Text

Response to Request for Additional Information Regarding Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits
	ML21113A044
Person / Time
Site:	Beaver Valley
Issue date:	04/22/2021
From:	Grabnar J; Energy Harbor Nuclear Corp
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	EPID L-2020-LLA-0233, L-21-106
	Download: ML21113A044 (222)
	v • d • e

energy harbor Energy Harbor Nuclear Corp.

Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 John J. Grabnar 724-682-5234 Site Vice President, Beaver Valley Nuclear April 22, 2021 L-21-106 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 Response to Request for Additional Information Regarding Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits (EPID L-2020-LLA-0233)

By correspondence dated October 30, 2020 (Accession No. ML20304A215), Energy Harbor Nuclear Corp. submitted to the Nuclear Regulatory Commission (NRC) a request to amend Technical Specification 5.6.4, "Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," Item b, by replacing the currently referenced analytical methods with more recent analytical methods found acceptable by the Nuclear Regulatory Commission (NRC) staff for calculating reactor vessel neutron fluence and reactor coolant system pressure and temperature limits when updating the reactor coolant system Pressure and Temperature Limits Report.

By email dated March 9, 2021 (Accession No. ML21068A350), the NRC staff requested additional information regarding the October 30, 2020 request. The Energy Harbor Nuclear Corp. response to the NRC request for information is attached.

The following topical reports are enclosed as they present the technical basis for the reactor coolant system heatup and cooldown limit curves and are referenced in the attached response to the NRC request for information.

A. Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation B. Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations Based on Capsule Y Data

Beaver Valley Power Station, Unit Nos. 1 and 2 L-21-106 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Manager - Fleet Licensing, at (330) 696-7208.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April ...2.2..._, 2021.

Sincerely, Grabnar, John 19072 Site Vice President, Beaver Valley Grahna.-,J o/rn 7 9072 ~~,~ ;;i:~2vi"li~i~!ocumcnt John J. Grabnar

Attachment:

Response to Request for Additional Information

Enclosures:

A. WCAP-18102-NP, Revision 2, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, March 2021.

B. WCAP-18559-NP, Revision 1, Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations Based on Capsule Y Data, March 2021.

cc: NRC Region I Administrator NRC Resident Inspector NRR Project Manager Director BRP/DEP Site BRP/DEP Representative

Attachment L-21-106 Response to Request for Additional Information Page 1 of 7 NRC staff requests for additional information are provided below in bold text and are followed by the Energy Harbor Nuclear Corp. response. References cited in the responses are listed following the last response.

1. Provide consistent fast neutron (E > 1.0 MeV) fluence values and corresponding EFPY for the period applicable to the LAR for beltline (and extended beltline) materials.

Response

Table 1 provides a comparison of the fluence results for Beaver Valley Power Station (BVPS), Unit 1, from the fluence model based on the methodology in WCAP-18124-NP-A, Revision 0, (RAPTOR, Reference 1, hereafter referred to as WCAP-18124-NP-A),

that is, the methodology supporting the license amendment request (LAR), to the fluence results in WCAP-18102-NP, Revision 2, (Reference 2, hereafter referred to as WCAP-18102-NP), which were developed with the fluence methodology in WCAP-14040-A, Revision 4, (DORT, Reference 3, hereafter referred to as WCAP-14040-A).

The comparison is shown at 50 effective full power years (EFPY) to correspond to WCAP-18102-NP and the current BVPS, Unit 1, PTLR (Reference 4). Fifty EFPY was selected at the time of creation of WCAP-18102-NP as it represented a realistic projection of the EFPY of BVPS, Unit 1, at the end of the 60-year license renewal period. Current projections confirm that BVPS, Unit 1, is not anticipated to exceed 50 EFPY prior to the end of the 60-year license renewal period. However, 50 EFPY is not intended to be an applicability period for this LAR. This LAR is intended to supplement the currently approved fluence methodology in Technical Specification 5.6.4, that is, WCAP-14040-A, to allow the use of WCAP-18124-NP-A as an alternative methodology for future PTLR revisions. The PTLR revisions will define the applicability period as appropriate, when submitted. Table 2 of Enclosure B of the submittal was provided to address the requirement in the WCAP-18124-NP-A safety evaluation report to benchmark the fluence results in the extended beltline prior to implementation.

However, the LAR does not change the BVPS, Unit 1 or Unit 2, PTLRs (References 4 and 5).

While some beltline fluence values show an increase using the RAPTOR methodology, the fluence values in WCAP-18102-NP were generated with an NRC-approved methodology documented in WCAP-14040-A and are in compliance with Regulatory Guide 1.190 (Reference 6). Therefore, the fluence values remain acceptable as the BVPS, Unit 1, licensing basis and no revision to the PTLR is required.

For BVPS, Unit 2, the latest fluence projections are contained in WCAP-18559-NP, Revision 1 (Reference 7, hereafter referred to as WCAP-18559-NP).

Attachment L-21-106 Page 2 of 7 Table 1 BVPS, Unit 1, Beltline and Extended Beltline Fluence Values at 50 EFPY DORT Surface Fluence(1) RAPTOR Surface Fluence(2)

(n/cm2, E > 1.0 MeV)(3,4) (n/cm2, E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell 5.88E+19 6.07E+19 Plates Lower Shell Plates 5.89E+19 5.99E+19 Intermediate to Lower 5.88E+19 5.99E+19 Shell Girth Weld Intermediate Shell 1.13E+19 1.11E+19 Longitudinal Welds Lower Shell Longitudinal 1.14E+19 1.11E+19 Welds Reactor Vessel Extended Beltline Materials Upper Shell Forging 7.18E+18 6.56E+18 Upper to Intermediate 7.18E+18 6.56E+18 Shell Girth Weld Inlet Nozzle to Upper 2.10E+17 1.36E+17 Shell Weld - Lowest Extent Outlet Nozzle to Upper 1.61E+17 9.69E+16 Shell Weld - Lowest Extent Lower Shell to Bottom 1.53E+16 1.20E+16 Head Weld Notes:

1. Taken from WCAP-18102-NP (Reference 2), Section 2.

2. Values interpolated from 48 EFPY and 54 EFPY fluence values in WCAP-17896-NP, Revision 0, (Reference 8), Section 6.

3. Neutrons per square centimeter (n/cm2)

4. Total energy greater than 1.0 mega electron-volt (E > 1.0 MeV) 2.a. Describe how the P-T limit curves for BVPS, Units 1 and 2, consider all ferritic pressure boundary components of the reactor vessel that are predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed operating period.

Attachment L-21-106 Page 3 of 7

Response

WCAP-18102-NP (Reference 2) and WCAP-18559-NP (Reference 7) consider embrittlement for all reactor vessel materials with fluence values projected to be greater than 1 x 1017 n/cm2 (E > 1 MeV) at the end of the license operating period as shown in the calculation of adjusted reference temperature (ART) values in these documents.

The acceptability of the inlet and outlet nozzles with respect to the pressure and temperature limit curves for BVPS, Units 1 and 2, are addressed in WCAP-18102-NP, Appendix B, and WCAP-18559-NP, Section 9, respectively. Both units use the evaluation in PWROG-15109-NP-A, Revision 0, (Reference 9), which generically addresses the concerns in NRC Regulatory Issue Summary (RIS) 2014-11 (Reference

10) that the pressure and temperature limit curves account for the higher stresses in the nozzle corner region. The NRC safety evaluation (Reference 11) of PWROG-15109-NP-A, Sections 4.10 and 4.11, found the generic nozzle pressure and temperature curves acceptable and bounded by the beltline pressure and temperature limit curves for domestic PWRs. This determination is valid up to a nozzle fluence of 4.28 x 1017 n/cm2.

The acceptability of all ferritic material outside the reactor vessel is addressed in WCAP-18102-NP, Appendix C for BVPS, Unit 1, and WCAP-18559-NP, Appendix B, for BVPS, Unit 2. These sections review the design requirements of the components to ensure they are consistent with RIS 2014-11, which states the following:

As specified in Sections I and IV.A of 10 CFR Part 50, Appendix G, ferritic RCPB [reactor coolant pressure boundary] components outside of the reactor vessel must meet the applicable requirements of ASME Code,Section III, Rules for Construction of Nuclear Facility Components.

The current licensing basis for the BVPS, Unit 2, pressure and temperature limit curves is WCAP-16528-NP, Revision 1, (Reference 12, hereafter referred to as WCAP-16528-NP), which does not explicitly address this topic. However, the fluence period (54 EFPY) considered in the cited WCAP-18559-NP is greater than or equal to the fluence periods considered in WCAP-16528-NP. Thus, the justification based on fluence in WCAP-18559-NP, that the nozzles and other reactor coolant pressure boundary components outside of the beltline are not limiting with respect to pressure and temperature limits at 54 EFPY, also applies to the current PTLR EFPY period.

2.b. If the current P-T limit curves do not consider all ferritic pressure boundary components of the reactor vessel that are predicted to experience a neutron fluence exposure greater than 1x1017 n/cm2 (E > 1 MeV) at the end of the licensed operating period, provide appropriately revised P-T limit curves for review.

Response

See response to request for additional information 2.a.

Attachment L-21-106 Page 4 of 7

3. Provide the technical basis for BVPS, Unit 2, analogous to the information for BVPS, Unit 1, contained in WCAP-18102-NP, Rev. 0. Provide inputs and analysis of the methodology and results for the generation of heatup and cooldown curves for normal operation for BVPS, Unit 2.

Response

WCAP-18559-NP (Reference 7) provides the requested information. The pressure and temperature limit curves developed in WCAP-18559-NP would be incorporated into the PTLR after WCAP-18124-NP-A (Reference 1) is identified as an NRC approved methodology in Technical Specification 5.6.4.

4. The staff therefore requests that, unless it plans to submit an exemption, the licensee:

1. Revise the ART calculations to include ART and margin values for all RV beltline and extended beltline materials for BVPS, Units 1 and 2.

2. If WCAP-18102-NP, Rev. 0 is maintained as a reference for BVPS, Unit 1, correct the associated tables and revise the document to remove the reference to TLR-RES/DE/CIB-2013-01.

Response

4.1. WCAP-18102-NP was revised to Revision 2 (Reference 2). This revision removes TLR-RES/DE/CIB-2013- 01 from the 1/4T and 3/4T ART calculations for BVPS, Unit 1. The results for the extended beltline components are still bounded by the 1/4T and 3/4T ART values used to generate the pressure and temperature curves (244.5 degrees Fahrenheit or °F at 1/4T and 209.5°F at 3/4T) and thus the pressure and temperature limit curves remain unchanged.

BVPS, Unit 2, does not use TLR-RES/DE/CIB-2013-01 in ART calculations in the current licensing basis in WCAP-16528-NP (Reference 12), nor in WCAP-18559-NP (Reference 7), which is cited elsewhere in this letter and contains the latest ART calculations at 54 EFPY.

4.2. WCAP-18102-NP has been updated to Revision 2 (Reference 2) and no longer includes a reference to TLR-RES/DE/CIB-2013-01.

5. Please indicate (a) whether the uncertainty analysis referred to in LAR Enclosure B deviated from that presented in WCAP-18124-NP-A, section 4.4, and if so, please indicate how. Additionally, for each of the parameters specified below, please indicate (b) whether each of the below potential contributors to uncertainty were taken into account in the uncertainty estimation or provide justification for not explicitly including them.

i. Approximations (such as homogenization) made in modeling core insulation, biological shield, and supplementary shield

Attachment L-21-106 Page 5 of 7 ii. Uncertainties in width of gap between RPV and biological shield iii. Homogenized density of water and structural materials between top of core and (including) upper grid plate iv. Order of angular quadrature and anisotropic scattering treatment Response 5(a):

The uncertainty analysis performed in support of LAR Enclosure B (Reference 13) is consistent with that presented in WCAP-18124-NP-A, Section 4.4 (Reference 1), but also included more parameters related to extended beltline materials, which were not investigated in WCAP-18124-NP-A, Section 4.4.

The uncertainty parameters provided in the bullets following have been evaluated for the RPV extended beltline region in support of LAR Enclosure B.

Geometrical meshing (Mesh)

Legendre order of scattering - P3 vs. P5 (PN)

Differencing scheme (DS) - directional theta-weighting (dtw) vs. theta-weighting (tw)

(DS)

Quadrature Sets - level symmetric quadrature sets: S8, S12, S14, S16, S18, and S20 (SN)

Stainless steel reactor internals thickness tolerances (Internals-radial)

Water annuli thickness between core barrel and RPV (Water annuli)

RPV thickness (RPV)

Reactor coolant temperature variation during normal operation (Water temperature)

Peripheral assembly source magnitude (Peripheral source)

Peripheral assembly burnup (Peripheral burnup)

Axial power distribution variation over the course of a fuel cycle (Axial PD)

Relative spatial distribution of the source from relative pin powers of peripheral assemblies (Pin power)

Core periphery modeling (Core periphery)

Active core elevation cycle-to-cycle variations (Fuel elevation)

Material mixture modeling of upper and lower RPV internals (Internals-axial)

Axial variation of coolant temperature modeling (Water temperature modeling)

Cycle-average axial power distribution modeling using core-average and peripheral assembly-average distributions (Axial PD periphery)

Concrete biological shield composition sensitivity - fixed water content (Concrete)

Attachment L-21-106 Page 6 of 7

Concrete biological shield composition sensitivity - modified water content (Concrete-H2O)

The total analytical uncertainty for the fluence in the Beaver Valley Unit 2 nozzle shell, nozzle shell longitudinal welds, and the nozzle shell to intermediate circumferential weld of the reactor pressure vessel, which are located at 211.60 centimeters (cm) above the core midplane or 28.8 cm above the top of the active core, through the above-mentioned process is estimated to be 25 percent, but is conservatively assumed to be 30 percent.

Response 5(b):

As discussed in the response to Item (a) above, most of the listed items in Item (b) have been explicitly addressed in the uncertainty analysis, such as the biological shield composition, homogenized material mixture above and below the active core, orders of angular quadrature sets (SN) and anisotropic scattering treatment (PN).

Of the four potential contributors to the uncertainty identified in Item (b), the only parameter that was not explicitly addressed in the uncertainty analysis is the uncertainties in width of the gap between the RPV and biological shield. Because the plant-specific fast neutron calculations performed in accordance with the WCAP-18124-NP-A methodology use the nominal or the as-built dimensions (when available) as input for the transport model, the exact width of the gap between the RPV and the biological shield is an input to the analysis. Therefore, the uncertainty analysis in the width of gap between the RPV and biological shield is considered unnecessary. In Reference 14, the RPV and biological shield cavity gap width has been increased by 10, 20, and 30 cm to demonstrate the streaming effect due to different RPV and biological shield gap width.

However, this level of uncertainty does not exist in the transport calculations for Beaver Valley Units 1 and 2 that used plant-specific gap width.

References

1. Westinghouse Report WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018, Accession Number ML18204A010.

2. Westinghouse Report WCAP-18102-NP, Revision 2, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, March 2021.

3. Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004, Accession Number ML050120209.

4. Beaver Valley Power Station, Unit No. 1, Pressure and Temperature Limits Report, Revision 10, Accession Number ML19105A881.

5. Beaver Valley Power Station, Unit No. 2, Pressure and Temperature Limits Report, Revision 7, Accession Number ML14133A107.

Attachment L-21-106 Page 7 of 7

6. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001, Accession Number ML010890301.

7. Westinghouse Report WCAP-18559-NP, Revision 1, Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations Based on Capsule Y Data, March 2021.

8. Westinghouse Report WCAP-17896-NP, Revision 0, Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2014, Accession Number ML14288A393.

9. Pressurized Water Reactor Owners Group (PWROG) Report PWROG-15109-NP-A, Revision 0, "PWR Pressure Vessel Nozzle Appendix G Evaluation," January 2020, Accession Number ML20024E573.

10. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S. Nuclear Regulatory Commission, October 2014, Accession Number ML14149A165.

11. NRC Safety Evaluation "Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report PWROG-15109-NP, Revision 0, 'PWR Pressure Vessel Nozzle Appendix G Evaluation' (EPID: L-2018-TOP-0009)," October 31, 2019, Accession Numbers ML19301D063 and ML19301D160.

12. Westinghouse Report WCAP-16528-NP, Revision 1, Beaver Valley Power Station Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, June 2008.

13. Energy Harbor Nuclear Corp. Letter L-20-274, Beaver Valley Power Station, Unit Nos. 1 and 2, BVPS-1 Docket No. 50-334, License No. DPR-66, BVPS-2 Docket No. 50-412, License No. NPF-73, Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits, October 30, 2020, Accession Number ML20304A215.

14. USNRC Public Meeting Presentation, Computation of Neutron Fluence Information Exchange, January 2017, Accession Number ML17038A135 and ML17038A136.

Enclosure A L-21-106 WCAP-18102-NP, Revision 2, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, March 2021 (119 pages follow)

Westinghouse Non-Proprietary Class 3 WCAP-18102-NP March 2021 Revision 2 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-18102-NP Revision 2 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation D. Brett Lynch*

Reactor Vessel/Containment Vessel (RV/CV) Design and Analysis Jianwei Chen*

Radiation Engineering and Analysis March 2021 Reviewers: Benjamin E. Mays* Approved: Lynn A. Patterson*, Manager License Renewal, Radiation RV/CV Design and Analysis Analysis, and Nuclear Operations Eugene T. Hayes* Jesse J. Klingensmith*, Manager Radiation Engineering and Analysis Radiation Engineering and Analysis

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 ii RECORD OF REVISION Rev. Description 0 Original Issue Updated Appendix E by removing reference to TLR-RES/DE/CIB-2013-01 and associated text. Also, Table E-1 was updated to utilize only calculated values of 1

RTNDT in the RTPTS calculations instead of setting calculated values less than 25°F equal to 0°F. Changes are indicated using change bars.

Updated Section 7 and Appendix B to remove the use of TLR-RES/DE/CIB-2013-

01. Also, Appendix B was updated to reference PWROG-15109-NP-A to address 2 the effect of the higher stresses on the inlet and outlet nozzles on the P-T curves as described in Regulatory Issue Summary 2014-11. Changes are indicated using change bars.

WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS LIST OF TABLES ...................................................................................................................................... v LIST OF FIGURES ................................................................................................................................... viii EXECUTIVE

SUMMARY

.......................................................................................................................... ix 1 INTRODUCTION ........................................................................................................................ 1-1 2 CALCULATED NEUTRON FLUENCE ..................................................................................... 2-1

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 2-1 2.3 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-3 3 FRACTURE TOUGHNESS PROPERTIES ................................................................................. 3-1 4 SURVEILLANCE DATA ............................................................................................................. 4-1 5 CHEMISTRY FACTORS ............................................................................................................. 5-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ................ 6-1 6.1 OVERALL APPROACH ................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ........................................... 6-5 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 7-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES....................... 8-1 9 REFERENCES ............................................................................................................................. 9-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt) ............................................... A-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES ....................................... B-1 APPENDIX C OTHER REACTOR COOLANT PRESSURE BOUNDARY FERRITIC COMPONENTS ....................................................................................................... C-1 APPENDIX D BEAVER VALLEY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION ......................................................................................................... D-1 APPENDIX E PRESSURIZED THERMAL SHOCK EVALUATION ........................................... E-1 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iv APPENDIX F VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ................................................ F-1 APPENDIX G SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE ................................ G-1 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 v LIST OF TABLES Table 2-1 Pressure Vessel Material Weld Axial Locations............................................................... 2-5 Table 2-2 Reactor Core Thermal Power Level for Beaver Valley Unit 1 ........................................ 2-5 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface................................................................................................ 2-6 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface................................................................................................ 2-7 Table 2-4 Calculated Iron Displacements per Atom at the Pressure Vessel Clad/Base Metal Interface

......................................................................................................................................... 2-8 Table 2-5 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Plates ............... 2-9 Table 2-6 Calculated Iron Displacements per Atom at the Pressure Vessel Plates ........................ 2-10 Table 2-7 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds ............................................................................... 2-11 Table 2-8 Calculated Iron Displacements per Atom at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds............................................................................................. 2-12 Table 2-9 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Longitudinal Welds

....................................................................................................................................... 2-13 Table 2-10 Calculated Iron Displacements per Atom at the Pressure Vessel Longitudinal Welds.. 2-14 Table 2-11 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of the Surveillance Capsules

....................................................................................................................................... 2-15 Table 2-12 Summary of Calculated Surveillance Capsule Lead Factors ......................................... 2-16 Table 2-13 Calculational Uncertainties ............................................................................................ 2-17 Table 3-1 Summary of Beaver Valley Unit 1 Reactor Vessel Base Metal Material Initial RTNDT Determination Methodologies ......................................................................................... 3-2 Table 3-2 Summary of the Best-Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Beaver Valley Unit 1 Reactor Vessel Materials ............................................................... 3-3 Table 3-3 Summary of Beaver Valley Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values ............................................................................................. 3-4 Table 4-1 Beaver Valley Unit 1 Surveillance Capsule Data............................................................. 4-2 Table 4-2 St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data for Weld Heat # 90136

......................................................................................................................................... 4-3 Table 5-1 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Lower Shell Plate B6903-1 Using Surveillance Capsule Data..................................................................................... 5-2 Table 5-2 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 305424 Using Surveillance Capsule Data ............................................................................................... 5-2 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 vi Table 5-3 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 90136 Using Surveillance Capsule Data ............................................................................................... 5-3 Table 5-4 Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors .................. 5-4 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY ...................................... 7-3 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 1/4T Location .............................................. 7-4 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 3/4T Location .............................................. 7-6 Table 7-4 Summary of the Limiting ART Values Used in the Generation of the Beaver Valley Unit 1 Heatup and Cooldown Curves at 50 EFPY...................................................................... 7-8 Table 8-1 Beaver Valley Unit 1 50 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Notch, and w/o Margins for Instrumentation Errors) .................................................................................................... 8-5 Table 8-2 Beaver Valley Unit 1 50 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Notch, and w/o Margins for Instrumentation Errors) ......................................................................... 8-7 Table A-1 KIt Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements and w/o Margins for Instrument Errors) .................................................. A-2 Table A-2 KIt Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors) .................................................. A-3 Table B-1 ART Calculations for the Beaver Valley Unit 1 Reactor Vessel Nozzle Materials at 50 EFPY............................................................................................................................... B-3 Table B-2 Summary of the Limiting ART Values for the Beaver Valley Unit 1 Inlet and Outlet Nozzle Materials ......................................................................................................................... B-4 Table D-1 Mean Chemical Composition and Operating Temperature for St. Lucie Unit 1 and Millstone Unit 2 .............................................................................................................. D-4 Table D-2 Operating Temperature Adjustments for the St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data .............................................................................................. D-5 Table D-3 Calculation of Weld Heat # 90136 Interim Chemistry Factor for the Credibility Evaluation Using St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data ....................... D-5 Table D-4 Best-Fit Evaluation for Surveillance Weld Metal Heat # 90136 Using St. Lucie Unit 1 and Millstone Unit 2 Data ..................................................................................................... D-6 Table D-5 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit 1 Surveillance Capsule Data ........................................................................ D-7 Table D-6 Beaver Valley Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line .......... D-8 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 vii Table E-1 RTPTS Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY

........................................................................................................................................ E-3 Table F-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors .................................. F-10 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles ...................................... F-11 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Cj Factors at the Surveillance Capsule Center, Core Midplane Elevation .................................................................... F-16 Table F-4a Measured Sensor Activities and Reaction Rates of Surveillance Capsule V ................ F-18 Table F-4b Measured Sensor Activities and Reaction Rates of Surveillance Capsule U ................ F-19 Table F-4c Measured Sensor Activities and Reaction Rates of Surveillance Capsule W ............... F-20 Table F-4d Measured Sensor Activities and Reaction Rates of Surveillance Capsule Y ................ F-21 Table F-4e Measured Sensor Activities and Reaction Rates of Surveillance Capsule X ................ F-22 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center ......................................................................................... F-23 Table F-6 Comparison of Calculated and Best-Estimate Exposure Rates at the Surveillance Capsule Center ............................................................................................................................ F-26 Table F-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions ....................................................................................... F-27 Table F-8 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... F-27 Table G-1 Surveillance Capsule Withdrawal Schedule ................................................................... G-1 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 viii LIST OF FIGURES Figure 2-1 Beaver Valley Unit 1 r, Reactor Geometry at the Core Midplane; Octant with No Surveillance Capsules .................................................................................................... 2-18 Figure 2-2 Beaver Valley Unit 1 r, Reactor Geometry at the Core Midplane; Octant with Surveillance Capsules .................................................................................................... 2-19 Figure 2-3 Beaver Valley Unit 1 r,z Reactor Geometry................................................................... 2-20 Figure 8-1 Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 50 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc) ...................................................................................................... 8-3 Figure 8-2 Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, and -100°F/hr) Applicable for 50 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc) .......................................................................... 8-4 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 ix EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure-temperature (P-T) limit curves for normal operation of the Beaver Valley Unit 1 reactor vessel. The heatup and cooldown P-T limit curves were generated using the limiting Adjusted Reference Temperature (ART) values for Beaver Valley Unit 1 increased by a small margin for conservatism. The limiting ART values were those of Lower Shell Plate B6903-1 (Position 1.1) at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations. The P-T limit curves were generated using the KIc methodology detailed in the 1998 Edition through the 2000 Addenda of the ASME (American Society of Mechanical Engineers) Code,Section XI, Appendix G. The P-T limit curve generation methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4.

The P-T limit curves were generated for 50 effective full-power years (EFPY) using heatup rates of 60 and 100°F/hr, and cooldown rates of 0, -20, -40, -60, and -100°F/hr. The curves were developed with the flange requirements and without margins for instrumentation errors. They can be found in Figures 8-1 and 8-2.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 50 EFPY.

Appendix B contains consideration of the reactor vessel inlet and outlet nozzles. As discussed in Appendix B, the reactor vessel inlet and outlet nozzles for Beaver Valley Unit 1 are not limiting for P-T limits, and thus the P-T limit curves generated based on the limiting cylindrical beltline material (Lower Shell Plate B6903-1) at 50 EFPY are applicable to Beaver Valley Unit 1.

Appendix C contains discussion of the other ferritic Reactor Coolant Pressure Boundary (RCPB) components relative to P-T limits. As discussed in Appendix C, all of the other ferritic RCPB components meet the applicable requirements of Section III of the ASME Code.

Appendix D contains an updated credibility evaluation for Beaver Valley Unit 1 considering all applicable sister plant surveillance program data and updated Beaver Valley Unit 1 surveillance capsule fluence values.

Appendix E contains a Pressurized Thermal Shock (PTS) evaluation for the reactor vessel materials at 50 EFPY for Beaver Valley Unit 1. In the previous Beaver Valley Unit 1 analysis of record, the limiting reactor vessel plate material, Lower Shell Plate B6903-1, was predicted to exceed the RTPTS screening criteria of 270°F for plates at 39.6 EFPY of plant operation. However, as discussed in Appendix E, this material, while still the limiting material, is now predicted to remain under the RTPTS screening limit through 50 EFPY (end of license extension [EOLE]).

Appendix F contains an evaluation of the neutron dosimetry contained in the Beaver Valley Unit 1 surveillance capsules withdrawn to date.

Appendix G contains an updated surveillance capsule withdrawal schedule for Beaver Valley Unit 1.

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Heatup and cooldown P-T limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced RTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, RTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (RTNDT(U)). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The U.S. Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. 1]. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values (RTNDT(U) + RTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (increased by a small margin for conservatism) and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [Ref. 2]. Specifically, the KIc methodology of the 1998 Edition through the 2000 Addenda of ASME Code,Section XI, Appendix G [Ref. 3] was used. The KIc curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the KIc curve so that allowable stress intensity factors can be obtained for the material as a function of temperature. Allowable operating limits are then determined using the allowable stress intensity factors.

The following statement excludes the fluence method. For the purpose of this plant-specific evaluation, the P-T limit curve generation method of WCAP-14040-A, Revision 4 is identical to the P-T limit curve generation method of WCAP-14040-NP-A, Revision 2 [Ref. 21] with the addition of the allowance for Beaver Valley Unit 1 to also utilize ASME Code Case N-640 and ASME Code Section XI, Appendix G (1995 Edition through 1996 Addenda). The fluence method utilized is detailed in Section 2.

The purpose of this report is to present the calculations and the development of the Beaver Valley Unit 1 heatup and cooldown P-T limit curves for 50 EFPY. This report documents the calculated ART values and the development of the P-T limit curves for normal operation. The calculated ART values for 50 EFPY are documented in Section 7 of this report. The fluence projections used in calculation of the ART values are provided in Section 2 of this report.

The P-T limit curves herein were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [Ref. 4] have been incorporated in the P-T limit curves. As discussed in Appendix B, the inlet and outlet nozzles are not limiting for P-T limits, and thus the P-T limit curves generated in Section 8 are applicable for Beaver Valley Unit 1 at 50 EFPY. Discussion of the other ferritic RCPB components relative to P-T limits is contained in Appendix C.

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Westinghouse Non-Proprietary Class 3 1-2 Appendix D contains an updated credibility evaluation for surveillance data applicable to Beaver Valley Unit 1, and Appendix E contains a Pressurized Thermal Shock evaluation for the Beaver Valley Unit 1 reactor vessel materials at 50 EFPY.

Appendix F contains an evaluation of the neutron dosimetry contained in the Beaver Valley Unit 1 surveillance capsules withdrawn to date.

Appendix G contains an updated surveillance capsule withdrawal schedule for Beaver Valley Unit 1.

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Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinates (SN) transport analysis was performed for the Beaver Valley Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. An evaluation of the dosimetry sensor sets from the first five surveillance capsules is provided in Appendix F. The dosimetry analysis shows that the +/-20% (1) acceptance criteria specified in Regulatory Guide 1.190 [Ref. 6] is met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 EFPY.

All of the calculations described in this section and in Appendix F were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (Specifically, ENDF/B-VI). Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 6].

Additionally, the fluence calculations herein were performed in accordance with WCAP-14040-A, Revision 4 [Ref. 2], which is a topical report having an NRC approved method that complies with NRC Regulatory Guide 1.190. The method used for the fluence calculations is the same as that employed during the analysis of Beaver Valley Unit 1 Capsule Y documented in WCAP-15571, Revision 0 [Ref. 22].

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Beaver Valley Unit 1 reactor vessel, a series of fuel-cycle-specific forward transport calculations were completed using the following three-dimensional fluence rate synthesis technique:

(r, z)

(r, , z) = (r, ) x (r) where (r, , z) is the synthesized three-dimensional neutron fluence rate distribution, (r, ) is the transport solution in r, geometry, (r, z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and (r ) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r, two-dimensional calculation. This synthesis procedure was completed for each operating cycle at Beaver Valley Unit 1.

Plan views of the r, geometry of the Beaver Valley Unit 1 reactor at the core midplane are shown in Figures 2-1 and 2-2. In each of these figures, a single octant is depicted showing the arrangement of surveillance capsules, where Figure 2-1 shows an octant with no surveillance capsules, and Figure 2-2 shows an octant with surveillance capsules. The maximum exposure of the pressure vessel occurs in octants with no surveillance capsules. In developing these analytical models, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids and guide tubes. The geometric mesh description of the r, reactor model consisted of 189 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 2-2 radial by 75 azimuthal intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r, calculations was set at a value of 0.001.

The r,z model used for the Beaver Valley Unit 1 calculations is shown in Figure 2-3. The model extends radially from the centerline of the reactor core out to a location interior to the neutron shield tank and over an axial span from an elevation approximately five feet below to five feet above the active fuel. As in the case of the r, models, nominal design dimensions and full-power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of this reactor model consisted of 189 radial by 223 axial intervals. As in the case of the r, calculations, mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 189 radial mesh intervals included in the r,z model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The data utilized for the core power distributions in plant-specific transport analyses 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 source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code [Ref. 7] and the BUGLE-96 cross-section library [Ref. 8]. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion and angular discretization was modeled with an S16 order of angular quadrature. Energy- and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

In Table 2-1, axial locations of the Beaver Valley Unit 1 pressure vessel material welds in terms of the transport models are provided. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the midplane of the active fuel stack.

Cycle-specific calculations were performed for Cycles 1 through 24, with core thermal powers given in Table 2-2.

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Westinghouse Non-Proprietary Class 3 2-3 Neutron exposure data pertinent to the pressure vessel clad/base metal interface are given in Tables 2-3 and 2-4 for fast neutron fluence rate and fluence (E > 1.0 MeV) and iron displacements per atom (dpa),

respectively. In each case the data are provided for each operating cycle of the Beaver Valley Unit 1 reactor.

The vessel exposure data are presented in terms of the maximum exposure experienced by the pressure vessel at azimuthal angles of 0°, 15°, 30°, and 45° relative to the core cardinal axes as well as the maximum exposure anywhere on the reactor pressure vessel.

Calculated fast neutron fluence (E > 1.0 MeV) and dpa for the pressure vessel plates are provided in Tables 2-5 and 2-6, respectively. Calculated fast neutron fluence (E > 1.0 MeV) and dpa for the pressure vessel circumferential welds are provided in Tables 2-7 and 2-8, while the equivalent data for the longitudinal welds are provided in Tables 2-9 and 2-10, respectively.

In Tables 2-3 through 2-10, calculated exposure values are projected to 32, 36, 40, 48, 50, and 60 EFPY.

Projections were based on the burnup weighted average of Cycles 22 through 24 power distributions and reactor operating conditions with the a rated core power of 2900 MWt. The projected results will remain valid as long as future plant operation is consistent with these assumptions.

In Table 2-11, calculated fast neutron fluence (E > 1.0 MeV) for the surveillance capsule for the Beaver Valley Unit 1 reactor is provided. In Table 2-12, a summary of the lead factors for each capsule at the time of removal from the reactor (or at end of Cycle 24 if still inserted) is provided.

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Beaver Valley Unit 1 reactor pressure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

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

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

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

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

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

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Westinghouse Non-Proprietary Class 3 2-4 knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Beaver Valley Unit 1 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Beaver Valley Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule or pressure vessel neutron exposures.

Table 2-13 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 2. 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 given in Appendix F support these uncertainty assessments for Beaver Valley Unit 1.

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Westinghouse Non-Proprietary Class 3 2-5 Table 2-1 Pressure Vessel Material Weld Axial Locations Axial Location(a)

Material Inches cm Lower Shell to Lower Closure Head Weld -121.10 -307.59 Lower Shell to Intermediate Shell Weld -20.50 -52.07 Intermediate Shell to Upper Shell Weld 80.20 203.71 Inlet Nozzle to Upper Shell Weld - Lowest Extent 100.46 255.17 Outlet Nozzle to Upper Shell Weld - Lowest Extent 102.71 260.88 Note:

(a) Axial locations are with respect to the core midplane at 0 cm.

Table 2-2 Reactor Core Thermal Power Level for Beaver Valley Unit 1 Core Power Cycle (MWt) 1 2652 2 2652 3 2652 4 2652 5 2652 6 2652 7 2652 8 2652 9 2652 10 2652 11 2652 12 2652 13 2652 14 2660(a) 15 2689 16 2689 17 2689 18 2799(b) 19 2900 20 2900 21 2900 22 2900 23 2900 24 2900 Notes:

(a) There was a mid-cycle uprate during Cycle 14.

(b) There were two mid-cycle uprates during Cycle 18.

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Westinghouse Non-Proprietary Class 3 2-6 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Cumulative Maximum Fast Neutron Fluence Rate (E > 1.0 MeV)

Operating (n/cm2-s) Elevation(a)

Cycle Time (cm)

(EFPY) 0° 15° 30° 45° Maximum 1 1.2 5.51E+10 2.53E+10 1.42E+10 9.20E+09 5.51E+10 -54.00 2 1.9 5.79E+10 2.69E+10 1.54E+10 1.01E+10 5.79E+10 104.00 3 2.7 6.28E+10 2.87E+10 1.58E+10 1.00E+10 6.28E+10 -54.00 4 3.6 4.70E+10 2.21E+10 1.21E+10 7.72E+09 4.70E+10 -54.00 5 4.8 4.56E+10 2.15E+10 1.18E+10 7.79E+09 4.56E+10 -54.00 6 5.9 3.52E+10 1.91E+10 1.19E+10 7.65E+09 3.52E+10 -54.00 7 7.1 4.30E+10 2.14E+10 1.16E+10 7.67E+09 4.30E+10 -60.00 8 8.2 4.23E+10 2.17E+10 1.19E+10 7.52E+09 4.23E+10 -54.00 9 9.6 3.79E+10 1.99E+10 1.19E+10 8.25E+09 3.79E+10 -54.00 10 10.8 2.96E+10 1.56E+10 1.07E+10 7.49E+09 2.96E+10 -60.00 11 11.8 2.97E+10 1.50E+10 1.11E+10 8.10E+09 2.97E+10 -54.00 12 12.9 3.08E+10 1.63E+10 1.16E+10 7.23E+09 3.08E+10 -54.00 13 14.3 3.21E+10 1.65E+10 1.11E+10 7.49E+09 3.21E+10 -58.00 14 15.6 3.25E+10 1.44E+10 8.45E+09 5.86E+09 3.25E+10 -60.00 15 16.9 2.77E+10 1.38E+10 9.34E+09 6.49E+09 2.77E+10 -118.00 16 18.4 3.24E+10 1.63E+10 9.60E+09 6.72E+09 3.24E+10 -54.00 17 19.6 3.18E+10 1.57E+10 8.99E+09 5.99E+09 3.18E+10 -60.00 18 21.0 3.71E+10 1.78E+10 9.87E+09 6.51E+09 3.71E+10 -54.00 19 22.5 3.22E+10 1.68E+10 1.01E+10 7.29E+09 3.22E+10 -54.00 20 23.8 3.78E+10 1.80E+10 1.02E+10 7.19E+09 3.78E+10 50.00 21 25.2 4.02E+10 1.81E+10 9.69E+09 6.61E+09 4.02E+10 50.00 22 26.6 3.74E+10 1.73E+10 1.00E+10 7.20E+09 3.74E+10 50.00 23 28.0 3.93E+10 1.81E+10 1.01E+10 7.05E+09 3.93E+10 -54.00 24 29.3 3.44E+10 1.69E+10 9.71E+09 6.84E+09 3.44E+10 50.00 Note:

(a) The elevation represents the distance between the maximum exposure location and the core midplane at 0 cm.

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Westinghouse Non-Proprietary Class 3 2-7 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Cumulative Maximum Fast Neutron Fluence (E > 1.0 MeV)

Operating (n/cm2) Elevation(a)

Cycle Time (cm)

(EFPY) 0° 15° 30° 45° Maximum 1 1.2 2.02E+18 9.27E+17 5.18E+17 3.37E+17 2.02E+18 -54.00 2 1.9 3.23E+18 1.49E+18 8.39E+17 5.47E+17 3.23E+18 -54.00 3 2.7 4.79E+18 2.21E+18 1.23E+18 7.96E+17 4.79E+18 -54.00 4 3.6 6.16E+18 2.85E+18 1.58E+18 1.02E+18 6.16E+18 -54.00 5 4.8 7.87E+18 3.66E+18 2.03E+18 1.31E+18 7.87E+18 -54.00 6 5.9 9.10E+18 4.33E+18 2.44E+18 1.58E+18 9.10E+18 -54.00 7 7.1 1.08E+19 5.17E+18 2.90E+18 1.88E+18 1.08E+19 -54.00 8 8.2 1.23E+19 5.92E+18 3.31E+18 2.15E+18 1.23E+19 -54.00 9 9.6 1.39E+19 6.79E+18 3.83E+18 2.50E+18 1.39E+19 -54.00 10 10.8 1.50E+19 7.37E+18 4.24E+18 2.79E+18 1.50E+19 -54.00 11 11.8 1.59E+19 7.83E+18 4.58E+18 3.03E+18 1.59E+19 -54.00 12 12.9 1.70E+19 8.41E+18 4.99E+18 3.29E+18 1.70E+19 -54.00 13 14.3 1.84E+19 9.12E+18 5.47E+18 3.62E+18 1.84E+19 -54.00 14 15.6 1.98E+19 9.72E+18 5.82E+18 3.86E+18 1.98E+19 -54.00 15 16.9 2.09E+19 1.03E+19 6.21E+18 4.13E+18 2.09E+19 -54.00 16 18.4 2.24E+19 1.10E+19 6.65E+18 4.43E+18 2.24E+19 -54.00 17 19.6 2.36E+19 1.16E+19 6.99E+18 4.67E+18 2.36E+19 -54.00 18 21.0 2.53E+19 1.24E+19 7.43E+18 4.95E+18 2.53E+19 -54.00 19 22.5 2.68E+19 1.32E+19 7.90E+18 5.29E+18 2.68E+19 -54.00 20 23.8 2.84E+19 1.40E+19 8.33E+18 5.60E+18 2.84E+19 -54.00 21 25.2 3.01E+19 1.48E+19 8.75E+18 5.89E+18 3.01E+19 -54.00 22 26.6 3.17E+19 1.55E+19 9.19E+18 6.20E+18 3.17E+19 -54.00 23 28.0 3.34E+19 1.63E+19 9.62E+18 6.50E+18 3.34E+19 -54.00 24 29.3 3.48E+19 1.70E+19 1.00E+19 6.78E+18 3.48E+19 -54.00 32.0 3.80E+19 1.85E+19 1.09E+19 7.38E+18 3.80E+19 -54.00 36.0 4.27E+19 2.07E+19 1.21E+19 8.27E+18 4.27E+19 -54.00 40.0 4.73E+19 2.29E+19 1.34E+19 9.15E+18 4.73E+19 -54.00 Future 48.0 5.66E+19 2.72E+19 1.59E+19 1.09E+19 5.66E+19 -54.00 50.0 5.89E+19 2.83E+19 1.65E+19 1.14E+19 5.89E+19 -54.00 60.0 7.06E+19 3.38E+19 1.96E+19 1.36E+19 7.06E+19 -54.00 Note:

(a) The elevation represents the distance between the maximum exposure location and the core midplane at 0 cm.

