 M-605-1 (Longitudinal) 970 2.... 9 12 2 19 t830 0.140 36.2 29.4 12.5 1 Intermediate Shell Plate 23 .0 02 M Tases)2630 1.000 74.2 102.7 20 23 M-605- 1 (Transverse) 830 0.140 16.6 15.8 12 13 Surveillance Weld Material (Ha 33)2630 1.000 34.1 26.5 19 9 (Heat # 83637) ]830_ 0.140 -_-_ -0.0(r) --- 0 (d)Heat Affected Zone Material 2630 1.000 ---79.5 --- 0 (d)Standard Reference Material 263- 1.000 --- 131.2 ---30 Notes: (a) Based on Regulatory Guide 1.99, Revision 2, methodology using the capsule fluence and mean weight percent values of copper and nickel of the surveillance material.(b) Calculated by CVGraph Version 6.0 using measured Charpy data (See Appendix C).(c) These ARTNDT were calculated to be negative values. Physically, this should not occur; therefore, conservative values of zero degrees F are shown in this table.(d) USE values were calculated to have increased.

Physically, this should not occur; therefore, conservative values of zero percent are shown in this table.WCAP- 17939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-15 Table 5-11 Tensile Properties of the St. Lucie Unit 2 Capsule 970 Reactor Vessel Surveillance Materials Irradiated to 2.25 x 10'9 n/cm 2 (E > 1.0 MeV)0.2% Fracture Test Ultimate Fracture Fracture Uniform Total Reduction Material Sample Yield Strength Load Strength True Elongation Elongation in Area Maeral Number Temp. ed Sregh La Strength Stress()(OF) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%)Intermediate Shell 2L5 150 71.9 93.4 3.37 68.7 160 12.1 24.3 57 Plate Plate 2KU 280 68.0 89.0 3.43 69.9 162 11.7 21.4 57 M-605-1 (Transverse) 2JM 550 61.7 89.1 3.46 70.5 156 10.9 18.7 55 3J4 72 67.1 86.4 2.87 58.5 192 10.6 23.0 70 Surveillance Weld Material 3K7 150 68.9 85.0 2.73 55.5 152 10.4 25.0 64 (Heat # 83637)3L5 550 65.4 84.5 2.84 57.8 167 9.1 20.8 65 4K5 72 64.3 82.2 2.72 55.4 172 7.9* 8.9* 68 Heat Affected Ze Ater 4J5 240 65.7 83.2 2.59 52.8 168 9.4 22.6 69 Zone MaterialIIIIII 4JK 550 64.6 84.1 2.73 55.5 172 8.8 20.9 68*Refer to Figure 5-25; specimen broke outside of gage section, so strain is not an accurate measurement.

These values are omitted from Figure 5-22.WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-16 IS PLATE M-605-1 (LONGITUDINAL)(.0: Hypeilic Tangent Curve Printed on 1/9/2015 7:16 AM Curve Plant Capsule Material O1i Heat #1 St. Luce 2 Uniffad SA533BI LT A-8490-2 2 St. Lucie 2 83* SA533BI LT A-8490-2 3 St. Lucie 2 970 SA533B1 LT A-8490-2 0.1 160 140 120 100 8o 60 40 20-300 I I I-200 -100 0 100 200 300 400 500 600 Temperature (e F)Curve Fluence LSE USE d-USE T @301 dT @M0 T @50 d-T @50 1 2-2 134 0 -8.4 0 332 0 2 2-2 1L9 -15 36.7 45-1 673 34A 3 2-2 108 -26 124-3 132.7 1733 1401 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-17 IS PLATE M-605-1 (LONGITUDINAL)

CVGraph 6-0: Hypetolic Tangent Cui-e Printed on 4110f2015 7:41 AM Curve Plant Canule Mate"ia Ori Heat#1 St Lucie 2 Unftrad SA533B1 LT A-8490-2 2 St.Lude2 830 SA533B1 LT A-8490-2 3 St. Lucie 2 97 SA533BI LT A-8490-2 100 A 1 so- i[800 40 640 40 20 A I I-300 -200 -100 0 I00 200 300 400 500 600 Temperature

(* F)Curve Fluence ISE USE d-USE T @35 d-T@35 1 0 1 85.48 0 13-1 0 2 0 1 85.49 0.01 512 38-1 3 0 1 87 152 146,3 1332 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-18 Westinghouse Non-Proprietary Class 3 5-18 IS PLATE M-6051 (LONGITUDINAL)

CVGrap.i 6-0: Hypabolic Tangent Curve Printed on 12/9/2014 8:-8 AM Curve Plant Capsule MatffWi Ori Heat #1 St Lucie 2 Uninad SA533BI LT A-8490-2 2 St Lucie 2 830 SA533BI LT A-8490.2 3 St- Lucie2 97' SA533BI LT A-8490-2 1,00 -81 ___ ______ _90 3 So __ __t _ !a,/16,, 70 S40 30 50 20 -' " --300 -200 -100 0 100 200 300 400 50 600 Temperature (1 F)Curve Yluence [SE USE d-USE T,0 d-T @50 1 0 100 0 883 0 2 0 100 0 122 33.7 3 0 10D 0 180.5 92-2 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-19 Westinghouse Non-Proprietary Class 3 5-19 IS PLATE M-605-I (TRANSVERSE)

CVGaph 6.0-: Hypebolic Tangent Curve Printed on 12W/92014 8:25 AM Curve Plant Capsule Mate"i* Oi6 Heat #1 St. bade 2 Uninad SA533B1 TL A-8490-2 2 St. Lucie 2 83 SA533B1 Tb A-8490-2 3 St. Lucie 2 2630 SA533B1 TL A-8490-2 4 St ladcie 2 97- SA533B1 TL A-8490-2 120 100 80~60 0 "-300-200 -100 0 100 20 300 4W0 500 600 Temperature (0 F)Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -20 IS PLATE M-605-1 (TRANSVERSE)

CVC-aph 6.0: Hyperbolic Tangent Cbirve Printed on 12i912014 8:25 AM CurveIFluence_]IISE__]ULSEId-USEI T @30 1Id-T @30 1 T g58 d-T @M I 2.2 103 0 30.4 0 71-5 0 2 2-2 102 -1 59.8 29-4 108.7 37.2 3 212 79 -24 133-1 102-7 182.1 110.6 4 22 78 -25 158 127.6 219.8 148.3 Figure 5-4(a) Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

-Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-21 IS PLATE M-605-1 (TRANSVERSE)

