View source for L-21-106, Response to Request for Additional Information Regarding Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits

{{Adams
| number = ML21113A044
| issue date = 04/22/2021
| title = Response to Request for Additional Information Regarding Amendment Request to Update Analytical Methods Used to Develop Reactor Coolant System Pressure and Temperature Limits
| author name = Grabnar J
| author affiliation = Energy Harbor Nuclear Corp
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000334, 05000412
| license number = DPR-066, NPF-073
| contact person = 
| case reference number = EPID L-2020-LLA-0233, L-21-106
| document type = Letter type:L, Response to Request for Additional Information (RAI)
| page count = 222
| project = EPID:L-2020-LLA-0233
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter: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 &deg;F at 1/4T and 209.5&deg;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
                                        &#xa9; 2021 Westinghouse Electric Company LLC All Rights Reserved
 *** 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&deg;F equal to 0&deg;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&deg;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&deg;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&deg;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&deg;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&deg;F/hr, and cooldown rates of 0, -20, -40, -60, and -100&deg;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&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 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&deg;, 15&deg;, 30&deg;, and 45&deg; 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.
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 WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                   March 2021 Revision 2
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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&deg;             15&deg;              30&deg;           45&deg;         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.
WCAP-18102-NP                                                                                                   March 2021 Revision 2
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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&deg;             15&deg;              30&deg;           45&deg;         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.
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                                            2-8 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&deg;            15&deg;            30&deg;          45&deg;         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&deg; Intermediate            45&deg; 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&deg; Intermediate            45&deg; 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&deg;)       (25&deg;)        (25&deg;)         (25&deg;)        (15&deg;)      (45&deg;/25&deg;/15&deg;)  (35&deg;/25&deg;)      (35&deg;/15&deg;)
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&deg; location after Cycle 19 and to a 15&deg; location after Cycle 22.
(b) Capsule T was moved to a 25&deg; location after Cycle 10.
(c) Capsule Z was moved to a 15&deg; 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&deg;)                                      Withdrawn EOC 1                                     1.47 U (65&deg;)                                      Withdrawn EOC 4                                     1.00 W (245&deg;)                                      Withdrawn EOC 6                                     1.05 Y (295&deg;)                                     Withdrawn EOC 13                                     1.14 X (285&deg;)                                     Withdrawn EOC 22                                     1.57 S (45&deg;/295&deg;/285&deg;)    (a)
In Reactor                                    0.74(d)
T (55&deg;/65&deg;)(b)                                       In Reactor                                    0.94(d)
Z (305&deg;/165&deg;)(c)                                      In Reactor                                    1.20(d)
Notes:
(a) Capsule S was moved to the 295&deg; location after Cycle 19 and to the 285&deg; location after Cycle 22.
(b) Capsule T was moved to the 65&deg; location after Cycle 10.
(c) Capsule Z was moved to the 165&deg; 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%
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                                              2-18
                          - 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
*** 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                                              2-19
                          - 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
*** 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                                      2-20 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
N I
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
*** 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                                            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.
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                                               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.
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                                            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              (&deg;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
 *** 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                                                 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               (&deg;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
(&deg;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].
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                                            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.
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                                                              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) (&deg;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].
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                                                                     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) (&deg;F)
(&deg;F)                (&deg;F) 97&deg;                         0.5174                               72.34                            541                 -1.7 St. Lucie Unit 1 104&deg;                        0.7885                                67.4                           544.6                 1.9 Data 284&deg;                        1.243                                 68.0                           546.3                 3.6 97&deg;                         0.324                                65.93                           544.3                 1.6 Millstone Unit 2 104&deg;                        0.949                                52.12                           547.6                 4.9 Data 83&deg;                         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&deg;F for Beaver Valley Unit 1. 542.7&deg;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.
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                                            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.
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                                              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                                                                                (&deg;F)                (&deg;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) &#xf7; (FF2) = (1585.86) &#xf7; (11.154) = 142.2&deg;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                                                                              (&deg;F)                (&deg;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) &#xf7; (FF ) = (1039.87) &#xf7; (5.577) = 186.5&deg;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&deg;F / 181.6&deg;F = 1.06.