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Westinghouse Non-Proprietary Class 3 2-8 Table 2-4 Calculated Iron Displacements per Atom at the Pressure Vessel Clad/Base Metal Interface Cumulative Maximum Iron Displacements per Atom Operating (dpa) Elevation(a)

Cycle Time (cm)

(EFPY) 0° 15° 30° 45° Maximum 1 1.2 3.23E-03 1.50E-03 8.29E-04 5.40E-04 3.23E-03 -54.00 2 1.9 5.18E-03 2.42E-03 1.34E-03 8.77E-04 5.18E-03 -54.00 3 2.7 7.69E-03 3.58E-03 1.97E-03 1.28E-03 7.69E-03 -54.00 4 3.6 9.88E-03 4.62E-03 2.53E-03 1.64E-03 9.88E-03 -54.00 5 4.8 1.26E-02 5.93E-03 3.24E-03 2.11E-03 1.26E-02 -54.00 6 5.9 1.46E-02 7.01E-03 3.91E-03 2.54E-03 1.46E-02 -54.00 7 7.1 1.73E-02 8.37E-03 4.64E-03 3.02E-03 1.73E-02 -54.00 8 8.2 1.97E-02 9.59E-03 5.31E-03 3.44E-03 1.97E-02 -54.00 9 9.6 2.23E-02 1.10E-02 6.14E-03 4.02E-03 2.23E-02 -54.00 10 10.8 2.41E-02 1.19E-02 6.78E-03 4.47E-03 2.41E-02 -54.00 11 11.8 2.56E-02 1.27E-02 7.32E-03 4.86E-03 2.56E-02 -54.00 12 12.9 2.73E-02 1.36E-02 7.98E-03 5.28E-03 2.73E-02 -54.00 13 14.3 2.95E-02 1.48E-02 8.75E-03 5.80E-03 2.95E-02 -54.00 14 15.6 3.17E-02 1.57E-02 9.31E-03 6.19E-03 3.17E-02 -54.00 15 16.9 3.36E-02 1.67E-02 9.93E-03 6.62E-03 3.36E-02 -54.00 16 18.4 3.59E-02 1.79E-02 1.06E-02 7.11E-03 3.59E-02 -54.00 17 19.6 3.79E-02 1.88E-02 1.12E-02 7.48E-03 3.79E-02 -54.00 18 21.0 4.05E-02 2.01E-02 1.19E-02 7.94E-03 4.05E-02 -54.00 19 22.5 4.29E-02 2.14E-02 1.26E-02 8.48E-03 4.29E-02 -54.00 20 23.8 4.55E-02 2.26E-02 1.33E-02 8.97E-03 4.55E-02 -54.00 21 25.2 4.83E-02 2.39E-02 1.40E-02 9.44E-03 4.83E-02 -54.00 22 26.6 5.09E-02 2.51E-02 1.47E-02 9.94E-03 5.09E-02 -54.00 23 28.0 5.36E-02 2.64E-02 1.54E-02 1.04E-02 5.36E-02 -54.00 24 29.3 5.59E-02 2.75E-02 1.60E-02 1.09E-02 5.59E-02 -54.00 32.0 6.09E-02 2.99E-02 1.74E-02 1.18E-02 6.09E-02 -54.00 36.0 6.84E-02 3.35E-02 1.94E-02 1.33E-02 6.84E-02 -54.00 40.0 7.59E-02 3.70E-02 2.14E-02 1.47E-02 7.59E-02 -54.00 Future 48.0 9.08E-02 4.41E-02 2.54E-02 1.75E-02 9.08E-02 -54.00 50.0 9.45E-02 4.59E-02 2.64E-02 1.82E-02 9.45E-02 -54.00 60.0 1.13E-01 5.47E-02 3.14E-02 2.17E-02 1.13E-01 -54.00 Note:

(a) The elevation represents the distance between the maximum exposure location and the core midplane at 0 cm.

WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-9 Table 2-5 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Plates Maximum Fast Neutron Fluence Cumulative (E > 1.0 MeV)

Operating (n/cm2)

Cycle Time (EFPY) Upper Intermediate Lower Shell Shell Shell 1 1.2 1.79E+17 2.01E+18 2.02E+18 2 1.9 3.87E+17 3.22E+18 3.23E+18 3 2.7 5.48E+17 4.78E+18 4.79E+18 4 3.6 6.83E+17 6.14E+18 6.16E+18 5 4.8 8.81E+17 7.85E+18 7.87E+18 6 5.9 9.89E+17 9.09E+18 9.10E+18 7 7.1 1.16E+18 1.08E+19 1.08E+19 8 8.2 1.28E+18 1.22E+19 1.23E+19 9 9.6 1.44E+18 1.39E+19 1.39E+19 10 10.8 1.54E+18 1.50E+19 1.50E+19 11 11.8 1.63E+18 1.59E+19 1.59E+19 12 12.9 1.73E+18 1.70E+19 1.70E+19 13 14.3 1.86E+18 1.84E+19 1.84E+19 14 15.6 2.06E+18 1.97E+19 1.98E+19 15 16.9 2.24E+18 2.09E+19 2.09E+19 16 18.4 2.45E+18 2.24E+19 2.24E+19 17 19.6 2.64E+18 2.36E+19 2.36E+19 18 21.0 2.84E+18 2.52E+19 2.53E+19 19 22.5 3.04E+18 2.67E+19 2.68E+19 20 23.8 3.24E+18 2.83E+19 2.84E+19 21 25.2 3.47E+18 3.01E+19 3.01E+19 22 26.6 3.67E+18 3.17E+19 3.17E+19 23 28.0 3.88E+18 3.34E+19 3.34E+19 24 29.3 4.07E+18 3.48E+19 3.48E+19 32.0 4.48E+18 3.79E+19 3.80E+19 36.0 5.08E+18 4.26E+19 4.27E+19 40.0 5.68E+18 4.72E+19 4.73E+19 Future 48.0 6.88E+18 5.65E+19 5.66E+19 50.0 7.18E+18 5.88E+19 5.89E+19 60.0 8.68E+18 7.04E+19 7.06E+19 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-10 Table 2-6 Calculated Iron Displacements per Atom at the Pressure Vessel Plates Cumulative Maximum Iron Displacements per Atom Operating (dpa)

Cycle Time Upper Intermediate Lower (EFPY) Shell Shell Shell 1 1.2 2.98E-04 3.23E-03 3.23E-03 2 1.9 6.43E-04 5.17E-03 5.18E-03 3 2.7 9.11E-04 7.68E-03 7.69E-03 4 3.6 1.14E-03 9.86E-03 9.88E-03 5 4.8 1.47E-03 1.26E-02 1.26E-02 6 5.9 1.65E-03 1.46E-02 1.46E-02 7 7.1 1.92E-03 1.73E-02 1.73E-02 8 8.2 2.14E-03 1.96E-02 1.97E-02 9 9.6 2.41E-03 2.23E-02 2.23E-02 10 10.8 2.57E-03 2.41E-02 2.41E-02 11 11.8 2.72E-03 2.55E-02 2.56E-02 12 12.9 2.89E-03 2.73E-02 2.73E-02 13 14.3 3.11E-03 2.95E-02 2.95E-02 14 15.6 3.43E-03 3.16E-02 3.17E-02 15 16.9 3.73E-03 3.35E-02 3.36E-02 16 18.4 4.08E-03 3.58E-02 3.59E-02 17 19.6 4.40E-03 3.78E-02 3.79E-02 18 21.0 4.73E-03 4.04E-02 4.05E-02 19 22.5 5.05E-03 4.28E-02 4.29E-02 20 23.8 5.39E-03 4.54E-02 4.55E-02 21 25.2 5.77E-03 4.82E-02 4.83E-02 22 26.6 6.10E-03 5.08E-02 5.09E-02 23 28.0 6.46E-03 5.35E-02 5.36E-02 24 29.3 6.77E-03 5.58E-02 5.59E-02 32.0 7.45E-03 6.08E-02 6.09E-02 36.0 8.45E-03 6.83E-02 6.84E-02 40.0 9.44E-03 7.57E-02 7.59E-02 Future 48.0 1.14E-02 9.06E-02 9.08E-02 50.0 1.19E-02 9.43E-02 9.45E-02 60.0 1.44E-02 1.13E-01 1.13E-01 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-11 Table 2-7 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds Maximum Fast Neutron Fluence (E > 1.0 MeV)

Cumulative (n/cm2)

Operating Lower Shell Lower Shell Cycle Intermediate Time Outlet Inlet to to Lower (EFPY) Shell to Nozzle(a) Nozzle(a) Intermediate Closure Upper Shell Shell Head 1 1.2 3.29E+15 4.36E+15 1.79E+17 2.01E+18 4.53E+14 2 1.9 6.99E+15 9.29E+15 3.87E+17 3.22E+18 7.35E+14 3 2.7 9.78E+15 1.30E+16 5.48E+17 4.78E+18 1.13E+15 4 3.6 1.24E+16 1.65E+16 6.83E+17 6.14E+18 1.47E+15 5 4.8 1.64E+16 2.17E+16 8.81E+17 7.85E+18 1.86E+15 6 5.9 1.85E+16 2.45E+16 9.89E+17 9.09E+18 2.16E+15 7 7.1 2.18E+16 2.89E+16 1.16E+18 1.08E+19 2.60E+15 8 8.2 2.44E+16 3.23E+16 1.28E+18 1.22E+19 2.95E+15 9 9.6 2.76E+16 3.64E+16 1.44E+18 1.39E+19 3.35E+15 10 10.8 2.98E+16 3.93E+16 1.54E+18 1.50E+19 3.63E+15 11 11.8 3.17E+16 4.18E+16 1.63E+18 1.59E+19 3.85E+15 12 12.9 3.39E+16 4.47E+16 1.73E+18 1.70E+19 4.11E+15 13 14.3 3.68E+16 4.84E+16 1.86E+18 1.84E+19 4.45E+15 14 15.6 4.15E+16 5.46E+16 2.06E+18 1.97E+19 4.84E+15 15 16.9 4.60E+16 6.05E+16 2.24E+18 2.09E+19 5.19E+15 16 18.4 5.12E+16 6.73E+16 2.45E+18 2.24E+19 5.61E+15 17 19.6 5.59E+16 7.33E+16 2.64E+18 2.36E+19 5.98E+15 18 21.0 6.04E+16 7.92E+16 2.84E+18 2.52E+19 6.41E+15 19 22.5 6.51E+16 8.53E+16 3.04E+18 2.67E+19 6.80E+15 20 23.8 6.97E+16 9.14E+16 3.24E+18 2.83E+19 7.21E+15 21 25.2 7.49E+16 9.80E+16 3.47E+18 3.01E+19 7.67E+15 22 26.6 7.96E+16 1.04E+17 3.67E+18 3.17E+19 8.08E+15 23 28.0 8.44E+16 1.10E+17 3.88E+18 3.34E+19 8.52E+15 24 29.3 8.89E+16 1.16E+17 4.07E+18 3.48E+19 8.91E+15 32.0 9.83E+16 1.29E+17 4.48E+18 3.79E+19 9.74E+15 36.0 1.12E+17 1.47E+17 5.08E+18 4.26E+19 1.10E+16 40.0 1.26E+17 1.65E+17 5.68E+18 4.72E+19 1.22E+16 Future 48.0 1.54E+17 2.01E+17 6.88E+18 5.65E+19 1.46E+16 50.0 1.61E+17 2.10E+17 7.18E+18 5.88E+19 1.53E+16 60.0 1.95E+17 2.55E+17 8.68E+18 7.04E+19 1.83E+16 Note:

(a) Exposure for outlet and inlet nozzles is at the lowest extent of the weld.

WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-12 Table 2-8 Calculated Iron Displacements per Atom at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds Maximum Iron Displacements per Atom Cumulative (dpa)

Operating Lower Shell Lower Shell Cycle Intermediate Time Outlet Inlet to to Lower (EFPY) Shell to Nozzle(a) Nozzle(a) Intermediate Closure Upper Shell Shell Head 1 1.2 1.01E-05 1.19E-05 2.98E-04 3.23E-03 3.12E-06 2 1.9 2.03E-05 2.38E-05 6.43E-04 5.17E-03 5.06E-06 3 2.7 2.94E-05 3.43E-05 9.11E-04 7.68E-03 7.76E-06 4 3.6 3.70E-05 4.33E-05 1.14E-03 9.86E-03 1.01E-05 5 4.8 4.76E-05 5.57E-05 1.47E-03 1.26E-02 1.27E-05 6 5.9 5.45E-05 6.36E-05 1.65E-03 1.46E-02 1.48E-05 7 7.1 6.39E-05 7.47E-05 1.92E-03 1.73E-02 1.78E-05 8 8.2 7.18E-05 8.39E-05 2.14E-03 1.96E-02 2.02E-05 9 9.6 8.16E-05 9.53E-05 2.41E-03 2.23E-02 2.30E-05 10 10.8 8.76E-05 1.02E-04 2.57E-03 2.41E-02 2.49E-05 11 11.8 9.28E-05 1.08E-04 2.72E-03 2.55E-02 2.64E-05 12 12.9 9.89E-05 1.15E-04 2.89E-03 2.73E-02 2.82E-05 13 14.3 1.07E-04 1.25E-04 3.11E-03 2.95E-02 3.05E-05 14 15.6 1.16E-04 1.35E-04 3.43E-03 3.16E-02 3.30E-05 15 16.9 1.24E-04 1.45E-04 3.73E-03 3.35E-02 3.53E-05 16 18.4 1.34E-04 1.57E-04 4.08E-03 3.58E-02 3.82E-05 17 19.6 1.43E-04 1.68E-04 4.40E-03 3.78E-02 4.07E-05 18 21.0 1.54E-04 1.80E-04 4.73E-03 4.04E-02 4.35E-05 19 22.5 1.64E-04 1.92E-04 5.05E-03 4.28E-02 4.62E-05 20 23.8 1.75E-04 2.05E-04 5.39E-03 4.54E-02 4.90E-05 21 25.2 1.87E-04 2.19E-04 5.77E-03 4.82E-02 5.20E-05 22 26.6 1.98E-04 2.31E-04 6.10E-03 5.08E-02 5.48E-05 23 28.0 2.09E-04 2.44E-04 6.46E-03 5.35E-02 5.78E-05 24 29.3 2.19E-04 2.56E-04 6.77E-03 5.58E-02 6.04E-05 32.0 2.41E-04 2.81E-04 7.45E-03 6.08E-02 6.60E-05 36.0 2.72E-04 3.19E-04 8.45E-03 6.83E-02 7.43E-05 40.0 3.04E-04 3.56E-04 9.44E-03 7.57E-02 8.25E-05 Future 48.0 3.68E-04 4.30E-04 1.14E-02 9.06E-02 9.90E-05 50.0 3.83E-04 4.48E-04 1.19E-02 9.43E-02 1.03E-04 60.0 4.63E-04 5.41E-04 1.44E-02 1.13E-01 1.24E-04 Note:

(a) Exposure for outlet and inlet nozzles is at the lowest extent of the weld.

WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-13 Table 2-9 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Longitudinal Welds Maximum Fast Neutron Fluence Cumulative (E > 1.0 MeV)

Operating (n/cm2)

Cycle Time (EFPY) 45° Intermediate 45° Lower Shell Shell 1 1.2 3.36E+17 3.37E+17 2 1.9 5.46E+17 5.47E+17 3 2.7 7.95E+17 7.96E+17 4 3.6 1.02E+18 1.02E+18 5 4.8 1.31E+18 1.31E+18 6 5.9 1.58E+18 1.58E+18 7 7.1 1.88E+18 1.88E+18 8 8.2 2.14E+18 2.15E+18 9 9.6 2.50E+18 2.50E+18 10 10.8 2.78E+18 2.79E+18 11 11.8 3.03E+18 3.03E+18 12 12.9 3.29E+18 3.29E+18 13 14.3 3.61E+18 3.62E+18 14 15.6 3.85E+18 3.86E+18 15 16.9 4.12E+18 4.13E+18 16 18.4 4.42E+18 4.43E+18 17 19.6 4.66E+18 4.67E+18 18 21.0 4.94E+18 4.95E+18 19 22.5 5.28E+18 5.29E+18 20 23.8 5.58E+18 5.60E+18 21 25.2 5.87E+18 5.89E+18 22 26.6 6.18E+18 6.20E+18 23 28.0 6.48E+18 6.50E+18 24 29.3 6.77E+18 6.78E+18 32.0 7.37E+18 7.38E+18 36.0 8.25E+18 8.27E+18 40.0 9.13E+18 9.15E+18 Future 48.0 1.09E+19 1.09E+19 50.0 1.13E+19 1.14E+19 60.0 1.35E+19 1.36E+19 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-14 Table 2-10 Calculated Iron Displacements per Atom at the Pressure Vessel Longitudinal Welds Maximum Iron Displacements per Cumulative Atom Operating (dpa)

Cycle Time (EFPY) 45° Intermediate 45° Lower Shell Shell 1 1.2 5.39E-04 5.40E-04 2 1.9 8.76E-04 8.77E-04 3 2.7 1.28E-03 1.28E-03 4 3.6 1.63E-03 1.64E-03 5 4.8 2.10E-03 2.11E-03 6 5.9 2.53E-03 2.54E-03 7 7.1 3.02E-03 3.02E-03 8 8.2 3.44E-03 3.44E-03 9 9.6 4.01E-03 4.02E-03 10 10.8 4.46E-03 4.47E-03 11 11.8 4.85E-03 4.86E-03 12 12.9 5.27E-03 5.28E-03 13 14.3 5.79E-03 5.80E-03 14 15.6 6.18E-03 6.19E-03 15 16.9 6.61E-03 6.62E-03 16 18.4 7.10E-03 7.11E-03 17 19.6 7.47E-03 7.48E-03 18 21.0 7.93E-03 7.94E-03 19 22.5 8.47E-03 8.48E-03 20 23.8 8.95E-03 8.97E-03 21 25.2 9.42E-03 9.44E-03 22 26.6 9.92E-03 9.94E-03 23 28.0 1.04E-02 1.04E-02 24 29.3 1.09E-02 1.09E-02 32.0 1.18E-02 1.18E-02 36.0 1.32E-02 1.33E-02 40.0 1.46E-02 1.47E-02 Future 48.0 1.75E-02 1.75E-02 50.0 1.82E-02 1.82E-02 60.0 2.17E-02 2.17E-02 WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-15 Table 2-11 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of the Surveillance Capsules Cumulative Neutron (E > 1.0 MeV) Fluence (n/cm2)

Operating Time V U W Y X S(a) T(b) Z(c)

Cycle (EFPY) (15°) (25°) (25°) (25°) (15°) (45°/25°/15°) (35°/25°) (35°/15°)

1 1.2 2.97E+18 1.99E+18 1.99E+18 1.99E+18 2.97E+18 1.03E+18 1.32E+18 1.32E+18 2 1.9 -- 3.26E+18 3.26E+18 3.26E+18 4.84E+18 1.69E+18 2.17E+18 2.17E+18 3 2.7 -- 4.81E+18 4.81E+18 4.81E+18 7.17E+18 2.46E+18 3.19E+18 3.19E+18 4 3.6 -- 6.18E+18 6.18E+18 6.18E+18 9.25E+18 3.15E+18 4.08E+18 4.08E+18 5 4.8 -- -- 7.90E+18 7.90E+18 1.19E+19 4.04E+18 5.22E+18 5.22E+18 6 5.9 -- -- 9.52E+18 9.52E+18 1.40E+19 4.86E+18 6.29E+18 6.29E+18 7 7.1 -- -- -- 1.13E+19 1.67E+19 5.77E+18 7.44E+18 7.44E+18 8 8.2 -- -- -- 1.29E+19 1.91E+19 6.56E+18 8.47E+18 8.47E+18 9 9.6 -- -- -- 1.49E+19 2.19E+19 7.68E+18 9.85E+18 9.85E+18 10 10.8 -- -- -- 1.64E+19 2.38E+19 8.52E+18 1.09E+19 1.09E+19 11 11.8 -- -- -- 1.76E+19 2.52E+19 9.28E+18 1.21E+19 1.24E+19 12 12.9 -- -- -- 1.92E+19 2.71E+19 1.01E+19 1.37E+19 1.42E+19 13 14.3 -- -- -- 2.10E+19 2.93E+19 1.10E+19 1.55E+19 1.65E+19 14 15.6 -- -- -- -- 3.12E+19 1.18E+19 1.68E+19 1.84E+19 15 16.9 -- -- -- -- 3.30E+19 1.26E+19 1.83E+19 2.02E+19 16 18.4 -- -- -- -- 3.54E+19 1.35E+19 2.00E+19 2.26E+19 17 19.6 -- -- -- -- 3.74E+19 1.42E+19 2.13E+19 2.45E+19 18 21.0 -- -- -- -- 3.99E+19 1.51E+19 2.29E+19 2.70E+19 19 22.5 -- -- -- -- 4.24E+19 1.61E+19 2.47E+19 2.95E+19 20 23.8 -- -- -- -- 4.49E+19 1.78E+19 2.63E+19 3.20E+19 21 25.2 -- -- -- -- 4.74E+19 1.94E+19 2.80E+19 3.46E+19 22 26.6 -- -- -- -- 4.99E+19 2.11E+19 2.96E+19 3.70E+19 23 28.0 -- -- -- -- -- 2.36E+19 3.13E+19 3.95E+19 24 29.3 -- -- -- -- -- 2.58E+19 3.28E+19 4.18E+19 32.0 -- -- -- -- -- 3.06E+19 3.60E+19 4.65E+19 36.0 -- -- -- -- -- 3.77E+19 4.07E+19 5.36E+19 40.0 -- -- -- -- -- 4.48E+19 4.54E+19 6.07E+19 Future 48.0 -- -- -- -- -- 5.90E+19 5.49E+19 7.49E+19 50.0 -- -- -- -- -- 6.25E+19 5.72E+19 7.84E+19 60.0 -- -- -- -- -- 8.02E+19 6.91E+19 9.61E+19 Notes:

(a) Capsule S was moved to a 25° location after Cycle 19 and to a 15° location after Cycle 22.

(b) Capsule T was moved to a 25° location after Cycle 10.

(c) Capsule Z was moved to a 15° location after Cycle 10.

WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-16 Table 2-12 Summary of Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor V (165°) Withdrawn EOC 1 1.47 U (65°) Withdrawn EOC 4 1.00 W (245°) Withdrawn EOC 6 1.05 Y (295°) Withdrawn EOC 13 1.14 X (285°) Withdrawn EOC 22 1.57 S (45°/295°/285°) (a)

In Reactor 0.74(d)

T (55°/65°)(b) In Reactor 0.94(d)

Z (305°/165°)(c) In Reactor 1.20(d)

Notes:

(a) Capsule S was moved to the 295° location after Cycle 19 and to the 285° location after Cycle 22.

(b) Capsule T was moved to the 65° location after Cycle 10.

(c) Capsule Z was moved to the 165° location after Cycle 10.

(d) The lead factors for the capsules remaining in the reactor are calculated based on End of Cycle (EOC) 24, the last completed operating cycle.

WCAP-18102-NP March 2021 Revision 2

- - This record was final approved on 3/29/2021 6:47:13 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-17 Table 2-13 Calculational Uncertainties Uncertainty Description Vessel Inner Capsule Radius PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

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Westinghouse Non-Proprietary Class 3 2-18

- C<<*

- C.ba,S1111 Downco-illglm SllleldTnWalr lmatlan - Stol,..115"11

".So 52 oq 33.8 67.5 101.2 135.0 168.B 202.5 236.2 270.0

[cm]

Figure 2-1 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Midplane; Octant with No Surveillance Capsules WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 2-19

- C<<*

- C.bonSIIII Ar Downco-Reglm ShltldTorltWalr lmaffan - Stol,...,stell ca 33,8 67.5 101.2 135.0 168.8 202.5 236.2 270.0

[cm]

Figure 2-2 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Midplane; Octant with Surveillance Capsules WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 2-20 J

II)

N tt)

~ 1 = = = = = = = ===1-

~ - - - - - - - - ----

110 c-J U>

,......,Ill Etti

~I O'I c,;

u, I

co N

tt) u, c-J 0

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m---------

O'I

"'I tt) an 7(),0 59.6 119.2 178.8 238.4 298.o R

[cm]

Figure 2-3 Beaver Valley Unit 1 r,z Reactor Geometry WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requirements for P-T limit curve development are specified in 10 CFR 50, Appendix G [Ref. 4]. The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:

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

The Beaver Valley Unit 1 beltline materials traditionally included the intermediate and lower shell plate and weld materials; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. 9], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the development of P-T limit curves. The materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met. As seen from Tables 2-5 and 2-7 of this report, the extended beltline materials include the upper shell forging, upper to intermediate shell girth weld, the nozzle to upper shell welds, and the nozzle forging materials. The inlet and outlet nozzles are considered a part of the extended beltline, as the exposure at the nozzle welds is conservatively used to represent the exposure at the nozzles. Per NRC RIS 2014-11, the nozzle materials must be evaluated for their potential effect on P-T limit curves regardless of exposure - See Appendix B for more details.

As part of this P-T limit curve development effort, the methodology and evaluations used to determine the initial RTNDT values for the Beaver Valley Unit 1 reactor vessel beltline and extended beltline base metal materials were reviewed and updated, as appropriate. Table 3-1 contains a summary of these methodologies. Summary of the best-estimate copper (Cu) and nickel (Ni) contents, in units of weight percent (wt. %), as well as initial RTNDT values for the reactor vessel beltline and extended beltline materials are provided in Table 3-2 for Beaver Valley Unit 1. Table 3-3 contains a summary of the initial RTNDT values of the reactor vessel flange and replacement reactor vessel closure head. These values serve as input to the P-T limit curves flange-notch per Appendix G of 10 CFR 50 - See Section 6.3 for details.

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Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of Beaver Valley Unit 1 Reactor Vessel Base Metal Material Initial RTNDT Determination Methodologies Reactor Vessel Material Methodology Upper Shell Forging BTP 5-3, Paragraph B1.1 (3)(a) [Ref. 10]

Intermediate and Lower Shell Plates ASME Code,Section III, Subsection NB-2300(b) [Ref. 11]

Inlet and Outlet Nozzle Forgings BWRVIP-173-A, Alternate Approach 2(c) [Ref. 12]

Notes:

(a) The Beaver Valley Unit 1 Certified Material Test Report (CMTR) does not list the orientation of the Charpy V-Notch tests results for the upper shell forging material. However, even though BTP 5-3, Paragraph B1.1(3) methodology for SA-508, Class 2 material must be used due to this lack of Charpy V-Notch orientation, the initial RTNDT for this material remains drop-weight limited and is confirmed (See Table 3-2) due to excellent Charpy V-notch test results (in the assumed strong-orientation).

(b) The reactor vessel beltline plate material initial RTNDT values were determined in accordance with the methodology of ASME Code,Section III, Subsection NB-2300 [Ref. 11] utilizing CVGraph, Version 6.02 as documented in Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1 [Ref. 12].

(c) The initial RTNDT values of the Beaver Valley Unit 1 inlet and outlet nozzles were determined in accordance with BWRVIP-173-A [Ref. 13] - See Appendix B for more details.

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Westinghouse Non-Proprietary Class 3 3-3 Table 3-2 Summary of the Best-Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Beaver Valley Unit 1 Reactor Vessel Materials(a)

Fracture Chemical Toughness Reactor Vessel Material Composition Heat Number Property and Identification Number Wt. % Wt. % Initial RTNDT(c)

Cu Ni (°F)

Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-1(b) 0.14 0.62 26.8 Intermediate Shell Plate B6607-2 C4381-2 (b) 0.14 0.62 53.6 Lower Shell Plate B6903-1 C6317-1 (b) 0.21 0.54 13.1 Lower Shell Plate B7203-2 C6293-2 (b) 0.14 0.57 0.4 Intermediate to Lower Shell Girth Weld 11-714 90136 0.27 0.07 -56 Intermediate Shell Longitudinal Welds 305424 0.28 0.63 -56 19-714 A&B Lower Shell Longitudinal Welds 305414 0.34 0.61 -56 20-714 A&B Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 123V339VA1 0.12 0.68 40 305414 (d) 0.34 0.61 -56 (3951 & 3958)

AOFJ 0.03 0.93 10 Upper Shell to Intermediate Shell Girth Weld 10-714 FOIJ 0.03 0.94 10 EODJ 0.02 1.04 10 HOCJ 0.02 0.93 10 Inlet Nozzle B6608-1 95443-1 0.10 0.82 48.5 Inlet Nozzle B6608-2 95460-1 0.10 0.82 -15.2 Inlet Nozzle B6608-3 95712-1 0.08 0.79 11.4 EODJ 0.02 1.04 10 FOIJ 0.03 0.94 10 HOCJ 0.02 0.93 10 Inlet Nozzle Welds DBIJ 0.02 0.97 10 1-717B, 1-717D, 1-717F EOEJ 0.01 1.03 10 ICJJ 0.03 0.99 10 JACJ 0.04 0.97 10 Outlet Nozzle B6605-1 95415-1 0.13 0.77 -26.2 Outlet Nozzle B6605-2 95415-2 0.13 0.77 3.3 Outlet Nozzle B6605-3 95444-1 0.09 0.79 10.1 ICJJ 0.03 0.99 10 Outlet Nozzle Welds IOBJ 0.02 0.97 10 1-717A, 1-717C, 1-717E JACJ 0.04 0.97 10 HOCJ 0.02 0.93 10 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 3-4 Table 3-2 Summary of the Best-Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Beaver Valley Unit 1 Reactor Vessel Materials(a)

Fracture Chemical Toughness Reactor Vessel Material Composition Heat Number Property and Identification Number Wt. % Wt. % Initial RTNDT(c)

Cu Ni (°F)

Outlet Nozzle Welds EODJ 0.02 1.04 10 1-717A, 1-717C, 1-717E (continued) FOIJ 0.03 0.94 10 Surveillance Weld Data (e)

Beaver Valley Unit 1 305424 0.26 0.61 ---

St. Lucie Unit 1 0.23 0.07 ---

90136 Millstone Unit 2 0.30 0.06 ---

Fort Calhoun 305414 0.35 0.60 ---

Notes:

(a) All values originally documented in WCAP-15571, Supplement 1, Revision 2 [Ref. 14], unless otherwise noted.

(b) The reactor vessel beltline plate material heat numbers were taken from the Beaver Valley Unit 1 CMTRs.

(c) The initial RTNDT values for all the reactor vessel welds are generic. The initial RTNDT values for the base metal materials were updated or confirmed as discussed in Table 3-1.

(d) The chemistry values for weld Heat # 305414, as reported in WCAP-15571, Supplement 1, Revision 2 [Ref. 14], were rounded up for consistency with the reactor vessel beltline weld material that shares this same heat number.

(e) Surveillance data exists for weld Heat # 90136, # 305424, and # 305414 from multiple sources; see Section 4 for more details. The data for Beaver Valley Unit 1 weld metal Heat # 90136 was taken from WCAP-17896-NP [Ref. 5]. The data for St. Lucie Unit 1 weld metal Heat # 90136 was taken from the St. Lucie Unit 1 License Amendment Request for Extended Power Uprates, Attachment 5, Table 2.1.2-4 [Ref. 15]. The data for Millstone Unit 2 weld metal Heat # 90136 was taken from Table 4-1 of WCAP-16012 [Ref. 16]. The data for Fort Calhoun weld metal Heat # 305414 was taken from Table 5.2-4b of the 2012 Beaver Valley Unit 1 P-T Limits revision report [Ref. 17].

Table 3-3 Summary of Beaver Valley Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values Initial RTNDT Reactor Vessel Material Methodology

(°F)

ASME Code,Section III, Subsection NB-2300 Replacement Closure Head -4(a)

[Ref. 11]

BWRVIP-173-A, Alternate Approach 2 Vessel Flange 10(b)

[Ref. 12]

Notes:

(a) The initial RTNDT value of the replacement reactor vessel (RV) closure head was taken from WCAP-16799-NP, Revision 1 [Ref. 18] and was determined in accordance with the methodology of ASME Code,Section III, Subsection NB-2300 [Ref. 11].

(b) The initial RTNDT value of the vessel flange was updated, utilizing the methodology of BWRVIP-173-A, Alternate Approach 2 [Ref. 13], from the value documented in WCAP-16799-NP, Revision 1 [Ref. 18].

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Westinghouse Non-Proprietary Class 3 4-1 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2, calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from surveillance programs at other plants which include a reactor vessel beltline or extended beltline material should also be considered when calculating Position 2.1 chemistry factors. Data from a surveillance program at another plant is often called sister plant data.

The surveillance capsule plate material for Beaver Valley Unit 1 is from Lower Shell Plate B6903-1. The surveillance capsule weld material for Beaver Valley Unit 1 is Heat # 305424, which is applicable to the intermediate shell longitudinal welds. Table 4-1 summarizes the Beaver Valley Unit 1 surveillance data for the plate material and weld material (Heat # 305424) that will be used in the calculation of the Position 2.1 chemistry factor values for these materials. The results of the last withdrawn and tested surveillance capsule, Capsule X, were documented in WCAP-17896-NP [Ref. 5]. Appendix D concludes that the Beaver Valley Unit 1 surveillance plate and weld (Heat # 305424) material are non-credible; therefore, a full margin term will be utilized in the ART calculations contained in Section 7.

The Beaver Valley Unit 1 reactor vessel intermediate to lower shell girth weld seam was fabricated using weld Heat # 90136. Weld Heat # 90136 is contained in the St. Lucie Unit 1 and Millstone Unit 2 surveillance programs. Thus, the St. Lucie Unit 1 and Millstone Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor value for Beaver Valley Unit 1 weld Heat # 90136. Table 4-2 summarizes the applicable surveillance capsule data pertaining to weld Heat # 90136. The combined surveillance data is deemed credible per Appendix D; however, as a result of the Millstone Unit 2 surveillance data including both weld Heat # 90136 and 10137, the Position 2.1 chemistry factor calculations for weld Heat # 90136 will utilize a full margin term for conservatism. See Appendix D for details.

The Beaver Valley Unit 1 reactor vessel upper shell to intermediate shell girth weld seam and lower shell longitudinal weld seams were fabricated using weld Heat # 305414. Weld Heat # 305414 is contained in the Fort Calhoun surveillance program. Thus, in WCAP-15571, Supplement 1, Revision 2, the Fort Calhoun data was used to calculate the Position 2.1 chemistry factor value for Beaver Valley Unit 1 weld Heat #

305414, which is used herein. Furthermore, Appendix D of WCAP-17896-NP [Ref. 5] concluded that the weld Heat # 305414 data is non-credible; therefore, a full margin term will be utilized in the ART calculations contained in Section 7.

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Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Beaver Valley Unit 1 Surveillance Capsule Data Capsule Fluence(a) Measured 30 ft-lb Transition Material Capsule (x 1019 n/cm2, E > 1.0 MeV) Temperature Shift(b) (°F)

V 0.297 127.9 U 0.618 118.3 Lower Shell Plate B6903-1 (Longitudinal) W 0.952 147.7 Y 2.10 141.7 X 4.99 175.8 V 0.297 138.0 U 0.618 132.1 Lower Shell Plate B6903-1 (Transverse) W 0.952 180.2 Y 2.10 166.9 X 4.99 179.0 V 0.297 159.8 U 0.618 164.9 Surveillance Weld Material (Heat # 305424) W 0.952 186.3 Y 2.10 178.5 X 4.99 237.8 Notes:

(a) Data was taken from Appendix F, Section F.1.1.

(b) Data was taken from Table 5-10 of WCAP-17896-NP [Ref. 5].

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Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data for Weld Heat # 90136 Capsule Fluence(a) Inlet Temperature (x 1019 n/cm2, E > 1.0 Measured 30 ft-lb Transition Material Capsule(a) Temperature(b) Adjustment(c)

MeV) Temperature Shift(a) (°F)

(°F) (°F) 97° 0.5174 72.34 541 -1.7 St. Lucie Unit 1 104° 0.7885 67.4 544.6 1.9 Data 284° 1.243 68.0 546.3 3.6 97° 0.324 65.93 544.3 1.6 Millstone Unit 2 104° 0.949 52.12 547.6 4.9 Data 83° 1.74 56.09 548.0 5.3 Notes:

(a) For surveillance weld heat # 90136, data pertaining to the St. Lucie Unit 1 were taken from the St. Lucie Unit 1 License Amendment Request for Extended Power Uprates, Attachment 5, Table 2.1.1-3 [Ref. 15]. Data pertaining to Millstone Unit 2 were taken from Table 5-10 of WCAP-16012 [Ref. 16].

(b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal.

(c) Temperature adjustment = 1.0*(Tcapsule - Tplant), where Tplant = 542.7°F for Beaver Valley Unit 1. 542.7°F is the cycle-by-cycle average downcomer temperature for Beaver Valley Unit 1 for Cycle 1 through Cycle 22. The temperature adjustment procedure is applied to the weld RTNDT data for each of the St. Lucie Unit 1 and Millstone Unit 2 capsules in the Position 2.1 chemistry factor calculation - See Section 5 for more details.

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Westinghouse Non-Proprietary Class 3 5-1 5 CHEMISTRY FACTORS The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2, Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

The best-estimate copper and nickel weight percent values for the Beaver Valley Unit 1 reactor vessel materials are provided in Table 3-2 of this report.

The Position 2.1 chemistry factors are calculated for the materials that have available surveillance program results. The calculation is performed using the method described in Regulatory Guide 1.99, Revision 2.

The Beaver Valley Unit 1 surveillance data as well as the applicable sister plant data was summarized in Section 4 of this report, and will be utilized in the Position 2.1 chemistry factor calculations in this Section.

The Position 2.1 chemistry factor calculations are presented in Tables 5-1 through 5-3 for Beaver Valley Unit 1 reactor vessel materials that have associated surveillance data. These values were calculated using the surveillance data summarized in Section 4 of this report. Note that the Position 2.1 chemistry factor for weld Heat # 305414 was previously reported in WCAP-15571, Supplement 1, Revision 2 [Ref. 14] and is therefore not recalculated. All of the surveillance data is adjusted for irradiation temperature and chemical composition differences in accordance with the guidance presented at an industry meeting held by the NRC on February 12 and 13, 1998 [Ref. 19]. Margin will be applied to the ART calculations in Section 7 according to the conclusions of the credibility evaluation for each of the surveillance materials, as documented in Section 4.

The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-4 for Beaver Valley Unit 1.

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Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Lower Shell Plate B6903-1 Using Surveillance Capsule Data Capsule f(a)

LS Plate B6903-1 RTNDT(c) FF*RTNDT Capsule (x 1019 n/cm2, E > 1.0 FF(b) FF2 Data (°F) (°F)

MeV)

V 0.297 0.6677 127.9 85.40 0.446 U 0.618 0.8652 118.3 102.35 0.749 Longitudinal W 0.952 0.9862 147.7 145.66 0.973 Orientation Y 2.10 1.2018 141.7 170.30 1.444 X 4.99 1.4020 175.8 246.46 1.965 V 0.297 0.6677 138.0 92.14 0.446 U 0.618 0.8652 132.1 114.29 0.749 Transverse W 0.952 0.9862 180.2 177.72 0.973 Orientation Y 2.10 1.2018 166.9 200.58 1.444 X 4.99 1.4020 179.0 250.95 1.965 SUM: 1585.86 11.154 CFLS Plate B6903-1 = (FF

RTNDT) ÷ (FF2) = (1585.86) ÷ (11.154) = 142.2°F Notes:

(a) f = fluence.

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

(c) RTNDT values are the measured 30 ft-lb shift values. All values are taken from Table 4-1 of this report.