CVGraph 6.0: Hyperbolic Tangent Cw Printed on 4/1012015 7:47 AM Curve PLMrt Capsule Mterial On_ Heat #I St. Lade 2 Uninad SA53JBI IL A-8490-2 2 St. Lucie 2 83* SA533B1 TL A-9490-2 3 St. Lucie 2 263' SA533B1 IL A-8490-2 4 St. Lucie 2 97 SA533B1 TI A-8490-2 9o 70 6.4"40 30 20 10-300 -200 -100 0 100 200 300 400 IV0 600 Temperature (0 F)Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-22 Westinghouse Non-Proprietaiy Class 3 5-22 IS PLATE M-605-1 (TRANSVERSE)

CVGfap 6-0: Hyperbolic Tangent Curve Printed on 4A1012015 7:47 AM Cure Fluence LSE ISE d-U SE T @35 d-T@35 1 0 1 77-6 0 38-2 0 2 0 1 7537 23 75,6 37A 3 0 1 73 A460 151.3 113.1 4 0 1 64 6 1007 152.5 Figure 5-5(a) Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

-Continued WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -23 Westinghouse Non-Proprietary Class 3 5-23 IS PLATE M"605-I (TRANSVERSE)

CVGraph 6.0_ Hyperbolic Tangent CM, Printed on 1219/2014 8:19 AM Curve Pn Capsule Materal OL Het #1 St, Lnde 2 Unirad SA533BI TL A-8490-2 2 St. Lude 2 83* SA533B1 TL A-8490-2 3 St. Lucie 2 2630 SA533BI TL A-8490-2 4 St. Lucie 2 97- SA533B1 TL A-4900-2 s.as C" 110 100 90 8o 70 60 5o 40 30 20 10 0 G 2 3-41 --.......YA A& AL ..........Le-300 -200 -00 0 J00 200 Temperature (0 F)300 400 500 600 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -24 IS PLATE M-605-1 (TRANSVERSE)

CVGraph 6.0: Hyperbolic Tangent Curve Printed on 12/9/2014 8:19AM Cwle Yuence LSE USE d-ETSE T @50 d-T @50 1 0 100 0 871 0 2 0 100 0 169.6 82.5 3 0 100 0 1752 88.1 4 0 100 0 187-3 100.2 Figure 5-6(a) Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

-Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-25 SURVEILLANCE PROGRAM WELD METAL CVGr-ph 6.0: Hlpedsolic Tangent Curwe Prnted on 12/9/2014 8"28 AM Cunt. Plant Capsule Material O06u Heat #1 St, Lade 2 Uniumd SAW NA 83637 2 St. LCie 2 83* SAW NA 93637 3 St. Lud2 263r SAW NA 83637 4 St- lade 2 97- SAW NA 83637'-B c~.)140 120 100 60 40 20-300 -200 -100 0 100 200 30 400 500 600 Temperature (0 F)Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -26 SURVEILLANCE PROGRAM WELD METAL CVGraph 6.0: Hyperbolic Tangent Curve Printed on 12/912014 8:28 AM Curve fluence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 1 2.2 115 0 -50.5 0 -12.5 0 2 2.2 100 34.7 15.8 -5.8 6.7 3 2.2 105 24 26-5 19A 31.9 4 22 95 25.7 24.8 16.2 28.7 Figure 5-7(a) Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) -Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-27 SURVEILLANCE PROGRAM WELD METAL CVGrph 6.0- Hyperbolic Tangent Curve Pfrned on 12/9)2014 8:46 AM Curve Plant Capsule Mzterig 1ii Heat#1 St. Lauie 2 Uni-ad SAW NA 83637 2 St. bide 2 S3* SAW NA 83637 3 St Lade 2 2630 SAW NA 83637 4 St, Luile 2 97- SAW 'NA 3637 140 120 S100.~80 60 040 20-300 -200 -100 0 100 200 300 400 500 600 Temperature

(* F)Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-28 SURVEILLANCE PROGRAM %ELD AMETAL CVGiaph 6.0: Hfypezbolc Tangent Cmuve Punted on 1219/22014 8:46 AM Cumr Fluence LSE USE d-USE T @35 d-T ig35S 1 1 8828 0 -27.1 0 2 1 8186 42 -16 111 3 1 114.16 25.88 "-1-1 26 4 1 814A5 -6.83 -2 251 Figure 5-8(a) Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) -Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -29 SURVEILLANCE PROGRAM WELD METAL CVGraph 6-0: Hypedboic Tangent Cume Printed on 12J9F2014 8:51 AM Curv PIMant Capsule Materia Ori. Heat#1 St. Lucie 2 Lb," SAW NA 83637 2 St. Lcie 2 830 SAW NA 83637 3 St. Lucie 2 2630 SAW NA 83637 4 St. Lucie 2 97* SAW NA 83637 110 100 90 80 so cc Qj 70 60 50 40 30 20 10 0 8 ~ I--.-.---- ---2--~ L J L 1-300 -200 -100 0 100 200 300 Temperature (0 F)400 S00 600 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-30 SURVEILLANCE PROGRANI WELD METAL CVGraph 6.0: Hypebolic Tangent Curve Printed on 12NM2014 8:51 AM Cun-ve fuence LSE USE d-USE T @50 d-T @50 1 0 100 0 13 0 2 0 too 0 10o8 9-5 3 0 100 0 16A 15.1 4 0 100 0 8.8 7.5 Figure 5-9(a) Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637) -Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-31 HEAT AFFECTED ZONE CVGraph 6.0: Hypefblic Tangent Cave Pfinted on 1/9/'2015 7:21 AM crave Plant Capswe Mate6 O Heat#1 St- Lade 2 Unirad SA533BI NA A-8490-2 2 St Lucie 2 830 SA533BI NA A-8490-2 3 St Lucie 2 2630 SA533B1 NA A-8490-2 4 St. Luie 2 97- SA533BI NA A-8490-2 rI~I 200 115 125 100 75 50 25 I't 0 f ----T I I! I ! I I I I! _ --300 100 0 100 200 300 400 so 600 Temperature (o F)Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-32 HEAT AFFECTED ZONE CVGraph 6.0: Hyperbolic Tangent Curve Printed on 1/9/2)015 7:21 AM Curve Huence LSE USE d-USE T @30 d-T @30 T @50 d-T @50 1 2-2 105 0 -33.1 0 17.9 0 2 2.2 119 14 -51.5 -18.4 20.5 2.6 3 2-2 130 25 46.4 79.5 109.1 91.2 4 2-2 93 137 -103.9 64.1 46.2 Figure 5-10(a)Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material -Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-33 HEAT AFFECTED ZONE CVGraph 6.0: Hyperolic Tangent Curve Printed on 119/2015 7:29 AM CIrve plant Capsule Material Of1 Hea#1 StLucie2 Uhirrad SA533BI NA A-8490-2 2 St. Lucie 2 830 SA533B1 NA A-8490-2 3 St. Lucie 2 26r SA533B1 NA A-8490-2 4 St. Luie 2 970 SA533B1 NA A-8490-2 100 6et 40 20 0 F _ 1 1 I 1 1 1 1_ I ___-304 -200 -100 0 100) 200 300 400 so 600 Temperature ( F)Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-34 HEAT AFFECTED ZOINE CVC&aph 6.0: Hyperbolic Tangent Cunve Printed on 1/9/2015 7:29 AM Curve fluence LSE USE d-USE T @35 d-T T35 1 1 73.10 0 8.5 0 2 1 71.91 -1.25 '113 2.8 3 1 83.1 9_94 76.6 68.1 4 1 77.97 4.81 81.6 73.1 Figure 5-11(a) Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material -Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-35 HOEAT AYFECTED ZONE CVGrah 6.0: Hyperbolic Tangent Curve Printed on 119t2015 8:21 AM curve Plant Capsule Material Oni Heat #1 St. Lcie 2 Unin-Ad SA533BI NA A-8490-2 2 St. Lucie 2 830 SA533BI NA A-490-2 3 St Lucie 2 2630 SA533BI NA A-8490-2 4 St Lucie 2 9r SA533BI NA A-8490-2 110 100 90 so t 70 40 30 20 10 010-300-200 -100 0 100 200 300 400 500 600 Temperature (0 F)Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-36 HEAT AFFECTED ZONTE CVGmaph 6.0:

Tangent Curve Printed on 119W2015 8:21 AM Curve Fnuence LSE USE d-USE 1 @50 d-T @50 1 0 100 0 52.1 0 2 0 100 0 61.5 9A 3 0 100 0 101.7 49.6 4 0 100 0 92A. 403 Figure 5-12(a) Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material -Continued WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-37 Westinghouse Non-Proprietary Class 3 5-37 STANDARD REFERENCE MATERIAL CVGph 6.0-- Hyperbolic Tangent Cuwe Printed on I129014 9-09 AM Curve Plant Cap5U1e Material OrL Heat #1 St- Lucie 2 LUniiad SA533B1 LT HSST-0MY 2 St Lucie 2 2630 SA533BI LT HSST-01MY 140 120 100'8so~60 40 20 0=-300-200 -100 0 100 200 300 400 500 600 Temperature

(* F)Can't. Fluence LSE USE d-USE T @3o d-T @30 T d-T @50 1 2.2 122 0 25.9 0 53.2 0 2 2.2 86 -36 157.1 131-2 200.9 147.7 Figure 5-13 Charpy V-Notch Impact Energy vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Standard Reference Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-38 Westinghouse Non-Proprietary Class 3 5-38 STANDARD REFERENCE MATERIAL CVGAph 6.0: HypeAolc Tangent Cune Pnnted on 410/20157:50 AM Curve f Plaut Cap~de Mam1OfL Heat#I St, Lucie2 UnillAd SA533BI LT, HSST-QIMY 1 2 1 St Luie2 " 2630 1 SA533BI I LT HSSTr-OIMY I ýýj!00 80 00 0-3(-2 2-- , 0 , 00 ,.200-100 0 100 200 300 400 500 600 Temperature ( 1F)Curve Fluence i SE USE 4-USE T @35 d-T @35, 1 0, 1 77,11 0. 38 0 2 0 1 75 -2.11 180.7 142-7 Figure 5-14 Charpy V-Notch Lateral Expansion vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Standard Reference Material WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-39 STANDARD REFERENCE MATERIAL6-0: Hyperbolic Tzagent CurvePinted an 1219f2014 9:14 AM Curve PlaMA Capsule Material Ori HMat#1 St Lucie 2 Uniiad SA533B1 LT HSST-OIMY 2 St. lucie 2 2630 SA533B1 LT HSST-OIMY 110 100-A 2 90 i. 70 --50~40 30 20 ' --" 2 0 10-300 -200 -100 0 100 200 300 400 500 600 Temperature (0 F)Curve ISE VUSE d-TSE T @50 d-T @50 1 0 100 0 79 0 2 0 100 0 19061 111.1 Figure 5-15 Charpy V-Notch Percent Shear vs. Temperature for the St. Lucie Unit 2 Reactor Vessel Standard Reference Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -40 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Longitudinal Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-41 24P, 150-F 26E, 170°F 1 21U, 200-F S 21 Y, 200-F Figure 5-17 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -42 1 355, F 37P, 0°F 32A, 20°F 32B, 70°F Figure 5-18 Charpy Impact Specimen Fracture Surfaces for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -43 Figure 5-19 Charpy Impact Specimen Fracture Surfaces for the St. Lucie Unit 2 Reactor Vessel Heat-Affected Zone Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-44 ultimate tensile Strength 02% Yield St-ength 40-0-20.0 100 200 M00 400 500 o0 Tumostwc CF)Legend: A and o and n are unirradiated A and o and o are irradiated to 2.25 x 101 9 2 (E > 1.0 MeV)Am Reduction 60.0 40.0 50.0 Total Elongation 20.0 -10.0 ---Uniform Elongation 0 to() _0;a0 4;0 500 000 Figure 5-20 Tensile Properties for St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-45 120.0 Ultinate Tensile Strength a _ _ _ _ _ _ _ _ _ _ _60.0 A 0.2% Yield Strength 4010 0 300 200 1300 4000 600 Thsealore

{'F)Legend: Aand o and m are unirradiated A and o and o are irradiated to 2.25 x 10'9 n/cm 2 (E > 1.0 MeV)1t.0 Arem Reduction 60.0 30.0 30D Total Elongation 20.0 1-Uniform Elongation 0 too "0 30 4;030 600 T=mpw-oe (')Figure 5-21 Tensile Properties for the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-46 IWO .Vltlmate Tensile Srength 0.2% Yied Strength 20.0 O00 4 0 100 W0 Tuiip--x (*F)400 w00 Legend: A and a and m are unirradiated A and o and o are irradiated to 2.25 x 10 1 9 n/cm 2 (E > 1.0 MeV)I Area Reduction-n 50.A i L 4 Total Elongation Uf0.0 A0. A I Uniformi Elongation 0 100 200 300 Tc4mpcrtw' (T1)400 600 Figure 5-22 Tensile Properties for the St. Lucie Unit 2 Reactor Vessel Heat Affected Zone Material WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5 -47 Westinghouse Non-Proprietary Class 3 5-47 2L5 -Tested at 150*F 2KU -Tested at 280°F 2JM -Tested at 550°F Figure 5-23 Fractured Tensile Specimens from St. Lucie Unit 2 Reactor Vessel Intermediate Shell Plate M-605-1 (Transverse Orientation)