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                                               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)                            (&deg;F)                 (&deg;F) 97&deg;                        0.5174                 0.8160       82.6 (72.34)            67.44          0.666 St. Lucie Unit 1 104&deg;                       0.7885                 0.9333        81.1 (67.4)            75.68          0.871 Data 284&deg;                       1.243                  1.0606        83.8 (68.0)            88.85          1.125 97&deg;                        0.324                  0.6902       67.5 (65.93)            46.61          0.476 Millstone Unit 2 104&deg;                       0.949                  0.9853       57.0 (52.12)            56.18          0.971 Data(d) 83&deg;                         1.74                  1.1523       61.4 (56.09)            70.74          1.328 SUM:          405.50          5.437 CFWeld Heat # 90136 = (FF
* RTNDT) &#xf7; (FF ) = (405.50) &#xf7; (5.437) = 74.6&deg;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&deg;F / 106.6&deg;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
 *** 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                                            5-4 Table 5-4          Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material                                                             Chemistry Factor (&deg;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 Welds 20-714 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
    *** 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                                              5-5 Table 5-4          Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material                                                             Chemistry Factor (&deg;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
       *** 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                                            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 &deg;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 &deg;F/hr.
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                                            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&deg;F and 0.379 ft2/hr at 550&deg;F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-&deg;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.
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                                            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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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&deg;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&deg;F is associated with the vessel flange of the Beaver Valley Unit 1 vessel, so the minimum allowable temperature of this region is 130&deg;F at pressures greater than 621 psig (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.
WCAP-18102-NP                                                                                                    March 2021 Revision 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&deg;F when the initial RTNDT is a measured value, and 17&deg;F when a generic value is available. The standard deviation for the RTNDT margin term, , is 17&deg;F for plates or forgings when surveillance data is not used or is non-credible, and 8.5&deg;F (half the value) for plates or forgings when credible surveillance data is used.
For welds,  is equal to 28&deg;F when surveillance capsule data is not used or is non-credible, and is 14&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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&deg;F. The increased limiting ART values, using the Axial Flaw methodology, for Lower Shell Plate B6903-1 are summarized in Table 7-4.
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                                                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.
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                                                                 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                                             (&deg;F)     (n/cm2, E > 1.0 MeV)          FF           (&deg;F)          (&deg;F)       (&deg;F)       (&deg;F)   (&deg;F)       (&deg;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 Welds 19-714 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
                           *** 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                                                                            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                                                (&deg;F)    (n/cm2, E > 1.0 MeV)            FF             (&deg;F)             (&deg;F)      (&deg;F)     (&deg;F)       (&deg;F)          (&deg;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&deg;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&deg;F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal  = 28&deg;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&deg;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.
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                                                                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                                            (&deg;F)     (n/cm2, E > 1.0 MeV)                         (&deg;F)         (&deg;F)        (&deg;F)      (&deg;F)   (&deg;F)       (&deg;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 Welds 19-714 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
                           *** 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                                                                           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                                                (&deg;F)    (n/cm2, E > 1.0 MeV)                             (&deg;F)            (&deg;F)       (&deg;F)     (&deg;F)       (&deg;F)          (&deg;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&deg;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&deg;F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal  = 28&deg;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&deg;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
                             *** 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                                             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&deg;F                                                        209.5&deg;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&deg;F to add additional conservatism.
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                                            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&deg;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&deg;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&deg;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&deg;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&deg;F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
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                                            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.