Table 5-2 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 305424 Using Surveillance Capsule Data Capsule f(a)

Weld Metal RTNDT(c) FF*RTNDT Capsule (x 1019 n/cm2, E > 1.0 FF(b) FF2 Heat # 305424 (°F) (°F)

MeV)

V 0.297 0.6677 169.4 (159.8) 113.10 0.446 U 0.618 0.8652 174.8 (164.9) 151.23 0.749 Beaver Valley W 0.952 0.9862 197.5 (186.3) 194.76 0.973 Unit 1 Data Y 2.10 1.2018 189.2 (178.5) 227.40 1.444 X 4.99 1.4020 252.1 (237.8) 353.39 1.965 SUM: 1039.87 5.577 CFWeld Heat # 305424 = (FF

RTNDT) ÷ (FF ) = (1039.87) ÷ (5.577) = 186.5°F 2

Notes:

(a) f = fluence.

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

(c) RTNDT values are the measured 30 ft-lb shift values. The RTNDT values are adjusted using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-1 of this report). Ratio applied to the Beaver Valley Unit 1 surveillance data = CFVessel Weld / CFSurv. Weld = 191.7°F / 181.6°F = 1.06.

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Westinghouse Non-Proprietary Class 3 5-3 Table 5-3 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 90136 Using Surveillance Capsule Data Weld Metal Capsule f(a) RTNDT(c) FF*RTNDT Capsule FF(b) FF2 Heat # 90136 (x 1019 n/cm2, E > 1.0 MeV) (°F) (°F) 97° 0.5174 0.8160 82.6 (72.34) 67.44 0.666 St. Lucie Unit 1 104° 0.7885 0.9333 81.1 (67.4) 75.68 0.871 Data 284° 1.243 1.0606 83.8 (68.0) 88.85 1.125 97° 0.324 0.6902 67.5 (65.93) 46.61 0.476 Millstone Unit 2 104° 0.949 0.9853 57.0 (52.12) 56.18 0.971 Data(d) 83° 1.74 1.1523 61.4 (56.09) 70.74 1.328 SUM: 405.50 5.437 CFWeld Heat # 90136 = (FF

RTNDT) ÷ (FF ) = (405.50) ÷ (5.437) = 74.6°F 2

Notes:

(a) f = fluence.

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

(c) RTNDT values are the measured 30 ft-lb shift values. The RTNDT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-2 of this report). The temperature adjustments are listed in Table 4-2. Ratio applied to the St. Lucie Unit 1 surveillance data = CFVessel Weld /

CFSurv. Weld = 124.3°F / 106.6°F = 1.17. A ratio of 1.00 was conservatively applied to the Millstone Unit 2 surveillance data, since CFVessel Weld < CFSurv. Weld.

(d) Millstone Unit 2 surveillance data contains specimens from both weld Heat # 90136 and weld Heat # 10137. However, this inclusion of an additional heat is not expected to negatively impact the subsequent reactor vessel integrity calculation results, as additional conservatisms are in place. See Appendix D for more details.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 5-4 Table 5-4 Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (°F)

Heat Number and Identification Number Position 1.1(a) Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-1 100.5 ---

Intermediate Shell Plate B6607-2 C4381-2 100.5 ---

Lower Shell Plate B6903-1 C6317-1 147.2 142.2(b)

Lower Shell Plate B7203-2 C6293-2 98.7 -

Intermediate to Lower Shell Girth Weld 11-714 90136 124.3 74.6(c)

Intermediate Shell Longitudinal Welds 305424 191.7 186.5(d)19-714 A&B Lower Shell Longitudinal Welds20-714 A&B 305414 210.5 216.9(e)

Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 123V339VA1 84.2 ---

305414 210.5 216.9(e)

AOFJ 41.0 ---

Upper Shell to Intermediate Shell Girth FOIJ 41.0 ---

Weld 10-714 EODJ 27.0 ---

HOCJ 27.0 ---

Inlet Nozzle B6608-1 95443-1 67.0 ---

Inlet Nozzle B6608-2 95460-1 67.0 ---

Inlet Nozzle B6608-3 95712-1 51.0 ---

EODJ 27.0 ---

FOIJ 41.0 ---

HOCJ 27.0 ---

Inlet Nozzle Welds 1-717B, 1-717D, 1-717F DBIJ 27.0 ---

EOEJ 20.0 ---

ICJJ 41.0 ---

JACJ 54.0 ---

Outlet Nozzle B6605-1 95415-1 95.3 ---

Outlet Nozzle B6605-2 95415-2 95.3 ---

Outlet Nozzle B6605-3 95444-1 58.0 ---

ICJJ 41.0 ---

IOBJ 27.0 ---

Outlet Nozzle Welds 1-717A, 1-717C, 1-717E JACJ 54.0 ---

HOCJ 27.0 ---

EODJ 27.0 ---

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 5-5 Table 5-4 Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (°F)

Heat Number and Identification Number Position 1.1(a) Position 2.1 Outlet Nozzle Welds 1-717A, 1-717C, 1-717E FOIJ 41.0 ---

(continued)

Surveillance Weld Data Beaver Valley Unit 1 305424 181.6 ---

St. Lucie Unit 1 106.6 ---

90136 Millstone Unit 2 135.5 ---

Fort Calhoun 305414 212.0 ---

Notes:

(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-2 of this report and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

(b) Position 2.1 chemistry factor was taken from Table 5-1 of this report. As discussed in Section 4, the surveillance plate data is not credible.

(c) Position 2.1 chemistry factor was taken from Table 5-3 of this report. As discussed in Section 4, the surveillance weld data for Heat # 90136 is credible; however, no reduction in the margin term will be taken.

(d) Position 2.1 chemistry factor was taken from Table 5-2 of this report. As discussed in Section 4, the surveillance weld data for Heat # 305424 is not credible.

(e) Position 2.1 chemistry factor was taken from WCAP-15571, Supplement 1, Revision 2 [Ref. 14]. As discussed in Section 4, the surveillance weld data for Heat # 305414 is not credible.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 6-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIc, for the metal temperature at that time. KIc is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3]. The KIc curve is given by the following equation:

K Ic =33.2+20.734*e[ 0.02(T RTNDT )] (1)

where, KIc (ksiin.) = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This KIc curve is based on the lower bound of static critical KI values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* KIm + KIt < KIc (2)

where, KIm = stress intensity factor caused by membrane (pressure) stress KIt = stress intensity factor caused by the thermal gradients KIc = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 6-2 For membrane tension, the corresponding KI for the postulated defect is:

K Im = Mm x ( pRi / t ) (3) where, Mm for an inside axial surface flaw is given by:

Mm = 1.85 for t < 2, Mm = 0.926 t for 2 t 3.464 ,

Mm = 3.21 for t > 3.464 and, Mm for an outside axial surface flaw is given by:

Mm = 1.77 for t < 2, Mm = 0.893 t for 2 t 3.464 ,

Mm = 3.09 for t > 3.464 Similarly, Mm for an inside or an outside circumferential surface flaw is given by:

Mm = 0.89 for t < 2, Mm = 0.443 t for 2 t 3.464 ,

Mm = 1.53 for t > 3.464 where:

p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).

For bending stress, the corresponding KI for the postulated axial or circumferential defect is:

KIb = Mb

Maximum Bending Stress, where Mb is two-thirds of Mm (4)

The maximum KI produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:

KIt = 0.953x10-3 x CR x t2.5 (5) where CR is the cooldown rate in °F/hr., or for a postulated axial or circumferential outside surface defect KIt = 0.753x10-3 x HU x t2.5 (6) where HU is the heatup rate in °F/hr.

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Westinghouse Non-Proprietary Class 3 6-3 The through-wall temperature difference associated with the maximum thermal KI can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal KI.

(a) The maximum thermal KI relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

(b) Alternatively, the KI for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness axial or circumferential inside surface defect using the relationship:

KIt = (1.0359 C 0 + 0.6322 C1 + 0.4753C 2 + 0.3855C 3)

a (7) or similarly, KIt during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:

KIt = (1.043C 0 + 0.630C1 + 0.481C 2 + 0.401C 3)

a (8) where the coefficients C0, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

( x ) = C 0 + C1( x / a ) + C 2 ( x / a ) 2 + C 3( x / a ) 3 (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).

Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves [Ref. 2] Section 2.6 (equations 2.6.2-4 and 2.6.3-1). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft2/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-°F.

At any time during the heatup or cooldown transient, KIc is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, paragraph G-2120), the appropriate value for RTNDT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these, the allowable pressures are calculated.

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Westinghouse Non-Proprietary Class 3 6-4 For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the T (temperature) across the vessel wall developed during cooldown results in a higher value of KIc at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIc exceeds KIt, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location, and therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIc for the inside 1/4T flaw during heatup is lower than the KIc for the flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIc values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present.

These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

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Westinghouse Non-Proprietary Class 3 6-5 Following the generation of pressure-temperature curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Ref. 4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is calculated to be 621 psig. The initial RTNDT values of the replacement reactor vessel closure head and vessel flange are documented in Table 3-3. The limiting unirradiated RTNDT of 10°F is associated with the vessel flange of the Beaver Valley Unit 1 vessel, so the minimum allowable temperature of this region is 130°F at pressures greater than 621 psig (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.

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Westinghouse Non-Proprietary Class 3 7-1 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + RTNDT + Margin (10)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [Ref. 11]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.

RTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

RTNDT = CF

f (0.28 - 0.10 log f) (11)

To calculate RTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth:

f(depth x) = fsurface

e (-0.24x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 7.875 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the RTNDT at the specific depth.

The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves [Ref. 2].

Table 7-1 contains the surface fluence values at 50 EFPY, which were used for the development of the P-T limit curves contained in this report. Table 7-1 also contains the 1/4T and 3/4T calculated fluence values and fluence factors, per Regulatory Guide 1.99, Revision 2. The values in this table will be used to calculate the 50 EFPY ART values for the Beaver Valley Unit 1 reactor vessel materials.

Margin is calculated as M = 2 I2 + 2 . The standard deviation for the initial RTNDT margin term (I) is 0°F when the initial RTNDT is a measured value, and 17°F when a generic value is available. The standard deviation for the RTNDT margin term, , is 17°F for plates or forgings when surveillance data is not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data is used.

For welds, is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14°F (half the value) when credible surveillance capsule data is used. The value for need not exceed 0.5 times the mean value of RTNDT.

Contained in Tables 7-2 and 7-3 are the 50 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Beaver Valley Unit 1 heatup and cooldown curves.

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Westinghouse Non-Proprietary Class 3 7-2 The inlet and outlet nozzle forging materials for Beaver Valley Unit 1 have projected fluence values that exceed the 1 x 1017 n/cm2 fluence threshold at 50 EFPY per Table 2-7; therefore, per NRC RIS 2014-11

[Ref. 9], neutron radiation embrittlement must be considered herein for these materials. The nozzle ART calculations conservatively utilize the maximum fluence value for each nozzle material, as documented in Appendix B. Thus, ART calculations for the inlet and outlet nozzle forging materials utilizing the 1/4T and 3/4T fluence values are excluded from Tables 7-2 and 7-3, respectively.

The limiting ART values for Beaver Valley Unit 1 to be used in the generation of the P-T limit curves are based on Lower Shell Plate B6903-1 (Position 1.1). For conservatism, limiting ART values were rounded to the nearest whole number, then increased by 0.5°F. The increased limiting ART values, using the Axial Flaw methodology, for Lower Shell Plate B6903-1 are summarized in Table 7-4.

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Westinghouse Non-Proprietary Class 3 7-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY Surface 1/4T f 3/4T f Fluence, f(a) 1/4T 3/4T Reactor Vessel Region (n/cm2, (n/cm2, (n/cm2, FF FF E > 1.0 MeV) E > 1.0 MeV)

E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell Plates 5.88 x 1019 3.666 x 1019 1.3370 1.425 x 1019 1.0982 Lower Shell Plates 5.89 x 1019 3.672 x 1019 1.3374 1.427 x 1019 1.0987 Intermediate to Lower Shell 5.88 x 1019 3.666 x 1019 1.3370 1.425 x 1019 1.0982 Girth Weld Intermediate Shell 1.13 x 1019 7.04 x 1018 0.9018 2.74 x 1018 0.6469 Longitudinal Welds Lower Shell Longitudinal 1.14 x 1019 7.11 x 1018 0.9042 2.76 x 1018 0.6492 Welds Reactor Vessel Extended Beltline Materials Upper Shell Forging 7.18 x 1018 4.48 x 1018 0.7764 1.74 x 1018 0.5366 Upper to Intermediate Shell 7.18 x 1018 4.48 x 1018 0.7764 1.74 x 1018 0.5366 Girth Weld Inlet Nozzle to Upper Shell 2.10 x 1017(b) 1.31 x 1017 0.1313 5.09 x 1016 0.0679 Weld - Lowest Extent Outlet Nozzle to Upper 1.61 x 1017(b) 1.00 x 1017 0.1099 3.90 x 1016 0.0556 Shell Weld - Lowest Extent Notes:

(a) 50 EFPY fluence values were taken from Tables 2-5, 2-7, and 2-9.

(b) The fluence for the inlet and outlet nozzle to upper shell welds was also used as the fluence for the inlet and outlet nozzle materials. The actual nozzle forging fluence values, at the location of a postulated flaw along the nozzle corner region, are expected to be lower since they are further away from the active core.

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Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 1/4T Location Reactor Vessel Material and ID CF 1/4T Fluence 1/4T RTNDT(U)(a) RTNDT I(a) (c) Margin ART(d)

Heat Number Number (°F) (n/cm2, E > 1.0 MeV) FF (°F) (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-1 100.5 3.666 x 1019 1.3370 26.8 134.4 0 17 34.0 195.2 Intermediate Shell Plate B6607-2 C4381-2 100.5 3.666 x 1019 1.3370 53.6 134.4 0 17 34.0 222.0 Lower Shell Plate B6903-1 C6317-1 147.2 3.672 x 10 19 1.3374 13.1 196.9 0 17 34.0 244.0 Using Beaver Valley Unit 1 C6317-1 142.2 3.672 x 1019 1.3374 13.1 190.2 0 17 34.0 237.3 surveillance data Lower Shell Plate B7203-2 C6293-2 98.7 3.672 x 1019 1.3374 0.4 132.0 0 17 34.0 166.4 Intermediate to Lower Shell Girth 90136 124.3 3.666 x 1019 1.3370 -56 166.2 17 28 65.5 175.7 Weld 11-714 Using St. Lucie Unit 1 and Millstone 90136 74.6 3.666 x 1019 1.3370 -56 99.7 17 28 65.5 109.3 Unit 2 surveillance data Intermediate Shell Longitudinal 305424 191.7 0.704 x 1019 0.9018 -56 172.9 17 28 65.5 182.4 Welds19-714 A&B Using Beaver Valley Unit 1 305424 186.5 0.704 x 1019 0.9018 -56 168.2 17 28 65.5 177.7 surveillance data Lower Shell Longitudinal Welds 305414 210.5 0.711 x 1019 0.9042 -56 190.3 17 28 65.5 199.9 20-714 A&B Using Fort Calhoun surveillance data 305414 216.9 0.711 x 1019 0.9042 -56 196.1 17 28 65.5 205.6 Reactor Vessel Extended Beltline Materials(e)

Upper Shell Forging B6604 123V339VA1 84.2 0.448 x 1019 0.7764 40 65.4 0 17 34.0 139.4 Upper Shell to Intermediate Shell 305414 210.5 0.448 x 1019 0.7764 -56 163.4 17 28 65.5 172.9 Girth Weld 10-714 (3951 & 3958) 305414 Using Fort Calhoun surveillance data 216.9 0.448 x 1019 0.7764 -56 168.4 17 28 65.5 177.9 (3951 & 3958)

AOFJ 41.0 0.448 x 1019 0.7764 10 31.8 17 15.9 46.6 88.4 Upper Shell to Intermediate Shell FOIJ 41.0 0.448 x 1019 0.7764 10 31.8 17 15.9 46.6 88.4 Girth Weld 10-714 (continued) EODJ 27.0 0.448 x 1019 0.7764 10 21.0 17 10.5 39.9 70.9 HOCJ 27.0 0.448 x 1019 0.7764 10 21.0 17 10.5 39.9 70.9 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 7-5 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 1/4T Location Reactor Vessel Material and ID CF 1/4T Fluence 1/4T RTNDT(U)(a) RTNDT I(a) (c) Margin ART(d)

Heat Number Number (°F) (n/cm2, E > 1.0 MeV) FF (°F) (°F) (°F) (°F) (°F) (°F)

EODJ 27.0 0.0131 x 1019 0.1313 10 3.5 17 1.8 34.2 47.7 FOIJ 41.0 0.0131 x 1019 0.1313 10 5.4 17 2.7 34.4 49.8 HOCJ 27.0 0.0131 x 10 19 0.1313 10 3.5 17 1.8 34.2 47.7 Inlet Nozzle Welds 1-717B, 1-717D, DBIJ 27.0 0.0131 x 10 19 0.1313 10 3.5 17 1.8 34.2 47.7 1-717F EOEJ 20.0 0.0131 x 10 19 0.1313 10 2.6 17 1.3 34.1 46.7 ICJJ 41.0 0.0131 x 10 19 0.1313 10 5.4 17 2.7 34.4 49.8 JACJ 54.0 0.0131 x 10 19 0.1313 10 7.1 17 3.5 34.7 51.8 ICJJ 41.0 0.0100 x 1019 0.1099 10 4.5 17 2.3 34.3 48.8 IOBJ 27.0 0.0100 x 1019 0.1099 10 3.0 17 1.5 34.1 47.1 Outlet Nozzle Welds 1-717A, 1-717C, JACJ 54.0 0.0100 x 10 19 0.1099 10 5.9 17 3.0 34.5 50.5 1-717E HOCJ 27.0 0.0100 x 10 19 0.1099 10 3.0 17 1.5 34.1 47.1 EODJ 27.0 0.0100 x 10 19 0.1099 10 3.0 17 1.5 34.1 47.1 FOIJ 41.0 0.0100 x 10 19 0.1099 10 4.5 17 2.3 34.3 48.8 Notes:

(a) The plate and forging material initial RTNDT values are measured values. The initial RTNDT values for all of the reactor vessel welds are generic; hence I = 17°F for all reactor vessel welds.

(b) Not used.

(c) As discussed in Section 4, the surveillance plate and weld Heat # 305414 and # 305424 data were deemed non-credible. The surveillance weld data for Heat # 90136 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. 1], the base metal

= 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal = 28°F for Position 1.1 and 2.1 with non-credible surveillance data.

Since a full margin term will be used for Heat # 90136, = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, need not exceed 0.5*RTNDT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(U) + RTNDT + Margin.

(e) As discussed in Section 7, the inlet and outlet nozzle forging material ART calculations utilizing a 1/4T fluence value are excluded from this table.

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Westinghouse Non-Proprietary Class 3 7-6 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 3/4T Location Reactor Vessel Material and ID CF 3/4T Fluence RTNDT(U)(a) RTNDT I(a) (c) Margin ART(d)

Heat Number 3/4T FF Number (°F) (n/cm2, E > 1.0 MeV) (°F) (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-1 100.5 1.425 x 1019 1.0982 26.8 110.4 0 17 34.0 171.2 Intermediate Shell Plate B6607-2 C4381-2 100.5 1.425 x 10 19 1.0982 53.6 110.4 0 17 34.0 198.0 Lower Shell Plate B6903-1 C6317-1 147.2 1.427 x 10 19 1.0987 13.1 161.7 0 17 34.0 208.8 Using Beaver Valley Unit 1 C6317-1 142.2 1.427 x 1019 1.0987 13.1 156.2 0 17 34.0 203.3 surveillance data Lower Shell Plate B7203-2 C6293-2 98.7 1.427 x 1019 1.0987 0.4 108.4 0 17 34.0 142.8 Intermediate to Lower Shell Girth 90136 124.3 1.425 x 1019 1.0982 -56 136.5 17 28 65.5 146.0 Weld 11-714 Using St. Lucie Unit 1 and Millstone 90136 74.6 1.425 x 1019 1.0982 -56 81.9 17 28 65.5 91.4 Unit 2 surveillance data Intermediate Shell Longitudinal 305424 191.7 0.274 x 1019 0.6469 -56 124.0 17 28 65.5 133.5 Welds19-714 A&B Using Beaver Valley Unit 1 305424 186.5 0.274 x 1019 0.6469 -56 120.6 17 28 65.5 130.2 surveillance data Lower Shell Longitudinal Weld 305414 210.5 0.276 x 1019 0.6492 -56 136.6 17 28 65.5 146.2 20-714 A&B Using Fort Calhoun surveillance data 305414 216.9 0.276 x 1019 0.6492 -56 140.8 17 28 65.5 150.3 Reactor Vessel Extended Beltline Materials(e)

Upper Shell Forging B6604 123V339VA1 84.2 0.174 x 1019 0.5366 40 45.2 0 17 34.0 119.2 Upper Shell to Intermediate Shell 305414 210.5 0.174 x 1019 0.5366 -56 113.0 17 28 65.5 122.5 Girth Weld 10-714 (3951 & 3958) 305414 Using Fort Calhoun surveillance data 216.9 0.174 x 1019 0.5366 -56 116.4 17 28 65.5 125.9 (3951 & 3958)

AOFJ 41.0 0.174 x 1019 0.5366 10 22.0 17 11.0 40.5 72.5 Upper Shell to Intermediate Shell FOIJ 41.0 0.174 x 10 19 0.5366 10 22.0 17 11.0 40.5 72.5 Girth Weld 10-714 (continued) EODJ 27.0 0.174 x 1019 0.5366 10 14.5 17 7.2 37.0 61.4 HOCJ 27.0 0.174 x 1019 0.5366 10 14.5 17 7.2 37.0 61.4 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 7-7 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 3/4T Location Reactor Vessel Material and ID CF 3/4T Fluence RTNDT(U)(a) RTNDT I(a) (c) Margin ART(d)

Heat Number 3/4T FF Number (°F) (n/cm2, E > 1.0 MeV) (°F) (°F) (°F) (°F) (°F) (°F)

EODJ 27.0 0.00509 x 1019 0.0679 10 1.8 17 0.9 34.0 45.9 FOIJ 41.0 0.00509 x 10 19 0.0679 10 2.8 17 1.4 34.1 46.9 HOCJ 27.0 0.00509 x 10 19 0.0679 10 1.8 17 0.9 34.0 45.9 Inlet Nozzle Welds 1-717B, 1-717D, DBIJ 27.0 0.00509 x 10 19 0.0679 10 1.8 17 0.9 34.0 45.9 1-717F EOEJ 20.0 0.00509 x 10 19 0.0679 10 1.4 17 0.7 34.0 45.4 ICJJ 41.0 0.00509 x 10 19 0.0679 10 2.8 17 1.4 34.1 46.9 JACJ 54.0 0.00509 x 10 19 0.0679 10 3.7 17 1.8 34.2 47.9 ICJJ 41.0 0.00390 x 10 19 0.0556 10 2.3 17 1.1 34.1 46.4 IOBJ 27.0 0.00390 x 10 19 0.0556 10 1.5 17 0.8 34.0 45.5 Outlet Nozzle Welds 1-717A, JACJ 54.0 0.00390 x 1019 0.0556 10 3.0 17 1.5 34.1 47.1 1-717C, 1-717E HOCJ 27.0 0.00390 x 1019 0.0556 10 1.5 17 0.8 34.0 45.5 EODJ 27.0 0.00390 x 10 19 0.0556 10 1.5 17 0.8 34.0 45.5 FOIJ 41.0 0.00390 x 10 19 0.0556 10 2.3 17 1.1 34.1 46.4 Notes:

(a) The plate and forging material initial RTNDT values are measured values. The initial RTNDT values for all of the reactor vessel welds are generic; hence I = 17°F for all reactor vessel welds.

(b) Not used.

(c) As discussed in Section 4, the surveillance plate and weld Heat # 305414 and # 305424 data were deemed non-credible. The surveillance weld data for Heat # 90136 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. 1], the base metal

= 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal = 28°F for Position 1.1 and 2.1 with non-credible surveillance data.

Since a full margin term will be used for Heat # 90136, = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, need not exceed 0.5*RTNDT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(U) + RTNDT + Margin.

(e) As discussed in Section 7, the inlet and outlet nozzle forging material ART calculations utilizing a 3/4T fluence value are excluded from this table.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 7-8 Table 7-4 Summary of the Limiting ART Values Used in the Generation of the Beaver Valley Unit 1 Heatup and Cooldown Curves at 50 EFPY 1/4T Limiting ART(a) 3/4T Limiting ART(a) 244.5°F 209.5°F Lower Shell Plate B6903-1 (Position 1.1)

Note:

(a) The ART values used for P-T limit curve development in this report are the limiting 1/4T and 3/4T ART values calculated in Tables 7-2 and 7-3 rounded to the nearest whole number, then increased by 0.5°F to add additional conservatism.

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Westinghouse Non-Proprietary Class 3 8-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4.

Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 50 EFPY, with the flange requirements and using the Axial Flaw methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of -0, -20, -40, -60, and -100°F/hr applicable for 50 EFPY, with the flange requirements and using the Axial Flaw methodology. The heatup and cooldown curves were generated using the 1998 Edition through the 2000 Addenda ASME Code Section XI, Appendix G.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 8-1 and 8-2. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 Edition through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

1.5 KIm < KIc (13)

where, KIm is the stress intensity factor covered by membrane (pressure) stress, KIc = 33.2 + 20.734 e [0.02 (T - RTNDT)],

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation in order to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 6 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic leak tests for the Beaver Valley Unit 1 reactor vessel at 50 EFPY is 301°F; this temperature value is calculated based on Equation (13). The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

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Westinghouse Non-Proprietary Class 3 8-2 Figures 8-1 and 8-2 define all of the above limits for ensuring prevention of non-ductile failure for the Beaver Valley Unit 1 reactor vessel for 50 EFPY with the flange requirements and without instrumentation uncertainties. The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 8-1 and 8-2 are presented in Tables 8-1 and 8-2. The P-T limit curves shown in Figures 8-1 and 8-2 were generated based on the limiting ART values for the cylindrical beltline and extended beltline reactor vessel materials. These ART values were slightly increased to add additional margin; this approach is conservative. As discussed in Appendix B, the P-T limits developed for the cylindrical beltline region remain bounding with consideration of the reactor vessel inlet and outlet nozzles for Beaver Valley Unit 1 at 50 EFPY.

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Westinghouse Non-Proprietary Class 3 8-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B6903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY: 1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 loperlim Version 54 Run: 19454 Operlim xlsm Version 54 I ,

2250 I

ILeak Test Limit~,

2000 J J 1750 IUnacceptable Operation Ii I IAccept~ble I Operation 1500 IHeatup Rate~

60 Dea. F/Hr

§, 60 Deg. F/Hr

-fl>

a.

IHeaL ateJJ I

Q)

L..

1250 I

J fl> 100Deo.F/H// Critical Limit.I fl> 100 Deq. F/Hr Q)

L..

a..

/; I

'C Q)

+-' 1000

~

J 0

cu u

~/

750

~

Criticality Limit based on 500 inservice hydrostatic test *-

temperature (301°F) for the service period up to 50 EFPY HBoltup Temperatur

.J 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-1 Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 50 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ K1c)

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 8-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B6903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY: 1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 Operlim Version:5.4 Run:19454 Operlim.xlsm Version: 5.4 2250 2000 Unacceptable Acceptable Operation Operation 1750 1500 Calculated Pressure (psig) 1250 1000 750 Cooldown Rates

ºF/Hr Steady-State 500

-20

-40

-60

-100 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-2 Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, and -100°F/hr) Applicable for 50 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc)

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Westinghouse Non-Proprietary Class 3 8-5 Table 8-1 Beaver Valley Unit 1 50 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/

KIc, w/ Flange Notch, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 301 0 60 0 301 0 60 602 301 1190 60 552 301 947 65 602 305 1241 65 552 305 990 70 602 310 1303 70 552 310 1042 75 602 315 1358 75 552 315 1099 80 602 320 1417 80 552 320 1162 85 602 325 1483 85 552 325 1232 90 602 330 1555 90 552 330 1310 95 602 335 1636 95 552 335 1395 100 602 340 1724 100 552 340 1488 105 602 345 1821 105 552 345 1592 110 603 350 1929 110 552 350 1706 115 604 355 2048 115 552 355 1832 120 606 360 2179 120 552 360 1971 125 609 365 2324 125 552 365 2124 130 612 370 2483 130 552 370 2292 135 616 135 552 375 2464 140 621 140 553 145 627 145 555 150 633 150 557 155 640 155 561 160 648 160 565 165 657 165 570 170 667 170 575 175 678 175 582 180 691 180 590 185 704 185 598 190 719 190 608 195 736 195 619 200 755 200 631 205 775 205 645 210 798 210 660 215 823 215 677 220 851 220 696 225 882 225 717 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 8-6 Table 8-1 Beaver Valley Unit 1 50 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/

KIc, w/ Flange Notch, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 230 915 230 741 235 953 235 766 240 994 240 795 245 1040 245 827 250 1085 250 861 255 1132 255 900 260 1184 260 943 265 1241 265 990 270 1303 270 1042 275 1358 275 1099 280 1417 280 1162 285 1483 285 1232 290 1555 290 1310 295 1636 295 1395 300 1724 300 1488 305 1821 305 1592 310 1929 310 1706 315 2048 315 1832 320 2179 320 1971 325 2324 325 2124 330 2483 330 2292 335 2464 Leak Test Limit T (°F) P (psig) 283 2000 301 2485 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 8-7 Table 8-2 Beaver Valley Unit 1 50 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -20°F/hr. -40°F/hr. -60°F/hr. -100°F/hr.

T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 607 60 563 60 518 60 426 65 621 65 608 65 564 65 519 65 426 70 621 70 609 70 565 70 520 70 427 75 621 75 610 75 566 75 521 75 428 80 621 80 611 80 567 80 522 80 429 85 621 85 613 85 569 85 523 85 431 90 621 90 614 90 570 90 525 90 432 95 621 95 616 95 572 95 527 95 434 100 621 100 618 100 574 100 529 100 436 105 621 105 621 105 576 105 531 105 439 110 621 110 621 110 579 110 534 110 442 115 621 115 621 115 582 115 537 115 445 120 621 120 621 120 585 120 541 120 449 125 621 125 621 125 589 125 545 125 453 130 621 130 621 130 593 130 549 130 458 130 680 130 637 135 598 135 554 135 464 135 684 135 641 140 603 140 559 140 470 140 689 140 646 145 609 145 566 145 477 145 694 145 652 150 615 150 572 150 485 150 700 150 658 155 623 155 580 155 494 155 706 155 665 160 630 160 588 160 504 160 713 160 672 165 639 165 598 165 515 165 721 165 680 170 649 170 609 170 527 170 729 170 689 175 660 175 620 175 541 175 739 175 700 180 672 180 633 180 556 180 749 180 711 185 685 185 648 185 573 185 761 185 723 190 700 190 664 190 593 190 774 190 737 195 717 195 682 195 614 195 788 195 752 200 735 200 702 200 637 200 803 200 769 205 755 205 724 205 664 205 821 205 788 210 778 210 748 210 693 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 8-8 Table 8-2 Beaver Valley Unit 1 50 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -20°F/hr. -40°F/hr. -60°F/hr. -100°F/hr.

T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 210 840 210 808 215 802 215 775 215 725 215 861 215 831 220 830 220 805 220 761 220 884 220 856 225 860 225 838 225 801 225 910 225 884 230 894 230 875 230 846 230 938 230 915 235 931 235 916 235 895 235 970 235 949 240 973 240 961 240 949 240 1004 240 987 245 1018 245 1011 245 1010 245 1043 245 1029 250 1069 250 1067 250 1067 250 1085 250 1075 255 1125 255 1125 255 1125 255 1132 255 1127 260 1183 260 1183 260 1183 260 1184 260 1183 265 1241 265 1241 265 1241 265 1241 265 1241 270 1305 270 1305 270 1305 270 1305 270 1305 275 1375 275 1375 275 1375 275 1375 275 1375 280 1452 280 1452 280 1452 280 1452 280 1452 285 1537 285 1537 285 1537 285 1537 285 1537 290 1632 290 1632 290 1632 290 1632 290 1632 295 1736 295 1736 295 1736 295 1736 295 1736 300 1851 300 1851 300 1851 300 1851 300 1851 305 1979 305 1979 305 1979 305 1979 305 1979 310 2120 310 2120 310 2120 310 2120 310 2120 315 2275 315 2275 315 2275 315 2275 315 2275 320 2448 320 2448 320 2448 320 2448 320 2448 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

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

Nuclear Regulatory Commission, May 1988.

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

3. Appendix G to the 1998 Edition through the 2000 Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

4. Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.

5. Westinghouse Report WCAP-17896-NP, Revision 0, Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2014.

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

7. RSICC Computer Code Collection CCC-650, DOORS3.2: One-, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System, April 1998.

8. 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.

9. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S. Nuclear Regulatory Commission, October 14, 2014. [Agencywide Document Management System (ADAMS)

Accession Number ML14149A165]

10. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, Fracture Toughness Requirements, Revision 2, U.S. Nuclear Regulatory Commission, March 2007.

11. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Section NB-2300, Fracture Toughness Requirements for Material.

12. Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1, Determination of Unirradiated RTNDT Values of the Four Beaver Valley Unit 1 Reactor Vessel Beltline Plate Materials Using a Hyperbolic Tangent Curve Fit, dated July 6, 2015.

13. BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.

14. Westinghouse Report WCAP-15571 Supplement 1, Revision 2, Analysis of Capsule Y from the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2011.

15. Florida Power & Light Letter L-2010-078, Attachment 5, License Amendment Request Extended Power Uprates Licensing Report Florida Power & Light St. Lucie Nuclear Plant, Unit 1, April 2010.

[ADAMS Accession Number ML101160193]

16. Westinghouse Report WCAP-16012, Revision 0, Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program, February 2003.

17. FirstEnergy Nuclear Operating Company Letter L-12-077, Pressure and Temperature Limits Report Revision, dated April 5, 2012.

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Westinghouse Non-Proprietary Class 3 9-2

18. Westinghouse Report WCAP-16799-NP, Revision 1, Beaver Valley Power Station Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, June 2007.

19. K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number ML110070570]

20. Not used.

21. Westinghouse Report WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996.

22. Westinghouse Report WCAP-15571, Revision 0, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, November 2000.

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)

Tables A-1 and A-2 contain the thermal stress intensity factors (KIt) for the maximum heatup and cooldown rates at 50 EFPY for Beaver Valley Unit 1. The reactor vessel cylindrical shell radii to the 1/4T and 3/4T locations are as follows:

1/4T Radius = 80.625 inches

3/4T Radius = 84.562 inches WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 A-2 Table A-1 KIt Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements and w/o Margins for Instrument Errors)

Water Vessel Temperature 1/4T Thermal Stress Vessel Temperature 3/4T Thermal Stress Temp. at 1/4T Location for Intensity Factor at 3/4T Location for Intensity Factor

(°F) 100°F/hr Heatup (°F) (ksi in.) 100°F/hr Heatup (°F) (ksi in.)

60 56.130 -0.987 55.065 0.493 65 58.927 -2.377 55.425 1.455 70 62.129 -3.521 56.315 2.377 75 65.562 -4.586 57.748 3.208 80 69.262 -5.475 59.641 3.929 85 73.079 -6.273 61.944 4.558 90 77.089 -6.948 64.601 5.101 95 81.193 -7.553 67.562 5.578 100 85.435 -8.069 70.788 5.991 105 89.755 -8.531 74.238 6.353 110 94.171 -8.928 77.881 6.671 115 98.650 -9.285 81.690 6.951 120 103.196 -9.594 85.642 7.198 125 107.790 -9.875 89.717 7.418 130 112.433 -10.118 93.898 7.612 135 117.114 -10.341 98.171 7.785 140 121.829 -10.535 102.523 7.940 145 126.574 -10.715 106.944 8.080 150 131.343 -10.873 111.424 8.206 155 136.136 -11.020 115.955 8.320 160 140.945 -11.151 120.529 8.423 165 145.773 -11.275 125.142 8.519 170 150.613 -11.385 129.788 8.606 175 155.467 -11.491 134.462 8.687 180 160.330 -11.586 139.161 8.762 185 165.204 -11.678 143.881 8.833 190 170.083 -11.763 148.620 8.899 195 174.972 -11.845 153.374 8.961 200 179.864 -11.920 158.143 9.020 205 184.764 -11.995 162.923 9.077 210 189.666 -12.064 167.713 9.131 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 A-3 Table A-2 KIt Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors)

Water Vessel Temperature at 1/4T -100°F/hr Cooldown Temp. Location for -100°F/hr 1/4T Thermal Stress

(°F) Cooldown (°F) Intensity Factor (ksi in.)

210 232.426 13.510 205 227.352 13.454 200 222.278 13.398 195 217.204 13.342 190 212.131 13.286 185 207.057 13.230 180 201.983 13.175 175 196.909 13.119 170 191.836 13.063 165 186.762 13.008 160 181.688 12.952 155 176.615 12.897 150 171.541 12.842 145 166.468 12.786 140 161.395 12.731 135 156.322 12.676 130 151.249 12.622 125 146.176 12.567 120 141.103 12.512 115 136.031 12.457 110 130.958 12.403 105 125.886 12.349 100 120.813 12.295 95 115.741 12.240 90 110.669 12.187 85 105.597 12.133 80 100.526 12.079 75 95.454 12.025 70 90.382 11.972 65 85.311 11.919 60 80.241 11.865 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES As described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. B-1], reactor vessel non-beltline materials may define pressure-temperature (P-T) limit curves that are more limiting than those calculated for the reactor vessel cylindrical shell beltline materials. Reactor vessel nozzles, penetrations, and other discontinuities have complex geometries that can exhibit significantly higher stresses than those for the reactor vessel beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperatures (RTNDT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.

The methodology contained in WCAP-14040-A, Revision 4 [Ref. B-2] was used in the main body of this report to develop P-T limit curves for the limiting Beaver Valley Unit 1 cylindrical shell beltline material; however, WCAP-14040-A, Revision 4 does not consider ferritic materials in the area adjacent to the beltline, specifically the stressed inlet and outlet nozzles. Due to the geometric discontinuity, the inside corner regions of these nozzles are the most highly stressed ferritic component outside the beltline region of the reactor vessel; therefore, these components are analyzed in this Appendix.

B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The ART values for the inlet and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 7 of this report, and weight percent (wt. %) copper (Cu) and nickel (Ni), initial RTNDT value, and projected neutron fluence as inputs.

The ART values for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table B-1 and a summary of the limiting inlet and outlet nozzle ART values for Beaver Valley Unit 1 is presented in Table B-2.

Nozzle Material Properties Copper and nickel weight percent values and the subsequent Position 1.1 CF values, were previously documented in WCAP-15571, Supplement 1, Revision 2 [Ref. B-4] and are taken directly from this source for this analysis. The initial RTNDT values were determined for each of the Beaver Valley Unit 1 reactor vessel inlet and outlet nozzle forging materials using the BWRVIP-173-A [Ref. B-5] Alternative Approach 2 methodology, contained in Appendix B of that report. For all six of the Beaver Valley Unit 1 inlet and outlet nozzle materials, CVGraph Version 6.02 was utilized to plot the material-specific Charpy V-Notch impact energy data from the Certified Material Test Reports (CMTRs) to determine the transition temperatures at 35 ft-lb and 50 ft-lb, as specified in the Alternative Approach 2 methodology. The 35 ft-lb and 50 ft-lb temperatures were then evaluated per the Alternative Approach 2 methodology presented in BWRVIP-173-A, to determine the initial RTNDT values for the inlet and outlet nozzle materials for Beaver Valley Unit 1. The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in the CMTRs; thus, it was conservatively assumed that the orientation was the strong direction for each nozzle forging. Therefore, the 50 ft-lb transition temperatures for the inlet nozzles were increased by 30°F to provide conservative estimates for specimens oriented in the weak direction per the Alternative Approach 2 methodology in BWRVIP-173-A.

The material properties of the Beaver Valley Unit 1 inlet and outlet nozzle forging materials are documented in Table B-1.

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Westinghouse Non-Proprietary Class 3 B-2 Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Beaver Valley Unit 1 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside corner of the nozzle, for conservatism.

Per Table 2-7, the inlet nozzles are determined to receive a projected maximum fluence of 2.10 x 1017 n/cm2 (E > 1 MeV) at the lowest extent of the nozzles at 50 EFPY. Similarly, the outlet nozzles are projected to achieve a maximum fluence value of 1.61 x 1017 n/cm2 (E > 1 MeV) at the lowest extent of the nozzles at 50 EFPY. Per NRC RIS 2014-11 [Ref. B-1], embrittlement of reactor vessel materials, with projected fluence values greater than 1 x 1017 n/cm2, must be considered.

The neutron fluence values used in the ART calculations for the Beaver Valley Unit 1 inlet and outlet nozzle forging materials are summarized in Table B-1.