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietwy Class 3 5-48 Westinghouse Non-Proprietaiy Class 3 5-48 3J4 -Tested at 72"F 3K7 -Tested at 150*F 3L5 -Tested at 550°F Figure 5-24 Fractured Tensile Specimens from the St. Lucie Unit 2 Reactor Vessel Surveillance Program Weld Metal (Heat # 83637)WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietaq Class 3 5-49 Westinghouse Non-Proprietaiy Class 3 5-49 4K5 -Tested at 72"F 4J5 -Tested at 240'F 4JK -Tested at 550"F Figure 5-25 Fractured Tensile Specimens from the St. Lucie Unit 2 Reactor Vessel Heat Affected Zone Material WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-50 U)U)1UU 90, 801 70, 60 50 40'30.20.10,...... ....................

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WCAP- 17939-NP May 2015 WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-51 5 IA U)0 10 20 Strain [%]30 Tensile Specimen 2JM Tested at 550°F Figure 5-27 Engineering Stress-Strain Curve for St. Lucie Unit 2 Intermediate Shell Plate M-605-1 Tensile Specimen 2JM (Transverse Orientation)

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0 10 20t Strain [%/0 Tensile Specimen 4J5 Tested at 240°F Figure 5-30 Engineering Stress-Strain Curves for St. Lucie Unit 2 Heat Affected Zone Material Tensile Specimens 4K5 and 4J5 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 5-55 Westinghouse Non-Proprietary Class 3 5-55 0 10 [%0 30 strain [M1 Tensile Specimen 4JK Tested at 550°F Figure 5-31 Engineering Stress-Strain Curve for St. Lucie Unit 2 Heat Affected Zone Material Tensile Specimen 4JK WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates S, transport analysis performed for the St. Lucie Unit 2 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 970, withdrawn at the end of the twentieth plant operating cycle, is provided.

In addition, the sensor sets from the previously withdrawn capsules (830 and 2630) 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 formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY) at 3020 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, "Analysis and Interpretation of Light-Water Reactor Surveillance Results," [Reference 17] recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference.

The energy-dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693-94, "Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom" [Reference 18]. 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" [Reference 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 using 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" [Reference 19]. 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 [Reference 20]. As an improvement, instead of the fluence rate synthesis technique, three-dimensional transport calculations were performed.

WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-2 6.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the St. Lucie Unit 2 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 830, 970, 1040, 2630, 277', and 2840 as shown in Figure 4-1. These full-core positions correspond to the following octant symmetric locations represented in Figure 6-1: 70 from the core cardinal axes (for the 830, 970, 2630 and 2770 capsules) and 140 from the core cardinal axes (for the 1040 and 2840 capsules).

The stainless steel specimen containers are 1.5-inch by 0.75-inch and are approximately 98 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane., thus spanning the approximate central eight feet of the 11.4-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 St. Lucie Unit 2 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 St. Lucie Unit 2 transport calculations, the r,0,z models depicted (given as r,0 section view) in Figures 6-1 and 6-2 were utilized since, with the exception of the capsules, the reactor is octant symmetric.

The rz section view depicted in Figure 6-3 shows the model having an axial span from an elevation one foot below the bottom of the active fuel and one foot above the top of the active fuel. These r,0,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.

For the reactor pressure vessel, the vessel averaged inner radius 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,0,z reactor models consisted of 151 radial by 119 azimuthal by 135 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 rO,z calculations was set at a value of 0.001.The core power distributions used in the plant-specific transport analysis were provided by Florida Power& Light Company for each of the first 21 fuel cycles at St. Lucie Unit 2. Specifically, the data utilized 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 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 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-3 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 [Reference 21] and the BUGLE-96 cross-section library [Reference 22]. 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 P 3 Legendre expansion, and angular discretization was modeled with an S 8 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 6-4. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 70 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 projected into the future.Similar information, in terms of both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 6-2, for the reactor vessel inner radius at four azimuthal locations.

The vessel data given in Table 6-2 were taken at the clad/base metal interface and thus represent maximum calculated exposure levels on the vessel. From the data provided in Table 6-2, it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the 2 0 th fuel cycle (i.e., after 25.55 EFPY at 3020 MWt of plant operation) was 1.73E+19 n/cm 2.These data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end of the 20' fuel cycle, as well as future projections to 32, 36, 40, 48, 55, and 60 EFPY at 3020 MWt. The calculations account for uprates from 2560 MWt to 2700 MWt that occurred at the end of Cycle 1, and from 2700 MWt to 3020 MWt that occurred at the end of Cycle 19. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 21 were representative of future plant operation.

The future projections are also based on the current reactor power level of 3020 MWt.The calculated fast neutron exposures for the three surveillance capsules withdrawn from the St. Lucie Unit 2 reactor are provided in Table 6-3. These assigned neutron exposure levels are based on the plant-and fuel-cycle-specific neutron transport calculations performed for the St. Lucie Unit 2 reactor. From the data provided in Table 6-3, Capsule 970 received a fluence (E > 1.0 MeV) of 2.25E+19 n/cm 2 after exposure through the end of the 2 0 th fuel cycle (i.e., after 25.55 EFPY at 3020 MWt of plant operation).

Updated lead factors for the St. Lucie Unit 2 surveillance capsules are provided in Table 6-4. The capsule lead factor is defined as the ratio of the calculated axial average 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 6-4, the lead factors for capsules that have been withdrawn from the reactor (83', 2630, and 970) were based on the calculated fluence values for the WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-4 irradiation period corresponding to the time of withdrawal for the individual capsules.