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                                            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&deg;F (Axial Flaw) 3/4T, 209.5&deg;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
          &sect;,                                                                              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&deg;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&deg;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
 *** 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                                    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&deg;F (Axial Flaw) 3/4T, 209.5&deg;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
                                                                                         &#xba;F/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&deg;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)
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                                            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&deg;F/hr Heatup            60&deg;F/hr Criticality         100&deg;F/hr Heatup         100&deg;F/hr Criticality T (&deg;F)        P (psig)      T (&deg;F)      P (psig)       T (&deg;F)     P (psig)       T (&deg;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
*** 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                                            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&deg;F/hr Heatup            60&deg;F/hr Criticality         100&deg;F/hr Heatup         100&deg;F/hr Criticality T (&deg;F)        P (psig)      T (&deg;F)      P (psig)       T (&deg;F)     P (psig)       T (&deg;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 (&deg;F)                                             P (psig) 283                                                 2000 301                                                 2485 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                                            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&deg;F/hr.              -40&deg;F/hr.              -60&deg;F/hr.            -100&deg;F/hr.
T (&deg;F)     P (psig)     T (&deg;F)     P (psig)    T (&deg;F)     P (psig)    T (&deg;F)    P (psig)   T (&deg;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
*** 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                                            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&deg;F/hr.              -40&deg;F/hr.              -60&deg;F/hr.            -100&deg;F/hr.
T (&deg;F)     P (psig)     T (&deg;F)     P (psig)    T (&deg;F)     P (psig)    T (&deg;F)    P (psig)   T (&deg;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
*** 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                                            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.
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                                            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.
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                                            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
 *** 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                                            A-2 Table A-1         KIt Values for Beaver Valley Unit 1 at 50 EFPY 100&deg;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
(&deg;F)        100&deg;F/hr Heatup (&deg;F)               (ksi in.)          100&deg;F/hr Heatup (&deg;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
*** 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                                            A-3 Table A-2         KIt Values for Beaver Valley Unit 1 at 50 EFPY 100&deg;F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors)
Water        Vessel Temperature at 1/4T              -100&deg;F/hr Cooldown Temp.           Location for -100&deg;F/hr               1/4T Thermal Stress
(&deg;F)              Cooldown (&deg;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&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
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                                                                  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)       (&deg;F)                                                (&deg;F)         (&deg;F)        (&deg;F)        (&deg;F)    (&deg;F)        (&deg;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&deg;F for Position 1.1 without surveillance data. However,  need not exceed 0.5*RTNDT.
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                                            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                          (&deg;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]
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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&deg;F for welds and 17&deg;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&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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 (&deg;F)(b) 97&deg;                                            541 Weld Metal Heat # 90136 104&deg;        0.23        0.07                  544.6 (St. Lucie Unit 1 Data) 284&deg;                                          546.3 97&deg;                                          544.3 Weld Metal Heat # 90136 104&deg;        0.30        0.06                  547.6 (Millstone Unit 2 Data) 83&deg;                                          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&deg;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&deg;F     (from Table 5-4)
Ratio = 121.2 &#xf7; 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&deg;F CFSurv. Weld (Millstone Unit 2)   =      135.5&deg;F (from Table 5-4)
Ratio = 121.2 &#xf7; 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.
WCAP-18102-NP                                                                                                    March 2021 Revision 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 (&deg;F)   Temperature (&deg;F)          Adjustment (&deg;F) 97&deg;                         541                                                 -4.3 Weld Metal Heat # 90136 104&deg;                        544.6                                                -0.7 (St. Lucie Unit 1 Data) 284&deg;                        546.3                                                 1.0 545.3 97&deg;                        544.3                                                -1.0 Weld Metal Heat # 90136 104&deg;                        547.6                                                 2.3 (Millstone Unit 2 Data) 83&deg;                        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
(&deg;F)                 (&deg;F)
Weld Metal Heat #            97&deg;                            0.5174               0.8160     77.6 (72.34)            63.29           0.666 90136 (St. Lucie            104&deg;                           0.7885               0.9333     76.0 (67.4)             70.97           0.871 Unit 1 Data)              284&deg;                           1.243                1.0606     78.7 (68.0)             83.43           1.125 Weld Metal Heat #            97&deg;                            0.324                0.6902     57.8 (65.93)            39.89           0.476 90136 (Millstone             104&deg;                           0.949                0.9853     48.4 (52.12)            47.72           0.971 Unit 2 Data)              83&deg;                             1.74                1.1523     52.3 (56.09)            60.29           1.328 SUM:           365.59          5.437 CFHeat # 90136 = (FF
* RTNDT) &#xf7; (FF2) = (365.59) &#xf7; (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.