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Westinghouse Non-Proprietary Class 3 B-3 Table B-1 ART Calculations for the Beaver Valley Unit 1 Reactor Vessel Nozzle Materials at 50 EFPY Fluence at Lowest Wt. % Wt. % CF(a) RTNDT(U)(c) RTNDT U (e) Margin ART Reactor Vessel Material Extent of Nozzle(b) FF(b)

Cu(a) Ni(a) (°F) (°F) (°F) (°F) (°F) (°F) (°F)

(n/cm2, E > 1.0 MeV)

Inlet Nozzle B6608-1 0.10 0.82 67.0 0.0210 x 1019 0.1773 48.5 11.9 0 5.9 11.9 72.3 Inlet Nozzle B6608-2 0.10 0.82 67.0 0.0210 x 10 19 0.1773 -15.2 11.9 0 5.9 11.9 8.6 Inlet Nozzle B6608-3 0.08 0.79 51.0 0.0210 x 1019 0.1773 11.4 9.0 0 4.5 9.0 29.5 Outlet Nozzle B6605-1 0.13 0.77 95.3 0.0161 x 1019 0.1501 -26.2 14.3 0 7.2 14.3 2.4 Outlet Nozzle B6605-2 0.13 0.77 95.3 0.0161 x 1019 0.1501 3.3 14.3 0 7.2 14.3 31.9 Outlet Nozzle B6605-3 0.09 0.79 58.0 0.0161 x 10 19 0.1501 10.1 8.7 0 4.4 8.7 27.5 Notes:

(a) Cu and Ni wt. % values, as well as CF values were obtained from WCAP-15571, Supplement 1, Revision 2 [Ref. B-4].

(b) Fluence values conservatively correspond to 50 EFPY fluence values at the lowest extent of the nozzle weld. FF values were calculated using Regulatory Guide 1.99, Revision 2.

(c) RTNDT(U) values were determined using the Alternative Approach 2 methodology as described in Appendix B of BWRVIP-173-A.

(d) Not used.

(e) Per Regulatory Guide 1.99, Revision 2, the base metal nozzle forging materials = 17°F for Position 1.1 without surveillance data. However, need not exceed 0.5*RTNDT.

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Westinghouse Non-Proprietary Class 3 B-4 Table B-2 Summary of the Limiting ART Values for the Beaver Valley Unit 1 Inlet and Outlet Nozzle Materials Nozzle Material and ID Limiting ART Value EFPY Number (°F)

Inlet Nozzle B6608-1 72.3 50 Outlet Nozzle B6605-2 31.9 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS PWROG-15109-NP-A [Ref. B-6] addresses the effects of the inlet and outlet nozzle P-T limit curves generically for the U.S. pressurized water reactor (PWR) operating fleet. The results of PWROG-15109-NP-A demonstrate that P-T limit curves developed with current NRC-approved methods (e.g. WCAP-14040-A) bound the generic nozzle P-T limit curves. The NRC has accepted PWROG-15109-NP-A for addressing the concerns of RIS 2014-11 for the inlet and outlet nozzles. The results and conclusions of PWROG-15109-NP-A are applicable as long as the plant-specific Beaver Valley Unit 1 fluence of the nozzle corners remains less than the screening criterion of 4.28 x 1017 n/cm2, as described in PWROG-15109-NP-A. Table 2-7 demonstrates Beaver Valley Unit 1 adherence to this screening criterion at 50 EFPY, thus PWROG-15109-NP-A is applicable.

Conclusion PWROG-15109-NP-A demonstrates that the nozzles will not be limiting with respect to the P-T limit curves at Beaver Valley Unit 1 at 50 EFPY. Therefore, the P-T limits provided in Section 8 for 50 EFPY remain limiting for the beltline and non-beltline reactor vessel components.

B.3 REFERENCES B-1 NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S.

Nuclear Regulatory Commission, October 14, 2014. [ADAMS Accession Number ML14149A165]

B-2 Westinghouse Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

B-3 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988.

B-4 Westinghouse Report WCAP-15571 Supplement 1, Revision 2, Analysis of Capsule Y from the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2011.

B-5 BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.

B-6 Pressurized Water Reactor Owners Group (PWROG) Report PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]

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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C OTHER REACTOR COOLANT PRESSURE BOUNDARY FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [Ref. C-1], requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement for all RCPB components, which is specified in NB-2332(b) of the Section III ASME Code, is the relevant requirement that would affect the pressure-temperature (P-T) limits. The lowest service temperature (LST) requirement of NB-2332(b) of the Section III ASME Code is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 1/2 inches [Ref. C-2]. Note that the Beaver Valley Unit 1 reactor coolant system does not have ferritic materials in the Class 1 piping, pumps or valves. Therefore, the LST requirements of NB-2332(b) are not applicable to the Beaver Valley Unit 1 P-T limits and the only ferritic RCPB components that are not part of the reactor vessel beltline or extended beltline consist of the replacement steam generators, the replacement reactor vessel closure head and the pressurizer.

The replacement steam generators (RSG) were designed and evaluated to the 1989 Edition Section III ASME Code and met all applicable requirements at the time of construction. Furthermore, the RSGs have not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for these components with regards to P-T limits.

The replacement reactor vessel closure head materials have been considered in the development of the P-T limits, see Section 6.3 of this report for further detail. The replacement reactor vessel closure head was constructed to the 1989 Edition Section III ASME Code and met all applicable requirements at the time of construction. Furthermore, the replacement reactor vessel closure head has not undergone neutron embrittlement that would affect P-T limits.

The pressurizer was constructed to the 1965 Edition through 1966 Winter Addenda Section III ASME Code and ASME Code Case-1401 and met all applicable requirements at the time of construction and is original to the plant. Furthermore, the pressurizer has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits.

C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995.

C-2 ASME B&PV Code Section III, Division I, NB-2332, Material for Piping Pumps, and Valves, Excluding Bolting Material.

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

The credibility of all surveillance program data previously applicable to the Beaver Valley Unit 1 reactor vessel was assessed in WCAP-17896-NP [Ref. D-2]. However, since this evaluation, additional weld Heat

90136 surveillance capsule data from the Millstone Unit 2 surveillance program has been deemed applicable, and the Beaver Valley Unit 1 surveillance capsule fluence values have been updated. Thus, this Appendix documents the necessary updates to the credibility evaluation of surveillance program data applicable to Beaver Valley Unit 1.

The Millstone Unit 2 surveillance program includes two distinct welds, Heat # 90136 and Heat # 10137.

In previous analyses, this weld surveillance data was treated as one combined weld and subsequently analyzed together. However, these two weld metal heats were not melted together into a tandem weld; they were individually deposited. It cannot be determined with full confidence how much of the overall surveillance weld is which weld metal heat and, furthermore, exactly which weld heat specimens are contained in which surveillance capsules in the Millstone Unit 2 program.

The Millstone Unit 2 (combined) surveillance weld data met the second and third credibility criteria of Regulatory Guide 1.99, Revision 2 [Ref. D-1]. Additionally, Table D-2 of WCAP-16012 [Ref. D-3]

indicates that all of the measured weld RTNDT values were within the 1-sigma scatter band; therefore, suggesting that there is good agreement between the measured capsule data and the embrittlement correlations. If the two heats of weld material were evaluated individually, one would expect that the scatter in the data would decrease since the irradiated material would embrittle differently for the two separate welds with different, as-measured, copper and nickel contents. However, since the (combined) weld material already passes the Regulatory Guide 1.99, Revision 2 credibility analysis, a re-evaluation of the material (as two separate heats) is not expected to significantly change the overall results of the subsequent reactor vessel integrity analyses. Thus, the surveillance weld metal will be considered to be only Heat #

90136 for the evaluations contained herein. All currently determined input data for Position 2.1 chemistry factor determination (See Section 5) and surveillance data credibility assessment documented in this Appendix will be used as-is, as documented in the Millstone Unit 2 surveillance capsule analyses of record.

For conservatism, no reduction in the margin term of Regulatory Guide 1.99, Revision 2 [Ref. D-1] and 10 CFR 50.61 [Ref. D-4] was taken to account for the additional uncertainties, despite the data remaining credible (see Section D.2). Additionally, the Beaver Valley Unit 1 intermediate to lower shell girth weld seam 11-714 (Heat # 90136) was assigned the most limiting calculated ART and RTPTS values during the evaluations contained in Section 7 and Appendix E, respectively. Thus, since the values determined using Position 2.1 are less conservative than the values determined using Position 1.1, the more conservative WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 D-2 Position 1.1 values were used. However, despite these additional conservatisms, the Beaver Valley Unit 1 intermediate to lower shell girth weld seam 11-714 (Heat # 90136) was not the limiting material in any evaluation.

D.2 EVALUATION The only required updates to the previously determined credibility conclusions are to update Criterion 3 of Regulatory Guide 1.99, Revision 2 [Ref. D-1] to include the combined surveillance capsule data set for weld Heat # 90136 from both the St. Lucie Unit 1 and Millstone Unit 2 surveillance programs and to update Criterion 3 of Regulatory Guide 1.99, Revision 2 [Ref. D-1] utilizing the Beaver Valley Unit 1 surveillance capsule fluence values documented in Appendix F. These evaluations are documented herein. Criterion #

1, 2, 4, and 5 conclusions remain unchanged from those documented in Appendix D of WCAP-17896-NP

[Ref. D-3]. Note also that the credibility assessment of Heat # 305414 data remains valid as documented in Appendix D of WCAP-17896-NP [Ref. D-2].

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2 [Ref. D-1] normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-5].

The functional form of the least-squares method as described in Regulatory Guide 1.99, Revision 2 will be utilized to determine a best-fit line for this data and to determine if the scatter of these RTNDT values about this line is less than 28°F for the weld.

Following is the calculation of the best-fit line as described in Reference D-1. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [Ref. D-6]. At this meeting, the NRC presented five cases. Of the five cases, Case 5 (Surveillance Data from Other Sources Only) most closely represents the situation for the Beaver Valley Unit 1 reactor vessel intermediate to lower shell girth weld seam 11-714 (Heat # 90136) as described below. Of the five cases, Case 1 (Surveillance data available from plant but no other source) most closely represents the situation for the Beaver Valley Unit 1 surveillance materials as described below.

Heat # 90136 (Case 5) - This weld heat pertains to the intermediate to lower shell girth weld seam 11-714 in the Beaver Valley Unit 1 reactor vessel. This weld heat is not contained in the Beaver Valley Unit 1 surveillance program. However, it is contained in the St. Lucie Unit 1 and Millstone Unit 2 surveillance programs. NRC Case 5 per Reference D-6 is entitled Surveillance Data from Other Sources Only and most closely represents the situation for Beaver Valley Unit 1 weld Heat # 90136.

Lower Shell Plate B6903-1 (Case 1) - This plate material will be evaluated using the NRC Case 1 guidelines as described above.

Weld Heat # 305424 (Case 1) - This weld heat pertains to the intermediate shell longitudinal welds in the Beaver Valley Unit 1 reactor vessel. NRC Case 1 per Reference D-6 regarding Surveillance data available from the plant but no other source most closely represents the situation for Beaver Valley Unit 1 weld Heat # 305424.

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Westinghouse Non-Proprietary Class 3 D-3 Credibility Assessment Case 5: Weld Heat # 90136 (St. Lucie Unit 1 Data Only)

Following the NRC Case 5 guidelines, the St. Lucie Unit 1 and Millstone Unit 2 surveillance weld metal (Heat # 90136) will be evaluated for credibility. Weld Heat # 90136 pertains to Beaver Valley Unit 1 reactor vessel intermediate to lower shell girth weld seam 11-714, but is not contained in the Beaver Valley Unit 1 surveillance program.

In accordance with the NRC Case 5 guidelines, the data from only St. Lucie Unit 1 will be analyzed first, since the irradiation environment for St. Lucie Unit 1 is judged closer to that of Beaver Valley Unit 1 as evidenced by the temperature adjustments documented in Table 4-2. This assessment was performed in Appendix D of WCAP-17896-NP [Ref. D-2] and concluded that the surveillance data for Heat # 90136 from St. Lucie Unit 1 only was credible. Therefore, in accordance with Case 5, the combined data from both St. Lucie Unit 1 and Millstone Unit 2 will now be assessed to determine the credibility conclusion for all applicable data for weld Heat # 90136.

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Westinghouse Non-Proprietary Class 3 D-4 Credibility Assessment Case 5: Weld Heat # 90136 (All data)

In accordance with the NRC Case 5 guidelines, the data from St. Lucie Unit 1 and Millstone Unit 2 will now be analyzed together. Data is adjusted to the mean chemical composition and operating temperature of the surveillance capsules. This is performed in Table D-1.

Table D-1 Mean Chemical Composition and Operating Temperature for St. Lucie Unit 1 and Millstone Unit 2 Cu Ni Inlet Temperature during Material Capsule Wt. %(a) Wt. %(a) Period of Irradiation (°F)(b) 97° 541 Weld Metal Heat # 90136 104° 0.23 0.07 544.6 (St. Lucie Unit 1 Data) 284° 546.3 97° 544.3 Weld Metal Heat # 90136 104° 0.30 0.06 547.6 (Millstone Unit 2 Data) 83° 548.0 MEAN 0.265 0.065 545.3 Note:

(a) Chemistry data obtained from Table 3-2.

(b) Temperature data obtained from Table 4-2.

Therefore, the St. Lucie Unit 1 and Millstone Unit 2 surveillance capsule data will be adjusted to the mean chemical composition and operating temperature calculated in Table D-1.

St. Lucie Unit 1 data CFMean = 121.2°F (calculated per Table 1 of Regulatory Guide 1.99, Revision 2 [Ref. D-1] using Cu Wt. % =

0.265 and Ni Wt. % = 0.065 per Table D-1)

CFSurv. Weld (St. Lucie Unit 1) = 106.6°F (from Table 5-4)

Ratio = 121.2 ÷ 106.6 = 1.14 (applied to St. Lucie Unit 1 surveillance data for weld Heat # 90136 in the credibility evaluation)

Millstone Unit 2 data CFMean = 121.2°F CFSurv. Weld (Millstone Unit 2) = 135.5°F (from Table 5-4)

Ratio = 121.2 ÷ 135.5 = 0.89 (applied to Millstone Unit 2 surveillance data for weld Heat # 90136 in the credibility evaluation)

The capsule-specific temperature adjustments are as shown in Table D-2.

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Westinghouse Non-Proprietary Class 3 D-5 Table D-2 Operating Temperature Adjustments for the St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data Inlet Temperature during Mean Operating Temperature Material Capsule Period of Irradiation (°F) Temperature (°F) Adjustment (°F) 97° 541 -4.3 Weld Metal Heat # 90136 104° 544.6 -0.7 (St. Lucie Unit 1 Data) 284° 546.3 1.0 545.3 97° 544.3 -1.0 Weld Metal Heat # 90136 104° 547.6 2.3 (Millstone Unit 2 Data) 83° 548.0 2.7 Using the chemical composition and operating temperature adjustments described and calculated above, an interim chemistry factor is calculated for weld Heat # 90136 using the St. Lucie Unit 1 and Millstone Unit 2 data. This calculation is shown in Table D-3 below.

Table D-3 Calculation of Weld Heat # 90136 Interim Chemistry Factor for the Credibility Evaluation Using St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data Capsule f(a) RTNDT(c) FF*RTNDT Material Capsule FF(b) FF2 (x 10 n/cm2, E > 1.0 MeV) 19

(°F) (°F)

Weld Metal Heat # 97° 0.5174 0.8160 77.6 (72.34) 63.29 0.666 90136 (St. Lucie 104° 0.7885 0.9333 76.0 (67.4) 70.97 0.871 Unit 1 Data) 284° 1.243 1.0606 78.7 (68.0) 83.43 1.125 Weld Metal Heat # 97° 0.324 0.6902 57.8 (65.93) 39.89 0.476 90136 (Millstone 104° 0.949 0.9853 48.4 (52.12) 47.72 0.971 Unit 2 Data) 83° 1.74 1.1523 52.3 (56.09) 60.29 1.328 SUM: 365.59 5.437 CFHeat # 90136 = (FF

RTNDT) ÷ (FF2) = (365.59) ÷ (5.437) = 67.2F Notes:

(a) f = fluence.

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

(c) RTNDT values are the measured 30 ft-lb shift values. Each RTNDT value has first been adjusted according to the temperature adjustments summarized in Table D-2. Then, the RTNDT values for each surveillance weld data point are adjusted by the ratios determined previously for weld Heat # 90136 (pre-adjusted values are listed in parentheses and were taken from Table 4-2).

The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. D-1] is presented in Table D-4.

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Westinghouse Non-Proprietary Class 3 D-6 Table D-4 Best-Fit Evaluation for Surveillance Weld Metal Heat # 90136 Using St. Lucie Unit 1 and Millstone Unit 2 Data CF Capsule f Measured Predicted Residual Material Capsule (Slopebest-fit) (x 1019 n/cm2, FF RTNDT RTNDT RTNDT <28°F (Weld)

(°F) E > 1.0 MeV) (°F) (°F) (°F)

Weld Metal Heat # 97° 67.2 0.5174 0.8160 77.6 54.8 22.7 Yes 90136 (St. Lucie 104° 67.2 0.7885 0.9333 76.0 62.7 13.3 Yes Unit 1 Data) 284° 67.2 1.243 1.0606 78.7 71.3 7.4 Yes Weld Metal Heat # 97° 67.2 0.324 0.6902 57.8 46.4 11.4 Yes 90136 (Millstone 104° 67.2 0.949 0.9853 48.4 66.2 17.8 Yes Unit 2 Data) 83° 67.2 1.74 1.1523 52.3 77.4 25.1 Yes The scatter of RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. D-1], should be less than 28°F for weld metal. Table D-4 indicates that 100% (six out of six) of the surveillance data points fall within the +/- 1 of 28°F scatter band for surveillance weld materials. Therefore, the surveillance weld material (Heat # 90136) is deemed credible per the third criterion when all available data is considered.

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Westinghouse Non-Proprietary Class 3 D-7 Credibility Assessment Case 1: Lower Shell Plate B6903-1 and Weld Heat # 305424 Following the NRC Case 1 guidelines, the Beaver Valley Unit 1 surveillance plate and weld metal (Heat #

305424) will be evaluated for credibility. Note that when evaluating the credibility of the surveillance weld data, the measured RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. The chemistry factors for the Beaver Valley Unit 1 surveillance plate and weld material contained in the surveillance program were calculated in accordance with Regulatory Guide 1.99, Revision 2, Position 2.1 and are presented in Table D-5. The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-6.

Table D-5 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit 1 Surveillance Capsule Data Capsule f(a)

RTNDT(c) FF*RTNDT Material Capsule (x 1019 n/cm2, E FF(b) FF2

(°F) (°F)

> 1.0 MeV)

V 0.297 0.6677 127.9 85.40 0.446 Lower Shell U 0.618 0.8652 118.3 102.35 0.749 Plate B6903-1 W 0.952 0.9862 147.7 145.66 0.973 (Longitudinal) Y 2.10 1.2018 141.7 170.30 1.444 X 4.99 1.4020 175.8 246.46 1.965 V 0.297 0.6677 138.0 92.14 0.446 Lower Shell U 0.618 0.8652 132.1 114.29 0.749 Plate B6903-1 W 0.952 0.9862 180.2 177.72 0.973 (Transverse) Y 2.10 1.2018 166.9 200.58 1.444 X 4.99 1.4020 179.0 250.95 1.965 SUM: 1585.86 11.154 CF B6903-1 = (FF

RTNDT) ÷ (FF2) = (1585.86) ÷ (11.154) = 142.2°F V 0.297 0.6677 159.8 106.70 0.446 Surveillance U 0.618 0.8652 164.9 142.67 0.749 Weld Metal W 0.952 0.9862 186.3 183.73 0.973 (Heat #305424) Y 2.10 1.2018 178.5 214.52 1.444 X 4.99 1.4020 237.8 333.38 1.965 SUM: 981.01 5.577 CF Surv. Weld = (FF

RTNDT) ÷ (FF ) = (981.01) ÷ (5.577) = 175.9°F 2

Notes:

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

(c) RTNDT values are the measured 30 ft-lb shift values.

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Westinghouse Non-Proprietary Class 3 D-8 Table D-6 Beaver Valley Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line Capsule CF Measured Predicted Residual <17°F Fluence Material Capsule (Slopebest-fit) FF RTNDT RTNDT RTNDT (Base Metal)

(x 1019 n/cm2,

(°F) (°F) (°F) (°F) <28°F (Weld)

E > 1.0 MeV)

V 142.2 0.297 0.6677 127.9 94.9 33.0 No U 142.2 0.618 0.8652 118.3 123.0 4.7 Yes Lower Shell Plate B6903-1 W 142.2 0.952 0.9862 147.7 140.2 7.5 Yes (Longitudinal)

Y 142.2 2.10 1.2018 141.7 170.9 29.2 No X 142.2 4.99 1.4020 175.8 199.4 23.6 No V 142.2 0.297 0.6677 138.0 94.9 43.1 No U 142.2 0.618 0.8652 132.1 123.0 9.1 Yes Lower Shell Plate B6903-1 W 142.2 0.952 0.9862 180.2 140.2 40.0 No (Transverse)

Y 142.2 2.10 1.2018 166.9 170.9 4.0 Yes X 142.2 4.99 1.4020 179.0 199.4 20.4 No V 175.9 0.297 0.6677 159.8 117.4 42.4 No U 175.9 0.618 0.8652 164.9 152.2 12.7 Yes Surveillance Weld Material W 175.9 0.952 0.9862 186.3 173.5 12.8 Yes (Heat # 305424)

Y 175.9 2.10 1.2018 178.5 211.4 32.9 No X 175.9 4.99 1.4020 237.8 246.6 8.8 Yes From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Table D-6 indicates that only four of the ten surveillance data points fall inside the +/- 1 of 17°F scatter band for surveillance base metals; therefore, the plate data is deemed non-credible per the third criterion.

Table D-6 indicates that only three of the five surveillance data points fall inside the +/- 1 of 28°F scatter band for surveillance weld materials; therefore, the surveillance weld data is deemed non-credible per the third criterion.

D.3 CONCLUSION In conclusion, the combined surveillance data from St. Lucie Unit 1 and Millstone Unit 2 for weld Heat #

90136 may be applied to the Beaver Valley Unit 1 reactor vessel weld. The Position 2.1 chemistry factor calculation, as applicable to the Beaver Valley Unit 1 reactor vessel weld, is contained in Section 5. This Position 2.1 CF value could be used with a reduced margin term in the ART calculations contained in Section 7 and the RTPTS calculations contained in Appendix E. However, consistent with the discussion in Section D.1 of this Appendix, the ART and RTPTS values calculated with the Position 2.1 CF value for weld Heat # 90136 will utilize a full margin term for conservatism. Additionally, the Beaver Valley Unit 1 surveillance plate and weld data remain non-credible, as concluded in WCAP-17896-NP [Ref. D-2].

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Westinghouse Non-Proprietary Class 3 D-9 D.4 REFERENCES D-1 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U. S.

Nuclear Regulatory Commission, May 1988.

D-2 Westinghouse Report WCAP-17896-NP, Revision 0, Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2014.

D-3 Westinghouse Report WCAP-16012, Revision 0, Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program, February 2003.

D-4 Code of Federal Regulations, 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

D-5 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, ASTM, July 1982.

D-6 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

[ADAMS Accession Number ML110070570]

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Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E PRESSURIZED THERMAL SHOCK EVALUATION E.1 PRESSURIZED THERMAL SHOCK CALCULATIONS Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the reactor pressure vessel (RPV) under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect anywhere within the vessel wall.

In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [Ref. E-1]) that established screening criteria on PWR vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS. RTPTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement.

These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTPTS) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 [Ref E-2].

These accepted methods were used with the clad/base metal interface fluence values of Section 2 to calculate the following RTPTS values for the Beaver Valley Unit 1 RPV materials at 50 EFPY (EOLE). The EOLE RTPTS calculations are summarized in Table E-1. The following changes and updates to the analysis of record for PTS at Beaver Valley Unit 1, WCAP-15571 Supplement 1, Revision 2 [Ref. E-3], have been incorporated into the calculations contained in Table E-1 of this letter report.

1. Incorporation of the Capsule X results as documented in WCAP-17896-NP, Revision 0 [Ref. E-4],

updated reactor vessel fluence values (See Section 2), surveillance capsule irradiated material testing results for Lower Shell Plate B6903-1 and Intermediate Shell Longitudinal Welds19-714 A&B (Heat # 305424) (See Sections 4 and 5), and revised credibility conclusions (See Appendix D).

2. Incorporation of sister plant surveillance capsule test results for weld Heat # 90136 from the Millstone Unit 2 reactor vessel surveillance capsule program (See Sections 4, 5 and Appendix D).

Due to the uncertainty in the incorporation of the surveillance data from Millstone Unit 2 (two wire heats were used in the Millstone 2 surveillance weld, with some specimens being Heat # 90136 and others from another weld wire [Heat # 10137]), a full-margin term was used for this material in the RTPTS calculations contained in Table 1, even though the revised credibility analysis confirmed that Heat # 90136 remained credible.

3. Incorporation of revised initial reference nil-ductility transition temperature (RTNDT(U)) values for the four Beaver Valley Unit 1 reactor vessel plate materials as documented in Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1 [Ref. E-5] (See Section 3). It was also concluded that the upper shell forging material for the Beaver Valley Unit 1 reactor vessel has an appropriate RTNDT(U) value even though Branch Technical Position (BTP) 5-3, Paragraph B1.1 (3) [Ref. E-6] methodology for WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 E-2 SA-508, Class 2 forging material must be used, due to lack of clear definition of Charpy V-notch orientation. The initial RTNDT value of this material remains drop-weight limited due to the excellent Charpy V-notch test results, as documented in its Certified Material Test Report (CMTR) (See Section 3).

4. Utilization of BWRVIP-173-A [Ref. E-7] to redefine the initial RTNDT values of the six Beaver Valley Unit 1 nozzle forging materials (See Section 3 and Appendix B).

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Westinghouse Non-Proprietary Class 3 E-3 Table E-1 RTPTS Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY Reactor Vessel Material and ID CF(a) EOLE Fluence(a) RTNDT(U)(a) RTNDT U (b) Margin RTPTS(c)

Heat Number FF Number (°F) (n/cm2, E > 1.0 MeV) (°F) (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-1 100.5 5.88 x 1019 1.4330 26.8 144.0 0 17 34.0 204.8 Intermediate Shell Plate B6607-2 C4381-2 100.5 5.88 x 10 19 1.4330 53.6 144.0 0 17 34.0 231.6 Lower Shell Plate B6903-1 C6317-1 147.2 5.89 x 10 19 1.4333 13.1 211.0 0 17 34.0 258.1 Using Beaver Valley Unit 1 C6317-1 142.2 5.89 x 1019 1.4333 13.1 203.8 0 17 34.0 250.9 surveillance data Lower Shell Plate B7203-2 C6293-2 98.7 5.89 x 1019 1.4333 0.4 141.5 0 17 34.0 175.9 Intermediate to Lower Shell Girth 90136 124.3 5.88 x 1019 1.4330 -56 178.1 17 28 65.5 187.6 Weld 11-714 Using St. Lucie Unit 1 and 90136 74.6 5.88 x 1019 1.4330 -56 106.9 17 28 65.5 116.4 Millstone Unit 2 surveillance data Intermediate Shell Longitudinal 305424 191.7 1.13 x 1019 1.0341 -56 198.2 17 28 65.5 207.8 Welds19-714 A&B Using Beaver Valley Unit 1 305424 186.5 1.13 x 1019 1.0341 -56 192.9 17 28 65.5 202.4 surveillance data Lower Shell Longitudinal Welds 305414 210.5 1.14 x 1019 1.0366 -56 218.2 17 28 65.5 227.7 20-714 A&B Using Fort Calhoun surveillance 305414 216.9 1.14 x 1019 1.0366 -56 224.8 17 28 65.5 234.4 data Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 123V339VA1 84.2 0.718 x 1019 0.9071 40 76.4 0 17 34.0 150.4 Upper Shell to Intermediate Shell 305414 210.5 0.718 x 1019 0.9071 -56 190.9 17 28 65.5 200.5 Girth Weld 10-714 (3951 & 3958)

Using Fort Calhoun surveillance 305414 216.9 0.718 x 1019 0.9071 -56 196.7 17 28 65.5 206.3 data (3951 & 3958)

AOFJ 41.0 0.718 x 1019 0.9071 10 37.2 17 18.6 50.4 97.6 Upper Shell to Intermediate Shell FOIJ 41.0 0.718 x 1019 0.9071 10 37.2 17 18.6 50.4 97.6 Girth Weld 10-714 (continued) EODJ 27.0 0.718 x 1019 0.9071 10 24.5 17 12.2 41.9 76.4 HOCJ 27.0 0.718 x 1019 0.9071 10 24.5 17 12.2 41.9 76.4 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 E-4 Table E-1 RTPTS Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY Reactor Vessel Material and ID CF(a) EOLE Fluence(a) RTNDT(U)(a) RTNDT U (b) Margin RTPTS(c)

Heat Number FF Number (°F) (n/cm2, E > 1.0 MeV) (°F) (°F) (°F) (°F) (°F) (°F)

Inlet Nozzle B6608-1 95443-1 67.0 0.0210 x 1019 0.1773 48.5 11.9 0 5.9 11.9 72.3 Inlet Nozzle B6608-2 95460-1 67.0 0.0210 x 1019 0.1773 -15.2 11.9 0 5.9 11.9 8.6 Inlet Nozzle B6608-3 95712-1 51.0 0.0210 x 1019 0.1773 11.4 9.0 0 4.5 9.0 29.5 EODJ 27.0 0.0210 x 1019 0.1773 10 4.8 17 2.4 34.3 49.1 FOIJ 41.0 0.0210 x 1019 0.1773 10 7.3 17 3.6 34.8 52.0 HOCJ 27.0 0.0210 x 1019 0.1773 10 4.8 17 2.4 34.3 49.1 Inlet Nozzle Welds 1-717B, DBIJ 27.0 0.0210 x 1019 0.1773 10 4.8 17 2.4 34.3 49.1 1-717D, 1-717F EOEJ 20.0 0.0210 x 1019 0.1773 10 3.5 17 1.8 34.2 47.7 ICJJ 41.0 0.0210 x 1019 0.1773 10 7.3 17 3.6 34.8 52.0 JACJ 54.0 0.0210 x 1019 0.1773 10 9.6 17 4.8 35.3 54.9 Outlet Nozzle B6605-1 95415-1 95.3 0.0161 x 1019 0.1501 -26.2 14.3 0 7.2 14.3 2.4 Outlet Nozzle B6605-2 95415-2 95.3 0.0161 x 1019 0.1501 3.3 14.3 0 7.2 14.3 31.9 Outlet Nozzle B6605-3 95444-1 58.0 0.0161 x 1019 0.1501 10.1 8.7 0 4.4 8.7 27.5 ICJJ 41.0 0.0161 x 1019 0.1501 10 6.2 17 3.1 34.6 50.7 IOBJ 27.0 0.0161 x 1019 0.1501 10 4.1 17 2.0 34.2 48.3 Outlet Nozzle Welds 1-717A, JACJ 54.0 0.0161 x 1019 0.1501 10 8.1 17 4.1 35.0 53.1 1-717C, 1-717E HOCJ 27.0 0.0161 x 1019 0.1501 10 4.1 17 2.0 34.2 48.3 EODJ 27.0 0.0161 x 1019 0.1501 10 4.1 17 2.0 34.2 48.3 FOIJ 41.0 0.0161 x 1019 0.1501 10 6.2 17 3.1 34.6 50.7 Notes:

(a) CF values were taken from Table 5-4, fluence values were taken from Tables 2-5, 2-7, and 2-9, and RTNDT(U) values were taken from Table 3-2.

(b) As discussed in Section 4, the surveillance plate and weld Heat # 305414 and # 305424 data were deemed non-credible. The surveillance weld data for Heat # 90136 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of 10 CFR 50.61 [Ref. E-1], the base metal = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal = 28°F for Position 1.1 and 2.1 with non-credible surveillance data. Since a full margin term will be used for Heat # 90136, = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, need not exceed 0.5*RTNDT.

(c) The 10 CFR 50.61 [Ref. E-1] methodology was utilized in the calculation of the PTS values.

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Westinghouse Non-Proprietary Class 3 E-5 E.2 PRESSURIZED THERMAL SHOCK CONCLUSIONS The Beaver Valley Unit 1 limiting RTPTS value for base metal or longitudinal weld materials at 50 EFPY is 258.1°F (see Table E-1), which corresponds to Lower Shell Plate B6903-1 (using Position 1.1). The Beaver Valley Unit 1 limiting RTPTS value for circumferentially oriented welds at 50 EFPY is 206.3°F (see Table E-1), which corresponds to the Upper Shell to Intermediate Shell Girth Weld 10-714 (Heat # 305414, using Position 2.1).

Therefore, all of the beltline and extended beltline materials in the Beaver Valley Unit 1 reactor vessel are below the RTPTS screening criteria of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through EOLE (50 EFPY).

In the PTS analysis of record for Beaver Valley Unit 1, WCAP-15571 Supplement 1, Revision 2 [Ref. E-3], the limiting reactor vessel plate material, Lower Shell Plate B6903-1, was predicted to exceed the RTPTS screening criteria of 270°F for plates at 39.6 EFPY of plant operation. However, with the reevaluation of the Beaver Valley Unit 1 reactor vessel beltline plate material initial RTNDT values [Ref. E-5], along with incorporation of the Capsule X results [Ref. E-4], this material, while still the limiting material, is now predicted to remain under the RTPTS screening limit through EOLE. With consideration of the revised initial RTNDT value for Lower Shell Plate B6903-1, along with incorporation of the Capsule X results, this material is now predicted to remain under the RTPTS screening limit through a minimum of 80 EFPY of plant operation.

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Westinghouse Non-Proprietary Class 3 E-6 E.3 REFERENCES E-1 Code of Federal Regulations, 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

E-2 Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U. S.

Nuclear Regulatory Commission, May 1988.

E-3 Westinghouse Report WCAP-15571 Supplement 1, Revision 2, Analysis of Capsule Y from the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2011.

E-4 Westinghouse Report WCAP-17896-NP, Revision 0, Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2014.

E-5 Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1, Determination of Unirradiated RTNDT Values of the Four Beaver Valley Unit 1 Reactor Vessel Beltline Plate Materials Using a Hyperbolic Tangent Curve Fit, dated July 6, 2015.

E-6 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, Fracture Toughness Requirements, Revision 2, U.S. Nuclear Regulatory Commission, March 2007.

E-7 BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.

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Westinghouse Non-Proprietary Class 3 F-1 APPENDIX F VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS F.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn from service to date at Beaver Valley Unit 1 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [Ref. F-1]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least-squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 2.2 of this report.

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

Irradiation Iron Atom Azimuthal Withdrawal Time Fluence (E>1.0 MeV) Displacements Capsule ID Location Time [EFPY] [n/cm2] [dpa]

V 15° End of Cycle 1 1.2 2.97E+18 4.93E-03 U 25° End of Cycle 4 3.6 6.18E+18 1.00E-02 W 25° End of Cycle 6 5.9 9.52E+18 1.54E-02 Y 25° End of Cycle 13 14.3 2.10E+19 3.40E-02 X 15° End of Cycle 22 26.6 4.99E+19 8.25E-02 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules V, U, W, Y, and X are summarized as follows:

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Westinghouse Non-Proprietary Class 3 F-2 Reaction of Sensor Material Interest Capsule V Capsule U Capsule W Capsule Y Capsule X Copper-63 63 Cu(n,)60Co X X X X X Iron-54 54 Fe(n,p) Mn 54 X X X X X Nickel-58 58 Ni(n,p)58Co X X X X X Uranium-238 238 U(n,f)137Cs X X LOST X X Neptunium-237 237 Np(n,f) 137Cs X X X X X Cobalt-Aluminum* 59 Co(n,)60Co X X X X X

The cobalt-aluminum measurements for this plant include both bare and cadmium-covered wire sensors.

Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table F-1.

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

the measured specific activity of each monitor,

the physical characteristics of each monitor,

the operating history of the reactor,

the energy response of each monitor, and

the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules V, U, W, Y, and X are documented in References F-2 through F-6, respectively. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

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

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

A R= n Pj N0 F Y P i =1 ref C j [1 - e

- t j

] [ e- t d , j ]

where:

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

A = Measured specific activity (dps/g).

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

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

Y = Number of product atoms produced per reaction.

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

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

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

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

tj = Length of irradiation period j (sec).

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

n = Total number of irradiation periods.

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

In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 2.2, accounts for the change in sensor reaction rates caused by variations in fluence rate level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low-WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-4 leakage fuel management, the additional Cj term should be employed. The impact of changing fluence rate levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low-leakage to low-leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.

The fuel-cycle-specific neutron fluence rate values along with the computed values for Cj are listed in Table F-3. These fluence rate values represent the cycle-dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

In performing the dosimetry evaluations for the surveillance capsules, the sensor reaction rates measured at the locations in the capsule holder were indexed to the geometric center of the capsules. This indexing procedure required correcting the measured reaction rates by the application of analytically determined spatial gradients. For the Beaver Valley Unit 1 surveillance capsules, the gradient correction factors for each sensor reaction were obtained from the reference forward transport calculations and were used in a multiplicative fashion to relate individual measured reaction rates to the corresponding value at the geometric center of the surveillance capsule. The correction factors applied to the Beaver Valley Unit 1 sensor reaction rates are summarized as follows:

Correction Factor Capsule V Capsule U Capsule W Capsule Y Capsule X 63 Cu(n,) Radial Gradient 0.956 0.956 0.956 0.956 0.956 54 Fe(n,p) Radial Gradient 1.050 1.051 1.051 1.051 1.050 58 Ni(n,p) Radial Gradient 1.158 1.163 1.163 1.163 1.158 238 U(n,f) Radial Gradient 1.000 1.000 N/A 1.000 1.000 237 Np(n,f) Radial Gradient 1.000 1.000 1.000 1.000 1.000 59 Co(n,) Radial Gradient 0.957 0.956 0.956 0.956 0.957 59 Co(n,) Cd Radial Gradient 1.157 1.155 1.155 1.155 1.157 Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Beaver Valley Unit 1 fission sensor reaction rates are summarized as follows:

Correction Factor Capsule V Capsule U Capsule W Capsule Y Capsule X 235 U Impurity/Pu Build-in 0.873 0.860 N/A 0.806 0.712 238 U(,f) 0.955 0.960 N/A 0.960 0.956 Net 238U Correction Factor 0.834 0.826 N/A 0.774 0.681 237 Np(,f) 0.982 0.983 0.984 0.984 0.983 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-5 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules V, U, W, Y, and X are given in Table F-4a through Table F-4e. In Table F-4a through Table F-4e, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma-ray-induced fission effects.

F.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best estimates for key exposure rate parameters such as (E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties. For example:

R i +/- R i = ( ig +/- ig )( g +/- g )

g relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross section, ig, each with an uncertainty . The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the Beaver Valley Unit 1 surveillance capsule dosimetry, the FERRET code [Ref. F-7] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters

((E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.

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

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

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

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

For the Beaver Valley Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 2.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section F.1.1.

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Westinghouse Non-Proprietary Class 3 F-6 The dosimetry reaction cross sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [Ref. F-8].

The uncertainties associated with the measured reaction rates, dosimetry cross sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance [Ref. F-9].

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Beaver Valley Unit 1 surveillance capsule sensor sets.

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

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

Reaction Uncertainty 63 Cu(n,)60Co 5%

54 Fe(n,p)54Mn 5%

58 Ni(n,p)58Co 5%

238 U(n,f)137Cs 10%

237 Np(n,f)137Cs 10%

59 Co(n,)60Co 5%

These uncertainties are given at the 1 level.

Dosimetry Cross-Section Uncertainties The reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library.

This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross sections have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

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Westinghouse Non-Proprietary Class 3 F-7 For sensors included in the Beaver Valley Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross sections are provided in the SNLRML documentation package.

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

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

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

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

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

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

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

Calculated Neutron Spectrum The neutron spectra input to the least-squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape).

Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

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

M gg ' = Rn2 + R g

R g '

Pgg '

where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Pgg = [1 - ] gg + e-H

where, (g g')2 H=

2 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term).

The value of is 1.0 when g = g, and is 0.0 otherwise.

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Westinghouse Non-Proprietary Class 3 F-8 The set of parameters defining the input covariance matrix for the Beaver Valley Unit 1 calculated spectra was as follows:

Fluence Rate Normalization Uncertainty (Rn) 15%

Fluuence Rate Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation ()

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

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 F.1.3 Comparisons of Measurements and Calculations Results of the least-squares evaluations of the dosimetry from the Beaver Valley Unit 1 surveillance capsules withdrawn to date are provided in Tables F-5 and F-6. In Table F-5, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least-squares adjusted reaction rates.

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

The data comparisons provided in Tables F-5 and F-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross sections. Further, these results indicate that the use of the least-squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 2.3 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the 1 level. From Table F-6, it is noted that the corresponding uncertainties associated with the least-squares adjusted exposure parameters have been reduced to 6% for neutron fluence rate (E > 1.0 MeV) and 6-7% for iron atom displacement rate. Again, the uncertainties from the least-squares evaluation are at the 1 level.