For the capsules remaining in the reactor (1040, 2770, and 2840), the lead factor corresponds to the calculated fluence values at the end of Cycle 20, the last completed fuel cycle for St. Lucie Unit 2.6.3 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 970, which was withdrawn from St. Lucie Unit 2 at the end of the twentieth fuel cycle, is summarized below.Reaction Rate (rps/atom)

ReactionMC Measured (M) Calculated (C)Ti-46(n,p)Sc-46 7.12E- 16 6.43E-16 1.11 Fe-54(n,p)Mn-54 3.77E- 15 3.65E-15 1.03 Ni-58(n,p)Co-58 4.98E- 15 4.77E-15 1.05 U-238(n,f)Cs-137 9.79E- 15 1.25E- 14 0.78 Average 0.99% standard deviation 14.7 The measured-to-calculated (M/C) reaction rate ratios for the Capsule 970 threshold reactions range from 0.78 to 1.11, and the average M/C ratio is 0.99 +/- 14.7% (1c). This direct comparison falls within the+/- 20% criterion specified in Regulatory Guide 1.190. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable-for St. Lucie Unit 2.6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the St. Lucie Unit 2 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages: 1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-5 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 St.Lucie Unit 2 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 St. Lucie Unit 2 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with St. Lucie Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confinr 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 St. Lucie Unit 2 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 20.Description Capsule and Vessel IR PCA Comparisons 3%H. B. Robinson Comparisons 3%Analytical Sensitivity Studies 11%Additional Uncertainty for Factors not Explicitly Evaluated 5%Net Calculational Uncertainty 13%The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random, and no systematic bias was applied to the analytical results.The plant-specific measurement comparisons described in Appendix A support these uncertainty assessments for St. Lucie Unit 2.WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-6 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center Cycle Cycle Total Fluence Rate (n/cm 2-s) Fluence (n/cm 2)[D Length Time (EFPY) (EFPY) 7-Degree 14-Degree 7-Degree 14-Degree 1 1.11 1.11 3.99E+10 2.86E+10 1.40E+18 1.00E+18 2 1.12 2.23 3.94E+10 2.72E+10 2.79E+18 1.96E+18 3 1.22 3.45 3.55E+10 2.48E+10 4.16E+18 2.92E+18 4 1.16 4.61 2.62E+10 2.16E+10 5.12E+18 3.71E+18 5 1.3 5.91 2.58E+10 2.16E+10 6.17E+18 4.60E+18 6 1.35 7.26 2.54E+10 1.85E+10 7.25E+18 5.39E+18 7 1.21 8.47 2.74E+10 1.96E+10 8.30E+18 6.14E+18 8 1.38 9.85 1.69E+10 1.40E+10 9.04E+18 6.75E+18 9 1.22 11.07 2.57E+10 2.16E+10 1.00E+19 7.58E+18 10 1.44 12.51 2.73E+10 2.21E+10 1.13E+19 8.59E+18 11 1.32 13.83 2.45E+10 1.95E+10 1.23E+19 9.39E+18 12 1.51 15.34 2.39E+10 1.75E+10 1.34E+19 1.02E+19 13 1.29 16.63 2.65E+10 1.96E+10 1.45E+19 1.10E+19 14 1.43 18.06 2.33E+10 1.75E+10 1.56E+19 1.18E+19 15 1.15 19.21 2.72E+10 1.98E+10 1.65E+19 1.25E+19 16 1.25 20.46 2.80E+10 2.06E+10 1.77E+19 1.33E+19 17 1.25 21.71 2.65E+10 1.97E+10 1.87E+19 1.41E+19 18 1.42 23.13 2.59E+10 1.98E+10 1.99E+19 1.50E+19 19 1.19 24.32 3.24E+10 2.44E+10 2.11E+19 1.59E+19 20 1.23 25.55 3.72E+10 2.72E+10 2.25E+19 1.70E+19 21 1.38 26.93 3.56E+10 2.63E+10 2.41E+19 1.81E+19 FutureM' 5.07 32.00 3.92E+10 2.90E+10 3.03E+19 2.28E+19 Future 4.00 36.00 3.92E+10 2.90E+10 3.53E+19 2.64E+19 Future 4.00 40.00 3.92E+10 2.90E+10 4.02E+19 3.01E+19 Future 8.00 48.00 3.92E+10 2.90E+10 5.01E+19 3.74E+19 Future 7.00 55.00 3.92E+10 2.90E+10 5.88E+19 4.38E+19 Future 5.00 60.00 3.92E+10 2.90E+10 6.49E+19 4.84E+19 Notes: 1. Fluence rate (and fluence) projections were increased by a factor of 10% to allow for variations in future core power distributions.

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-7 Table 6-1 (Continued)

Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center Cycle Cycle Total dpa/s dpa ED Length Time FPY) (EFPY) 7-Degree 14-Degree 7-Degree 14-Degree 1 1.11 1.11 5.81E-11 4.18E-11 2.04E-03 1.47E-03 2 1.12 2.23 5.74E-l1 3.99E-11 4.07E-03 2.87E-03 3 1.22 3.45 5.18E-11 3.64E-11 6.06E-03 4.28E-03 4 1.16 4.61 3.82E-11 3.17E-11 7.46E-03 5.44E-03 5 1.3 5.91 3.77E-11 3.17E-11 9.OOE-03 6.73E-03 6 1.35 7.26 3.70E- 11 2.72E- I1 1.06E-02 7.89E-03 7 1.21 8.47 4.OOE-11 2.87E-11 1.21E-02 8.99E-03 8 1.38 9.85 2.47E-I 1 2.06E-11 1.32E-02 9.89E-03 9 1.22 11.07 3.76E-11 3.16E-11 1.46E-02 1.11E-02 10 1.44 12.51 3.99E-11 3.23E-11 1.64E-02 1.26E-02 11 1.32 13.83 3.57E-11 2.85E-11 1.79E702 1.38E-02 12 1.51 15.34 3.49E-11 2.56E-11 1.96E-02 1.50E-02 13 1.29 16.63 3.87E-11 2.88E-11 2.12E-02 1.62E-02 14 1.43 18.06 3.40E- 11 2.57E- 11 2.27E-02 1.73E-02 15 1.15 19.21 3.97E-11 2.90E-11 2.41E-02 1.84E-02 16 1.25 20.46 4.09E- 11 3.01E-11 2.58E-02 1.96E-02 17 1.25 21.71 3.87E- 11 2.89E- 11 2.73E-02 2.07E-02 18 1.42 23.13 3.78E-11 2.90E-11 2.90E-02 2.20E-02 19 1.19 24.32 4.73E- 11 3.58E- 11 3.08E-02 2.33E-02 20 1.23 25.55 5.42E- 11 3.99E- I 1 3.29E-02 2.49E-02 21 1.38 26.93 5.19E-11 3.86E-11 3.51E-02 2.66E-02 Future°')

5.07 32.00 5.71 E- 11 4.24E- 11 4.43E-02 3.34E-02 Future 4.00 36.00 5.71E-11 4.24E-11 5.15E-02 3.87E-02 Future 4.00 40.00 5.71 E- 11 4.24E- 11 5.87E-02 4.41E-02 Future 8.00 48.00 5.71E-11 4.24E-11 7.31E-02 5.48E-02 Future 7.00 55.00 5.71E- 11 4.24E- 11 8.57E-02 6.42E-02 Future 5.00 60.00 5.71E-11 4.24E- 11 9.47E-02 7.09E-02 Notes: 1. dpa/s (and dpa) projections were increased by a factor of 10% to allow for variations in future core power distributions.