WCAP-18102-NP                                                                                                        March 2021 Revision 2
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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&deg;F (Weld)
(&deg;F)          E > 1.0 MeV)                     (&deg;F)         (&deg;F)         (&deg;F)
Weld Metal Heat #         97&deg;            67.2               0.5174        0.8160        77.6         54.8         22.7              Yes 90136 (St. Lucie        104&deg;            67.2               0.7885        0.9333        76.0         62.7         13.3              Yes Unit 1 Data)          284&deg;            67.2               1.243         1.0606        78.7         71.3          7.4              Yes Weld Metal Heat #         97&deg;            67.2               0.324         0.6902        57.8         46.4         11.4              Yes 90136 (Millstone         104&deg;            67.2               0.949         0.9853        48.4         66.2         17.8              Yes Unit 2 Data)           83&deg;            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&deg;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&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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
(&deg;F)               (&deg;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) &#xf7; (FF2) = (1585.86) &#xf7; (11.154) = 142.2&deg;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) &#xf7; (FF ) = (981.01) &#xf7; (5.577) = 175.9&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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&deg;F Fluence Material           Capsule       (Slopebest-fit)                       FF       RTNDT        RTNDT      RTNDT         (Base Metal)
(x 1019 n/cm2,
(&deg;F)                                         (&deg;F)          (&deg;F)        (&deg;F)        <28&deg;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&deg;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&deg;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].
WCAP-18102-NP                                                                                                    March 2021 Revision 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]
WCAP-18102-NP                                                                                                   March 2021 Revision 2
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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 Welds 19-714 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
 *** 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                                            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).
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                                                              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                                          (&deg;F)     (n/cm2, E > 1.0 MeV)                      (&deg;F)         (&deg;F)     (&deg;F)       (&deg;F)  (&deg;F)      (&deg;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 Welds 19-714 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                                            (&deg;F)     (n/cm2, E > 1.0 MeV)                         (&deg;F)          (&deg;F)       (&deg;F)     (&deg;F)       (&deg;F)         (&deg;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&deg;F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal  = 28&deg;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&deg;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.
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                                            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&deg;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&deg;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&deg;F for base metal and/or longitudinal welds, and 300&deg;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&deg;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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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&deg;          End of Cycle 1              1.2               2.97E+18                4.93E-03 U                 25&deg;          End of Cycle 4              3.6               6.18E+18                1.00E-02 W                 25&deg;          End of Cycle 6              5.9               9.52E+18                1.54E-02 Y                 25&deg;          End of Cycle 13            14.3               2.10E+19                3.40E-02 X                 15&deg;          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:
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
WCAP-18102-NP                                                                                                    March 2021 Revision 2
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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.
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                                            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
 *** 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                                            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.
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                                               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.
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                                          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
*** 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                                          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
*** 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                                          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
*** 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                                          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
*** 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                                          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
*** 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                                          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
*** 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                                          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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                         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
*** 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                                             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
 *** 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                                          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&deg;                                        1.47               1.2                    2.97E+18 (EOC 1)
Withdrawn U                   65&deg;                                        1.00               3.6                    6.18E+18 (EOC 4)
Withdrawn W                   245&deg;                                        1.05               5.9                    9.52E+18 (EOC 6)
Withdrawn Y                  295&deg;                                        1.14              14.3                    2.10E+19 (EOC 13)
Withdrawn X                  285&deg;                                        1.57              26.6                    4.99E+19 (EOC 22) 285&deg; S(d)                                   In Reactor              0.74(d)           Note (d)                 2.58E+19(d)
(45&deg;/295&deg;)
T(e)               65&deg; (55&deg;)           In Reactor              0.94(e)           Note (e)                 3.28E+19(e)
Z(f)             165&deg; (305&deg;)           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.