Further comparisons of the measurement results (from Tables F-5 and F-6) with calculations are given in Tables F-7 and F-8. These comparisons are given on two levels. In Table F-7, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-9 threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table F-8, calculations of fast neutron exposure rates in terms of (E > 1.0 MeV) and dpa/s are compared with the best-estimate results obtained from the least-squares evaluation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the +/-20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron threshold reactions range from 0.95 to 1.12. The overall average M/C ratio for the entire set of Beaver Valley Unit 1 data is 1.01 with an associated standard deviation of 8.2%.

In the comparisons of best-estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.91 to 1.03 for neutron fluence rate (E > 1.0 MeV) and from 0.93 to 1.05 for iron atom displacement rate. The overall average BE/C ratios for neutron fluence rate (E > 1.0 MeV) and for the iron atom displacement rate are 0.98 with a standard deviation of 4.5% and 0.99 with a standard deviation of 4.4%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 2.2 of this report are validated for use in the assessment of the condition of the materials comprising the beltline and extended beltline region of the Beaver Valley Unit 1 reactor pressure vessel.

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Westinghouse Non-Proprietary Class 3 F-10 Table F-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors Reaction of Target Atom 90% Response Product Fission Yield Monitor Material Interest Fraction Range (MeV)(a) Half-life (%)

Copper-63 63 Cu (n,) 0.6917 5.0 - 11.9 5.272 y n/a Iron-54 54 Fe (n,p) 0.0585 2.2 - 8.5 312.11 d n/a Nickel-58 58 Ni (n,p) 0.6808 1.7 - 8.4 70.82 d n/a Uranium-238 238 U (n,f) 1.0000 1.4 - 7.2 30.07 y 6.02 Neptunium-237 237 Np (n,f) 1.0000 0.4 - 4.8 30.07 y 6.17 Cobalt-Aluminum 59 Co (n,) 0.0015 non-threshold 5.272 y n/a Note:

(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Beaver Valley Unit 1 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

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Westinghouse Non-Proprietary Class 3 F-11 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

May-76 250 Oct-78 0 Mar-81 0 Aug-83 0 Jun-76 101578 Nov-78 0 Apr-81 915873 Sep-83 233976 Jul-76 109048 Dec-78 504540 May-81 1453681 Oct-83 1779388 Aug-76 112413 Jan-79 787965 Jun-81 1874754 Nov-83 1695803 Sep-76 431584 Feb-79 1433040 Jul-81 1080807 Dec-83 1893541 Oct-76 608570 Mar-79 336146 Aug-81 1850579 Jan-84 1552319 Nov-76 199669 Apr-79 0 Sep-81 1765793 Feb-84 1811471 Dec-76 410132 May-79 0 Oct-81 1651329 Mar-84 1263132 Jan-77 189115 Jun-79 0 Nov-81 1782396 Apr-84 1815222 Feb-77 0 Jul-79 0 Dec-81 1148933 May-84 1812753 Mar-77 708513 Aug-79 650167 Jan-82 0 Jun-84 1533814 Apr-77 965179 Sep-79 1419642 Feb-82 0 Jul-84 1737076 May-77 1688148 Oct-79 786544 Mar-82 0 Aug-84 1949986 Jun-77 1049724 Nov-79 692354 Apr-82 0 Sep-84 1674388 Jul-77 1489440 Dec-79 0 May-82 0 Oct-84 658805 Aug-77 1116291 Jan-80 0 Jun-82 0 Nov-84 0 Sep-77 0 Feb-80 0 Jul-82 975423 Dec-84 0 Oct-77 40194 Mar-80 0 Aug-82 1597914 Jan-85 1230666 Nov-77 1030814 Apr-80 0 Sep-82 994760 Feb-85 1495792 Dec-77 1828830 May-80 0 Oct-82 1633910 Mar-85 1567714 Jan-78 1570520 Jun-80 0 Nov-82 1868403 Apr-85 1519174 Feb-78 1672227 Jul-80 0 Dec-82 1810831 May-85 1568263 Mar-78 1903683 Aug-80 0 Jan-83 1734339 Jun-85 1888526 Apr-78 1385543 Sep-80 0 Feb-83 1598708 Jul-85 1815511 May-78 0 Oct-80 0 Mar-83 1939771 Aug-85 1799541 Jun-78 141161 Nov-80 216989 Apr-83 1885670 Sep-85 1627814 Jul-78 1621228 Dec-80 916651 May-83 1732947 Oct-85 1491565 Aug-78 0 Jan-81 1118512 Jun-83 585214 Nov-85 1668503 Sep-78 0 Feb-81 878386 Jul-83 0 Dec-85 1951848 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-12 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Jan-86 1949567 Jun-88 1577195 Nov-90 1899189 Apr-93 0 Feb-86 1543257 Jul-88 1941312 Dec-90 1437202 May-93 0 Mar-86 1955574 Aug-88 1816437 Jan-91 962970 Jun-93 522652 Apr-86 1776190 Sep-88 1615227 Feb-91 1698024 Jul-93 1967310 May-86 825172 Oct-88 1541992 Mar-91 1920600 Aug-93 1899044 Jun-86 0 Nov-88 1202942 Apr-91 734753 Sep-93 1903751 Jul-86 0 Dec-88 1242361 May-91 0 Oct-93 735881 Aug-86 170868 Jan-89 1392655 Jun-91 0 Nov-93 779517 Sep-86 1689361 Feb-89 1379326 Jul-91 223412 Dec-93 1935108 Oct-86 1845418 Mar-89 1720567 Aug-91 1900996 Jan-94 1306074 Nov-86 1790031 Apr-89 1457829 Sep-91 1406382 Feb-94 1769531 Dec-86 1955537 May-89 1348415 Oct-91 1348372 Mar-94 1920450 Jan-87 1901768 Jun-89 1542116 Nov-91 60240 Apr-94 1892679 Feb-87 1685908 Jul-89 1946984 Dec-91 1967980 May-94 1064786 Mar-87 1952434 Aug-89 1819779 Jan-92 1968906 Jun-94 39682 Apr-87 1506172 Sep-89 43522 Feb-92 1839523 Jul-94 670158 May-87 20911 Oct-89 0 Mar-92 1966962 Aug-94 1759891 Jun-87 1667777 Nov-89 0 Apr-92 1821532 Sep-94 1902497 Jul-87 1886816 Dec-89 82504 May-92 1878892 Oct-94 1968574 Aug-87 1841589 Jan-90 1717462 Jun-92 1902940 Nov-94 1902781 Sep-87 1752375 Feb-90 1766224 Jul-92 1965639 Dec-94 1724467 Oct-87 1945202 Mar-90 1862208 Aug-92 1604887 Jan-95 72153 Nov-87 1762196 Apr-90 1546013 Sep-92 1629699 Feb-95 0 Dec-87 469765 May-90 1751901 Oct-92 494631 Mar-95 1225380 Jan-88 0 Jun-90 1871935 Nov-92 1620623 Apr-95 1903021 Feb-88 0 Jul-90 1053499 Dec-92 1768591 May-95 1967916 Mar-88 1677626 Aug-90 1779924 Jan-93 1769550 Jun-95 1900816 Apr-88 1884007 Sep-90 1851557 Feb-93 1599541 Jul-95 1961257 May-88 1929940 Oct-90 1671278 Mar-93 1318121 Aug-95 1381737 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-13 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Sep-95 1902415 Feb-98 0 Jul-00 1726698 Dec-02 1997926 Oct-95 1968103 Mar-98 0 Aug-00 1970932 Jan-03 1998176 Nov-95 1854945 Apr-98 0 Sep-00 1907941 Feb-03 1573244 Dec-95 1457253 May-98 0 Oct-00 1974214 Mar-03 242589 Jan-96 1965199 Jun-98 0 Nov-00 1900553 Apr-03 22364 Feb-96 1810977 Jul-98 0 Dec-00 1970715 May-03 1827829 Mar-96 1164783 Aug-98 973181 Jan-01 1950143 Jun-03 1867655 Apr-96 0 Sep-98 1909045 Feb-01 1705263 Jul-03 1998416 May-96 1103853 Oct-98 1974419 Mar-01 1971346 Aug-03 1997525 Jun-96 1767857 Nov-98 1906308 Apr-01 1287692 Sep-03 1934204 Jul-96 1965530 Dec-98 1968795 May-01 1970566 Oct-03 2001020 Aug-96 847659 Jan-99 1709974 Jun-01 1684701 Nov-03 1817339 Sep-96 1901466 Feb-99 1019933 Jul-01 1958104 Dec-03 1998755 Oct-96 1965965 Mar-99 1940265 Aug-01 1929939 Jan-04 1985616 Nov-96 1898890 Apr-99 716664 Sep-01 0 Feb-04 1869377 Dec-96 1959481 May-99 1410478 Oct-01 1328004 Mar-04 1899108 Jan-97 1957638 Jun-99 1906371 Nov-01 1718236 Apr-04 1920338 Feb-97 1744174 Jul-99 1967597 Dec-01 1823296 May-04 1998373 Mar-97 1153954 Aug-99 1958611 Jan-02 1990211 Jun-04 1931522 Apr-97 1023580 Sep-99 1494468 Feb-02 1801795 Jul-04 1998792 May-97 1963402 Oct-99 1951486 Mar-02 1989618 Aug-04 1998284 Jun-97 1705565 Nov-99 1823131 Apr-02 1837779 Sep-04 1933040 Jul-97 11589 Dec-99 1970481 May-02 1998037 Oct-04 1055791 Aug-97 1752251 Jan-00 1968186 Jun-02 1932939 Nov-04 855204 Sep-97 1683909 Feb-00 746198 Jul-02 1997832 Dec-04 1999043 Oct-97 0 Mar-00 0 Aug-02 1997954 Jan-05 1994898 Nov-97 0 Apr-00 1013809 Sep-02 1933446 Feb-05 1805552 Dec-97 0 May-00 1881060 Oct-02 2001352 Mar-05 1998423 Jan-98 468225 Jun-00 1907972 Nov-02 1157615 Apr-05 1931487 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-14 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

May-05 1911916 Oct-07 418056 Mar-10 2151255 Aug-12 2154521 Jun-05 1933922 Nov-07 2085894 Apr-10 1924250 Sep-12 2084237 Jul-05 1998525 Dec-07 2155422 May-10 2153565 Oct-12 2153783 Aug-05 1998289 Jan-08 2154241 Jun-10 2084317 Nov-12 2087475 Sep-05 1932792 Feb-08 1925350 Jul-10 2152557 Dec-12 2154578 Oct-05 2000893 Mar-08 2106637 Aug-10 2153379 Jan-13 2154191 Nov-05 1933689 Apr-08 2084532 Sep-10 2083825 Feb-13 1758987 Dec-05 1998375 May-08 2154156 Oct-10 36743 Mar-13 2151090 Jan-06 1998236 Jun-08 2083935 Nov-10 1800831 Apr-13 2083989 Feb-06 734160 Jul-08 2150483 Dec-10 2127093 May-13 2154100 Mar-06 0 Aug-08 2154107 Jan-11 2154831 Jun-13 2084869 Apr-06 643633 Sep-08 2084264 Feb-11 1944444 Jul-13 2154058 May-06 1833886 Oct-08 2154327 Mar-11 2151515 Aug-13 2153987 Jun-06 1934547 Nov-08 2087649 Apr-11 1752956 Sep-13 1935163 Jul-06 1998137 Dec-08 2154107 May-11 2154209 Oct-13 0 Aug-06 1666197 Jan-09 2154018 Jun-11 2084557 Nov-13 1397606 Sep-06 1864587 Feb-09 1945788 Jul-11 2147118 Dec-13 2155010 Oct-06 2062087 Mar-09 2150315 Aug-11 2154308 Jan-14 528851 Nov-06 1991907 Apr-09 1236557 Sep-11 2084682 Feb-14 1891737 Dec-06 2058622 May-09 661421 Oct-11 2152575 Mar-14 2151295 Jan-07 2058939 Jun-09 2085427 Nov-11 2087006 Apr-14 2084711 Feb-07 1828585 Jul-09 2155082 Dec-11 2154406 May-14 2153565 Mar-07 2022146 Aug-09 2139233 Jan-12 2154094 Jun-14 2084535 Apr-07 2085903 Sep-09 2084445 Feb-12 2015414 Jul-14 2153613 May-07 2139605 Oct-09 2154064 Mar-12 2134478 Aug-14 2154036 Jun-07 2085743 Nov-09 2087074 Apr-12 499226 Sep-14 2084732 Jul-07 2155428 Dec-09 2153726 May-12 1389815 Oct-14 1848055 Aug-07 2152836 Jan-10 2153807 Jun-12 2085729 Nov-14 2087509 Sep-07 1514542 Feb-10 1944710 Jul-12 2151237 Dec-14 2153981 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-15 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Jan-15 2154143 Feb-15 1945558 Mar-15 2151145 Apr-15 1245394 May-15 473968 Jun-15 1922472 Jul-15 2155674 Aug-15 2155704 Sep-15 2085592 Oct-15 2154801 Nov-15 2088157 Dec-15 2155053 Jan-16 2154801 Feb-16 2015642 Mar-16 2151916 Apr-16 2084943 May-16 2154636 Jun-16 2084938 Jul-16 2153963 Aug-16 2149558 Sep-16 1365500 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-16 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Cj Factors at the Surveillance Capsule Center, Core Midplane Elevation Cycle (E > 1.0 MeV) [n/cm2-s]

Length Fuel Cycle [EFPS] Capsule V Capsule U Capsule W Capsule Y Capsule X 1 3.66E+07 8.10E+10 5.43E+10 5.43E+10 5.43E+10 8.10E+10 2 2.26E+07 5.62E+10 5.62E+10 5.62E+10 8.27E+10 3 2.49E+07 6.21E+10 6.21E+10 6.21E+10 9.38E+10 4 2.91E+07 4.73E+10 4.73E+10 4.73E+10 7.14E+10 5 3.76E+07 4.56E+10 4.56E+10 6.94E+10 6 3.51E+07 4.63E+10 4.63E+10 6.15E+10 7 3.95E+07 4.47E+10 6.80E+10 8 3.48E+07 4.70E+10 6.98E+10 9 4.35E+07 4.60E+10 6.46E+10 10 3.77E+07 3.86E+10 4.86E+10 11 3.05E+07 4.06E+10 4.78E+10 12 3.58E+07 4.45E+10 5.18E+10 13 4.31E+07 4.17E+10 5.23E+10 14 4.16E+07 4.54E+10 15 4.19E+07 4.30E+10 16 4.56E+07 5.25E+10 17 3.89E+07 5.02E+10 18 4.39E+07 5.69E+10 19 4.65E+07 5.42E+10 20 4.27E+07 5.82E+10 21 4.44E+07 5.82E+10 22 4.33E+07 5.55E+10 Time Weighted Average Fluence Rate 8.10E+10 5.46E+10 5.12E+10 4.66E+10 5.94E+10 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-17 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Cj Factors at the Surveillance Capsule Center, Core Midplane Elevation Cycle Cj Length Fuel Cycle [EFPS] Capsule V Capsule U Capsule W Capsule Y Capsule X 1 3.66E+07 1.000 0.994 1.060 1.165 1.364 2 2.26E+07 1.029 1.097 1.206 1.393 3 2.49E+07 1.137 1.212 1.332 1.579 4 2.91E+07 0.867 0.924 1.016 1.202 5 3.76E+07 0.891 0.979 1.168 6 3.51E+07 0.904 0.993 1.035 7 3.95E+07 0.960 1.145 8 3.48E+07 1.008 1.175 9 4.35E+07 0.987 1.088 10 3.77E+07 0.828 0.818 11 3.05E+07 0.871 0.805 12 3.58E+07 0.955 0.872 13 4.31E+07 0.895 0.881 14 4.16E+07 0.765 15 4.19E+07 0.724 16 4.56E+07 0.883 17 3.89E+07 0.845 18 4.39E+07 0.959 19 4.65E+07 0.913 20 4.27E+07 0.980 21 4.44E+07 0.981 22 4.33E+07 0.935 Average 1.00 1.00 1.00 1.00 1.00 WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-18 Table F-4a Measured Sensor Activities and Reaction Rates of Surveillance Capsule V Adjusted Reaction Measured Activity(a) Saturated Activity Rate(b)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,) 60Co Top-Mid 4.24E+04 3.68E+05 5.62E-17 Middle 4.42E+04 3.84E+05 5.86E-17 Bot-Mid 4.28E+04 3.72E+05 5.67E-17 Average 5.72E-17 54 Fe (n,p) 54Mn Top 5.84E+05 3.94E+06 6.25E-15 Top-Mid 5.35E+05 3.61E+06 5.73E-15 Middle 5.62E+05 3.79E+06 6.02E-15 Bot-Mid 5.32E+05 3.59E+06 5.70E-15 Bottom 5.37E+05 3.62E+06 5.75E-15 Average 5.89E-15 58 Ni (n,p) 58Co Top-Mid 9.57E+05 5.64E+07 8.07E-15 Middle 9.75E+05 5.74E+07 8.22E-15 Bot-Mid 9.06E+05 5.34E+07 7.64E-15 Average 7.98E-15 238 U (n,f) 137Cs (Cd) Middle 1.36E+05 5.39E+06 3.54E-14 Including 235U, 239Pu, and fission corrections: 2.95E-14 237 Np (n,f) 137Cs (Cd) Middle 9.70E+05 3.84E+07 2.45E-13 Including fission corrections: 2.41E-13 59 Co (n,) 60Co Top 7.24E+06 6.30E+07 4.11E-12 Bottom 7.24E+06 6.30E+07 4.11E-12 Average 4.11E-12 59 Co (n,) 60Co (Cd) Top 2.59E+06 2.72E+07 1.78E-12 Bottom 2.55E+06 2.68E+07 1.75E-12 Average 1.76E-12 Notes:

(a) Measured specific activities are indexed to a counting date of September 16, 1980.

(b) Reaction rates are referenced to the Cycle 1 rated reactor power of 2652 MWt.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-19 Table F-4b Measured Sensor Activities and Reaction Rates of Surveillance Capsule U Adjusted Reaction Measured Activity(a) Saturated Activity Rate(b)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,) 60Co Top-Mid 9.44E+04 3.01E+05 4.60E-17 Middle 1.01E+05 3.22E+05 4.92E-17 Bot-Mid 9.30E+04 2.97E+05 4.53E-17 Average 4.68E-17 54 Fe (n,p) 54Mn Top 1.21E+06 2.82E+06 4.47E-15 Top-Mid 1.13E+06 2.63E+06 4.17E-15 Middle 1.22E+06 2.84E+06 4.50E-15 Bot-Mid 1.16E+06 2.70E+06 4.28E-15 Bottom 1.14E+06 2.65E+06 4.21E-15 Average 4.32E-15 58 Ni (n,p) 58Co Top-Mid 4.81E+06 3.98E+07 5.70E-15 Middle 5.22E+06 4.32E+07 6.19E-15 Bot-Mid 4.92E+06 4.08E+07 5.84E-15 Average 5.91E-15 238 U (n,f) 137Cs (Cd) Middle 2.89E+05 3.81E+06 2.50E-14 Including 235U, 239Pu, and fission corrections: 2.06E-14 237 Np (n,f) 137Cs (Cd) Middle 2.14E+06 2.82E+07 1.80E-13 Including fission corrections: 1.77E-13 59 Co (n,) 60Co Top 1.17E+07 3.74E+07 2.44E-12 Bottom 1.16E+07 3.71E+07 2.42E-12 Average 2.43E-12 59 Co (n,) 60Co (Cd) Top 4.27E+06 1.65E+07 1.08E-12 Bottom 4.18E+06 1.61E+07 1.05E-12 Average 1.06E-12 Notes:

(a) Measured specific activities are indexed to a counting date of April 3, 1985.

(b) Reaction rates are referenced to the Cycles 1-4 average rated reactor power of 2652 MWt.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-20 Table F-4c Measured Sensor Activities and Reaction Rates of Surveillance Capsule W Adjusted Reaction Measured Activity(a) Saturated Activity Rate(b)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,) 60Co Top-Mid 1.14E+05 2.65E+05 4.03E-17 Middle 1.20E+05 2.78E+05 4.25E-17 Bot-Mid 1.18E+05 2.74E+05 4.18E-17 Average 4.15E-17 54 Fe (n,p) 54Mn Top 1.00E+06 2.55E+06 4.05E-15 Top-Mid 9.20E+05 2.35E+06 3.73E-15 Top-Mid 8.80E+05 2.25E+06 3.56E-15 Middle 9.44E+05 2.41E+06 3.82E-15 Bot-Mid 8.85E+05 2.26E+06 3.58E-15 Bot-Mid 8.92E+05 2.28E+06 3.61E-15 Bottom 9.19E+05 2.35E+06 3.72E-15 Average 3.73E-15 58 Ni (n,p) 58Co Top-Mid 2.17E+06 3.47E+07 4.96E-15 Bot-Mid 2.20E+06 3.51E+07 5.03E-15 Average 5.00E-15 237 Np (n,f) 137Cs (Cd) Middle 2.57E+06 2.14E+07 1.37E-13 Including fission corrections: 1.34E-13 59 Co (n,) 60Co Top 1.36E+07 3.16E+07 2.06E-12 Bottom 1.40E+07 3.25E+07 2.12E-12 Average 2.09E-12 59 Co (n,) 60Co (Cd) Top 4.93E+06 1.38E+07 9.03E-13 Bottom 4.99E+06 1.40E+07 9.14E-13 Average 9.08E-13 Notes:

(a) Measured specific activities are indexed to a counting date of August 10, 1988.

(b) Reaction rates are referenced to the Cycles 1-6 average rated reactor power of 2652 MWt.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-21 Table F-4d Measured Sensor Activities and Reaction Rates of Surveillance Capsule Y Adjusted Reaction Measured Activity(a) Saturated Activity Rate(b)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,) 60Co Top-Mid 1.56E+05 2.50E+05 3.81E-17 Middle 1.67E+05 2.67E+05 4.08E-17 Bot-Mid 1.56E+05 2.50E+05 3.81E-17 Average 3.90E-17 54 Fe (n,p) 54Mn Top 1.42E+06 2.40E+06 3.81E-15 Top-Mid 1.35E+06 2.29E+06 3.63E-15 Middle 1.43E+06 2.42E+06 3.84E-15 Bot-Mid 1.37E+06 2.32E+06 3.68E-15 Bottom 1.33E+06 2.25E+06 3.57E-15 Average 3.71E-15 58 Ni (n,p) 58Co Top-Mid 1.71E+07 3.47E+07 4.97E-15 Middle 1.84E+07 3.74E+07 5.35E-15 Bot-Mid 1.73E+07 3.51E+07 5.03E-15 Average 5.11E-15 238 U (n,f) 137Cs (Cd) Middle 7.79E+05 3.03E+06 1.99E-14 Including 235U, 239Pu, and fission corrections: 1.54E-14 237 Np (n,f) 137Cs (Cd) Middle 5.39E+06 2.10E+07 1.34E-13 Including fission corrections: 1.31E-13 59 Co (n,) 60Co Top 1.81E+07 2.90E+07 1.89E-12 Bottom 1.61E+07 2.58E+07 1.68E-12 Average 1.79E-12 59 Co (n,) 60Co (Cd) Top 6.78E+06 1.31E+07 8.57E-13 Bottom 6.79E+06 1.32E+07 8.58E-13 Average 8.58E-13 Notes:

(a) Measured specific activities are indexed to a counting date of March 24, 2000.

(b) Reaction rates are referenced to the Cycles 1-13 average rated reactor power of 2652 MWt.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-22 Table F-4e Measured Sensor Activities and Reaction Rates of Surveillance Capsule X Adjusted Reaction Measured Activity(a) Saturated Activity Rate(b)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,) 60Co Top-Mid 2.18E+05 2.75E+05 4.19E-17 Middle 2.31E+05 2.91E+05 4.44E-17 Bot-Mid 2.20E+05 2.77E+05 4.23E-17 Average 4.28E-17 54 Fe (n,p) 54Mn Top 1.53E+06 2.67E+06 4.24E-15 Top-Mid 1.46E+06 2.55E+06 4.04E-15 Middle 1.59E+06 2.77E+06 4.40E-15 Bot-Mid 1.55E+06 2.70E+06 4.29E-15 Bottom 1.64E+06 2.86E+06 4.54E-15 Average 4.30E-15 58 Ni (n,p) 58Co Top-Mid 5.73E+06 4.36E+07 6.25E-15 Middle 6.10E+06 4.65E+07 6.65E-15 Bot-Mid 5.87E+06 4.47E+07 6.40E-15 Average 6.43E-15 238 U (n,f) 137Cs (Cd) Middle 1.87E+06 4.47E+06 2.94E-14 Including 235U, 239Pu, and fission corrections: 2.00E-14 237 Np (n,f) 137Cs (Cd) Middle 1.09E+07 2.61E+07 1.66E-13 Including fission corrections: 1.63E-13 59 Co (n,) 60Co Top 2.82E+07 3.56E+07 2.32E-12 Bottom 3.01E+07 3.80E+07 2.48E-12 Average 2.40E-12 59 Co (n,) 60Co (Cd) Top 1.84E+07 2.81E+07 1.83E-12 Bottom 1.24E+07 1.89E+07 1.23E-12 Average 1.53E-12 Notes:

(a) Measured specific activities are indexed to a counting date of April 1, 2014.

(b) Reaction rates are referenced to the Cycles 1-22 average rated reactor power of 2718 MWt.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-23 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule V Reaction Rate [rps/atom]

Reaction Measured Calculated Best-Estimate M/C M/BE 63 Cu(n,)60Co 5.72E-17 5.50E-17 5.60E-17 1.04 1.02 54 Fe(n,p)54Mn 5.89E-15 6.02E-15 5.98E-15 0.98 0.99 58 Ni(n,p) Co 58 7.98E-15 8.24E-15 8.15E-15 0.97 0.98 238 U(n,f)137Cs (Cd) 2.95E-14 2.87E-14 2.88E-14 1.03 1.02 237 Np(n,f)137Cs (Cd) 2.41E-13 2.14E-13 2.28E-13 1.13 1.05 59 Co(n,)60Co 4.11E-12 4.47E-12 4.12E-12 0.92 1.00 59 Co(n,)60Co (Cd) 1.76E-12 1.72E-12 1.76E-12 1.03 1.00 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.

Capsule U Reaction Rate [rps/atom]

Reaction Measured Calculated Best-Estimate M/C M/BE 63 Cu(n,)60Co 4.68E-17 4.30E-17 4.53E-17 1.09 1.03 54 Fe(n,p)54Mn 4.32E-15 4.41E-15 4.46E-15 0.98 0.97 58 Ni(n,p)58Co 5.91E-15 5.99E-15 6.05E-15 0.99 0.98 238 U(n,f)137Cs (Cd) 2.06E-14 1.99E-14 2.05E-14 1.04 1.01 237 Np(n,f)137Cs (Cd) 1.77E-13 1.40E-13 1.61E-13 1.27 1.10 59 Co(n,)60Co 2.43E-12 2.59E-12 2.44E-12 0.94 1.00 59 Co(n,)60Co (Cd) 1.06E-12 1.01E-12 1.06E-12 1.06 1.00 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-24 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule W Reaction Rate [rps/atom]

Reaction Measured Calculated Best-Estimate M/C M/BE 63 Cu(n,)60Co 4.15E-17 4.09E-17 4.00E-17 1.02 1.04 54 Fe(n,p)54Mn 3.73E-15 4.16E-15 3.81E-15 0.90 0.98 58 Ni(n,p)58Co 5.00E-15 5.64E-15 5.14E-15 0.89 0.97 237 Np(n,f)137Cs (Cd) 1.34E-13 1.31E-13 1.27E-13 1.03 1.05 59 Co(n,)60Co 2.09E-12 2.42E-12 2.10E-12 0.86 1.00 59 Co(n,)60Co (Cd) 9.08E-13 9.41E-13 9.07E-13 0.97 1.00 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.

Capsule Y Reaction Rate [rps/atom]

Reaction Measured Calculated Best-Estimate M/C M/BE 63 Cu(n,)60Co 3.90E-17 3.78E-17 3.84E-17 1.03 1.02 54 Fe(n,p)54Mn 3.71E-15 3.81E-15 3.75E-15 0.97 0.99 58 Ni(n,p)58Co 5.11E-15 5.17E-15 5.10E-15 0.99 1.00 238 U(n,f)137Cs (Cd) 1.54E-14 1.71E-14 1.67E-14 0.90 0.92 237 Np(n,f)137Cs (Cd) 1.31E-13 1.19E-13 1.24E-13 1.11 1.06 59 Co(n,)60Co 1.79E-12 2.19E-12 1.80E-12 0.82 0.99 59 Co(n,)60Co (Cd) 8.58E-13 8.51E-13 8.54E-13 1.01 1.00 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-25 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule X Reaction Rate [rps/atom]

Reaction Measured Calculated Best-Estimate M/C M/BE 63 Cu(n,)60Co 4.28E-17 4.34E-17 4.28E-17 0.99 1.00 54 Fe(n,p)54Mn 4.30E-15 4.56E-15 4.45E-15 0.94 0.97 58 Ni(n,p)58Co 6.43E-15 6.22E-15 6.21E-15 1.03 1.03 238 U(n,f)137Cs (Cd) 2.00E-14 2.13E-14 2.10E-14 0.94 0.95 237 Np(n,f)137Cs (Cd) 1.63E-13 1.55E-13 1.59E-13 1.05 1.03 59 Co(n,)60Co 2.40E-12 3.18E-12 2.45E-12 0.76 0.98 59 Co(n,)60Co (Cd) 1.53E-12 1.22E-12 1.50E-12 1.26 1.02 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-26 Table F-6 Comparison of Calculated and Best-Estimate Exposure Rates at the Surveillance Capsule Center (E > 1.0 MeV) [n/cm2-s]

Calculated Best-Estimate Best-Estimate BE/C Capsule ID Uncertainty (1)

V 8.12E+10 8.19E+10 6 1.00 U 5.47E+10 5.68E+10 6 1.03 W 5.13E+10 4.69E+10 6 0.91 Y 4.67E+10 4.58E+10 6 0.98 X 5.95E+10 5.89E+10 6 0.99 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) exposure rates.

Iron Atom Displacement Rate [dpa/s]

Calculated Best-Estimate Best-Estimate BE/C Capsule ID Uncertainty (1)

V 1.33E-10 1.35E-10 7 1.01 U 8.70E-11 9.17E-11 7 1.05 W 8.16E-11 7.60E-11 7 0.93 Y 7.42E-11 7.33E-11 6 0.98 X 9.67E-11 9.60E-11 7 0.99 Note:

See Section F.1.2 for details describing the Best-Estimate (BE) exposure rates.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 F-27 Table F-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Rations Capsule Capsule Capsule Capsule Capsule Average % Std Dev Reaction V U W Y X 63 Cu(n,)60Co 1.04 1.09 1.02 1.03 0.99 1.03 3.5 54 Fe(n,p)54Mn 0.98 0.98 0.90 0.97 0.94 0.95 3.6 58 Ni(n,p)58Co 0.97 0.99 0.89 0.99 1.03 0.97 5.3 238 U(n,f)137Cs (Cd) 1.03 1.04 0.90 0.94 0.98 7.0 237 Np(n,f)137Cs (Cd) 1.13 1.27 1.03 1.11 1.05 1.12 8.4 Note:

The overall average M/C ratio for the set of 24 sensor measurements is 1.01 with an associated standard deviation of 8.2%.

Table F-8 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID (E > 1.0 MeV) dpa/s V 1.00 1.01 U 1.03 1.05 W 0.91 0.93 Y 0.98 0.98 X 0.99 0.99 Average 0.98 0.99

% Standard Deviation 4.5 4.4 WCAP-18102-NP March 2021 Revision 2

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

F-2 WCAP-9860, Revision 0, Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit No. 1 Reactor Vessel Radiation Surveillance Program, January 1981.

F-3 WCAP-10867, Revision 0, Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 1985.

F-4 WCAP-12005, Revision 0, Analysis of Capsule W from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, November 1988.

F-5 WCAP-15571 Supplement 1, Revision 2, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2011.

F-6 WCAP-17896-NP, Revision 0, Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2014.

F-7 F. Schmittroth, FERRET Data Analysis Code, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

F-8 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross Section Compendium, July 1994.

F-9 ASTM Standard E 944-13, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA), 2013.

WCAP-18102-NP March 2021 Revision 2

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Westinghouse Non-Proprietary Class 3 G-1 APPENDIX G SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE The following surveillance capsule removal schedule (Table G-1) meets the requirements of ASTM E185-82 [Ref. G-1] as required by 10 CFR 50, Appendix H [Ref. G-2]. Note that it is recommended for future capsule(s) to be removed from the Beaver Valley Unit 1 reactor vessel.

Table G-1 Surveillance Capsule Withdrawal Schedule Capsule Capsule Lead Withdrawal Capsule Fluence Capsule Status(a)

Location Factor(a) EFPY(b,c) (n/cm2, E > 1.0 MeV)(c)

Withdrawn V 165° 1.47 1.2 2.97E+18 (EOC 1)

Withdrawn U 65° 1.00 3.6 6.18E+18 (EOC 4)

Withdrawn W 245° 1.05 5.9 9.52E+18 (EOC 6)

Withdrawn Y 295° 1.14 14.3 2.10E+19 (EOC 13)

Withdrawn X 285° 1.57 26.6 4.99E+19 (EOC 22) 285° S(d) In Reactor 0.74(d) Note (d) 2.58E+19(d)

(45°/295°)

T(e) 65° (55°) In Reactor 0.94(e) Note (e) 3.28E+19(e)

Z(f) 165° (305°) In Reactor 1.20(f) Note (f) 4.18E+19(f)

Notes:

(a) Updated in Section 2; see Table 2-12.

(b) EFPY from plant startup.

(c) Updated in Section 2; see Table 2-11.

(d) Capsule S was moved to the Capsule Y location at the End of Cycle (EOC) 19, and then moved to the Capsule X location at the EOC 22. Reported fluence value and lead factor are accumulated through EOC 24. Capsule S should remain in the reactor. If additional metallurgical data is needed for Beaver Valley Unit 1, such as in support of a second license renewal to 80 total years of operation, withdrawal and testing of Capsule S should be considered.

(e) Capsule T was moved to the Capsule U location at the EOC 10. Reported fluence value and lead factor are accumulated through EOC 24. Capsule T should remain in the reactor and continue to accrue irradiation for potential future testing, if needed.

(f) Capsule Z was moved to the original Capsule V location at the EOC 10. Reported fluence value and lead factor are accumulated through EOC 24. Based on the current information, Capsule Z should be withdrawn after 39 EFPY, which corresponds to the peak vessel fluence at EOLE (50 EFPY), 5.89 x 1019 n/cm2 (E > 1.0 MeV).

G.1 REFERENCES G-1 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, ASTM, July 1982.

G-2 Code of Federal Regulations 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, Federal Register, Volume 60, No. 243, dated December 19, 1995.

WCAP-18102-NP March 2021 Revision 2

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Enclosure B L-21-106 WCAP-18559-NP, Revision 1, Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations Based on Capsule Y Data, March 2021 (92 pages follow)

Westinghouse Non-Proprietary Class 3 WCAP-18559-N P March 2021 Revision 1 Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations Based on Capsule Y Data

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-18559-NP Revision 1 Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations Based on Capsule Y Data Donald M. McNutt III*

Reactor Vessel/Containment Vessel Design and Analysis Jianwei Chen*

Radiation Engineering and Analysis March 2021 Reviewers: D. Brett Lynch* Approved: Lynn A. Patterson*, Manager RV/CV Design and Analysis RV/CV Design and Analysis Eugene T. Hayes* Jesse J Klingensmith*, Manager Radiation Engineering and Analysis Radiation Engineering and Analysis

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

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Westinghouse Non-Proprietary Class 3 ii RECORD OF REVISION Revision 0: Original Issue Revision 1: The purpose of this revision is to update the last paragraph of Appendix E to reflect that the LTOP enable temperature provided corresponds to a coolant temperature.

WCAP-18559-NP March 2021 Revision 1

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

SUMMARY

......................................................................................................................... vii 1 INTRODUCTION ........................................................................................................... 1-1 2 RADIATION ANALYSIS AND NEUTRON DOSIMETRY .......................................... 2-1

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 2-1 2.3 NEUTRON DOSIMETRY .............................................................................................. 2-4 2.4 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-4 3 FRACTURE TOUGHNESS PROPERTIES .................................................................... 3-1 4 SURVEILLANCE DATA ................................................................................................ 4-1 5 CHEMISTRY FACTORS ................................................................................................ 5-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ... 6-1 6.1 OVERALL APPROACH ................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ........................................... 6-5 6.4 BOLTUP TEMPERATURE REQUIREMENTS ............................................................. 6-5 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ............................. 7-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.......... 8-1 9 REFERENCES ................................................................................................................ 9-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)...................................................... A-1 APPENDIX B OTHER RCPB FERRITIC COMPONENTS ................................................................. B-1 APPENDIX C UPPER-SHELF ENERGY EVALUATION ................................................................... C-1 APPENDIX D PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION ................................................................................................. D-1 APPENDIX E LTOP ENABLE TEMPERATURE ................................................................................. E-1 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 2-1 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Locations ............................................................................................................ 2-7 Table 2-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations ............................................................................................................ 2-9 Table 2-3 Calculated Surveillance Capsule Lead Factors .............................................................. 2-11 Table 2-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface.............................................................................................. 2-12 Table 2-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface.............................................................................................. 2-13 Table 2-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface ............................................................................................................... 2-14 Table 2-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface ......................................................................................................................... 2-15 Table 2-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Welds and Shells.............................................................................................................................. 2-16 Table 2-9 Calculated Maximum Iron Atom Displacements at Pressure Vessel Welds and Shells . 2-17 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Beaver Valley Unit 2 Reactor Vessel Materials ............................................................... 3-2 Table 3-2 Summary of Beaver Valley Unit 2 Reactor Vessel Closure Head Flange and Vessel Flange Initial RTNDT Values ......................................................................................................... 3-4 Table 4-1 Beaver Valley Unit 2 Surveillance Capsule Data............................................................. 4-2 Table 5-1 Calculation of Beaver Valley Unit 2 Chemistry Factors Using Surveillance Capsule Data

......................................................................................................................................... 5-2 Table 5-2 Summary of Beaver Valley Unit 2 Positions 1.1 and 2.1 Chemistry Factors .................. 5-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Beaver Valley Unit 2 Reactor Vessel Materials at 54 EFPY ...................................... 7-3 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 1/4T Location .......... 7-4 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 3/4T Location .......... 7-7 Table 8-1 Summary of the Limiting ART Values Used in the Generation of the Beaver Valley Unit 2 Heatup and Cooldown Curves at 54 EFPY................................................................... 8-2 Table 8-2 Beaver Valley Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) .............................................................................................. 8-6 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 v Table 8-3 Beaver Valley Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) .............................................................................................. 8-8 Table 8-4 Beaver Valley Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/ Margins for Instrumentation Errors) ............................................................................................ 8-12 Table 8-5 Beaver Valley Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/ Margins for Instrumentation Errors) ............................................................................................ 8-14 Table A-1 KIt Values for Beaver Valley Unit 2 at 54 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors) ................................................. A-2 Table A-2 KIt Values for Beaver Valley Unit 2 at 54 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors) ................................................. A-3 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for Beaver Valley Unit 2

........................................................................................................................................ C-2 Table D-1 RTPTS Calculations for the Beaver Valley Unit 2 Reactor Vessel Materials at 54 EFPY D-2 Table D-2 Evaluation of Beaver Valley Unit 2 ERG Limit Category .............................................. D-6 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 vi LIST OF FIGURES Figure 2-1 Arrangement of Surveillance Capsules in the Beaver Valley Unit 2 Reactor Vessel ..... 2-18 Figure 2-2 Beaver Valley Unit 2 Plan View of the Reactor Geometry at the Core Midplane 15.0° Neutron Pad Configuration ............................................................................................ 2-19 Figure 2-3 Beaver Valley Unit 2 Plan View of the Reactor Geometry at the Core Midplane 26° Neutron Pad Configuration ............................................................................................ 2-20 Figure 2-4 Beaver Valley Unit 2 Section View of the Reactor Geometry without Surveillance Capsule .......................................................................................................................... 2-21 Figure 2-5 Beaver Valley Unit 2 Section View of the Reactor Geometry with Surveillance Capsule ...

....................................................................................................................................... 2-22 Figure 8-1 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc) ...................................................................................................... 8-4 Figure 8-2 Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc) .......................................................................................... 8-5 Figure 8-3 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and with Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc) ........................................................................................................................... 8-10 Figure 8-4 Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and with Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc) .................................................................................................... 8-11 Figure C-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence ..................................................................................... C-4 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 vii EXECUTIVE

SUMMARY

This report presents the evaluation of the Beaver Valley Unit 2 reactor pressure vessel (RPV) with respect to reactor vessel integrity (RVI), particularly with consideration of the Beaver Valley Unit 2 Capsule Y testing results and RAPTOR-M3G fluence analysis as documented in WCAP-18558-NP. Beaver Valley Unit 2 has been approved for license extension to a total of 60 years of operation. The purpose of this report is to document RVI evaluations applicable through 60 years of operation. The evaluations in this report projected for 60 years of operation are applicable through 54 effective full-power years (EFPY), which is deemed end-of-license extension (EOLE).