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-8 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cycle Cycle Total Fluence Rate (n/cm 2-s)ID Length Time (EFPY) LegPT ) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.11 1.11 3.08E+10 2.00E+10 1.62E+10 1.26E+10 3.08E+10 2 1.12 2.23 3.22E+10 1.94E+10 1.61E+10 1.27E+10 3.22E+10 3 1.22 3.45 2.89E+ 10 1.76E+10 1.39E+10 1.06E+10 2.89E+ 10 4 1.16 4.61 1.97E+10 1.59E+10 1.48E+10 1.17E+10 1.97E+10 5 1.3 5.91 1.92E+10 1.58E+10 1.47E+10 1.13E+10 1.92E+10 6 1.35 7.26 2.07E+10 1.35E+10 1.16E+10 9.79E+09 2.07E+10 7 1.21 8.47 2.22E+10 1.41E+10 1.15E+10 1.01E+10 2.22E+10 8 1.38 9.85 1.36E+10 1.09E+ 10 1.24E+10 1.01E+10 1.36E+10 9 1.22 11.07 1.84E+10 1.55E+10 1.31E+10 9.87E+09 1.89E+10 10 1.44 12.51 1.99E+10 1.58E+10 1.31E+10 1.06E+10 2.01E+10 11 1.32 13.83 1.85E+10 1.40E+10 1.36E+10 1.15E+10 1.85E+10 12 1.51 15.34 1.88E+10 1.24E+10 1.18E+10 1.07E+10 1.88E+10 13 1.29 16.63 2.06E+10 1.39E+10 1.24E+10 1.13E+10 2.06E+10 14 1.43 18.06 1.80E+10 1.25E+10 1.17E+10 1.06E+10 1.80E+10 15 1.15 19.21 2.10E+10 1.39E+10 1.14E+10 9.72E+09 2.10E+10 16 1.25 20.46 2.14E+10 1.44E+10 1.17E+10 9.87E+09 2.14E+10 17 1.25 21.71 2.01E+10 1.38E+10 1.12E+10 9.65E+09 2.01E+10 18 1.42 23.13 1.93E+10 1.40E+10 1.15E+10 1.02E+10 1.93E+10 19 1.19 24.32 2.46E+10 1.73E+10 1.57E+10 1.29E+10 2.46E+10 20 1.23 25.55 2.88E+10 1.93E+10 1.76E+10 1.53E+10 2.88E+10 21 1.38 26.93 2.75E+10 1.87E+10 1.74E+10 1.49E+ 10 2.75E+10 FutureM 1 5.07 32.00 3.03E+10 2.06E+10 1.91E+10 1.64E+10 3.03E+10 Future 4.00 36.00 3.03E+10 2.06E+10 1.91E+10 1.64E+10 3.03E+10 Future 4.00 40.00 3.03E+10 2.06E+10 1.91E+10 1.64E+10 3.03E+10 Future 8.00 48.00 3.03E+10 2.06E+10 1.91E+10 1.64E+10 3.03E+10 Future 7.00 55.00 3.03E+10 2.06E+10 1.91E+10 1.64E+10 3.03E+10 Future 5.00 60.00 3.03E+10 2.06E+10 1.91E+10 1.64E+10 3.03E+10 Notes: 1. Fluence rate projections were increased by a factor of 10% to allow for variations in future core power distributions.

WCAP- 17939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-9 Table 6-2 (Continued)

Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cycle Cycle Total Fluence (n/cm 2)Cce Length Time ED (EFPY) TeF 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.11 1.11 1.08E+18 7.01E+17 5.67E+17 4.42E+17 1.08E+18 2 1.12 2.23 2.21E+18 1.38E+18 1.13E+18 8.86E+17 2.21E+18 3 1.22 3.45 3.31E+18 2.05E+18 1.67E+18 1.29E+18 3.31E+18 4 1.16 4.61 4.03E+18 2.64E+18 2.21E+18 1.72E+18 4.03E+18 5 1.3 5.91 4.82E+18 3.28E+18 2.81E+18 2.19E+18 4.82E+18 6 1.35 7.26 5.68E+18 3.85E+18 3.30E+18 2.59E+18 5.68E+18 7 1.21 8.47 6.52E+18 4.38E+18 3.73E+18 2.97E+18 6.52E+18 8 1.38 9.85 7.11E+18 4.85E+18 4.26E+18 3.41E+18 7.11E+18 9 1.22 11.07 7.78E+18 5.42E+18 4.75E+18 3.77E+18 7.78E+18 10 1.44 12.51 8.69E+18 6.13E+18 5.34E+18 4.25E+18 8.69E+18 11 1.32 13.83 9.46E+18 6.72E+18 5.91E+18 4.73E+18 9.46E+18 12 1.51 15.34 1.04E+19 7.31E+18 6.47E+18 5.24E+18 1.04E+19 13 1.29 16.63 1.12E+19 7.87E+18 6.97E+18 5.70E+18 1.12E+19 14 1.43 18.06 1.20E+19 8.44E+18 7.50E+18 6.18E+18 1.20E+19 15 1.15 19.21 1.28E+19 8.94E+18 7.91E+18 6.53E+18 1.28E+19 16 1.25 20.46 1.36E+19 9.51E+18 8.37E+18 6.92E+18 1.36E+19 17 1.25 21.71 1.44E+19 1.01E+19 8.81E+18 7.30E+18 1.44E+19 18 1.42 23.13 1.53E+19 1.07E+19 9.33E+18 7.76E+18 1.53E+19 19 1.19 24.32 1.62E+19 1.13E+19 9.92E+18 8.24E+18 1.62E+19 20 1.23 25.55 1.73E+19 1.21E+19 1.06E+19 8.84E+18 1.73E+19 21 1.38 26.93 1.85E+19 1.29E+19 1.14E+19 9.49E+18 1.85E+19 FutureM' 5.07 32.00 2.33E+19 1.62E+19 1.44E+19 1.21E+19 2.33E+19 Future 4.00 36.00 2.72E+19 1.88E+19 1.68E+19 1.42E+19 2.72E+19 Future 4.00 40.00 3.10E+19 2.14E+19 1.93E+19 1.63E+19 3.10E+19 Future 8.00 48.00 3.86E+19 2.66E+19 2.41E+19 2.04E+19 3.86E+19 Future 7.00 55.00 4.53E+19 3.11E+19 2.83E+19 2.41E+19 4.53E+19 Future 5.00 60.00 5.01E+19 3.44E+19 3.13E+19 2.66E+19 5.01E+19 Notes: 1. Fluence rate (and fluence) projections were increased by a factor of 10% to allow for variations in future core power distributions.