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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&deg;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&deg;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&deg; Neutron Pad Configuration ............................................................................................ 2-19 Figure 2-3          Beaver Valley Unit 2 Plan View of the Reactor Geometry at the Core Midplane 26&deg; 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&deg;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&deg;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&deg;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&deg;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&deg;F/hr, and cooldown rates of 0 (steady-state), 20, 40, 60, and 100&deg;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.
WCAP-18559-NP                                                                                                    March 2021 Revision 1
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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.
WCAP-18559-NP                                                                                                    March 2021 Revision 1
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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&deg;, 107&deg;, 110&deg;, 287&deg;, 290&deg;, and 340&deg;, respectively. These full-core positions correspond to the following octant symmetric locations represented in Figure 2-3: 17&deg; from the core cardinal axes (for the 107&deg;, 287&deg; WCAP-18559-NP                                                                                                    March 2021 Revision 1
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Westinghouse Non-Proprietary Class 3                                            2-2 and 343&deg; dual surveillance capsule holder locations found in octants with a 26&deg; neutron pad segment) and 20&deg; from the core cardinal axes (for the 110&deg;, 290&deg; and 340&deg; dual surveillance capsule holder locations found in octants with a 26&deg; 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&deg; and 20&deg;, 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&deg;, 15&deg;, 30&deg;, and 45&deg;, 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.
WCAP-18559-NP                                                                                                    March 2021 Revision 1
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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.
WCAP-18559-NP                                                                                                    March 2021 Revision 1
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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.
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                                            2-6 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.
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                                            2-7 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
*** 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                                            2-8 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
*** 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                                            2-9 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&deg;              15&deg;              30&deg;              45&deg;        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&deg;              15&deg;              30&deg;              45&deg;        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&deg;              15&deg;              30&deg;              45&deg;        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&deg;              15&deg;               30&deg;             45&deg;        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&deg; 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&deg; 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.
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                                           3-2 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                         (&deg;F)            (ft-lb)
(&deg;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
           *** 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                                            3-3 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                         (&deg;F)            (ft-lb)
(&deg;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&deg;F.
(c) The generic RTNDT(U) values were determined in accordance with NUREG-0800 [7] and 10 CFR 50.61 [10] ; thus, I = 17&deg;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].
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                                            3-4 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
(&deg;F)
Closure Head Flange                            -10 Vessel Flange                                0 Notes:
(a) Values taken from WCAP-16528-NP [12].
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                                            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.
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                                            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 (&deg;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].
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                                            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.
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                                            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
(&deg;F)             (&deg;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) &#xf7; (FF2) = (1034.90) &#xf7; (16.779) = 61.7&deg;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) &#xf7; (FF2) = (134.68) &#xf7; (8.389) = 16.1&deg;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&deg;F / 38.0&deg;F). This would result in a lower calculated Position 2.1 CF.
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                                            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 (&deg;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                       ---
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                                            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 (&deg;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.
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                                            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
 *** 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                                            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 &deg;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 &deg;F/hr.
WCAP-18559-NP                                                                                                    March 2021 Revision 1
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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&deg;F and 0.379 ft2/hr at 550&deg;F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-&deg;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.
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                                            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.
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                                            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&deg;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&deg;F is associated with the vessel flange of the Beaver Valley Unit 2 vessel, so the minimum allowable temperature of this region is 120&deg;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&deg;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&deg;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&deg;F per Table 3-2, the minimum boltup temperature for the Beaver Valley Unit 2 reactor vessel is 60&deg;F without margins for instrument uncertainties. This limit is shown in 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                                            6-6 Figures 8-1 and 8-2. The minimum boltup temperature with margins for instrument uncertainty is 77&deg;F.
This limit is shown in Figures 8-3 and 8-4.
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                                            7-1 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&deg;F when the initial RTNDT is a measured value, and 17&deg;F when a generic value is available. The standard deviation for the RTNDT margin term, , is 17&deg;F for plates or forgings when surveillance data is not used or is non-credible, and 8.5&deg;F (half the value) for plates or forgings when credible surveillance data is used.