A summary of results for the Beaver Valley Unit 2 RVI evaluation is provided below. Based on the results presented herein, it is concluded that the Beaver Valley Unit 2 RPV will continue to meet RPV integrity regulatory requirements through the extended period of operation.

Although, RVI requirements continue to be met, the incorporation of the Beaver Valley Unit 2 Capsule Y testing results and RAPTOR-M3G fluence analysis did reduce the applicability term of the EOLE pressure-temperature (P-T) limit curves generated in WCAP-16528-NP. In order to address this, heatup and cooldown P-T limit curves were generated using the revised limiting Adjusted Reference Temperature (ART) values for Beaver Valley Unit 2 through 54 EFPY. The limiting ART values were those of Intermediate Shell Plate B9004-2 at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations.

The P-T limit curves were generated for 54 EFPY using the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G. The P-T limit curve generation methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4. Heatup rates of 60 and 100°F/hr, and cooldown rates of 0 (steady-state), 20, 40, 60, and 100°F/hr were used to generate the P-T limit curves, with the flange requirements and both with and without margins for instrumentation errors. The Beaver Valley Unit 2 End of License Extension (EOLE) corresponding to 60 years of operation is 54 EFPY. The EOLE P-T limit curves without margins for instrumentation errors can be found in Figures 8-1 and 8-2, and the EOLE P-T limit curves with margins for instrumentation errors can be found in Figures 8-3 and 8-4.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 54 EFPY.

Appendix B contains discussion of the other ferritic Reactor Coolant Pressure Boundary (RCPB) components relative to P-T limits. As discussed in Appendix B, all of the other ferritic RCPB components meet the applicable requirements of Section III of the ASME Code.

Appendix C contains an upper-shelf energy (USE) evaluation for all Beaver Valley Unit 2 reactor vessel beltline and extended beltline materials. Per Appendix C, all beltline and extended beltline materials are projected to maintain USE values above the 50 ft-lb screening criterion per 10 CFR 50 Appendix G at 54 EFPY.

Appendix D contains a pressurized thermal shock (PTS) evaluation for all Beaver Valley Unit 2 reactor vessel beltline and extended beltline materials. Per Appendix D, all beltline and extended beltline materials WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 viii have projected RTPTS values below the screening criteria set forth in 10 CFR 50.61. Additionally, Beaver Valley Unit 2 will remain in Category I of the Emergency Response Guidelines through 54 EFPY.

Appendix E contains the evaluation of the low temperature overpressure protection (LTOP) enable temperature according to ASME Code Case N-641.

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION The purpose of this report is to evaluate the Beaver Valley Unit 2 beltline and extended beltline materials with respect to reactor vessel integrity (RVI) with consideration of the testing results of the reactor vessel surveillance program Capsule Y, documented in WCAP-18558-NP [13]. These materials are evaluated to determine their reference temperature for pressurized thermal shock (RTPTS), upper-shelf energy (USE),

and adjusted reference temperature (ART) values at end of license extension (EOLE), which corresponds to 54 effective full-power years (EFPY). The applicability period of the 54 EFPY pressure-temperature (P-T) limit curves calculated in WCAP-16528-NP [12] is also analyzed to determine the applicability period of the curves as a result of Beaver Valley Unit 2 Capsule Y data and updated fluence projections.

Reference nil-ductility transition temperature (RTNDT) increases and the USE decreases as the material is exposed to fast-neutron irradiation. To find the most limiting RTNDT and USE at any time period in the reactor's life, Regulatory Guide 1.99 (RG 1.99), Revision 2 [1] is used to calculate the RTNDT and USE percent decease due to the associated radiation exposure. The resulting RTNDT values are used to adjust the unirradiated RTNDT (RTNDT(U)) in order to satisfy the requirements of 10 CFR Part 50.61 [10], the PTS Rule, and to verify/calculate P-T limit curves in accordance with the requirements of 10 CFR Part 50, Appendix G [4]. (Note, the methodology to calculate RTPTS is stipulated in 10 CFR Part 50.61; however, it is identical to RG 1.99.) The resulting limiting USE values are used to satisfy the requirements of 10 CFR 50, Appendix G.

Due to the resulting reduction in the P-T limit curves applicability term, this report also presents the calculations and the development of the new Beaver Valley Unit 2 heatup and cooldown P-T limit curves for 54 EFPY. The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (plus an additional margin to account for future perturbations such as an uprate or surveillance capsule results) and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [2]. Specifically, the KIc methodology of the 1998 through the 2000 Addenda Edition of ASME Code,Section XI, Appendix G [3] was used. The KIc curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the KIc curve so that allowable stress intensity factors can be obtained for the material as a function of temperature. Allowable operating limits are then determined using the allowable stress intensity factors.

The P-T limit curves herein were generated both with and without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [4] have been incorporated in the P-T limit curves.

Discussion of the other ferritic RCPB components relative to P-T limits is contained in Appendix B.

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Westinghouse Non-Proprietary Class 3 2-1 2 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

2.1 INTRODUCTION

A discrete ordinates (Sn) transport analysis was performed for the Beaver Valley Unit 2 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, neutron exposure parameters in terms of fast neutron (E > 1.0 MeV) fluence 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 Y, withdrawn at the end of the 20th plant operating cycle, is provided. Comparisons of the results from the dosimetry evaluations with the analytical predictions served to validate the plant-specific neutron transport calculations. These validated calculations subsequently form the basis for projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY).

The use of fast neutron (E > 1.0 MeV) fluence 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, 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 E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, [16] recommends reporting displacements per iron atom along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy-dependent dpa function to be used for this evaluation is specified in ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom [17]. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has been promulgated in Revision 2 to Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials [1].

All of the calculations and dosimetry evaluations described in this section were based on nuclear cross-section data derived from ENDF/B-VI. 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 [18]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET [19].

2.2 DISCRETE ORDINATES ANALYSIS The arrangement of the surveillance capsules in the Beaver Valley Unit 2 reactor vessel is shown in Figure 2-1. Six irradiation capsules attached to the neutron pads are included in the reactor design that constitutes the reactor vessel surveillance program. Capsules U, V, W, X, Y, and Z are located at azimuthal angles of 343°, 107°, 110°, 287°, 290°, and 340°, respectively. These full-core positions correspond to the following octant symmetric locations represented in Figure 2-3: 17° from the core cardinal axes (for the 107°, 287° WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 2-2 and 343° dual surveillance capsule holder locations found in octants with a 26° neutron pad segment) and 20° from the core cardinal axes (for the 110°, 290° and 340° dual surveillance capsule holder locations found in octants with a 26° neutron pad segment). The stainless steel specimen containers are 1.182-inch by 1-inch and are approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core.

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

The presence of these materials has a significant effect on both the spatial distribution of neutron exposure 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 Beaver Valley Unit 2 reactor vessel and surveillance capsules, plant-specific 3D forward transport calculations were carried out to directly solve for the space- and energy-dependent neutron exposure rate, (r,,z,E).

For the Beaver Valley Unit 2 transport calculations, the models depicted in Figure 2-2 and Figure 2-3 were utilized. The reactor is octant symmetric with two different neutron pad and surveillance capsule configurations: octants with a 26-degree neutron pad segment and surveillance capsules located at 17° and 20°, and octants with a 15-degree neutron pad segment and no surveillance capsules.

Each octant model contained a representation of the reactor core, the reactor internals, 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 generally employed for the various structural components. In addition, 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, etc.

Section views of the models are shown in Figure 2-4 and Figure 2-5. Figure 2-4 shows the RAPTOR-M3G octant model without surveillance capsules, while Figure 2-5 shows the RAPTOR-M3G octant model with surveillance capsules. Both models extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation more than five feet below the active fuel to more than five feet above the active fuel.

Each of the two RAPTOR-M3G models consisted of 306 radial mesh, 235 azimuthal mesh, and 345 axial mesh. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the RAPTOR-M3G calculations was set at a value of 0.001.

The core power distributions used in the plant-specific transport analysis for the first 20 fuel cycles at Beaver Valley Unit 2 included cycle-dependent fuel assembly initial enrichments, burnups, radial and axial power distributions. Actual operating characteristics through Cycle 20 have been evaluated; projections WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 2-3 beyond Cycle 20 were based on the linear average of Cycles 17, 18, and 19 spatial power distributions, water temperatures, and reactor power level as directed by Beaver Valley. The cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions were 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 exposure rate, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the RAPTOR-M3G discrete ordinates code and the BUGLE-96 cross-section library [5]. 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 P3 Legendre expansion, and angular discretization was modeled with an S16 order of angular quadrature.

Results of the discrete ordinates transport analyses pertinent to the surveillance capsule evaluations are provided in Table 2-1 through Table 2-3. In Table 2-1, the calculated fast neutron fluence rate and fluence (E > 1.0 MeV) are provided at the geometric center of the capsules, as a function of irradiation time for the Beaver Valley Unit 2 reactor. Similar data presented in terms of iron atom displacement rate and integrated iron atom displacements are given in Table 2-2.

In Table 2-3, lead factors associated with surveillance capsules are provided as a function of operating time for the Beaver Valley Unit 2 reactor. The lead factor is defined as the ratio of the neutron fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the maximum neutron fluence (E > 1.0 MeV) at the pressure vessel clad/base metal interface.

Neutron exposure data pertinent to the pressure vessel clad/base metal interface are given in Table 2-4 and Table 2-5 for neutron fluence (E > 1.0 MeV) rate and fluence (E > 1.0 MeV), respectively, and in Table 2-6 and Table 2-7 for dpa/s and dpa, respectively. In each case, the data are provided for each operating cycle of the Beaver Valley Unit 2 reactor. Neutron fluence (E > 1.0 MeV) and dpa are also projected to future operating times extending to 60 EFPY. The vessel exposure data are presented in terms of the maximum exposure experienced by the pressure vessel at azimuthal angles of 0°, 15°, 30°, and 45°, and at the azimuthal location providing the maximum exposure relative to the core cardinal axes.

In Table 2-8 and Table 2-9, maximum projected fluences and dpa, respectively, of the various pressure vessel materials are given.

These data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end of Cycle 20 and projections to 60 EFPY. The projections beyond Cycle 20 were based on the linear average of Cycles 17, 18, and 19 spatial power distributions, water temperatures, and reactor power level.

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Westinghouse Non-Proprietary Class 3 2-4 2.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures reported in Section 2.2 is demonstrated by a direct comparison against the measured sensor reaction rates and 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 Beaver Valley Unit 2, Capsule Y, is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed from Beaver Valley Unit 2 up to date based on both direct and least-squares evaluation comparisons is documented in WCAP-18558-NP [13].

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule Y, that was withdrawn from the reactor at the conclusion of Cycle 20, is summarized below.

Reaction Rate (rps/atom)

Reaction Measured Calculated M/C 63 Cu (n,) 60Co 5.35E-17 5.69E-17 0.94 54 54 Fe (n,p) Mn 5.57E-15 6.34E-15 0.88 58 58 Ni (n,p) Co 8.11E-15 8.90E-15 0.91 Average of M/C Results 0.91 Standard Deviation (%) 3.3 The measured-to-calculated (M/C) reaction rate ratios for the Capsule Y threshold reactions range from 0.88 to 0.94, and the average M/C ratio is 0.91 +/- 3.3% (1). This direct comparison falls within the +/- 20%

criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 2.2; therefore, the calculations are deemed applicable for Beaver Valley Unit 2.

2.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Beaver Valley Unit 2 surveillance capsules 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. Simulator Benchmark Comparisons: Comparisons of calculations with measurements from simulator benchmarks, including the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 Experiment.

2. Operating Reactor and Calculational Benchmarks: Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment. Also considered are comparisons of calculations to results published in the NRC fluence calculation benchmark.

3. Analytic Uncertainty Analysis: 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.

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Westinghouse Non-Proprietary Class 3 2-5

4. Plant-Specific Benchmarking: Comparisons of the plant-specific calculations with all available dosimetry results from the Beaver Valley Unit 2 surveillance program.

The first phase of the methods qualification (simulator benchmark 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 (operating reactor and calculational benchmark 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 Beaver Valley Unit 2 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Beaver Valley Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures described in Section 2.2. As such, the validation of the Beaver Valley Unit 2 analytical model based on the measured plant dosimetry is completely described in WCAP-18558-NP [13].

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Westinghouse Report WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET [19].

Description Capsule and Vessel IR Simulator Benchmark Comparisons 3%

Operating Reactor and Calculational Benchmarks 5%

Analytic Uncertainty Analysis 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 WCAP-18558-NP [13]

support these uncertainty assessments for Beaver Valley Unit 2.

The NRC-issued Safety Evaluation for WCAP-18124-NP-A appears in Section A of [19]. The NRC identified two Limitations and Conditions associated with the application of RAPTOR-M3G and FERRET, which are reproduced here for convenience:

1. Applicability of WCAP-18124-NP, Revision 0 is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided.

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Westinghouse Non-Proprietary Class 3 2-6 Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to the response parameters of interest (e.g. pressure-temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and the reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

2. Least squares adjustment is acceptable if the adjustments to the M/C ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the discrepancy should be disqualified.

The neutron exposure values applicable to the surveillance capsules and the maximum reactor pressure vessel neutron exposure values used to derive the surveillance capsule lead factors are completely covered by the benchmarking and uncertainty analyses in WCAP-18124-NP-A. Note, however, that this report does contain neutron exposure values for materials that are outside the qualification basis of WCAP-18124-NP-A (i.e. extended beltline materials). For the materials considered to be located in the extended beltline region, Beaver Valley has provided justification for the use of WCAP-18124-NP-A for the extended beltline to the NRC in [14].

Limitation #2 applies in situations where the least-squares analysis is used to adjust the calculated values of neutron exposure. In this report, the least-squares analysis is provided only as a supplemental check on the results of the dosimetry evaluation. The least-squares analysis was not used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure. Therefore, Limitation #2 does not apply.

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Westinghouse Non-Proprietary Class 3 2-7 Table 2-1 Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Locations Cycle Total Fluence Rate (n/cm2-s)

Length Time Cycle (EFPY) (EFPY) 17-Degree 20-Degree 1 1.25 1.25 1.54E+11 1.35E+11 2 1.01 2.27 1.25E+11 1.11E+11 3 1.24 3.50 1.41E+11 1.28E+11 4 1.26 4.77 1.37E+11 1.24E+11 5 1.23 6.00 1.33E+11 1.16E+11 6 1.25 7.24 1.24E+11 1.15E+11 7 1.27 8.52 1.22E+11 1.10E+11 8 1.33 9.85 1.17E+11 1.06E+11 9 1.22 11.07 1.10E+11 9.64E+10 10 1.50 12.57 1.13E+11 1.01E+11 11 1.44 14.02 1.17E+11 1.03E+11 12 1.41 15.42 1.09E+11 9.49E+10 13 1.41 16.83 1.14E+11 9.88E+10 14 1.36 18.19 1.10E+11 9.75E+10 15 1.26 19.45 1.06E+11 9.43E+10 16 1.45 20.90 1.17E+11 1.02E+11 17 1.41 22.31 1.21E+11 1.06E+11 18 1.33 23.64 1.17E+11 1.01E+11 19 1.44 25.08 1.22E+11 1.06E+11 20 1.38 26.46 1.24E+11 1.08E+11 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 2-8 Table 2-1 (continued) Calculated Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Locations Total Fluence (n/cm2)

Time Cycle (EFPY) U V W X Y Z A 1 1.25 6.10E+18 6.10E+18 5.35E+18 6.10E+18 5.35E+18 5.35E+18 2 2.27 1.01E+19 8.91E+18 1.01E+19 8.91E+18 8.91E+18 3 3.50 1.56E+19 1.39E+19 1.56E+19 1.39E+19 1.39E+19 4 4.77 2.11E+19 1.88E+19 2.11E+19 1.88E+19 1.88E+19 5 6.00 2.62E+19 2.33E+19 2.62E+19 2.33E+19 2.33E+19 6 7.24 2.79E+19 3.11E+19 2.79E+19 2.79E+19 7 8.52 3.23E+19 3.60E+19 3.23E+19 3.23E+19 8 9.85 3.68E+19 4.09E+19 3.68E+19 3.68E+19 9 11.07 4.52E+19 4.05E+19 4.05E+19 4.22E+18 10 12.57 5.05E+19 4.53E+19 4.53E+19 9.55E+18 11 14.02 5.58E+19 5.00E+19 5.00E+19 1.49E+19 12 15.42 5.42E+19 5.42E+19 1.97E+19 13 16.83 5.86E+19 5.86E+19 2.48E+19 14 18.19 6.27E+19 6.27E+19 2.95E+19 15 19.45 6.65E+19 6.65E+19 3.37E+19 16 20.90 7.12E+19 7.12E+19 3.91E+19 17 22.31 7.59E+19 7.59E+19 4.45E+19 18 23.64 8.02E+19 8.02E+19 4.94E+19 19 25.08 8.50E+19 8.50E+19 5.49E+19 20 26.46 8.97E+19 8.97E+19 6.03E+19 32 1.08E+20 8.13E+19 36 1.21E+20 9.64E+19 40 1.34E+20 1.12E+20 48 1.61E+20 1.42E+20 54 1.80E+20 1.65E+20 60 2.00E+20 1.87E+20 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 2-9 Table 2-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations Cycle Total Iron Atom Displacement Rate Length Time (dpa/s)

Cycle (EFPY) (EFPY) 17-Degree 20-Degree 1 1.25 1.25 3.13E-10 2.69E-10 2 1.01 2.27 2.50E-10 2.19E-10 3 1.24 3.50 2.81E-10 2.51E-10 4 1.26 4.77 2.74E-10 2.43E-10 5 1.23 6.00 2.67E-10 2.29E-10 6 1.25 7.24 2.48E-10 2.26E-10 7 1.27 8.52 2.45E-10 2.16E-10 8 1.33 9.85 2.35E-10 2.09E-10 9 1.22 11.07 2.20E-10 1.90E-10 10 1.50 12.57 2.25E-10 1.98E-10 11 1.44 14.02 2.34E-10 2.03E-10 12 1.41 15.42 2.18E-10 1.87E-10 13 1.41 16.83 2.30E-10 1.95E-10 14 1.36 18.19 2.20E-10 1.92E-10 15 1.26 19.45 2.13E-10 1.85E-10 16 1.45 20.90 2.35E-10 2.02E-10 17 1.41 22.31 2.41E-10 2.09E-10 18 1.33 23.64 2.36E-10 2.00E-10 19 1.44 25.08 2.45E-10 2.09E-10 20 1.38 26.46 2.48E-10 2.12E-10 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-10 Table 2-2 (continued) Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations Total Iron Atom Displacements (dpa)

Time Cycle (EFPY) U V W X Y Z A 1 1.25 1.24E-02 1.24E-02 1.06E-02 1.24E-02 1.06E-02 1.06E-02 2 2.27 2.04E-02 1.76E-02 2.04E-02 1.76E-02 1.76E-02 3 3.50 3.13E-02 2.75E-02 3.13E-02 2.75E-02 2.75E-02 4 4.77 4.23E-02 3.71E-02 4.23E-02 3.71E-02 3.71E-02 5 6.00 5.26E-02 4.60E-02 5.26E-02 4.60E-02 4.60E-02 6 7.24 5.49E-02 6.24E-02 5.49E-02 5.49E-02 7 8.52 6.36E-02 7.22E-02 6.36E-02 6.36E-02 8 9.85 7.24E-02 8.21E-02 7.24E-02 7.24E-02 9 11.07 9.06E-02 7.97E-02 7.97E-02 8.47E-03 10 12.57 1.01E-01 8.91E-02 8.91E-02 1.91E-02 11 14.02 1.12E-01 9.83E-02 9.83E-02 2.98E-02 12 15.42 1.07E-01 1.07E-01 3.95E-02 13 16.83 1.15E-01 1.15E-01 4.97E-02 14 18.19 1.24E-01 1.24E-01 5.91E-02 15 19.45 1.31E-01 1.31E-01 6.76E-02 16 20.90 1.40E-01 1.40E-01 7.84E-02 17 22.31 1.49E-01 1.49E-01 8.91E-02 18 23.64 1.58E-01 1.58E-01 9.91E-02 19 25.08 1.67E-01 1.67E-01 1.10E-01 20 26.46 1.77E-01 1.77E-01 1.21E-01 32 2.13E-01 1.63E-01 36 2.39E-01 1.93E-01 40 2.65E-01 2.24E-01 48 3.16E-01 2.84E-01 54 3.55E-01 3.30E-01 60 3.94E-01 3.76E-01 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-11 Table 2-3 Calculated Surveillance Capsule Lead Factors Cycle Total Lead Factor Length Time Cycle (EFPY) (EFPY) U V W X Y Z A 1 1.25 1.25 2.95 2.95 2.58 2.95 2.58 2.58 2 1.01 2.27 3.16 2.79 3.16 2.79 2.79 3 1.24 3.50 3.39 3.03 3.39 3.03 3.03 4 1.26 4.77 3.42 3.06 3.42 3.06 3.06 5 1.23 6.00 3.41 3.04 3.41 3.04 3.04 6 1.25 7.24 3.09 3.45 3.09 3.09 7 1.27 8.52 3.10 3.45 3.10 3.10 8 1.33 9.85 3.12 3.47 3.12 3.12 9 1.22 11.07 3.43 3.08 3.08 0.32 10 1.50 12.57 3.45 3.09 3.09 0.65 11 1.44 14.02 3.45 3.09 3.09 0.92 12 1.41 15.42 3.07 3.07 1.12 13 1.41 16.83 3.02 3.02 1.28 14 1.36 18.19 3.03 3.03 1.42 15 1.26 19.45 3.03 3.03 1.54 16 1.45 20.90 3.01 3.01 1.65 17 1.41 22.31 3.01 3.01 1.76 18 1.33 23.64 2.98 2.98 1.84 19 1.44 25.08 2.97 2.97 1.92 20 1.38 26.46 2.97 2.97 1.99 32 2.93 2.21 36 2.91 2.32 40 2.90 2.41 48 2.88 2.54 54 2.86 2.61 60 2.85 2.67 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-12 Table 2-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface Fluence Rate (n/cm2-s)

Cycle Total Elevation Length Time of Max.

Cycle (EFPY) (EFPY) 0° 15° 30° 45° Maximum (cm) 1 1.25 1.25 5.23E+10 2.78E+10 2.06E+10 1.37E+10 5.23E+10 -112.93 2 1.01 2.27 3.58E+10 2.15E+10 1.60E+10 1.06E+10 3.58E+10 -58.96 3 1.24 3.50 3.62E+10 2.33E+10 1.83E+10 1.28E+10 3.62E+10 -5.00 4 1.26 4.77 3.91E+10 2.34E+10 1.70E+10 1.05E+10 3.91E+10 -58.96 5 1.23 6.00 3.97E+10 2.30E+10 1.58E+10 1.04E+10 3.97E+10 -60.96 6 1.25 7.24 3.35E+10 2.12E+10 1.81E+10 1.27E+10 3.35E+10 -58.96 7 1.27 8.52 3.52E+10 2.11E+10 1.74E+10 1.28E+10 3.52E+10 -58.96 8 1.33 9.85 3.24E+10 2.00E+10 1.55E+10 1.04E+10 3.24E+10 -58.96 9 1.22 11.07 3.53E+10 1.92E+10 1.47E+10 1.09E+10 3.53E+10 -60.96 10 1.50 12.57 3.11E+10 1.91E+10 1.49E+10 1.06E+10 3.11E+10 -58.96 11 1.44 14.02 3.45E+10 2.00E+10 1.45E+10 9.38E+09 3.45E+10 -58.96 12 1.41 15.42 3.27E+10 1.87E+10 1.38E+10 9.80E+09 3.27E+10 -58.96 13 1.41 16.83 3.91E+10 1.99E+10 1.40E+10 9.56E+09 3.91E+10 -58.96 14 1.36 18.19 3.19E+10 1.87E+10 1.42E+10 9.83E+09 3.19E+10 94.94 15 1.26 19.45 3.09E+10 1.80E+10 1.39E+10 9.86E+09 3.09E+10 92.94 16 1.45 20.90 3.80E+10 2.00E+10 1.50E+10 1.13E+10 3.80E+10 92.94 17 1.41 22.31 3.55E+10 2.04E+10 1.54E+10 1.13E+10 3.55E+10 92.94 18 1.33 23.64 3.96E+10 2.04E+10 1.45E+10 1.02E+10 3.96E+10 94.94 19 1.44 25.08 3.91E+10 2.09E+10 1.54E+10 1.12E+10 3.91E+10 -58.96 20 1.38 26.46 3.71E+10 2.09E+10 1.45E+10 9.45E+09 3.71E+10 46.97 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-13 Table 2-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface Fluence (n/cm2)

Cycle Total Elevation Length Time of Max.

Cycle(a) (EFPY) (EFPY) 0° 15° 30° 45° Maximum (cm) 1 1.25 1.25 2.07E+18 1.10E+18 8.14E+17 5.44E+17 2.07E+18 -112.93 2 1.01 2.27 3.19E+18 1.78E+18 1.32E+18 8.78E+17 3.19E+18 -66.96 3 1.24 3.50 4.59E+18 2.68E+18 2.03E+18 1.38E+18 4.59E+18 -60.96 4 1.26 4.77 6.15E+18 3.62E+18 2.71E+18 1.80E+18 6.15E+18 -60.96 5 1.23 6.00 7.69E+18 4.51E+18 3.32E+18 2.20E+18 7.69E+18 -60.96 6 1.25 7.24 9.01E+18 5.34E+18 4.04E+18 2.70E+18 9.01E+18 -58.96 7 1.27 8.52 1.04E+19 6.19E+18 4.74E+18 3.22E+18 1.04E+19 -58.96 8 1.33 9.85 1.18E+19 7.03E+18 5.39E+18 3.65E+18 1.18E+19 -58.96 9 1.22 11.07 1.32E+19 7.77E+18 5.96E+18 4.07E+18 1.32E+19 -58.96 10 1.50 12.57 1.46E+19 8.68E+18 6.66E+18 4.58E+18 1.46E+19 -58.96 11 1.44 14.02 1.62E+19 9.59E+18 7.32E+18 5.00E+18 1.62E+19 -58.96 12 1.41 15.42 1.77E+19 1.04E+19 7.94E+18 5.44E+18 1.77E+19 -58.96 13 1.41 16.83 1.94E+19 1.13E+19 8.56E+18 5.86E+18 1.94E+19 -58.96 14 1.36 18.19 2.07E+19 1.21E+19 9.17E+18 6.28E+18 2.07E+19 -58.96 15 1.26 19.45 2.20E+19 1.28E+19 9.72E+18 6.68E+18 2.20E+19 -58.96 16 1.45 20.90 2.37E+19 1.37E+19 1.04E+19 7.19E+18 2.37E+19 -58.96 17 1.41 22.31 2.52E+19 1.46E+19 1.11E+19 7.70E+18 2.52E+19 -58.96 18 1.33 23.64 2.69E+19 1.55E+19 1.17E+19 8.13E+18 2.69E+19 -58.96 19 1.44 25.08 2.86E+19 1.64E+19 1.24E+19 8.64E+18 2.86E+19 -58.96 20 1.38 26.46 3.02E+19 1.74E+19 1.30E+19 9.05E+18 3.02E+19 -58.96 32 3.68E+19 2.09E+19 1.57E+19 1.10E+19 3.68E+19 -58.96 36 4.16E+19 2.35E+19 1.76E+19 1.23E+19 4.16E+19 -58.96 40 4.64E+19 2.61E+19 1.95E+19 1.37E+19 4.64E+19 -58.96 48 5.59E+19 3.13E+19 2.33E+19 1.65E+19 5.59E+19 -58.96 54 6.30E+19 3.52E+19 2.61E+19 1.86E+19 6.30E+19 -58.96 60 7.02E+19 3.91E+19 2.90E+19 2.06E+19 7.02E+19 -58.96 Note:

(a) Values beyond Cycle 20 are projected based on the linear average of Cycles 17, 18, and 19.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-14 Table 2-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface Displacement Rate (dpa/s)

Cycle Total Elevation Length Time of Max.

Cycle (EFPY) (EFPY) 0° 15° 30° 45° Maximum (cm) 1 1.25 1.25 8.27E-11 4.37E-11 3.14E-01 2.11E-11 8.27E-11 -112.93 2 1.01 2.27 5.67E-11 3.37E-11 2.45E-01 1.65E-11 5.67E-11 -58.96 3 1.24 3.50 5.73E-11 3.65E-11 2.80E-01 1.97E-11 5.73E-11 -5.00 4 1.26 4.77 6.18E-11 3.66E-11 2.61E-01 1.63E-11 6.18E-11 -58.96 5 1.23 6.00 6.28E-11 3.61E-11 2.43E-01 1.61E-11 6.28E-11 -60.96 6 1.25 7.24 5.30E-11 3.33E-11 2.77E-01 1.95E-11 5.30E-11 -58.96 7 1.27 8.52 5.57E-11 3.32E-11 2.66E-01 1.98E-11 5.57E-11 -58.96 8 1.33 9.85 5.13E-11 3.13E-11 2.38E-01 1.61E-11 5.13E-11 -58.96 9 1.22 11.07 5.58E-11 3.02E-11 2.26E-01 1.69E-11 5.58E-11 -60.96 10 1.50 12.57 4.93E-11 3.00E-11 2.28E-01 1.64E-11 4.93E-11 -58.96 11 1.44 14.02 5.45E-11 3.14E-11 2.23E-01 1.45E-11 5.45E-11 -58.96 12 1.41 15.42 5.17E-11 2.94E-11 2.12E-01 1.51E-11 5.17E-11 -1.00 13 1.41 16.83 6.18E-11 3.13E-11 2.15E-01 1.48E-11 6.18E-11 -58.96 14 1.36 18.19 5.04E-11 2.93E-11 2.18E-01 1.52E-11 5.04E-11 92.94 15 1.26 19.45 4.89E-11 2.83E-11 2.13E-01 1.52E-11 4.89E-11 92.94 16 1.45 20.90 6.00E-11 3.16E-11 2.31E-01 1.74E-11 6.00E-11 92.94 17 1.41 22.31 5.62E-11 3.21E-11 2.36E-01 1.75E-11 5.62E-11 92.94 18 1.33 23.64 6.26E-11 3.21E-11 2.22E-01 1.58E-11 6.26E-11 92.94 19 1.44 25.08 6.18E-11 3.29E-11 2.37E-01 1.74E-11 6.18E-11 -58.96 20 1.38 26.46 5.87E-11 3.29E-11 2.23E-01 1.46E-11 5.87E-11 46.97 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-15 Table 2-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Displacements (dpa)

Cycle Total Elevation Length Time of Max.

Cycle(a) (EFPY) (EFPY) 0° 15° 30° 45° Maximum (cm) 1 1.25 1.25 3.27E-03 1.73E-03 1.24E-03 8.37E-04 3.27E-03 -112.93 2 1.01 2.27 5.05E-03 2.79E-03 2.02E-03 1.35E-03 5.05E-03 -66.96 3 1.24 3.50 7.27E-03 4.21E-03 3.11E-03 2.12E-03 7.27E-03 -60.96 4 1.26 4.77 9.74E-03 5.68E-03 4.15E-03 2.77E-03 9.74E-03 -60.96 5 1.23 6.00 1.22E-02 7.07E-03 5.09E-03 3.40E-03 1.22E-02 -60.96 6 1.25 7.24 1.43E-02 8.39E-03 6.18E-03 4.17E-03 1.43E-02 -60.96 7 1.27 8.52 1.65E-02 9.72E-03 7.25E-03 4.96E-03 1.65E-02 -58.96 8 1.33 9.85 1.87E-02 1.10E-02 8.25E-03 5.64E-03 1.87E-02 -58.96 9 1.22 11.07 2.08E-02 1.22E-02 9.12E-03 6.29E-03 2.08E-02 -58.96 10 1.50 12.57 2.31E-02 1.36E-02 1.02E-02 7.07E-03 2.31E-02 -58.96 11 1.44 14.02 2.56E-02 1.51E-02 1.12E-02 7.73E-03 2.56E-02 -58.96 12 1.41 15.42 2.79E-02 1.64E-02 1.22E-02 8.40E-03 2.79E-02 -58.96 13 1.41 16.83 3.07E-02 1.77E-02 1.31E-02 9.05E-03 3.07E-02 -58.96 14 1.36 18.19 3.28E-02 1.90E-02 1.41E-02 9.71E-03 3.28E-02 -58.96 15 1.26 19.45 3.47E-02 2.01E-02 1.49E-02 1.03E-02 3.47E-02 -58.96 16 1.45 20.90 3.74E-02 2.16E-02 1.60E-02 1.11E-02 3.74E-02 -58.96 17 1.41 22.31 3.99E-02 2.30E-02 1.70E-02 1.19E-02 3.99E-02 -58.96 18 1.33 23.64 4.25E-02 2.43E-02 1.79E-02 1.26E-02 4.25E-02 -58.96 19 1.44 25.08 4.53E-02 2.58E-02 1.90E-02 1.33E-02 4.53E-02 -58.96 20 1.38 26.46 4.78E-02 2.73E-02 2.00E-02 1.40E-02 4.78E-02 -58.96 32 5.82E-02 3.29E-02 2.40E-02 1.69E-02 5.82E-02 -58.96 34 6.58E-02 3.70E-02 2.70E-02 1.91E-02 6.58E-02 -58.96 40 7.33E-02 4.11E-02 2.99E-02 2.12E-02 7.33E-02 -58.96 48 8.83E-02 4.92E-02 3.57E-02 2.55E-02 8.83E-02 -58.96 54 9.96E-02 5.53E-02 4.01E-02 2.86E-02 9.96E-02 -58.96 60 1.11E-01 6.14E-02 4.45E-02 3.18E-02 1.11E-01 -58.96 Note:

(a) Values beyond Cycle 20 are projected based on the linear average of Cycles 17, 18, and 19.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-16 Table 2-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Welds and Shells Fast Fluence (n/cm2) 26.46 Material EFPY 32 EFPY 36 EFPY 40 EFPY Outlet Nozzle to Nozzle Shell 2.73E+16 3.38E+16 3.85E+16 4.32E+16 Weld - Lowest Extent Inlet Nozzle to Nozzle Shell 6.01E+16 7.59E+16 8.74E+16 9.89E+16 Weld - Lowest Extent Upper Shell - Lowest Extent 1.82E+18 2.30E+18 2.65E+18 3.00E+18 Intermediate to Upper Shell Weld 1.82E+18 2.30E+18 2.65E+18 3.00E+18 Intermediate Shell 3.01E+19 3.67E+19 4.14E+19 4.61E+19 Intermediate Shell Long. Welds 8.90E+18 1.08E+19 1.21E+19 1.35E+19 Lower to Intermediate Shell Weld 3.01E+19 3.67E+19 4.14E+19 4.61E+19 Lower Shell 3.02E+19 3.68E+19 4.16E+19 4.64E+19 Lower Shell Long. Welds 9.05E+18 1.10E+19 1.23E+19 1.37E+19 Lower Shell to Lower Closure 8.98E+15 1.10E+16 1.24E+16 1.39E+16 Head Weld Fast Fluence (n/cm2)

Material 48 EFPY 54 EFPY 60 EFPY Outlet Nozzle to Nozzle Shell 5.26E+16 5.96E+16 6.67E+16 Weld - Lowest Extent Inlet Nozzle to Nozzle Shell 1.22E+17 1.39E+17 1.56E+17 Weld - Lowest Extent Upper Shell - Lowest Extent 3.69E+18 4.21E+18 4.73E+18 Intermediate to Upper Shell Weld 3.69E+18 4.21E+18 4.73E+18 Intermediate Shell 5.56E+19 6.28E+19 6.99E+19 Intermediate Shell Long. Welds 1.62E+19 1.83E+19 2.03E+19 Lower to Intermediate Shell Weld 5.56E+19 6.27E+19 6.98E+19 Lower Shell 5.59E+19 6.30E+19 7.02E+19 Lower Shell Long. Welds 1.65E+19 1.86E+19 2.06E+19 Lower Shell to Lower Closure 1.68E+16 1.89E+16 2.11E+16 Head Weld WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-17 Table 2-9 Calculated Maximum Iron Atom Displacements at Pressure Vessel Welds and Shells Iron Atom Displacements (dpa) 26.46 Material EFPY 32 EFPY 36 EFPY 40 EFPY Outlet Nozzle to Nozzle Shell 1.56E-04 1.93E-04 2.20E-04 2.46E-04 Weld - Lowest Extent Inlet Nozzle to Nozzle Shell 2.09E-04 2.59E-04 2.95E-04 3.31E-04 Weld - Lowest Extent Upper Shell - Lowest Extent 2.93E-03 3.70E-03 4.25E-03 4.81E-03 Intermediate to Upper Shell Weld 2.93E-03 3.70E-03 4.25E-03 4.81E-03 Intermediate Shell 4.76E-02 5.80E-02 6.55E-02 7.30E-02 Intermediate Shell Long. Welds 1.38E-02 1.67E-02 1.88E-02 2.09E-02 Lower to Intermediate Shell Weld 4.76E-02 5.80E-02 6.55E-02 7.30E-02 Lower Shell 4.78E-02 5.82E-02 6.58E-02 7.33E-02 Lower Shell Long. Welds 1.40E-02 1.69E-02 1.91E-02 2.12E-02 Lower Shell to Lower Closure 5.03E-05 6.14E-05 6.95E-05 7.76E-05 Head Weld Iron Atom Displacements (dpa)

Material 48 EFPY 54 EFPY 60 EFPY Outlet Nozzle to Nozzle Shell 3.00E-04 3.40E-04 3.80E-04 Weld - Lowest Extent Inlet Nozzle to Nozzle Shell 4.04E-04 4.58E-04 5.12E-04 Weld - Lowest Extent Upper Shell - Lowest Extent 5.92E-03 6.75E-03 7.58E-03 Intermediate to Upper Shell Weld 5.92E-03 6.75E-03 7.58E-03 Intermediate Shell 8.80E-02 9.93E-02 1.11E-01 Intermediate Shell Long. Welds 2.51E-02 2.82E-02 3.14E-02 Lower to Intermediate Shell Weld 8.80E-02 9.92E-02 1.10E-01 Lower Shell 8.83E-02 9.96E-02 1.11E-01 Lower Shell Long. Welds 2.55E-02 2.86E-02 3.18E-02 Lower Shell to Lower Closure 9.37E-05 1.06E-04 1.18E-04 Head Weld WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-18 Figure 2-1 Arrangement of Surveillance Capsules in the Beaver Valley Unit 2 Reactor Vessel WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31 :48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-19 Figure 2-2 Beaver Valley Unit 2 Plan View of the Reactor Geometry at the Core Midplane 15.0° Neutron Pad Configuration WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31 :48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-20

+-w----,-,~-..----,-1~-,----,Q-1. -,~-.*-o--~,es~ 2.02.s 21.1t~2- .~->>--------.---------t _

lcm.J Figure 2-3 Beaver Valley Unit 2 Plan View of the Reactor Geometry at the Core Midplane 26° Neutron Pad Configuration WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31 :48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-21

.L. J.,.....-----

N

~

~ 4.115H02 cm OJ.llOEi'OO Figure 2-4 Beaver Valley Unit 2 Section View of the Reactor Geometry without Surveillance Capsule WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31 :48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-22 itz......_..,__ _ _ _ _ __

I 4. 1115E +02 cm R I O.:lilCIE+Q01 , .11E+02 Figure 2-5 Beaver Valley Unit 2 Section View of the Reactor Geometry with Surveillance Capsule WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31 :48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requirements for RVI and P-T limit curve development are specified in 10 CFR 50, Appendix G [4].

The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:

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

The Beaver Valley Unit 2 beltline materials traditionally included Intermediate Shell Plates B9004-1 and -2, Lower Shell Plates B9005-1 and -2, Intermediate Shell Longitudinal Welds Seams 101-124 A and B, Lower Shell Longitudinal Weld Seams 101-142 A and B, and Intermediate to Lower Shell Girth Weld Seam 101-171; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [6], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E >

1.0 MeV) at the end of the licensed operating period should be considered to experience neutron fluence sufficient to cause embrittlement. The additional materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met.

As seen from Table 2-8 of this report, the extended beltline materials include Upper Shell Plates B9003-1,

-2, and -3, Upper Shell Longitudinal Weld Seams 101-122 A, B, and C, Upper to Intermediate Shell Girth Weld Seam 103-121, Inlet Nozzles B9011-1, -2, and -3, and Inlet Nozzle Weld Seams 105-121 A, B, and C.