WCAP- 17939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-10 Table 6-2 (Continued)

Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cycle Cycle Total dpa/s ED Length Time (EFPY) (EFPY) 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.11 1.11 4.68E-11 3.06E-11 2.47E-11 1.94E-11 4.68E-11 2 1.12 2.23 4.89E-11 2.97E-11 2.46E-11 1.95E-11 4.89E-11 3 1.22 3.45 4.39E-11 2.70E-11 2.13E-11 1.63E-11 4.39E-11 4 1.16 4.61 3.01E-11 2.44E-11 2.27E-11 1.81E-1I 3.01E-11 5 1.3 5.91 2.93E-11 2.43E-11 2.25E-11 1.74E-11 2.93E-11 6 1.35 7.26 3.15E-11 2.08E-11 1.78E-11 1.50E-11 3.15E-11 7 1.21 8.47 3.38E-11 2.16E-11 1.76E-11 1.55E-11 3.38E-11 8 1.38 9.85 2.07E-11 1.67E-11 1.90E-11 1.56E-11 2.07E-11 9 1.22 11.07 2.81E-11 2.38E-11 2.01E-11 1.52E-11 2.88E-11 10 1.44 12.51 3.03E-11 2.41E-I 1 2.OOE-11 1.63E-11 3.06E-11 11 1.32 13.83 2.83E-1I 2.15E-11 2.08E-I1 1.77E-11 2.83E-11 12 1.51 15.34 2.86E-11 1.91E-11 1.80E-11 1.65E-11 2.86E-11 13 1.29 16.63 3.14E-11 2.13E-11 1.90E-11 1.74E-11 3.14E-11 14 1.43 18.06 2.74E- 11 1.92E- 11 1.79E- 11 1.62E- 11 2.74E- 11 15 1.15 19.21 3.21E-11 2.13E-11 1.75E-11 1.49E-11 3.21E-11 16 1.25 20.46 3.26E-11 2.21E-11 1.79E-11 1.52E-l1 3.26E-11 17 1.25 21.71 3.06E-11 2.12E-11 1.71E-11 1.48E-11 3.06E-11 18 1.42 23.13 2.95E-11 2.14E-11 1.76E-11 1.57E-1l 2.95E-11 19 1.19 24.32 3.74E-11 2.65E-11 2.39E-11 1.98E-11 3.74E-11 20 1.23 25.55 4.39E-1l 2.95E-11 2.70E- 11 2.35E-11 4.39E-11 21 1.38 26.93 4.19E-11 2.87E-1l 2.66E-11 2.29E-11 4.19E-l1 Future~l 5.07 32.00 4.61E-11 3.15E-11 2.92E-11 2.52E-11 4.61E-11 Future 4.00 36.00 4.61E-11 3.15E-11 2.92E-11 2.52E-11 4.61E-l1 Future 4.00 40.00 4.61E-11 3.15E-11 2.92E-11 2.52E-11 4.61E-11 Future 8.00 48.00 4.61E-11 3.15E-11 2.92E-11 2.52E-11 4.61E-11 Future 7.00 55.00 4.61E-l1 3.15E-11 2.92E-11 2.52E-11 4.61E-11 Future 5.00 60.00 4.61E-11 3.15E-11 2.92E-11 2.52E-11 4.61E-11 Notes: I. dpa/s projections were increased by a factor of 10% to allow for variations in future core power distributions.

WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-11 Table 6-2 (Continued)

Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Cycle Cycle Total dpa Cce Length Time ID (EFPY) TeF 0-Degree 15-Degree 30-Degree 45-Degree Maximum 1 1.11 1.11 1.64E-03 1.07E-03 8.66E-04 6.79E-04 1.64E-03 2 1.12 2.23 3.35E-03 2.11E-03 1.73E-03 1.36E-03 3.35E-03 3 1.22 3.45 5.04E-03 3.15E-03 2.55E-03 1.99E-03 5.04E-03 4 1.16 4.61 6.14E-03 4.04E-03 3.38E-03 2.65E-03 6.14E-03 5 1.3 5.91 7.34E-03 5.04E-03 4.30E-03 3.36E-03 7.34E-03 6 1.35 7.26 8.65E-03 5.90E-03 5.04E-03 3.99E-03 8.65E-03 7 1.21 8.47 9.92E-03 6.71E-03 5.70E-03 4.57E-03 9.92E-03 8 1.38 9.85 1.08E-02 7.43E-03 6.52E-03 5.23E-03 1.08E-02 9 1.22 11.07 1.18E-02 8.31E-03 7.25E-03 5.79E-03 1.18E-02 10 1.44 12.51 1.32E-02 9.41E-03 8.16E-03 6.53E-03 1.32E-02 11 1.32 13.83 1.44E-02 1.03E-02 9.03E-03 7.27E-03 1.44E-02 12 1.51 15.34 1.58E-02 1.12E-02 9.89E-03 8.05E-03 1.58E-02 13 1.29 16.63 1.70E-02 1.21E-02 1.07E-02 8.76E-03 1.70E-02 14 1.43 18.06 1.83E-02 1.29E-02 1.15E-02 9.50E-03 1.83E-02 15 1.15 19.21 1.94E-02 1.37E-02 1.21E-02 1.00E-02 1.94E-02 16 1.25 20.46 2.07E-02 1.46E-02 1.28E-02 1.06E-02 2.07E-02 17 1.25 21.71 2.19E-02 1.54E-02 1.35E-02 1.12E-02 2.19E-02 18 1.42 23.13 2.32E-02 1.64E-02 1.43E-02 1.19E-02 2.32E-02 19 1.19 24.32 2.47E-02 1.74E-02 1.52E-02 1.27E-02 2.47E-02 20 1.23 25.55 2.64E-02 1.85E-02 1.62E-02 1.36E-02 2.64E-02 21 1.38 26.93 2.82E-02 1.98E-02 1.74E-02 1.46E-02 2.82E-02 Future°')