For welds,  is equal to 28&deg;F when surveillance capsule data is not used or is non-credible, and is 14&deg;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
 *** 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                                           7-3 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
     *** 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                                                              7-4 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       (&deg;F)    (n/cm , E > 1.0 MeV) 2 FF(b)         (&deg;F)         (&deg;F)          (&deg;F) (&deg;F)    (&deg;F)       (&deg;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
                               *** 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                                                              7-5 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       (&deg;F)    (n/cm2, E > 1.0 MeV)         FF(b)         (&deg;F)         (&deg;F)          (&deg;F) (&deg;F)    (&deg;F)       (&deg;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       (&deg;F)     (n/cm2, E > 1.0 MeV)        FF(b)          (&deg;F)           (&deg;F)        (&deg;F)     (&deg;F)      (&deg;F)        (&deg;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&deg;F for Position 1.1 or Position 2.1 with non-credible surveillance data and 
       = 8.5&deg;F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2 [1], the weld metal  = 28&deg;F for Position 1.1, and the weld metal  = 14&deg;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      (&deg;F)     (n/cm , E > 1.0 MeV) 2 FF(b)         (&deg;F)          (&deg;F)          (&deg;F) (&deg;F)    (&deg;F)       (&deg;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
                              *** 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                                                               7-8 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      (&deg;F)     (n/cm2, E > 1.0 MeV)        FF(b)         (&deg;F)          (&deg;F)          (&deg;F) (&deg;F)    (&deg;F)       (&deg;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      (&deg;F)     (n/cm2, E > 1.0 MeV)         FF(b)          (&deg;F)           (&deg;F)        (&deg;F)     (&deg;F)      (&deg;F)        (&deg;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&deg;F for Position 1.1 or Position 2.1 with non-credible surveillance data and 
        = 8.5&deg;F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2 [1], the weld metal  = 28&deg;F for Position 1.1, and the weld metal  = 14&deg;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                                            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&deg;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&deg;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&deg;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&deg;F and 500&deg;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&deg;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&deg;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&deg;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&deg;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&deg;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
 *** 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                                            8-2
: 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&deg;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&deg;F without uncertainties and 234&deg;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&deg;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&deg;F                         143&deg;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&deg;F (Axial Flaw) 3/4T,     143&deg;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&deg;F/Hr Heatup Rate 60&deg;F/Hr 1500                                                      Critical Limit Calculated Pressure (PSIG) 100&deg;F/Hr Heatup Rate 100&deg;F/Hr 1250 Acceptable Operation 1000 750 500                                                 Criticality Limit based on inservice hydrostatic test temperature (214&#xba;F) 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&deg;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&deg;F (Axial Flaw) 3/4T,    143&deg;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&deg;F/hr
                                                                -20&deg;F/hr
                                                                -40&deg;F/hr 500                               -60&deg;F/hr
                                                                -100&deg;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&deg;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&deg;F/hr Heatup             60&deg;F/hr Criticality         100&deg;F/hr Heatup          100&deg;F/hr Criticality T (&deg;F)        P (psig)      T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;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&deg;F/hr Heatup             60&deg;F/hr Criticality         100&deg;F/hr Heatup          100&deg;F/hr Criticality T (&deg;F)        P (psig)      T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;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 (&deg;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&deg;F/hr Cooldown          40&deg;F/hr Cooldown          60&deg;F/hr Cooldown         100&deg;F/hr Cooldown T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;F)      P (psig)      T (&deg;F)      P (psig)     T (&deg;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&deg;F/hr Cooldown          40&deg;F/hr Cooldown          60&deg;F/hr Cooldown         100&deg;F/hr Cooldown T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;F)      P (psig)      T (&deg;F)      P (psig)     T (&deg;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&deg;F (Axial Flaw) 3/4T,     143&deg;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&deg;F/Hr Heatup Rate 60&deg;F/Hr 1500                                                         Critical Limit Calculated Pressure (PSIG) 100&deg;F/Hr Heatup Rate 100&deg;F/Hr 1250 Acceptable Operation 1000 750 500                                                   Criticality Limit based on inservice hydrostatic test temperature (234&#xba;F) 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&deg;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&deg;F (Axial Flaw) 3/4T,    143&deg;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&deg;F/hr
                                                                 -20&deg;F/hr
                                                                 -40&deg;F/hr 250                              -60&deg;F/hr