A summary of the best-estimate copper (Cu) and nickel (Ni), contents, in units of weight percent (wt. %),

as well as initial RTNDT and initial USE values for the reactor vessel beltline and extended beltline materials are provided in Table 3-1 for Beaver Valley Unit 2. Table 3-2 contains a summary of the initial RTNDT values of the reactor vessel flange and reactor vessel closure head flange. These flange initial RTNDT values serve as input to the P-T limit curves flange-notch per Appendix G of 10 CFR 50 - See Section 6.3 for details.

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Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Beaver Valley Unit 2 Reactor Vessel Materials(a)

Chemical Fracture Toughness Property Composition(a)

Reactor Vessel Material Heat Flux Type Initial Upper-Initial and Identification Number Number (Lot) Wt. % Wt. % I (b) Shelf Energy RTNDT(b)

Cu Ni (°F) (ft-lb)

(°F)

Reactor Vessel Beltline Materials Intermediate Shell Plate B9004-1 --- --- 0.065 0.55 60 0 83 Intermediate Shell Plate B9004-2 --- --- 0.06 0.57 40 0 79 Lower Shell Plate B9005-1 --- --- 0.08 0.58 28 0 82 Lower Shell Plate B9005-2 --- --- 0.07 0.57 33 0 78 Intermediate to Lower Shell 0091 Circumferential Girth Weld 83642 0.046 0.086 -30 0 145 (3536)

Seam 101-171 Intermediate Shell Longitudinal Weld 0091 83642 0.046 0.086 -30 0 145 Seams 101-124 A & B (3536)

Lower Shell Longitudinal Weld 0091 83642 0.046 0.086 -30 0 145 Seams 101-142 A & B (3536)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B9003-1 A9406-1 --- 0.13 0.60 50 0 98 Upper Shell Plate B9003-2 B4431-2 --- 0.12 0.60 60 0 80 Upper Shell Plate B9003-3 A9406-2 --- 0.13 0.60 50 0 98 0091 51912 0.156(d) 0.059(d) -50 0 96 (3490) 0091 51912 0.156(d) 0.059(d) -70 0 114 (3536)

Upper Shell Longitudinal Weld Seams 101-122A, 101-122B, and 101-122C EAIB --- 0.02 0.98 10 (Gen.) (c) 17(c) 118 IAGA --- 0.03 0.98 -30 0 160 BOHB --- 0.05 1.00 10 (Gen.) (c) 17 (c) 97 BAOED --- 0.02 1.00 -50 0 150 0091 4P5174 0.09 1.00(g) -50 0 70 (1122)

Upper to Intermediate Shell Girth 0091 51922 0.05 1.00(g) -56 (Gen.) (c) 17(c) 102 Weld Seam 103-121 (3489)

AAGC --- 0.03 0.98 -70 0 111 KOIB --- 0.03 0.97 -60 0 110 2V2436-Inlet Nozzle B9011-1 --- 0.11 0.85 60 0 68.25(e)01-002 2V2437-Inlet Nozzle B9011-2 --- 0.13(f) 0.88 60 (Gen.)(c) 17(c) 75.4(e)02-001 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 3-3 Table 3-1 Summary of the Best-Estimate Chemistry, Initial RTNDT, and Initial USE Values for the Beaver Valley Unit 2 Reactor Vessel Materials(a)

Chemical Fracture Toughness Property Composition(a)

Reactor Vessel Material Heat Flux Type Initial Upper-Initial and Identification Number Number (Lot) Wt. % Wt. % I (b) Shelf Energy RTNDT(b)

Cu Ni (°F) (ft-lb)

(°F) 2V2445-Inlet Nozzle B9011-3 --- 0.13(f) 0.84 70 0 79.3(e)02-003 0091 4P5174 0.09 1.00(g) -50 0 70 (1122)

LOHB --- 0.03 1.03 -60 0 137 HABJC --- 0.02 1.02 -70 0 162 BABBD --- 0.02 1.04 -70 0 142 Inlet Nozzle Weld Seams FABGC --- 0.03 1.02 -80 0 119 105-121A, 105-121B, and 105-121C EOBC --- 0.02 0.96 -60 0 127 FAAFC --- 0.07 1.04 -60 0 119 CCJC --- 0.02 0.99 -60 0 109 FAGB --- 0.02 1.06 -30 0 114 BAOED --- 0.02 1.00 -50 0 150 Surveillance Materials 0091 Surveillance Weld 83642 0.065 0.065 --- --- ---

(3536)

Notes:

(a) Unless otherwise noted, information extracted from [9].

(b) Unless otherwise noted, all RTNDT(U) values are based on measured data; thus, I = 0°F.

(c) The generic RTNDT(U) values were determined in accordance with NUREG-0800 [7] and 10 CFR 50.61 [10] ; thus, I = 17°F.

(d) Chemistry obtained from Combustion Engineering Report CE-NPSD-1039, Rev. 2 [11].

(e) Value is equal to 65% of the USE measured in the tangential (strong) direction per [7].

(f) The Cu wt% was not available from the CMTR, so in accordance with [1], a standard deviation analysis (average + standard deviation) was done to determine the value based on Westinghouse 508 Class 2 Nozzle Forgings (178 data points).

(g) Default wt% Ni content per [1].

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Westinghouse Non-Proprietary Class 3 3-4 Table 3-2 Summary of Beaver Valley Unit 2 Reactor Vessel Closure Head Flange and Vessel Flange Initial RTNDT Values Initial RTNDT(a)

Reactor Vessel Material

(°F)

Closure Head Flange -10 Vessel Flange 0 Notes:

(a) Values taken from WCAP-16528-NP [12].

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Westinghouse Non-Proprietary Class 3 4-1 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2 [1], calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, surveillance data generated from surveillance programs at other plants which include a reactor vessel beltline or extended beltline material should also be considered when calculating Position 2.1 chemistry factors.

The surveillance capsule plate material for Beaver Valley Unit 2 is from Intermediate Shell Plate B9004-2.

Per [13], the surveillance plate data are deemed non-credible for Beaver Valley Unit 2; therefore, a reduced margin term cannot be utilized in the Intermediate Shell Plate B9004-2 ART or PTS calculations contained in Section 7 and Appendix D, respectively.

The Beaver Valley Unit 2 surveillance weld specimens were fabricated from weld material Heat # 83642 flux type 091, Lot # 3536. This weld material is applicable to all beltline welds, i.e., the intermediate shell longitudinal weld seams, the lower shell longitudinal weld seams, and the intermediate to lower shell girth weld seam. Per [13], the surveillance data are deemed credible for Beaver Valley Unit 2; therefore, a reduced margin term will be utilized in the ART and PTS calculations for these welds contained in Section 7 and Appendix D, respectively.

Table 4-1 summarizes the surveillance data available for the Beaver Valley Unit 2 plate and weld materials that will be used in the calculation of the Position 2.1 chemistry factor values.

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Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Beaver Valley Unit 2 Surveillance Capsule Data(a)

Capsule Fluence Measured 30 ft-lb Transition Measured Upper-Shelf Material Capsule (x 10 n/cm2, E > 1.0 MeV) 19 Energy Decrease (%)

Temperature Shift (°F)

U 0.610 24.0 0 Intermediate Shell V 2.62 56.0 11 Plate B9004-2 W 3.68 71.8 5 (Longitudinal) X 5.58 98.0 15 Y 8.97 143.2 29 U 0.610 18.2 0 Intermediate Shell V 2.62 46.5 1 Plate B9004-2 W 3.68 63.8 3 (Transverse) X 5.58 104.5 4 Y 8.97 123.5 19 U 0.610 5.8 9 Beaver Valley V 2.62 26.6 4 Unit 2 Surveillance W 3.68 6.0 2 Weld X 5.58 23.4 7 (Heat # 83642)

Y 8.97 36.6 5 Note:

(a) Surveillance data from [13].

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Westinghouse Non-Proprietary Class 3 5-1 5 CHEMISTRY FACTORS The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2, Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

The best-estimate copper and nickel weight percent values for the Beaver Valley Unit 2 reactor vessel materials are provided in Table 3-1 of this report.

The Position 2.1 chemistry factor calculation is presented in Table 5-1 for the Beaver Valley Unit 2 surveillance materials. These values were calculated using the surveillance data summarized in Section 4 of this report, which only includes surveillance program results from Beaver Valley Unit 2. The calculations are performed using the method described in Regulatory Guide 1.99, Revision 2. All of the surveillance weld data considers the chemical composition differences between the surveillance weld and the weld being evaluated, in accordance with the guidance presented at an industry meeting held by the NRC on February 12 and 13, 1998 [15]. Margin will be applied to the ART calculations in Section 7 and PTS calculations in Appendix D according to the conclusions of the credibility evaluation for the surveillance material, as documented in Section 4.

The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-2 for Beaver Valley Unit 2.

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Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 Calculation of Beaver Valley Unit 2 Chemistry Factors Using Surveillance Capsule Data Capsule f(a) RTNDT(c) FF*RTNDT Material Capsule FF(b) FF2 (x 10 n/cm2, E > 1.0 MeV) 19

(°F) (°F)

U 0.610 0.862 24.0 20.68 0.742 Intermediate Shell V 2.62 1.258 56.0 70.44 1.582 Plate B9004-2 W 3.68 1.338 71.8 96.06 1.790 (Longitudinal) X 5.58 1.423 98.0 139.49 2.026 Y 8.97 1.500 143.2 214.76 2.249 U 0.610 0.862 18.2 15.68 0.742 Intermediate Shell V 2.62 1.258 46.5 58.49 1.582 Plate B9004-2 W 3.68 1.338 63.8 85.36 1.790 (Transverse) X 5.58 1.423 104.5 148.74 2.026 Y 8.97 1.500 123.5 185.21 2.249 SUM: 1034.90 16.779 CF B9004-2 = (FF

RTNDT) ÷ (FF2) = (1034.90) ÷ (16.779) = 61.7°F U 0.610 0.862 5.8 5.00 0.742 Beaver Valley V 2.62 1.258 26.6 33.46 1.582 Unit 2 Surveillance Weld W 3.68 1.338 6.0 8.03 1.790 (Heat #83642) X 5.58 1.423 23.4 33.31 2.026 Y 8.97 1.500 36.6 54.89 2.249 SUM: 134.68 8.389 CF Surv. Weld = (FF

RTNDT) ÷ (FF2) = (134.68) ÷ (8.389) = 16.1°F Notes:

(a) f = fluence. Data taken from Table 4-1.

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

(c) RTNDT values taken from Table 4-1. The surveillance weld RTNDT values have been conservatively adjusted by a factor of 1.0. The calculated adjustment is the ratio of the Position 1.1 CFs of the vessel weld to the surveillance weld (CFVessel Weld /

CFSurv. Weld), which is 0.905 (34.4°F / 38.0°F). This would result in a lower calculated Position 2.1 CF.

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Westinghouse Non-Proprietary Class 3 5-3 Table 5-2 Summary of Beaver Valley Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Heat Number Chemistry Factor (°F) and Identification Number (Lot) Position 1.1(a) Position 2.1(b)

Reactor Vessel Beltline Materials Intermediate Shell Plate B9004-1 --- 40.5 ---

Intermediate Shell Plate B9004-2 --- 37.0 61.7 Lower Shell Plate B9005-1 --- 51.0 ---

Lower Shell Plate B9005-2 --- 44.0 ---

Intermediate to Lower Shell Girth Weld 83642 34.4 16.1 Seam 101-171 (3536) 83642 Lower Longitudinal Weld Seams 101-142 A & B 34.4 16.1 (3536)

Intermediate Longitudinal Weld 83642 34.4 16.1 Seams 101-124 A & B (3536)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B9003-1 --- 91.0 ---

Upper Shell Plate B9003-2 --- 83.0 ---

Upper Shell Plate B9003-3 --- 91.0 ---

51912 73.71 ---

(3490) 51912 73.71 ---

(3536)

Upper Shell Longitudinal Weld Seams 101-122 A, B, & C EAIB 27.0 ---

IAGA 41.0 ---

BOHB 68.0 ---

BAOED 27.0 ---

4P5174 122.0 ---

(1122)

Upper to Intermediate Shell Girth Weld 51922 68.0 ---

Seam 103-121 (3489)

AAGC 41.0 ---

KOIB 41.0 ---

Inlet Nozzle B9011-1 --- 77.0 ---

Inlet Nozzle B9011-2 --- 96.0 ---

Inlet Nozzle B9011-3 --- 96.0 ---

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Westinghouse Non-Proprietary Class 3 5-4 Table 5-2 Summary of Beaver Valley Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Heat Number Chemistry Factor (°F) and Identification Number (Lot) Position 1.1(a) Position 2.1(b) 4P5174 122.0 ---

(1122)

LOHB 41.0 ---

HABJC 27.0 ---

BABBD 27.0 ---

Inlet Nozzle Weld Seams 105-121 A, B, & C FABGC 41.0 ---

EOBC 27.0 ---

FAAFC 95.0 ---

CCJC 27.0 ---

FAGB 27.0 ---

BAOED 27.0 ---

Reactor Vessel Surveillance Material 83642 Surveillance Program Weld Metal 38.0 ---

(3536)

Notes:

(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of this report and Tables 1 and 2 of Regulatory Guide 1.99, Revision 2 [1].

(b) Position 2.1 chemistry factors were taken from Table 5-1 of this report. As discussed in Section 4, the surveillance plate data was deemed non-credible, while the surveillance weld data was deemed credible.

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Westinghouse Non-Proprietary Class 3 6-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALL APPROACH The ASME (American Society of Mechanical Engineers) approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIc, for the metal temperature at that time. KIc is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [3]. The KIc curve is given by the following equation:

K Ic =33.2+20.734*e[ 0.02(T RTNDT )] (1)

where, KIc (ksiin.) = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This KIc curve is based on the lower bound of static critical KI values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* KIm + KIt < KIc (2)

where, KIm = stress intensity factor caused by membrane (pressure) stress KIt = stress intensity factor caused by the thermal gradients KIc = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 6-2 For membrane tension, the corresponding KI for the postulated defect is:

K Im = Mm x ( pRi / t ) (3)

Axial Flaw Methodology where, Mm for an inside axial surface flaw is given by:

Mm = 1.85 for t < 2, Mm = 0.926 t for 2 t 3.464 ,

Mm = 3.21 for t > 3.464 and, Mm for an outside axial surface flaw is given by:

Mm = 1.77 for t < 2, Mm = 0.893 t for 2 t 3.464 ,

Mm = 3.09 for t > 3.464 Circumferential Flaw Methodology Similarly, Mm for an inside or an outside circumferential surface flaw is given by:

Mm = 0.89 for t < 2, Mm = 0.443 t for 2 t 3.464 ,

Mm = 1.53 for t > 3.464

where, p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).

For bending stress, the corresponding KI for the postulated axial or circumferential defect is:

KIb = Mb

Maximum Bending Stress, where Mb is two-thirds of Mm (4)

The maximum KI produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:

KIt = 0.953 x 10-3 x CR x t2.5 (5) where CR is the cooldown rate in °F/hr., or for a postulated axial or circumferential outside surface defect KIt = 0.753 x 10-3 x HU x t2.5 (6) where HU is the heatup rate in °F/hr.

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Westinghouse Non-Proprietary Class 3 6-3 The through-wall temperature difference associated with the maximum thermal KI can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal KI.

(a) The maximum thermal KI relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2).

(b) Alternatively, the KI for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness axial or circumferential inside surface defect using the relationship:

KIt = (1.0359 C 0 + 0.6322 C1 + 0.4753C 2 + 0.3855C 3)

a (7) or similarly, KIt during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:

KIt = (1.043C 0 + 0.630C1 + 0.481C 2 + 0.401C 3)

a (8) where the coefficients C0, C1, C2, and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

( x ) = C 0 + C1( x / a ) + C 2 ( x / a ) 2 + C 3( x / a ) 3 (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).

Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves [2] Section 2.6 (equations 2.6.2-4 and 2.6.3-1). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft2/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-°F.

At any time during the heatup or cooldown transient, KIc is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, Paragraph G-2120), the appropriate value for RTNDT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIt, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

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Westinghouse Non-Proprietary Class 3 6-4 For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the T (temperature) across the vessel wall developed during cooldown results in a higher value of KIc at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KIc exceeds KIt, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KIc for the inside 1/4T flaw during heatup is lower than the KIc for the flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIc values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present.

These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

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Westinghouse Non-Proprietary Class 3 6-5 Following the generation of pressure-temperature curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

Uncertainty Correction:

The P-T limits, as directly calculated by ASME methodology, typically represent the limiting material conditions at the reactor vessel beltline. However, pressure is NOT measured at the beltline and, as such, an adjustment for head loss between the measurement location and beltline location is necessary. In addition, uncertainties in the instruments can cause the pressure and temperature indications observed in the control room to differ from actual values. In order to account for the head loss pressure adjustment and instrument uncertainties, conservative corrections are made to the unadjusted P-T limits as follows:

(1) The pressure difference due to static and dynamic head loss (pressure associated with RCP starts and the pressure difference between the measured location and the point of interest, e.g.

pressurizer and the beltline) is subtracted for the pressure of ASME calculated P-T limit curves; (2) The uncertainty associated with the RCS pressure instrumentation is subtracted from the pressure of ASME calculated P-T limit curves; and (3) The uncertainty associated with the RCS temperature instrumentation is added to the temperature of ASME calculated P-T limit curves.

6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is calculated to be 621 psig. The initial RTNDT values of the reactor vessel closure head and vessel flange are documented in Table 3-2. The limiting unirradiated RTNDT of 0°F is associated with the vessel flange of the Beaver Valley Unit 2 vessel, so the minimum allowable temperature of this region is 120°F at pressures greater than 621 psig without margins for instrument uncertainties. This limit is shown in Figures 8-1 and 8-2. The minimum allowable temperature with uncertainties is 137°F at pressures greater than 523 psig during heatup and 536 psig during cooldown.

This limit is shown in Figures 8-3 and 8-4.

6.4 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RTNDT, in the closure flange region. This requirement is established in Appendix G to 10 CFR 50 [4]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [2], the minimum boltup temperature should be 60°F or the limiting unirradiated RTNDT of the closure flange region, whichever is higher. Since the limiting unirradiated RTNDT of this region is below 60°F per Table 3-2, the minimum boltup temperature for the Beaver Valley Unit 2 reactor vessel is 60°F without margins for instrument uncertainties. This limit is shown in WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 6-6 Figures 8-1 and 8-2. The minimum boltup temperature with margins for instrument uncertainty is 77°F.

This limit is shown in Figures 8-3 and 8-4.

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Westinghouse Non-Proprietary Class 3 7-1 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [1], the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + RTNDT + Margin (10)

Initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [8]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.

RTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

RTNDT = CF

f (0.28 - 0.10 log f) (11)

To calculate RTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth:

f(depth x) = fsurface

e (-0.24x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 7.875 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the RTNDT at the specific depth.

The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves [2].

Table 7-1 contains the surface fluence values at 54 EFPY, which were used for the development of the P-T limit curves contained in this report. Table 7-1 also contains the 1/4T and 3/4T calculated fluence values and fluence factors (FFs), per Regulatory Guide 1.99, Revision 2. The values in this table will be used to calculate the 54 EFPY ART values for the Beaver Valley Unit 2 reactor vessel materials.

Margin is calculated as M = 2 I2 + 2 . The standard deviation for the initial RTNDT margin term (I) is 0°F when the initial RTNDT is a measured value, and 17°F when a generic value is available. The standard deviation for the RTNDT margin term, , is 17°F for plates or forgings when surveillance data is not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data is used.

For welds, is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14°F (half the value) when credible surveillance capsule data is used. The value for need not exceed 0.5 times the mean value of RTNDT.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-2 Contained in Tables 7-2 and 7-3 are the 54 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Beaver Valley Unit 2 heatup and cooldown curves.

The outlet nozzle forging weld materials for Beaver Valley Unit 2 have projected fluence values that do not exceed the 1 x 1017 n/cm2 fluence threshold at 54 EFPY per Table 2-8. The projected fluence values for the outlet nozzle forging weld materials provide conservative estimates of the fluence values of the outlet nozzles at the lowest extent of the nozzle; therefore, per NRC RIS 2014-11 [6], neutron radiation embrittlement need not be considered herein for the nozzle forging or weld materials. Thus, ART calculations for the outlet nozzle forging and weld materials utilizing the 1/4T and 3/4T fluence values are excluded from Tables 7-2 and 7-3, respectively.

WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 7-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Beaver Valley Unit 2 Reactor Vessel Materials at 54 EFPY Surface 1/4T f (a) 3/4T f (a)

Fluence, f (a) Surface 1/4T 3/4T Reactor Vessel Material (n/cm2, (n/cm2, (n/cm2, FF(b) FF(b) FF(b)

E > 1.0 MeV) E > 1.0 MeV)

E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell Plate B9004-1 6.28 1.445 3.92 1.352 1.52 1.116 Intermediate Shell Plate B9004-2 6.28 1.445 3.92 1.352 1.52 1.116 Lower Shell Plate B9005-1 6.30 1.115 3.93 1.352 1.53 1.117 Lower Shell Plate B9005-2 6.30 1.115 3.93 1.352 1.53 1.117 Intermediate to Lower Shell Girth 6.27 1.444 3.91 1.351 1.52 1.116 Weld Seam 101-171 Intermediate Shell Longitudinal 1.83 1.166 1.14 1.037 0.443 0.774 Weld Seams 101-124 A & B Lower Shell Longitudinal Weld 1.86 1.170 1.16 1.041 0.451 0.778 Seams 101-142 A & B Reactor Vessel Extended Beltline Materials(d)

Upper Shell Plate B9003-1 0.463 0.786 0.289 0.660 0.112 0.440 Upper Shell Plate B9003-2 0.463 0.786 0.289 0.660 0.112 0.440 Upper Shell Plate B9003-3 0.463 0.786 0.289 0.660 0.112 0.440 Upper to Intermediate Shell Girth 0.463 0.786 0.289 0.660 0.112 0.440 Weld Seam 103-121 Upper Shell Longitudinal Weld 0.463 0.786 0.289 0.660 0.112 0.440 Seams 101-122 A, B, &C Inlet Nozzle B9011-1 0.0153 0.145 Inlet Nozzle B9011-2 0.0153 0.145 Note (c)

Inlet Nozzle B9011-3 0.0153 0.145 Inlet Nozzle Welds 0.0153 0.145 Seams 105-121 A, B, & C Notes:

(a) 54 EFPY fluence values are documented in Table 2-8. 1/4T and 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.875 inches) and equation f = fsurf

e-0.24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).

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

(c) For conservatism, only the surface fluence is considered for the nozzle materials.

(d) Per [14], the fluence values for the extended beltline materials from Table 2-8 have been multiplied by a factor of 1.10 to account for uncertainty in the RAPTOR-M3G program in the extended beltline.

WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 1/4T Location R.G. 1.99, CF(a) 1/4T Fluence(b) 1/4T RTNDT(U)(c) RTNDT I (e) Margin ART Reactor Vessel Material Rev. 2 Position (°F) (n/cm , E > 1.0 MeV) 2 FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell B9004-1 1.1 40.5 3.92 1.352 60 54.7 0 17.0 34.0 148.7 Intermediate Shell B9004-2 1.1 37.0 3.92 1.352 40 50.0 0 17.0 34.0 124.0 Using Non-credible Beaver Valley Unit 2 2.1 61.7 3.92 1.352 40 83.4 0 17.0 34.0 157.4 Surveillance Data Lower Shell B9005-1 1.1 51.0 3.93 1.352 28 69.0 0 17.0 34.0 131.0 Lower Shell B9005-2 1.1 44.0 3.93 1.352 33 59.5 0 17.0 34.0 126.5 Intermediate to Lower Shell Girth Weld 1.1 34.4 3.91 1.351 -30 46.5 0 23.2 46.5 63.0 Seam 101-171 Using Credible Beaver Valley Unit 2 2.1 16.1 3.91 1.351 -30 21.8 0 10.9 21.8 13.5 Surveillance Data (Heat # 86342)

Intermediate Shell Longitudinal Weld 1.1 34.4 1.14 1.037 -30 35.7 0 17.8 35.7 41.3 Seams 101-124 A and B Using Credible Beaver Valley Unit 2 2.1 16.1 1.14 1.037 -30 16.7 0 8.3 16.7 3.4 Surveillance Data (Heat # 86342)

Lower Shell Longitudinal Weld 1.1 34.4 1.16 1.041 -30 35.8 0 17.9 35.8 41.6 Seams 101-142 A and B Using Credible Beaver Valley Unit 2 2.1 16.1 1.16 1.041 -30 16.8 0 8.4 16.8 3.5 Surveillance Data (Heat # 86342)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B9003-1 1.1 91.0 0.289 0.660 50 60.1 0 17.0 34.0 144.1 Upper Shell Plate B9003-2 1.1 83.0 0.289 0.660 60 54.8 0 17.0 34.0 148.8 Upper Shell Plate B9003-3 1.1 91.0 0.289 0.660 50 60.1 0 17.0 34.0 144.1 Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 73.71 0.289 0.660 -50 48.7 0 24.3 48.7 47.4 (Heat # 51912 (3490))

Upper Shell Longitudinal Weld Seam 101-122 A, B, and C 1.1 73.71 0.289 0.660 -70 48.7 0 24.3 48.7 27.4 (Heat # 51912 (3536))

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 27.0 0.289 0.660 10 17.8 17 8.9 38.4 66.2 (Heat # EAIB)

WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 7-5 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 1/4T Location R.G. 1.99, CF(a) 1/4T Fluence(b) 1/4T RTNDT(U)(c) RTNDT I (e) Margin ART Reactor Vessel Material Rev. 2 Position (°F) (n/cm2, E > 1.0 MeV) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 41.0 0.289 0.660 -30 27.1 0 13.5 27.1 24.2 (Heat # IAGA)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 68.0 0.289 0.660 10 44.9 17 22.5 56.3 111.2 (Heat # BOHB)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 27.0 0.289 0.660 -50 17.8 0 8.9 17.8 -14.3 (Heat # BAOED)

Intermediate to Upper Shell Girth Weld 1.1 122.0 0.289 0.660 -50 80.6 0 28.0 56.0 86.6 Seam 103-121 (Heat # 4P5174 (1122))

Intermediate to Upper Shell Girth Weld 1.1 68.0 0.289 0.660 -56 44.9 17 22.5 56.3 45.2 Seam 103-121 (Heat # 51922 (3489))

Intermediate to Upper Shell Girth Weld 1.1 41.0 0.289 0.660 -70 27.1 0 13.5 27.1 -15.8 Seam 103-121 (Heat # AAGC)

Intermediate to Upper Shell Girth Weld 1.1 41.0 0.289 0.660 -60 27.1 0 13.5 27.1 -5.8 Seam 103-121 (Heat # KOIB)

Inlet Nozzle B9011-1 1.1 77.0 0.0153(d) 0.145 60 11.2 0 5.6 11.2 82.4 Inlet Nozzle B9011-2 1.1 96.0 0.0153 (d) 0.145 60 13.9 17 7.0 36.7 110.7 Inlet Nozzle B9011-3 1.1 96.0 0.0153 (d) 0.145 70 13.9 0 7.0 13.9 97.9 Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 122.0 0.0153(d) 0.145 -50 17.7 0 8.9 17.7 -14.6 (Heat # 4P5174 (1122))

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 41.0 0.0153(d) 0.145 -60 6.0 0 3.0 6.0 -48.1 (Heat # LOHB)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -70 3.9 0 2.0 3.9 -62.2 (Heat # HABJC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -70 3.9 0 2.0 3.9 -62.2 (Heat # BABBD)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 41.0 0.0153(d) 0.145 -80 6.0 0 3.0 6.0 -68.1 (Heat # FABGC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -60 3.9 0 2.0 3.9 -52.2 (Heat # EOBC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 95.0 0.0153(d) 0.145 -60 13.8 0 6.9 13.8 -32.4 (Heat # FAAFC)

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-6 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 1/4T Location R.G. 1.99, CF(a) 1/4T Fluence(b) 1/4T RTNDT(U)(c) RTNDT I (e) Margin ART Reactor Vessel Material Rev. 2 Position (°F) (n/cm2, E > 1.0 MeV) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -60 3.9 0 2.0 3.9 -52.2 (Heat # CCJC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.015(d) 0.145 -30 3.9 0 2.0 3.9 -22.2 (Heat # FAGB)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -50 3.9 0 2.0 3.9 -42.2 (Heat # BAOED)

Notes:

(a) Data is from Table 5-2.

(b) Data is from Table 7-1.

(c) Data is from Table 3-1 (d) For conservatism, the surface fluence is considered for the inlet nozzle materials.

(e) As discussed in Section 4, the intermediate shell plate material surveillance data was deemed to be non-credible, while the surveillance weld data for Heat # 86342 surveillance data was deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or Position 2.1 with non-credible surveillance data and

= 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-7 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 3/4T Location R.G. 1.99, CF(a) 3/4T Fluence(b) 3/4T RTNDT(U)(c) RTNDT I (e) Margin ART Reactor Vessel Material Rev. 2 Position (°F) (n/cm , E > 1.0 MeV) 2 FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell B9004-1 1.1 40.5 1.52 1.116 60 45.2 0 17.0 34.0 139.2 Intermediate Shell B9004-2 1.1 37.0 1.52 1.116 40 41.3 0 17.0 34.0 115.3 Using Non-credible Beaver Valley Unit 2 2.1 61.7 1.52 1.116 40 68.9 0 17.0 34.0 142.9 Surveillance Data Lower Shell B9005-1 1.1 51.0 1.53 1.117 28 57.0 0 17.0 34.0 119.0 Lower Shell B9005-2 1.1 44.0 1.53 1.117 33 49.1 0 17.0 34.0 116.1 Intermediate to Lower Shell Girth Weld 1.1 34.4 1.52 1.116 -30 38.4 0 19.2 38.4 46.8 Seam 101-171 Using Credible Beaver Valley Unit 2 2.1 16.1 1.52 1.116 -30 18.0 0 9.0 18.0 5.9 Surveillance Data (Heat # 86342)

Intermediate Shell Longitudinal Weld 1.1 34.4 0.443 0.774 -30 26.6 0 13.3 26.6 23.2 Seams 101-124 A and B Using Credible Beaver Valley Unit 2 2.1 16.1 0.443 0.774 -30 12.5 0 6.2 12.5 -5.1 Surveillance Data (Heat # 86342)

Lower Shell Longitudinal Weld 1.1 34.4 0.451 0.778 -30 26.8 0 13.4 26.8 23.5 Seams 101-142 A and B Using Credible Beaver Valley Unit 2 2.1 16.1 0.451 0.778 -30 12.5 0 6.3 12.5 -4.9 Surveillance Data (Heat # 86342)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B9003-1 1.1 91.0 0.112 0.440 50 40.1 0 17.0 34.0 124.1 Upper Shell Plate B9003-2 1.1 83.0 0.112 0.440 60 36.5 0 17.0 34.0 130.5 Upper Shell Plate B9003-3 1.1 91.0 0.112 0.440 50 40.1 0 17.0 34.0 124.1 Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 73.71 0.112 0.440 -50 32.5 0 16.2 32.5 14.9 (Heat # 51912 (3490))

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 73.71 0.112 0.440 -70 32.5 0 16.2 32.5 -5.1 (Heat # 51912 (3536))

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 27.0 0.112 0.440 10 11.9 17 5.9 36.0 57.9 (Heat # EAIB)

WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 7-8 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 3/4T Location R.G. 1.99, CF(a) 3/4T Fluence(b) 3/4T RTNDT(U)(c) RTNDT I (e) Margin ART Reactor Vessel Material Rev. 2 Position (°F) (n/cm2, E > 1.0 MeV) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 41.0 0.112 0.440 -30 18.1 0 9.0 18.1 6.1 (Heat # IAGA)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 68.0 0.112 0.440 10 29.9 17 15.0 45.3 85.2 (Heat # BOHB)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 1.1 27.0 0.112 0.440 -50 11.9 0 5.9 11.9 -26.2 (Heat # BAOED)

Intermediate to Upper Shell Girth Weld 1.1 122.0 0.112 0.440 -50 53.7 0 26.9 53.7 57.4 Seam 103-121 (Heat # 4P5174 (1122))

Intermediate to Upper Shell Girth Weld 1.1 68.0 0.112 0.440 -56 29.9 17 15.0 45.3 19.2 Seam 103-121 (Heat # 51922 (3489))

Intermediate to Upper Shell Girth Weld 1.1 41.0 0.112 0.440 -70 18.1 0 9.0 18.1 -33.9 Seam 103-121 (Heat # AAGC)

Intermediate to Upper Shell Girth Weld 1.1 41.0 0.112 0.440 -60 18.1 0 9.0 18.1 -23.9 Seam 103-121 (Heat # KOIB)

Inlet Nozzle B9011-1 1.1 77.0 0.0153(d) 0.145 60 11.2 0 5.6 11.2 82.4 Inlet Nozzle B9011-2 1.1 96.0 0.0153 (d) 0.145 60 13.9 17 7.0 36.7 110.7 Inlet Nozzle B9011-3 1.1 96.0 0.0153 (d) 0.145 70 13.9 0 7.0 13.9 97.9 Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 122.0 0.0153(d) 0.145 -50 17.7 0 8.9 17.7 -14.6 (Heat # 4P5174 (1122))

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 41.0 0.0153(d) 0.145 -60 6.0 0 3.0 6.0 -48.1 (Heat # LOHB)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -70 3.9 0 2.0 3.9 -62.2 (Heat # HABJC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -70 3.9 0 2.0 3.9 -62.2 (Heat # BABBD)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 41.0 0.0153(d) 0.145 -80 6.0 0 3.0 6.0 -68.1 (Heat # FABGC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -60 3.9 0 2.0 3.9 -52.2 (Heat # EOBC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 95.0 0.0153(d) 0.145 -60 13.8 0 6.9 13.8 -32.4 (Heat # FAAFC)

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-9 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 54 EFPY at the 3/4T Location R.G. 1.99, CF(a) 3/4T Fluence(b) 3/4T RTNDT(U)(c) RTNDT I (e) Margin ART Reactor Vessel Material Rev. 2 Position (°F) (n/cm2, E > 1.0 MeV) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -60 3.9 0 2.0 3.9 -52.2 (Heat # CCJC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -30 3.9 0 2.0 3.9 -22.2 (Heat # FAGB)

Inlet Nozzle Weld Seams 105-121 A, B, & C 1.1 27.0 0.0153(d) 0.145 -50 3.9 0 2.0 3.9 -42.2 (Heat # BAOED)

Notes:

(a) Data is from Table 5-2.

(b) Data is from Table 7-1.

(c) Data is from Table 3-1.

(d) For conservatism, the surface fluence is considered for the inlet nozzle materials.

(e) As discussed in Section 4, the intermediate shell plate material surveillance data was deemed to be non-credible, while the surveillance weld data for Heat # 86342 surveillance data was deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal = 17°F for Position 1.1 or Position 2.1 with non-credible surveillance data and

= 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2 [1], the weld metal = 28°F for Position 1.1, and the weld metal = 14°F for Position 2.1 with credible surveillance data. However, need not exceed 0.5*RTNDT.

WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 8-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4 [2].

The P-T limit curves are developed both with and without margin for instrumentation uncertainties. Several uncertainties have been applied to the P-T limit curves. The wide range temperature uncertainty to be applied is 17°F. The wide range pressure uncertainty is 53 psi. Additionally, pressure corrections during startup and shutdown are applied depending on the number of reactor coolant pumps (RCPs) running.

During plant startup, when the indicated reactor coolant system (RCS) temperature is less than or equal to 137°F, the P-T limit curves should account for two RCPs in operation which is a pressure correction of 45 psid. If the indicated RCS temperature is greater than 137°F, then three RCPs should be considered in operation and a pressure correction of 62 psid should be applied.

During plant shutdown, when the indicated reactor coolant system (RCS) temperature is between 0°F and 500°F, the P-T limit curves should account for one RCP in operation which is a pressure correction of 32 psid. If the indicated RCS temperature is greater than or equal to 500°F, then three RCPs should be considered in operation and a pressure correction of 62 psid should be applied.

Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, -20, -40, -60, and -100°F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. Figure 8-3 presents the limiting heatup curves with margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. Figure 8-4 presents the limiting cooldown curves with margins for possible instrumentation errors using cooldown rates of 0,

-20, -40, -60, and -100°F/hr applicable for 54 EFPY, with the flange requirements and using the Axial Flaw methodology. The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 8-1 through 8-4. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

1.5 KIm < KIc (13)

WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 8-2

where, KIm is the stress intensity factor covered by membrane (pressure) stress [see page 6-2, Equation (3)],

KIc = 33.2 + 20.734 e [0.02 (T - RTNDT)] [see page 6-1 Equation (1)],

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation in order to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 6 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic leak tests for the Beaver Valley Unit 2 reactor vessel at 54 EFPY is 214°F without uncertainties and 234°F with uncertainties. This temperature is the minimum permissible temperature at which design pressure can be reached during a hydrostatic test per Equation (13). The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

The limiting ART values for Beaver Valley Unit 2 to be used in the generation of the P-T limit curves are based on Intermediate Shell Plate B9004-2 (Position 2.1). In order to provide an additional margin of conservatism, the limiting calculated ART values were rounded. The increased limiting ART values, using the Axial Flaw methodology, for Intermediate Shell Plates B9004-2 are summarized in Table 8-1.

Table 8-1 Summary of the Limiting ART Values Used in the Generation of the Beaver Valley Unit 2 Heatup and Cooldown Curves at 54 EFPY 1/4T Limiting ART(a) 3/4T Limiting ART(a) 158°F 143°F Intermediate Shell Forging B9004-2 (Position 2.1)

Note:

(a) The ART values used for P-T limit curve development in this report are the limiting ART values calculated in Tables 7-2 and 7-3 and rounded up to add additional margin.

The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 8-1 and 8-2 are presented in Tables 8-2 and 8-3. The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 8-3 and 8-4 are presented in Tables 8-4 and 8-5. The P-T limit curves shown WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-3 in Figures 8-1 through 8-4 were generated based on the limiting ART values for the cylindrical beltline and extended beltline reactor vessel materials rounded up. Vacuum refill limits for the Reactor Coolant System (RCS) are displayed on Figures 8-1 through 8-4 by showing a minimum pressure of 0 psia.

Inlet and Outlet Nozzles P-T Limit Curves NRC Regulatory Issue Summary (RIS) 2014-11 [6] requires that the P-T limit curves account for the higher stresses in the nozzle corner region due to the potential for more restrictive P-T limits, even if the RTNDT for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.

PWROG-15109-NP-A [20] addresses this concern generically for the U.S. pressurized water reactor (PWR) operating fleet. The results of PWROG-15109-NP-A demonstrate that P-T limit curves developed with current NRC-approved methods (e.g. WCAP-14040-A) bound the generic nozzle P-T limit curves. The NRC has issued a Safety Evaluation (SE) [21] which demonstrates that this is an acceptable means to address the concerns of RIS 2014-11. The results and conclusions of PWROG-15109-NP-A are applicable as long as the plant-specific Beaver Valley Unit 2 fluence of the nozzle corners remains less than the screening criterion of 4.28 x 1017 n/cm2, as described in PWROG-15109-NP-A. Table 2-8 demonstrates Beaver Valley Unit 2 adherence to this screening criterion, thus PWROG-15109-NP-A is applicable.

In conclusion, PWROG-15109-NP-A demonstrates that the nozzles will not be limiting with respect to the P-T limit curves at Beaver Valley Unit 2. Therefore, the concerns the concerns of RIS 2014-11 are adequately addressed.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate B9004-2 using Reg. Guide 1.99 Position 2.1 non-credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1/4T, 158°F (Axial Flaw) 3/4T, 143°F (Axial Flaw) 2500 OperlimAnalysis Version:5.4 Run:11714 Operlim.xlsm Version: 5.4.1 Leak Test Limit 2250 2000 Unacceptable Operation 1750 Critical Limit 60°F/Hr Heatup Rate 60°F/Hr 1500 Critical Limit Calculated Pressure (PSIG) 100°F/Hr Heatup Rate 100°F/Hr 1250 Acceptable Operation 1000 750 500 Criticality Limit based on inservice hydrostatic test temperature (214ºF) for the Boltup service period up to 54 EFPY 250 Temp.