5.07 32.00 3.56E-02 2.48E-02 2.20E-02 1.86E-02 3.56E-02 Future 4.00 36.00 4.14E-02 2.88E-02 2.57E-02 2.18E-02 4.14E-02 Future 4.00 40.00 4.72E-02 3.28E-02 2.94E-02 2.50E-02 4.72E-02 Future 8.00 48.00 5.88E-02 4.07E-02 3.68E-02 3.13E-02 5.88E-02 Future 7.00 55.00 6.90E-02 4.77E-02 4.33E-02 3.69E-02 6.90E-02 Future 5.00 60.00 7.63E-02 5.27E-02 4.79E-02 4.09E-02 7.63E-02 Notes: 1. dpa/s (and dpa) projections were increased by a factor of 10% to allow for variations in future core power distributions.

WCAP- 17939-NP May 2015 WCAP- I17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-12 Table 6-3 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from St. Lucie Unit 2 Irradiation Fluence Iron Atom Capsule Time (E > 1.0 MeV) Displacements Cycle(s) [EFPY] In/cm 2] Idpa]830 1 1.11 1.40E+ 18 2.04E-03 2630 1-9 11.07 1.00E+19 1.46E-02 970 1-20 25.55 2.25E+19 3.29E-02 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-13 Table 6-4 Calculated Surveillance Capsule Lead Factors Capsule Location Status Lead Factor 830 Withdrawn EOC 1 1.30 2630 Withdrawn EOC 9 1.29 970 Withdrawn EOC 20 1.30 1040 In Reactor")

0.98 2770 In Reactor(1) 1.30 2840 In Reactor(1' 0.98 Notes: 1, Lead factors are based on the cumulative exposures from Cycles I through 20.WCAP- 17939-NP May 2015 WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-14-1I*ho Sad~sbbm$W couw*IE 0 O-oQt+o 3.1.+02 Figure 6-1 St. Lucie Unit 2 rO,z Reactor Geometry rO Plan View without Surveillance Capsules WCAP-17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 6-15-M'= 000*-o 1 W i ý u I 5~wo-'-'p-C'4 0 10 Ui O~4OO 15Xti2 Figure 6-2 St. Lucie Unit 2 rOz Reactor Geometry rO Plan View with 70 and 14° Surveillance Capsules WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietaiy Class 3 6-16-Ce JLi M*w or**AM sw M%"Asw*w An*aWl-Pi 3.530E+02 cmi Figure 6-3 St. Lucie Unit 2 r,0,z Reactor Geometry rz Axial View WCAP- 17939-NP May 2015 Revision 0 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. 10]. Note that it is recommended for future capsule(s) to be removed from the St. Lucie Unit 2 reactor vessel.Table 7-1 Surveillance Capsule Withdrawal Schedule 830 Withdrawn (EOC 1)1.30 1.11 1.40E+18 2630 Withdrawn (EOC 9) 1.29 11.07 1.00E+ 19 970 Withdrawn (EOC 20) 1.30 25.55 2.25E+19 2770 In Reactor 1.30 44.1(d) 4.53E+ 19(d)1040 In Reactor 0.98 (e) (e)2840 In Reactor 0.98 (e) (e)Notes: (a) Updated in Capsule 970 dosimetry analysis; see Table 6-4.(b) EFPY from plant startup.(c) Updated in Capsule 97' dosimetry analysis; see Table 6-3.(d) Capsule 2770 should be withdrawn at the next vessel refueling outage after 44.1 EFPY of plant operation, which is when the fluence on the capsule would equal the projected 60-year (55 EFPY)peak vessel fluence.(e) Capsules 1040 and 2840 currently have a lead factor slightly less than one. If additional metallurgical data is needed for St. Lucie Unit 2, such as in support of a second license renewal to 80 total years of operation (estimated at 74 EFPY), 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 St. Lucie will apply for a second license extension, and given that many cycles of irradiation will be required for Capsules 1040 and 2840 to accumulate fluence greater than the vessel wall fluence at 74 EFPY, it is suggested that a potential relocation decision be implemented prior to 40 total years of operation.

WCAP- 17939-NP May 2015 WCAP-17939-NP May 2015 Revision 0 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-L-MCM-00 1, Revision 0, Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of St. Lucie No. 2 Reactor Vessel Materials, November 1979.4. BAW- 1880, Revision 0, Analysis of Capsule W-83 Florida Power and Light Company St. Lucie Plant Unit No. 2, Reactor Vessel Materials Surveillance Program, September 1985.5. ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, ASTM, 1973.6. Appendix G of the ASME Boiler and Pressure Vessel (B&PV) Code, Section X1, Division 1, Fracture Toughness Criteria for Protection Against Failure.7. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM.8. NUREG/CR-6413; ORNL/TM-13133, Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials, April 1996.9. CE-NPSD- 1039, Revision 2, Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, June 1997.10. ASTM El185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM, 1982.11. ASTM E23-12c, Standard Test Methods for Notched Bar hnpact Testing of Metallic Materials, ASTM, 2012.12. ASTM E2298-13a, Standard Test Method for Instrumented hnpact Testing of Metallic Materials, ASTM, 2013.13. ASTM A370-13, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, ASTM, 2013.14. ASTM E8/E8M-13a, Standard Test Methods for Tension Testing of Metallic Materials, ASTM, 2013.15. ASTM E21-09. Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, ASTM, 2009.WCAP- 17939-NP May 2015 Revision 0 Westinghouse Non-Proprietary Class 3 8-2 16. WCAP-15040, Revision 1, Analysis of Capsule 263°from the Florida Power & Light Company St. Lucie Unit 2 Reactor Vessel Radiation Surveillance Program, February 2010.17. ASTM E853-13, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, ASTM, 2014 18. 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, 1994.19. Regulatory Guide 1.190, Calculational and Dosimetrj Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.20. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Ov'erpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.21. WCAP- 16083-NP, Revision 1, Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry, April 2013.22. 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.WCAP- 17939-NP May 2015 Revision 0