                                                                 -100&deg;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&deg;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&deg;F/hr Heatup             60&deg;F/hr Criticality         100&deg;F/hr Heatup          100&deg;F/hr Criticality T (&deg;F)        P (psig)      T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;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&deg;F/hr Heatup             60&deg;F/hr Criticality         100&deg;F/hr Heatup          100&deg;F/hr Criticality T (&deg;F)        P (psig)      T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;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 (&deg;F)                                               P (psig) 218                                                  2000 234                                                  2485 Note:
(a) The error includes an adjustment of 17&deg;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&deg;F and 62 psid if greater than 137&deg;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&deg;F/hr Cooldown          40&deg;F/hr Cooldown          60&deg;F/hr Cooldown         100&deg;F/hr Cooldown T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;F)      P (psig)      T (&deg;F)      P (psig)     T (&deg;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&deg;F/hr Cooldown          40&deg;F/hr Cooldown          60&deg;F/hr Cooldown         100&deg;F/hr Cooldown T (&deg;F)       P (psig)       T (&deg;F)      P (psig)      T (&deg;F)      P (psig)      T (&deg;F)      P (psig)     T (&deg;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&deg;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&deg;F and 62 psid if greater than or equal to 500&deg;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&deg;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
(&deg;F)        100&deg;F/hr Heatup (&deg;F)               (ksi in.)          100&deg;F/hr Heatup (&deg;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&deg;F/hr Cooldown Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors)
Water               Vessel Temperature at 1/4T              100&deg;F/hr Cooldown Temp.                  Location for 100&deg;F/hr                1/4T Thermal Stress
(&deg;F)                      Cooldown (&deg;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.
WCAP-18559-NP                                                                                                    March 2021 Revision 1
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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).
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                                            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))
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                                              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.
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                                                                                           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
                          *** 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                                            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.
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                                            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.
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                                                                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,
(&deg;F)                                 FF(b)        (&deg;F)           (&deg;F)      (&deg;F)       (&deg;F)    (&deg;F)      (&deg;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)
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                                                                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,
(&deg;F)                                 FF(b)        (&deg;F)           (&deg;F)      (&deg;F)       (&deg;F)    (&deg;F)      (&deg;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)
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                                                                      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,
(&deg;F)                                   FF(b)          (&deg;F)             (&deg;F)        (&deg;F)    (&deg;F)      (&deg;F)         (&deg;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&deg;F when surveillance data is non-credible or not used to determine the CF, and the weld metal  = 28&deg;F when surveillance data is not used to determine the CF and  = 14&deg;F when credible surveillance data is used to determine the CF. However,  need not exceed 0.5*RTNDT per regulatory guidance in [D-1].
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                                            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&deg;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&deg;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&deg;F for base metal and/or longitudinal welds, and 300&deg;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&deg;F for a longitudinal flaw and 300&deg;F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250&deg;F for a longitudinal flaw or 300&deg;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.
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                                           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&deg;F                                 Category I 200&deg;F < RTNDT < 250&deg;F                              Category II 250&deg;F < RTNDT < 300&deg;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&deg;F Data (Position 2.1)
Notes:
(a) Longitudinally oriented flaws are applicable only up to 250&deg;F; circumferentially oriented flaws are applicable up to 300&deg;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&deg;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).
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                                            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.
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                                            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&deg;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&deg;F, and Te is defined for inside circumferential surface flaws as RTNDT - 85&deg;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].
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                                            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&deg;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&deg;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&deg;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&deg;F for 54 EFPY. Since an instrument uncertainty of 17&deg;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&deg;F for 54 EFPY.
E.1        REFERENCES E-1        ASME Code Case N-641, Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements Section 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].
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)}}