0 Lower limit for RCS pressure is 0 psia

-250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-1 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-5 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate B9004-2 using Reg. Guide 1.99 Position 2.1 non-credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1/4T, 158°F (Axial Flaw) 3/4T, 143°F (Axial Flaw) 2500 OperlimAnalysis Version:5.4 Run:11714 Operlim.xlsm Version: 5.4.1 2250 2000 Unacceptable Operation 1750 1500 Calculated Pressure (PSIG) 1250 Acceptable Operation 1000 Cooldown 750 Rates 0°F/hr

-20°F/hr

-40°F/hr 500 -60°F/hr

-100°F/hr 250 0

Lower limit for RCS pressure is 0 psia

-250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-2 Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-6 Table 8-2 Beaver Valley Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 214 -14.7 60 -14.7 214 -14.7 60 621 214 1017 60 621 214 833 65 621 215 1025 65 621 215 839 70 621 220 1073 70 621 220 872 75 621 225 1127 75 621 225 908 80 621 230 1185 80 621 230 949 85 621 235 1250 85 621 235 994 90 621 240 1322 90 621 240 1044 95 621 245 1401 95 621 245 1099 100 621 250 1488 100 621 250 1160 105 621 255 1584 105 621 255 1227 110 621 260 1691 110 621 260 1301 115 621 265 1808 115 621 265 1384 120 621 270 1938 120 621 270 1474 120 726 275 2081 120 651 275 1574 125 741 280 2238 125 658 280 1685 130 757 285 2413 130 666 285 1807 135 776 - - 135 677 290 1941 140 797 - - 140 689 295 2090 145 820 - - 145 703 300 2253 150 846 - - 150 719 305 2434 155 875 - - 155 738 - -

160 907 - - 160 759 - -

165 943 - - 165 783 - -

170 982 - - 170 809 - -

175 1025 - - 175 839 - -

180 1073 - - 180 872 - -

185 1127 - - 185 908 - -

190 1185 - - 190 949 - -

195 1250 - - 195 994 - -

200 1322 - - 200 1044 - -

205 1401 - - 205 1099 - -

210 1488 - - 210 1160 - -

215 1584 - - 215 1227 - -

220 1691 - - 220 1301 - -

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-7 Table 8-2 Beaver Valley Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 225 1808 - - 225 1384 - -

230 1938 - - 230 1474 - -

235 2081 - - 235 1574 - -

240 2238 - - 240 1685 - -

245 2413 - - 245 1807 - -

- - - - 250 1941 - -

- - - - 255 2090 - -

- - - - 260 2253 - -

- - - - 265 2434 - -

Leak Test Limit T (°F) P (psig) 196 2000 214 2485 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-8 Table 8-3 Beaver Valley Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 621 60 621 60 617 60 577 60 495 65 621 65 621 65 621 65 584 65 503 70 621 70 621 70 621 70 592 70 512 75 621 75 621 75 621 75 600 75 522 80 621 80 621 80 621 80 610 80 533 85 621 85 621 85 621 85 621 85 546 90 621 90 621 90 621 90 621 90 560 95 621 95 621 95 621 95 621 95 575 100 621 100 621 100 621 100 621 100 592 105 621 105 621 105 621 105 621 105 612 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 120 826 120 796 120 767 120 738 120 683 125 846 125 817 125 789 125 763 125 713 130 868 130 841 130 815 130 790 130 745 135 892 135 867 135 843 135 821 135 781 140 918 140 895 140 874 140 854 140 822 145 947 145 927 145 908 145 892 145 866 150 980 150 962 150 946 150 933 150 916 155 1016 155 1001 155 989 155 979 155 971 160 1055 160 1044 160 1035 160 1030 160 1030 165 1099 165 1091 165 1087 165 1086 165 1086 170 1147 170 1144 170 1144 170 1144 170 1144 175 1201 175 1201 175 1201 175 1201 175 1201 180 1260 180 1260 180 1260 180 1260 180 1260 185 1325 185 1325 185 1325 185 1325 185 1325 190 1397 190 1397 190 1397 190 1397 190 1397 195 1477 195 1477 195 1477 195 1477 195 1477 200 1565 200 1565 200 1565 200 1565 200 1565 205 1662 205 1662 205 1662 205 1662 205 1662 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-9 Table 8-3 Beaver Valley Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 210 1769 210 1769 210 1769 210 1769 210 1769 215 1888 215 1888 215 1888 215 1888 215 1888 220 2020 220 2020 220 2020 220 2020 220 2020 225 2165 225 2165 225 2165 225 2165 225 2165 230 2325 230 2325 230 2325 230 2325 230 2325 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-10 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate B9004-2 using Reg. Guide 1.99 Position 2.1 non-credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1/4T, 158°F (Axial Flaw) 3/4T, 143°F (Axial Flaw) 2500 OperlimAnalysis Version:5.4 Run:11714 Operlim.xlsm Version: 5.4.1 2250 Leak Test Limit 2000 Unacceptable Operation 1750 Critical Limit 60°F/Hr Heatup Rate 60°F/Hr 1500 Critical Limit Calculated Pressure (PSIG) 100°F/Hr Heatup Rate 100°F/Hr 1250 Acceptable Operation 1000 750 500 Criticality Limit based on inservice hydrostatic test temperature (234ºF) for the Boltup service period up to 54 EFPY 250 Temp.

0 Lower limit for RCS pressure is 0 psia

-250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-3 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and with Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-11 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate B9004-2 using Reg. Guide 1.99 Position 2.1 non-credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1/4T, 158°F (Axial Flaw) 3/4T, 143°F (Axial Flaw) 2500 OperlimAnalysis Version:5.4 Run:11714 Operlim.xlsm Version: 5.4.1 2250 2000 Unacceptable Operation 1750 1500 Calculated Pressure (PSIG) 1250 Acceptable Operation 1000 750 Cooldown 500 Rates 0°F/hr

-20°F/hr

-40°F/hr 250 -60°F/hr

-100°F/hr 0

Lower limit for RCS pressure is 0 psia

-250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-4 Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange Requirements and with Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-12 Table 8-4 Beaver Valley Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/ Margins for Instrumentation Errors(a))

60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 77 -14.7 234 -14.7 77 -14.7 234 -14.7 77 523 234 934 77 523 234 740 82 523 237 958 82 523 237 757 87 523 242 1012 87 523 242 793 92 523 247 1070 92 523 247 834 97 523 252 1135 97 523 252 879 102 523 257 1207 102 523 257 929 107 523 262 1286 107 523 262 984 112 523 267 1373 112 523 267 1045 117 523 272 1469 117 523 272 1112 122 523 277 1576 122 523 277 1186 127 523 282 1693 127 523 282 1269 132 523 287 1823 132 523 287 1359 137 523 292 1966 137 523 292 1459 137 611 297 2123 137 536 297 1570 142 626 302 2298 142 543 302 1692 147 642 - - 147 551 307 1826 152 661 - - 152 562 312 1975 157 682 - - 157 574 317 2138 162 705 - - 162 588 322 2319 167 731 - - 167 604 - -

172 760 - - 172 623 - -

177 792 - - 177 644 - -

182 828 - - 182 668 - -

187 867 - - 187 694 - -

192 910 - - 192 724 - -

197 958 - - 197 757 - -

202 1012 - - 202 793 - -

207 1070 - - 207 834 - -

212 1135 - - 212 879 - -

217 1207 - - 217 929 - -

222 1286 - - 222 984 - -

227 1373 - - 227 1045 - -

232 1469 - - 232 1112 - -

237 1576 - - 237 1186 - -

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-13 Table 8-4 Beaver Valley Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/ Margins for Instrumentation Errors(a))

60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 242 1693 - - 242 1269 - -

247 1823 - - 247 1359 - -

252 1966 - - 252 1459 - -

257 2123 - - 257 1570 - -

262 2298 - - 262 1692 - -

- - - - 267 1826 - -

- - - - 272 1975 - -

- - - - 277 2138 - -

- - - - 282 2319 - -

Leak Test Limit T (°F) P (psig) 218 2000 234 2485 Note:

(a) The error includes an adjustment of 17°F for instrumented temperature and 53 psid for instrumented pressure. The pressure is also adjusted by 45 psid if the indicated temperature is less than or equal to 137°F and 62 psid if greater than 137°F.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-14 Table 8-5 Beaver Valley Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/ Margins for Instrumentation Errors(a))

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 77 -14.7 77 -14.7 77 -14.7 77 -14.7 77 -14.7 77 536 77 536 77 532 77 492 77 410 82 536 82 536 82 536 82 499 82 418 87 536 87 536 87 536 87 507 87 427 92 536 92 536 92 536 92 515 92 437 97 536 97 536 97 536 97 525 97 448 102 536 102 536 102 536 102 536 102 461 107 536 107 536 107 536 107 536 107 475 112 536 112 536 112 536 112 536 112 490 117 536 117 536 117 536 117 536 117 507 122 536 122 536 122 536 122 536 122 527 127 536 127 536 127 536 127 536 127 536 132 536 132 536 132 536 132 536 132 536 137 536 137 536 137 536 137 536 137 536 137 741 137 711 137 682 137 653 137 598 142 761 142 732 142 704 142 678 142 628 147 783 147 756 147 730 147 705 147 660 152 807 152 782 152 758 152 736 152 696 157 833 157 810 157 789 157 769 157 737 162 862 162 842 162 823 162 807 162 781 167 895 167 877 167 861 167 848 167 831 172 931 172 916 172 904 172 894 172 886 177 970 177 959 177 950 177 945 177 945 182 1014 182 1006 182 1002 182 1001 182 1001 187 1062 187 1059 187 1059 187 1059 187 1059 192 1116 192 1116 192 1116 192 1116 192 1116 197 1175 197 1175 197 1175 197 1175 197 1175 202 1240 202 1240 202 1240 202 1240 202 1240 207 1312 207 1312 207 1312 207 1312 207 1312 212 1392 212 1392 212 1392 212 1392 212 1392 217 1480 217 1480 217 1480 217 1480 217 1480 222 1577 222 1577 222 1577 222 1577 222 1577 WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-15 Table 8-5 Beaver Valley Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc, w/ Flange Requirements, and w/ Margins for Instrumentation Errors(a))

Steady-State 20°F/hr Cooldown 40°F/hr Cooldown 60°F/hr Cooldown 100°F/hr Cooldown T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 227 1684 227 1684 227 1684 227 1684 227 1684 232 1803 232 1803 232 1803 232 1803 232 1803 237 1935 237 1935 237 1935 237 1935 237 1935 242 2080 242 2080 242 2080 242 2080 242 2080 247 2240 247 2240 247 2240 247 2240 247 2240 252 2417 252 2417 252 2417 252 2417 252 2417 Note:

(a) The error includes an adjustment of 17°F for instrumented temperature and 53 psid for instrumented pressure.

The pressure is also adjusted by 32 psid if the indicated temperature is less than 500°F and 62 psid if greater than or equal to 500°F.

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988. [Agencywide Document Management System (ADAMS) Accession Number ML003740284]

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

[ADAMS Accession Number ML050120209]

3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

4. Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, November 29, 2019.

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

6. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S. Nuclear Regulatory Commission, October 2014. [ADAMS Accession Number ML14149A165]

7. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, Fracture Toughness Requirements, Revision 4, U.S. Nuclear Regulatory Commission, March 2019. [ADAMS Accession Number ML18338A516]

8. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Class 1 Components.

9. Westinghouse Report, WCAP-16527-NP, Supplement 1, Rev. 1, Analysis of Capsule X from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program, September 2011.

10. Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

11. Combustion Engineering Document, CE-NPSD-1039, Rev. 02, Best Estimate Copper and Nickel Values in CE Fabricate Reactor Vessel Welds, June 1997.

12. Westinghouse Report, WCAP-16528-NP, Rev. 1, Beaver Valley Power Station Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, June 2008.

13. Westinghouse Report, WCAP-18558-NP, Rev. 0, Analysis of Capsule Y from the Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program, June 2020.

14. Westinghouse Letter, LTR-REA-20-59-NP, Rev. 0, Justification of Using RAPTOR-M3G for Reactor Pressure Vessel Extended Beltline Materials at Beaver Valley Units 1 and 2, June 18, 2020. [Attached to document with ADAMS Accession Number ML20304A215]

15. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number ML110070570]

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-2

16. ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results, 2018.

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

18. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession Number ML010890301]

19. Westinghouse Report WCAP-18124-NP-A, Rev. 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018. [ADAMS Accession Number ML18204A010]

20. Pressurized Water Reactor Owners Group (PWROG) Report PWROG-15109-NP-A, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]

21. NRC Safety Evaluation Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report PWROG-15109-NP, Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation (EPID L-2018-TOP-0009), October 31, 2019. [ADAMS Accession Numbers ML19301D063 &

ML19301D160]

WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIt)

Tables A-1 and A-2 contain the thermal stress intensity factors (KIt) for the maximum heatup and cooldown rates at 54 EFPY for Beaver Valley Unit 2. The reactor vessel cylindrical shell radii to the 1/4T and 3/4T locations are as follows:

1/4T Radius = 80.629 inches

3/4T Radius = 84.566 inches WCAP-18559-NP March 2021 Revision 1

- - This record was final approved on 3/29/2021 4:31:48 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-2 Table A-1 KIt Values for Beaver Valley Unit 2 at 54 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors)

Water Vessel Temperature 1/4T Thermal Stress Vessel Temperature 3/4T Thermal Stress Temp. at 1/4T Location for Intensity Factor at 3/4T Location for Intensity Factor

(°F) 100°F/hr Heatup (°F) (ksi in.) 100°F/hr Heatup (°F) (ksi in.)

60 56.130 -0.987 55.065 0.493 65 58.927 -2.377 55.425 1.455 70 62.129 -3.521 56.315 2.377 75 65.562 -4.586 57.748 3.208 80 69.262 -5.475 59.641 3.929 85 73.079 -6.273 61.944 4.558 90 77.089 -6.948 64.601 5.101 95 81.193 -7.553 67.562 5.578 100 85.435 -8.069 70.788 5.991 105 89.755 -8.531 74.238 6.353 110 94.171 -8.928 77.881 6.671 115 98.650 -9.285 81.690 6.951 120 103.196 -9.594 85.642 7.198 125 107.791 -9.875 89.717 7.418 130 112.433 -10.118 93.898 7.612 135 117.114 -10.341 98.171 7.785 140 121.829 -10.535 102.523 7.940 145 126.574 -10.715 106.944 8.080 150 131.343 -10.873 111.424 8.206 155 136.136 -11.020 115.955 8.320 160 140.945 -11.151 120.530 8.423 165 145.773 -11.275 125.142 8.519 170 150.613 -11.385 129.788 8.606 175 155.467 -11.491 134.463 8.687 180 160.330 -11.586 139.161 8.762 185 165.204 -11.678 143.881 8.833 190 170.083 -11.763 148.620 8.899 195 174.972 -11.845 153.374 8.961 200 179.864 -11.920 158.143 9.020 205 184.764 -11.995 162.923 9.077 210 189.666 -12.064 167.713 9.131 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 A-3 Table A-2 KIt Values for Beaver Valley Unit 2 at 54 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors)

Water Vessel Temperature at 1/4T 100°F/hr Cooldown Temp. Location for 100°F/hr 1/4T Thermal Stress

(°F) Cooldown (°F) Intensity Factor (ksi in.)

210 232.426 13.510 205 227.352 13.454 200 222.278 13.398 195 217.204 13.342 190 212.131 13.286 185 207.057 13.230 180 201.983 13.175 175 196.909 13.119 170 191.835 13.063 165 186.762 13.008 160 181.688 12.952 155 176.615 12.897 150 171.541 12.842 145 166.468 12.786 140 161.395 12.731 135 156.322 12.676 130 151.249 12.622 125 146.176 12.567 120 141.103 12.512 115 136.031 12.457 110 130.958 12.403 105 125.886 12.349 100 120.813 12.295 95 115.741 12.240 90 110.669 12.187 85 105.597 12.133 80 100.525 12.079 75 95.454 12.025 70 90.382 11.972 65 85.311 11.919 60 80.241 11.865 WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B OTHER RCPB FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [B-1], requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement (LST) for all RCPB components, which is specified in NB-3211 and NB-2332(b) of the Section III ASME Code [B-2], is the relevant requirement that would affect the pressure-temperature (P-T) limits.

This requirement is applicable to ferritic materials outside of the reactor vessel with a nominal wall thickness greater than 2 1/2 inches, such as piping, pumps and valves. The Beaver Valley Unit 2 reactor coolant system components do not contain ferritic materials in the Class 1 piping, pumps and valves.

Therefore, the LST requirements of NB-2332(b) and NB-3211 are not applicable to the Beaver Valley Unit 2 P-T limits.

The other ferritic RCPB components that are not part of the reactor vessel beltline or extended beltline consist of the reactor vessel closure head, pressurizer, and steam generators.

RIS 2014-11 also addresses other ferritic components of the reactor coolant system relative to P-T limits, and states the following:

As specified in Sections I and IV.A of 10 CFR Part 50, Appendix G, ferritic RCPB components outside of the reactor vessel must meet the applicable requirements of ASME Code,Section III, Rules for Construction of Nuclear Facility Components.

The reactor vessel closure head flange materials have been considered in the development of P-T limits, see Section 6.3 of this report for further detail. Furthermore, the reactor vessel closure head was constructed to the 1971 Edition through 1972 Summer Addenda of Section III of the ASME Code and has met all applicable requirements at the time of construction. The steam generators and pressurizer were also constructed to the 1971 Edition through 1972 Summer Addenda of Section III of the ASME Code and have met all applicable requirements at the time of construction. These Beaver Valley Unit 2 primary system components are analyzed to the identified ASME Code Section III Editions and met all applicable requirements at the time of construction. In addition, these components have not undergone neutron embrittlement. Therefore, no further consideration is necessary for these components with regard to P-T limits.

B.1 REFERENCES B-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

B-2 ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Class 1 Components.

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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C UPPER-SHELF ENERGY EVALUATION Charpy upper-shelf energy (USE) is associated with the determination of acceptable reactor pressure vessel (RPV) toughness during the licensed operating period.

The requirements on USE are included in 10 CFR 50, Appendix G [C-1]. 10 CFR 50, Appendix G requires utilities to submit an analysis at least three years prior to the time that the USE of any RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.

There are two methods that can be used to predict the decrease in USE with irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide 1.99, Revision 2 [C-2].

For vessel materials that are not in the surveillance program or have non-credible data, the Charpy USE (Position 1.2) is assumed to decrease as a function of fluence and copper content, as indicated in Figure 2 of Regulatory Guide 1.99, Revision 2.

When two or more credible surveillance sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material. The surveillance data are then used in conjunction with the Regulatory Guide to predict the change in USE (Position 2.2) of the RPV material due to irradiation by plotting the reduced plant surveillance data on Figure 2 of the Guide (Figure C-1 of this report) and fitting the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.

The 54 EFPY (EOLE) Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2.

The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection.

The projected USE values were calculated to determine if the Beaver Valley Unit 2 beltline and extended beltline materials remain above the 50 ft-lb criterion at 54 EFPY (EOLE). These calculations are summarized in Table C-1. Note, even though surveillance data for Intermediate Shell Plate B9004-2 is non-credible, it is still used in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2. The data from the surveillance materials are determined to be non-credible by Credibility Criterion 3. Credibility Criterion 3 indicates that even if the surveillance data are not considered credible for determination of RTNDT, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82.

USE Conclusion For Beaver Valley Unit 2, the limiting USE value at 54 EFPY is 57.7 ft-lb (see Table C-1); this value corresponds to Lower Shell Plates B9005-1 and B9005-2. Therefore, all of the beltline and extended beltline materials in the Beaver Valley Unit 2 reactor vessel are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G) through EOLE (54 EFPY).

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Westinghouse Non-Proprietary Class 3 C-2 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for Beaver Valley Unit 2 EOLE 1/4T Projected Wt % Fluence(b) Initial USE(a) Projected USE EOLE Reactor Vessel Material Cu (a)

(x 10 n/cm ,

19 2 (ft-lb) Decrease (%) USE E > 1.0 MeV) (ft-lb)

Position 1.2(c)

Reactor Vessel Beltline Materials Intermediate Shell Plate B9004-1 0.065 3.92 83 26 61.4 Intermediate Shell Plate B9004-2 0.06 3.92 79 26 58.5 Lower Shell Plate B9005-1 0.08 3.93 82 26 60.7 Lower Shell Plate B9005-2 0.07 3.93 78 26 57.7 Intermediate Shell Longitudinal Weld 0.046 1.14 145 20 116.0 Seams 101-124 A and B (Heat # 83642)

Lower Shell Longitudinal Weld 0.046 1.16 145 20 116.0 Seams 101-142 A and B (Heat # 83642)

Intermediate to Lower Shell Girth Weld Seam 0.046 3.91 145 26 107.3 101-171 (Heat # 83642)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B9003-1 0.13 0.289 98 17 81.3 Upper Shell Plate B9003-2 0.12 0.289 80 16 67.2 Upper Shell Plate B9003-3 0.13 0.289 98 17 81.3 Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 0.156 0.289 96 23 73.9 (Heat # 51912 (3490))

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 0.156 0.289 114 23 87.8 (Heat # 51912 (3536))

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 0.02 0.289 118 14 101.5 (Heat # EAIB)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 0.03 0.289 160 14 137.6 (Heat # IAGA)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 0.05 0.289 97 14 83.4 (Heat # BOHB)

Upper Shell Longitudinal Weld Seams 101-122 A, B, and C 0.02 0.289 150 14 129.0 (Heat # BAOED)

Intermediate to Upper Shell Girth Weld Seam 0.09 0.289 70 17 58.1 103-121 (Heat # 4P5174 (1122))

Intermediate to Upper Shell Girth Weld Seam 0.05 0.289 102 14 87.7 103-121 (Heat # 51922 (3489))

Intermediate to Upper Shell Girth Weld Seam 0.03 0.289 111 14 95.5 103-121 (Heat # AAGC)

Intermediate to Upper Shell Girth Weld Seam 0.03 0.289 110 14 94.6 103-121 (Heat # KOIB)

Inlet Nozzle B9011-1 0.11 0.0153 68.25 9 62.1 Inlet Nozzle B9011-2 0.13 0.0153 75.4 9 68.6 Inlet Nozzle B9011-3 0.13 0.0153 79.3 9 72.2 Inlet Nozzle Weld Seams 105-121 A, B, & C 0.09 0.0153 70 10 63.0 (Heat # 4P5174 (1122))

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Westinghouse Non-Proprietary Class 3 C-3 Table C-1 Predicted Position 1.2 and 2.2 USE Values at 54 EFPY (EOLE) for Beaver Valley Unit 2 EOLE 1/4T Projected Wt % Fluence(b) Initial USE (a)

Projected USE EOLE Reactor Vessel Material Cu(a) (x 1019 n/cm2, (ft-lb) Decrease (%) USE E > 1.0 MeV) (ft-lb)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.03 0.0153 137 8 126.0 (Heat # LOHB)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.02 0.0153 162 8 149.0 (Heat # HABJC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.02 0.0153 142 8 130.6 (Heat # BABBD)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.03 0.0153 119 8 109.5 (Heat # FABGC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.02 0.0153 127 8 116.8 (Heat # EOBC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.07 0.0153 119 9 108.3 (Heat # FAAFC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.02 0.0153 109 8 100.3 (Heat # CCJC)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.02 0.0153 114 8 104.9 (Heat # FAGB)

Inlet Nozzle Weld Seams 105-121 A, B, & C 0.02 0.0153 150 8 138.0 (Heat # BAOED)

Position 2.2(d)

Intermediate Shell Plate B9004-2 0.06 3.92 79 24 60.0 Intermediate Shell Longitudinal Weld 0.046 1.14 145 11 129.1 Seams 101-124 A and B (Heat # 83642)

Lower Shell Longitudinal Weld 0.046 1.16 145 11 129.1 Seams 101-142 A and B (Heat # 83642)

Intermediate to Lower Shell Girth Weld Seam 0.046 3.91 145 15 123.3 101-171 (Heat # 83642)

Notes:

(a) Data taken from Table 3-1 of this report. If the base metal and weld Cu weight percentages are below the minimum value presented in Figure 2 of [C-2] (0.1 for base metal and 0.05 for welds), then the Cu weight percentages were conservatively rounded up to the minimum value.

(b) Values taken from Table 7-1. Fluence values above 1017 n/cm2 (E > 1.0 MeV) but below 2 x 1017 n/cm2 (E > 1.0 MeV) were rounded to 2 x 1017 n/cm2 (E > 1.0 MeV) when determining the % decrease because 2 x 1017 n/cm2 is the lowest fluence displayed in Figure 2 of Regulatory Guide 1.99, Revision 2 [C-2].

(c) Percentage USE decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2 [C-2] and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide and using the material-specific Cu wt. % values. The percent-loss lines were extended into the low fluence area of Regulatory Guide 1.99, Revision 2, Figure 2, i.e., below 1018 n/cm2, in order to determine the USE % decrease as needed.

(d) Percentage USE decrease values are based on Position 2.2 of [C-2]. Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of [C-2]) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE. Note the measured USE percent decrease is shown in Table 4-1.

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Westinghouse Non-Proprietary Class 3 C-4

% Copper Limiting Plate Percent USE Base Metal Decrease 29% from Capsule Y 0.35 Lon itudinal Orientation Plate Line w

U)

C:

Weld Line C.

0 I.,

C (1)

C>

ns 10.0

'E (1)

(.J I.,

(1)

a. Low er Shell Plate

.1:ai_..=-_ _.,.__ _t--t---+-t-+-+-1-:::::t:;. --=:;;;- - - + - --tt-- 1/4T Fluence = 3.93 x 10 19 n/crn2 and lnterrrediate Shell Plate 1/4T Fluence = 3.92 x 10 19 n/crn2 1--------,,::=-1..-=-----,f------,f---+----,-1--1-+-+-----+---1+-- and Upper Shell Plates and all Upper lnterrrediate Shell to Low er Shell Shell Welds Girth Weld 1/4T Fluence = 2 .89 x 10 18 n/crn2 1/4T Fluence = 3.91 x 10 19 n/crn2 I I I I I Surveillance Material:

Intermediate Shell Inlet Nozzles and Welds lnterrrediate Shell Longitudinal Welds *-1---- ----------,-- Plate B9004-2 1/4T Fluence = 2 x 101 7 n/crn2 1/4T Fluence = 1.14 x 10 19 n/crn2

Surveillance Material:

Note, fluence was rounded up from and Weld Heat # 83642 1.53 x 1017 n/crn2 to correspond to the Low er Shell Longitudinal Welds Re * *

alue. 1/4T Fluence = 1.16 x 10 19 n/crn2 1.0 +--_ _ _ __ _ _..___..____.__..__...__...__......._ _ _ ___.__ _.......__ __.__.....___,___.___.___._-+-_ _ _ _.......__ _ ...___ ,.____,____...__..__........._

1.00E+17 1.00E+18 1.00E+19 1.00E+20 Neutron Fluence, n/cm 2 (E > 1 MeV)

Figure C-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence WCAP-18559-NP March 2021 Revision 1

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Westinghouse Non-Proprietary Class 3 C-5 C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

C-2 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

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Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D PRESSURIZED THERMAL SHOCK AND EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION D.1 PRESSURIZED THERMAL SHOCK Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the reactor pressure vessel (RPV) under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high pressurization, significant degradation of vessel material toughness caused by radiation embrittlement, and the presence of a critical-size defect anywhere within the vessel wall.

In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [D-1]) that established screening criteria on Pressurized Water Reactor (PWR) vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS.

RTPTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license.

The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTPTS) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 [D-2].

These accepted methods were used with the clad/base metal interface fluence values of Section 2 to calculate the following RTPTS values for the Beaver Valley Unit 2 RPV materials at 54 EFPY (EOLE). The EOLE RTPTS calculations are summarized in Table D-1.

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Westinghouse Non-Proprietary Class 3 D-2 Table D-1 RTPTS Calculations for the Beaver Valley Unit 2 Reactor Vessel Materials at 54 EFPY Surface Fluence(b)

CF(a) Surface RTNDT(U)(c) RTNDT U (d) Margin RTPTS Reactor Vessel Material (x 1019 n/cm2,

(°F) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell B9004-1 40.5 6.28 1.445 60 58.5 0 17.0 34.0 152.5 Intermediate Shell B9004-2 37.0 6.28 1.445 40 53.5 0 17.0 34.0 127.5 Using Non-credible Beaver Valley Unit 2 61.7 6.28 1.445 40 89.1 0 17.0 34.0 163.1 Surveillance Data Lower Shell B9005-1 51.0 6.30 1.445 28 73.7 0 17.0 34.0 135.7 Lower Shell B9005-2 44.0 6.30 1.445 33 63.6 0 17.0 34.0 130.6 Intermediate to Lower Shell Girth Weld 34.4 6.27 1.444 -30 49.7 0 24.8 49.7 69.4 Seam 101-171 Using Credible Beaver Valley Unit 2 16.1 6.27 1.444 -30 23.3 0 11.6 23.3 16.5 Surveillance Data (Heat # 86342)

Intermediate Shell Longitudinal Weld 34.4 1.83 1.166 -30 40.1 0 20.1 40.1 50.2 Seams 101-124 A and B Using Credible Beaver Valley Unit 2 16.1 1.83 1.166 -30 18.8 0 9.4 18.8 7.5 Surveillance Data (Heat # 86342)

Lower Shell Longitudinal Weld 34.4 1.86 1.170 -30 40.2 0 20.1 40.2 50.5 Seams 101-142 A and B Using Credible Beaver Valley Unit 2 16.1 1.86 1.170 -30 18.8 0 9.4 18.8 7.7 Surveillance Data (Heat # 86342)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B9003-1 91.0 0.463 0.786 50 71.5 0 17.0 34.0 155.5 Upper Shell Plate B9003-2 83.0 0.463 0.786 60 65.2 0 17.0 34.0 159.2 Upper Shell Plate B9003-3 91.0 0.463 0.786 50 71.5 0 17.0 34.0 155.5 Upper Shell Longitudinal Weld Seam 101-122 73.71 0.463 0.786 -50 57.9 0 28.0 56.0 63.9 A, B, and C (Heat # 51912 (3490))

Upper Shell Longitudinal Weld Seam 101-122 73.71 0.463 0.786 -70 57.9 0 28.0 56.0 43.9 A, B, and C (Heat # 51912 (3536))

Upper Shell Longitudinal Weld Seam 101-122 27.0 0.463 0.786 10 21.2 17 10.6 40.1 71.3 A, B, and C (Heat # EAIB)

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Westinghouse Non-Proprietary Class 3 D-3 Table D-1 RTPTS Calculations for the Beaver Valley Unit 2 Reactor Vessel Materials at 54 EFPY Surface Fluence(b)

CF(a) Surface RTNDT(U)(c) RTNDT U (d) Margin RTPTS Reactor Vessel Material (x 1019 n/cm2,

(°F) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

E > 1.0 MeV)

Upper Shell Longitudinal Weld Seam 101-122 41.0 0.463 0.786 -30 32.2 0 16.1 32.2 34.4 A, B, and C (Heat # IAGA)

Upper Shell Longitudinal Weld Seam 101-122 68.0 0.463 0.786 10 53.4 17 26.7 63.3 126.7 A, B, and C (Heat # BOHB)

Upper Shell Longitudinal Weld Seam 101-122 27.0 0.463 0.786 -50 21.2 0 10.6 21.2 -7.6 A, B, and C (Heat # BAOED)

Intermediate to Upper Shell Girth Weld Seam 122.0 0.463 0.786 -50 95.8 0 28.0 56.0 101.8 103-121 (Heat # 4P5174 (1122))

Intermediate to Upper Shell Girth Weld Seam 68.0 0.463 0.786 -56 53.4 17 26.7 63.3 60.7 103-121 (Heat # 51922 (3489))

Intermediate to Upper Shell Girth Weld Seam 41.0 0.463 0.786 -70 32.2 0 16.1 32.2 -5.6 103-121 (Heat # AAGC)

Intermediate to Upper Shell Girth Weld Seam 41.0 0.463 0.786 -60 32.2 0 16.1 32.2 4.4 103-121 (Heat # KOIB)

Inlet Nozzle B9011-1 77.0 0.0153 0.145 60 11.2 0 5.6 11.2 82.4 Inlet Nozzle B9011-2 96.0 0.0153 0.145 60 13.9 17 7.0 36.7 110.7 Inlet Nozzle B9011-3 96.0 0.0153 0.145 70 13.9 0 7.0 13.9 97.9 Inlet Nozzle Welds 105-121 A, B, & C 122.0 0.0153 0.145 -50 17.7 0 8.9 17.7 -14.6 (Heat # 4P5174 (1122))

Inlet Nozzle Welds 105-121 A, B, & C 41.0 0.0153 0.145 -60 6.0 0 3.0 6.0 -48.1 (Heat # LOHB)

Inlet Nozzle Welds 105-121 A, B, & C 27.0 0.0153 0.145 -70 3.9 0 2.0 3.9 -62.2 (Heat # HABJC)

Inlet Nozzle Welds 105-121 A, B, & C 27.0 0.0153 0.145 -70 3.9 0 2.0 3.9 -62.2 (Heat # BABBD)

Inlet Nozzle Welds 105-121 A, B, & C 41.0 0.0153 0.145 -80 6.0 0 3.0 6.0 -68.1 (Heat # FABGC)

Inlet Nozzle Welds 105-121 A, B, & C 27.0 0.0153 0.145 -60 3.9 0 2.0 3.9 -52.2 (Heat # EOBC)

Inlet Nozzle Welds 105-121 A, B, & C 95.0 0.0153 0.145 -60 13.8 0 6.9 13.8 -32.4 (Heat # FAAFC)

Inlet Nozzle Welds 105-121 A, B, & C 27.0 0.0153 0.145 -60 3.9 0 2.0 3.9 -52.2 (Heat # CCJC)

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Westinghouse Non-Proprietary Class 3 D-4 Table D-1 RTPTS Calculations for the Beaver Valley Unit 2 Reactor Vessel Materials at 54 EFPY Surface Fluence(b)

CF(a) Surface RTNDT(U)(c) RTNDT U (d) Margin RTPTS Reactor Vessel Material (x 1019 n/cm2,

(°F) FF(b) (°F) (°F) (°F) (°F) (°F) (°F)

E > 1.0 MeV)

Inlet Nozzle Welds 105-121 A, B, & C 27.0 0.0153 0.145 -30 3.9 0 2.0 3.9 -22.2 (Heat # FAGB)

Inlet Nozzle Welds 105-121 A, B, & C 27.0 0.0153 0.145 -50 3.9 0 2.0 3.9 -42.2 (Heat # BAOED)

Notes:

(a) Data is from Table 5-2.

(b) Data is from Table 7-1.

(c) Data is from Table 3-1.

(d) The credibility conclusion for the surveillance material is discussed in Section 4. The intermediate shell plate material surveillance data was determined to be non-credible, while the weld Heat # 86342 surveillance data were determined to be credible. Per the guidance of 10 CFR 50.61 [D-1], the base metal = 17°F when surveillance data is non-credible or not used to determine the CF, and the weld metal = 28°F when surveillance data is not used to determine the CF and = 14°F when credible surveillance data is used to determine the CF. However, need not exceed 0.5*RTNDT per regulatory guidance in [D-1].

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Westinghouse Non-Proprietary Class 3 D-5 PTS Conclusion The Beaver Valley Unit 2 limiting RTPTS value for base metal or longitudinal weld materials at 54 EFPY is 163.1°F (see Table D-1), which corresponds to Intermediate Shell Plate B9004-2 when considering non-credible surveillance data (using Position 2.1). The Beaver Valley Unit 2 limiting RTPTS value for circumferentially oriented welds at 54 EFPY is 101.8°F (see Table D-1), which corresponds to the Intermediate to Upper Shell Girth Weld Seam 103-121, Heat # 4P5174 (1122) (using Position 1.1).

Therefore, all of the beltline and extended beltline materials in the Beaver Valley Unit 2 reactor vessel are below the RTPTS screening criteria of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through EOLE (54 EFPY).

D.2 EMERGENCY RESPONSE GUIDELINE LIMITS The Emergency Response Guideline (ERG) limits were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event [D-3]. Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.

The highest value of RTNDT for which the generic category ERG limits were developed is 250°F for a longitudinal flaw and 300°F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250°F for a longitudinal flaw or 300°F for a circumferential flaw, plant-specific ERG P-T limits must be developed.

The ERG category is determined by the magnitude of the limiting RTNDT value, which is calculated the same way as the RTPTS values are calculated in Section D.1 of this report. The material with the highest RTPTS defines the limiting material. Table D-2 identifies ERG category limits and the limiting material RTNDT values at 54 EFPY for Beaver Valley Unit 2.

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Westinghouse Non-Proprietary Class 3 D-6 Table D-2 Evaluation of Beaver Valley Unit 2 ERG Limit Category ERG Pressure-Temperature Limits [D-3]

Applicable RTNDT Value(a) ERG P-T Limit Category RTNDT < 200°F Category I 200°F < RTNDT < 250°F Category II 250°F < RTNDT < 300°F Category III b Limiting RTNDT Value (b)

Reactor Vessel Material RTNDT Value @ 54 EFPY Intermediate Shell Plate B9004-2 using Non-credible Beaver Valley Unit 2 Surveillance 163.1°F Data (Position 2.1)

Notes:

(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.

(b) Value taken from Table D-1.

Per the ERG limit guidance document [D-3], some vessels do not change categories for operation through the end of license. However, when a vessel does change ERG categories between the beginning and end of operation, a plant-specific assessment must be performed to determine at what operating time the category changes. Thus, the ERG classification need not be changed until the operating cycle during which the maximum vessel value of actual or estimated real-time RTNDT exceeds the limit on its current ERG category.

Per Table D-2, the limiting material for Beaver Valley Unit 2 (Intermediate Shell Plate B9004-2 [Position 2.1]) has an RTNDT less than 200°F through 54 EFPY. Therefore, Beaver Valley Unit 2 remains in ERG Category I through EOLE (54 EFPY).

Conclusion of ERG P-T Limit Categorization As summarized above, Beaver Valley Unit 2 will remain in ERG Category I through EOLE (54 EFPY).

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Westinghouse Non-Proprietary Class 3 D-7 D.3 REFERENCES D-1 Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Federal Register, November 29, 2019.

D-2 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

D-3 Westinghouse Owners Group Document HF04BG, Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 3, March 2014.

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Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E LTOP ENABLE TEMPERATURE ASME Code Case N-641 [0] presents alternative procedures for calculating pressure-temperature relationships and low temperature overpressure protection (LTOP) system effective temperatures, Te, and allowable pressures. The procedures provided in Code Case N-641 take into account alternative fracture toughness properties, circumferential and axial reference flaws, and plant-specific LTOP effective temperature calculations.

Per ASME Code Case N-641, the LTOP system shall be effective below the higher temperature determined in accordance with (1) and (2) below. Alternatively, LTOP systems shall be effective below the higher temperature determined in accordance with (1) and (3) below.

(1) a coolant temperature(a) of 200°F (2) a coolant temperature(a) corresponding to a reactor vessel metal temperature(b), for all vessel beltline materials, where Te is defined for inside axial surface flaws as RTNDT + 40°F, and Te is defined for inside circumferential surface flaws as RTNDT - 85°F.

(3) a coolant temperature(a) corresponding to a reactor vessel metal temperature(b), for all vessel beltline materials, where Te is calculated on a plant-specific basis for axial and circumferential reference flaws using the following equation:

Te = RTNDT + 50 ln [((F

Mm (pRi / t)) - 33.2) / 20.734]

Where, F = 1.1, accumulation factor for safety relief valves Mm = the value of Mm determined in accordance with G-2214.1, in.

p= vessel design pressure, ksig Ri = vessel inner radius, in.

t= vessel wall thickness, in.

Notes:

(a) The coolant temperature is the reactor coolant inlet temperature.

(b) The vessel metal temperature is the temperature at a distance one-fourth of the vessel section thickness from the clad/base metal interface in the vessel beltline region. RTNDT is the highest adjusted reference temperature (for weld or base metal in the beltline region) at a distance one-fourth of the vessel section thickness from the vessel clad/base metal interface as determined by Regulatory Guide 1.99, Revision 2 [0].

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Westinghouse Non-Proprietary Class 3 E-2 Using the ASME Code Case N-641 equations and the following inputs, the Beaver Valley Unit 2 LTOP system minimum enable temperatures using Cases 2 and 3 were determined for heatup and cooldown.

RTNDT = 158°F (see Table 8-1)

F = 1.1 Mm = 2.599 in.

p = 2.485 ksig Ri = 78.66 inches t =7.875 inches The LTOP system shall be effective below the higher temperature determined in accordance with (1) and (2) above, which has been determined to be a coolant temperature of 219°F for 54 EFPY. Alternatively, LTOP systems shall be effective below the higher temperature determined in accordance with (1) and (3) above, which has been determined to be a coolant temperature of 209°F for 54 EFPY. Therefore, the cooldown and heatup minimum required enable temperatures for the Beaver Valley Unit 2 reactor vessel (without uncertainties) is a coolant temperature of 209°F for 54 EFPY. Since an instrument uncertainty of 17°F is to be included, the minimum required enable temperatures for the Beaver Valley Unit 2 reactor vessel (with margins for instrument uncertainty) is a coolant temperature of 226°F for 54 EFPY.

E.1 REFERENCES E-1 ASME Code Case N-641, Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1, ASME International, January 17, 2000.

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

[ADAMS Accession Number ML003740284].

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v t e Letter Sequence Response to RAI
EPID:L-2020-LLA-0233, (Approved, Closed)
Initiation Acceptance Supplement Administration Questions 9 March 2021 Answers 22 April 2021 Results Approval, Approval Other: L-20-274, Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits, ML21320A359
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